Text

Forwards Proof & Review ABWR TS & Bases for Sections 3.4 RCS,3.5 ECCS & 5.0 Administrative Controls (No Bases)
	ML20056D479
Person / Time
Site:	05200001
Issue date:	07/23/1993
From:	Crutchfield D; Office of Nuclear Reactor Regulation
To:	Marriott P; GENERAL ELECTRIC CO.
References
	NUDOCS 9308160300
	Download: ML20056D479 (172)
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NUCLEAR REGULATORY COMMISSION

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July 23,1993 Docket No.52-001 Mr. Patrick W. Marriott, Manager Licensing & Consulting Services GE Nuclear Energy 175 Curtner Avenue San Jose, California 95125

Dear Mr. Harriott:

SUBJECT:

ADVANCED BOILING WATER REACTOR (ABWR) TECHNICAL SPECIFICATIONS

~

Enclosed are the proof and review ABWR technical specifications and their bases for the following sections:

3.4 Reactor Coolant Systems 3.5 ECCS 5.0 Administrative Controls (No Bases)

As we discussed at our management meeting on June 10, 1993, the Nuclear Regulatory Commission (NRC) staff will be providing GE Nuclear Energy (GE),

until August 31, selected sections of proof and review ABWR technical specifi-cations. These sections are based on the NRC staff review of the GE mark-up of the BWR-6 and BWR-4 Standard Technical Specifications; the sections, as provided, are acceptable to the NRC staff. As discussed, we anticipate that

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GE will interface very closely with the staff to resolve any issues on these sections prior to August 31, 1993. Under this arrangement, we anticipate that formal comments to proof and review ABWR technical specifications made by Sectember 20, 1993, will be few.

The electronic text of these sections is available on the NRC Technical Specifications Branch electronic bulletin board (OTSB-BBS) in Wordperfect 5.1 format. The data telephone number for the OTSB-BBS is (301) 504-1778, and the system operator is Tom Dunning who is available for assistance at (301) 504-1189. Also, in accordance with our agreements, GE will maintain these sections in Wordperfect 5.1 format and will produce subsequent issues of the ABWR technical specifications in Wordperfect 5.1 format.

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Mr. Patrick W. Marriott July 23, 1993 If you have any questions about technical specifications please contact Mark Reinhart with the Nuclear Reactor Regulation Technical Specifications Branch.

He may be reached at (301) 504-1185.

i Sincerely, (Original signed by)

Dennis M. Crutchfield, Associate Director for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation

Enclosure:

As stated cc w/ enclosure:

See next page DISTRIBUTION:

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OFC:

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DATE:

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CPostusny:sg JNWi 07/%$@3 07/N93 07/)f/93 DATE:

0FFICIAL RECORD COPY:

DOCUMENT NAME:

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Mr. Patrick W. Marriott July 23, 1993 If you have any questions about technical specifications please contact Mark Reinhart with the Nuclear Raactor Regulation Technical Specifications Branch.

He may be reached at (301) 504-1185.

Sincerely, c5f $.

Dennis M. Crutch te d, As ciate Director for Advanced Reactors nd License Renewal Office of Nuclear Reactor Regulation

Enclosure:

As stated cc w/ enclosure:

See next page 0

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Mr. Patrick W. Marriott Docket No.52-001 General Electric Company cc:

Mr. Robert Mitchell Mr. Joseph Quirk General Electric Company GE Nuclear Energy 175 Curtner Avenue General Electric Company San Jose, California 95125 175 Curtner Avenue, Mail Code 782 San Jose, California 95125 Mr. L. Gifford, Program Manager Regulatory Programs Mr. Raymond Ng GE Nuclear Energy 1776 Eye Street, N.W.

12300 Twinbrook Parkway Suite 300 Suite 315 Washir.gton, D.C.

20006 Rockville, Maryland 20852 Director, Criteria & Standards Division 4

Office of Radiation Programs U.S. Environmental Protection Agency 401 M Street, S.W.

Washington, D.C.

20460 j

Mr. Sterling Franks U.S. Department of Energy NE-42 Washington, D.C.

20585 Marcus A. Rowden, Esq.

Fried, Frank, Harris, Shriver & Jacobson 1001 Pennsylvania Avenue, N.W.

Suite 800 Washington, D.C.

20004 l

Jay M. Gutierrez, Esq.

Newman & Holtzinger, P.C.

1615 L Street, N.W.

Suite 1000 I

Washington, D.C.

20036

)

Mr. Steve Goldberg Budget Examiner 725 17th Street, N.W.

Room 8002 Washington, D.C.

20503 Mr. Frank A. Ross U.S. Department of Energy, NE-42 Office of LWR Safety and Technology 19901 Germantown Road Germantown, Maryland 20874 l

l

(

i Reactor Intcrnal Pumps 3.4.1 l

3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Reactor Internal Pumps (RIPS) - Operating LC0 3.4.1 At least nine RIPS shall be OPERABLE.

APPLICABILITY:

MODES I and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Less than nine RIPS A.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months OPERABLE.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify at least nine RIPS are OPERABLE at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months any THERMAL POWER level.

Enclosure ABWR TS 3.4-1 P&R, 07/22/93

U L

S/RVs 3.4.2 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.2 Safety / Relief Valves (S/RVs)

LC0 3.4.2 The safety function of [seven] S/RVs shall be OPERABLE, AND The relief function of [seven] additional S/RVs shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One [ required] S/RV A.1 Restore [requirea]

14 days inoperable.

S/RV to OPERABLE status.

B. Required Attic and B.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Con.ietion e

Time of Condition A AND not met.

B.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months 08

[Two] or more

[ required] S/RVs inoperable.

ABWR TS 3.4-1 P&R, 07/22/93

b e

S/RVs 3.4.2 i

i SURVEILLANCE REQUIREMENTS l

SURVEILLANCE FREQUENCY SR 3.4.2.1 Verify the safety function lift setpoints

[In accordance of the [ required] S/RVs are as follows:

with the Inservice Number of Setpoint Testing Program S/RVs (osia) or [18] months]

[8]

[1165 i 34.9]

[6]

[1180 1 35.4]

[6]

[1190 1 35.7)

Following testing, lift settings shall be within i 1%.

SR 3.4.2.2

NOTE--------------------

f Valve actuation may be excluded.

Verify each [ required] relief function S/RV

[18] months actuates on an actual or simulated automatic initiation signal.

SR 3.4.2.3

NOTE--------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam dome pressure is 2 [950] psig.

Verify each [ required] S/RV opens when

[18] months on manually actuated.

a STAGGERED TEST BASIS for-each valve solenoid l

I

+

i-l l

ABWR TS 3.4-2 P&R, 07/22/93

c-8 RCS Operational LEAKAGE 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 RCS Operational LEAKAGE LCO 3.4.3 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE; b.

s 5 gpm unidentified LEAKAGE; [and]

c.

s [30] gpm total LEAKAGE averaged over the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period; [and d.

s 2 gpm increase in unidentified LEAKAGE within the previous [4] hour period in MODE 1].

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS 1

CONDITION REQUIRED ACTION COMPLETION TIME A.

Unidentified LEAKAGE A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months i

not within limit.

within limits.

QB Total LEAKAGE no't within limit.

i i

B.

Unidentified LEAKAGE B.1 Reduced LEAKAGE to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months increase not within within limit.

t limit.

03 (continued) i ABWR TS 3.4-1 P&R, 07/22/93

RCS Operaticnal' LEAKAGE 3.4.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME p

1 B.

(continued) 8.2 Verify source of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months unidentified LEAKAGE increase is not service sensitive type 304 or type 316 austenitic stainless steel.

C.

Required Action and C.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND i

or B not met.

C.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months j

QB l

Pressure boundary LEAKAGE exists.

l 1

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY i

SR 3.4.3.1 Verify RCS unidentified and total LEAKAGE B hours and unidentified LEAKAGE increase are within limits.

ABWR TS 3.4-2 P&R, 07/22/93

V RCS PIV Leakag 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Pressure Isolation Valve (PIV) Leakage LCO 3.4.4 The leakage from each RCS PIV shall be within limit.

APPLICABILITY:

MODES I and 2, MODE 3, except valves in the residual heat removal (RHR) shutdown cooling flowpath are not required to meet the requirements of this LC0 when in the shutdown cooling mode of operation.

ACTIONS

= ---- --

=-----NOTES------------------------------------

1.

Separate Condition entry is allowed for each flow path.

2.

Enter applicable Conditions and Required Actions for systems made inoperable by PIVs.

CONDITION REQUIRED ACTION COMPLETION TIME A.

Leakage from one or

= --

NOTE-------------

more RCS PIVs not Each valve used to satisfy within limit.

Required Action A.1 and i

Required Action A.2 shall I

have been verified to meet SR 3.4.6.1 and be in the i

reactor coolant pressure boundary [or the high pressure portion of the system).

(continued) l ABWR TS 3.4-1 P&R, 07/22/5J

RCS PIV Leakage 3.4.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.I Isolate the high 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months pressure portion of the affected system from the low pressure portion by use of one closed manual, deactivated automatic, or check valve.

AND A.2 Isolate the high 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months pressure portion of

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the affected system from the low pressure portion by use of a second closed manual, deactivated automatic, or check valve.

B.

Required Action and B.I Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met.

AND B.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ABWR TS 3.4-2 P&R, 07/22/93

6 RCS PIV Leakage 3.4.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1

=

==--NOTE--------------------

Not required to be performed in MODE 3.

==

Verify equivalent leakage of each RCS PIV

[In accordance is s 0.5 gpm per nominal inch of valve size with Inservice up to a maximum of 5 gpm, at an RCS Testing Program pressure 2 (1040] psig and s [1060] psig.

or [18] months) em e

ABWR TS 3.4-3 P&R, 07/22/93

6 RCS Leakage Detection Instrumentation 3.4.5 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.5 RCS Leakage Detection Instrumentation LCO 3.4.5 The following RCS leakage detection instrumentation shall be OPERABLE:

a.

Drywell floor drain sump monitoring system; [and]

b.

One channel of either drywell atmospheric particulate or atmospheric gaseous monitoring system; [and c.

Drywell air cooler condensate flow rate monitoring system].

APPLICABILITY:

MODES I, 2, and 3.

ACTIONS

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CONDITION REQUIRED ACTION COMPLETION TIME A.

Drywell floor drain

NOTE-------------

sump monitoring system LC0 3.0.4 is not applicable.

inoperable.

A.I Restore drywell floor 30 days drain sump monitoring system to OPERABLE status.

(continued)

ABWR TS 3.4-I P&R, 07/22/93

b RCS Leakage Detection Instrumentation i

3.4.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.

Required drywell

NOTE------------

I atmospheric monitoring LCO 3.0.4 is not system inoperable.

applicable.

______==-

i B.1 Analyze grab samples Once per of drywell 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months atmosphere.

I AND B.2 Restore required 30 days drywell atmospheric monitoring system to OPERABLE scatus.

C. Drywell air cooler

NOTE-------------

condensate flow rate Not applicable when the monitoring system required drywell atmospheric inoperable.

monitoring system is inoperable.

C.1 Perform SR 3.4.7.1.

Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months u

D. Required drywell

NOTE-------------

atmospheric LCO 3.0.4 is not applicable.

monitoring system inoperable.

i D.1 Restore required 30 days AND drywell atmospheric monitoring system to I

Drywell air cooler OPERABLE status.

condensate flow rate monitoring QR system inoperable.

(continued)

ABWR TS 3.4-2 P&R, 07/22/93

6 RCS Leakage D tection Instrumentation 3.4.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. (continued) 0.2 Restore drywell air 30 days cooler condensate flow rate monitoring system to OPERABLE status.

E.

Required Action and E.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion l

Time of Condition A, AND B, [C, or D] not met.

E.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months F.

All required leakage F.1 Enter LCO 3.0.3.

Immediately

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detection systems inoperable.

SURVEILLANCE REQUIREMENTS i

SURVEILLANCE FREQUENCY l

SR 3.4.5.1 Perform CHANNEL CHECK of required drywell 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months atmospheric monitoring system.

SR 3.4.5.2 Perform CHANNEL FUNCTIONAL TEST of required 31 days leakage detection instrumentation.

SR 3.4.5.3 Perform CHANNEL CALIBRATION of required

[18] months l

leakage detection instrumentation.

1 i

l l

l ABWR TS 3.4-3 P&R, 07/22/93 L

RCS Spacific Activity 3.4.6 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.6 RCS Specific Activity LCO 3.4.6 The specific activity of the reactor coolant shall be limited to:

DOSE EQUIVALENT I-131 specific activity s [0.2) pCi/gm; a.

and b.

Gross specific activity 5 100/E pCi/gm.

APPLICABILITY:

MODE 1, MODES 2 and 3 with any main steam line not isolated.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Reactor coolant A.1 Determine DOSE Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months specific activity EQUIVALENT I-131.

> [0.2] pCi/gm and s 4.0 Ci/gm DOSE AND EQUIVALENT I-131.

A.2 Restore DOSE 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months EQUIVALENT I-131 to within limits.

i B.

Required Action and 8.1 Determine DOSE Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion EQUIVALENT I-131.

l Time of Condition A I

not met.

AND 08 B.2.1 Isolate all main 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months steam lines.

Reactor coolant specific activity QB

> [4.0] Ci/gm DOSE EQUIVALENT I-131.

(continued)

ABWR TS 3.4-1 P&R, 07/22/93

4 RCS Sp:cific Activity 3.4.6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B.

(continued)

B.2.2.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND B.2.2.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months i

C.

Reactor coolant C.1 Isolate all main 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months f

specific activity steam lines.

> 100/E pC1/gm.

08

~

C.2.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND C.2.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE-FREQUENCY SR 3.4.6.1 Verify reactor coolant gross specific 7 days activity is s 100/E pCi/gm.

SR 3.4.6.2

NOTE--------------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE 31 days r

EQUIVALENT 1-131 specific activity is s [0.2] pCi/gm.

(continued)~

ABWR TS 3.4-2 P&R, 07/22/93 1

o RCS Specific Activity 3.4.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.6.3

NOTE------==-

== =------

Not required to be performed until 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for 2 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Determine E from a sample taken in MODE I 184 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last suberitical for 2 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

.mM-L o

}

4 1

1 1

ABWR TS 3.4-3 P&R, 07/22/93

6 RHR Shutdown Cooling System--Shutdown 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System--Shutdown LCO 3.4.7 Two RHR shutdown cooling subsystems shall be OPERABLE, and, with no recirculation pump in operation, at least one RHR shutdown cooling subsystem shall be in operation.

NOTES-------------------

====--

1.

Both RHR shutdown cooling subsystems and recirculation pumps may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided one RHR shutdown cooling subsystem is OPERABLE.

2.

One RHR shutdown cooling subsystem may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for performance of Surveillances provided the remaining RHR shutdown cooling subsystem is OPERABLE.

APPLICABILITY:

MODE 3 (with reactor steam dome pressure < [the RHR cut in permissive pressure)) and MODE 4.

ACTIONS

NOTE-------------------------------------

LC0 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or two RHR A.1 Initiate action to Immediately shutdown cooling restore RHR shutdown subsystems inoperable.

cooling subsystem (s) to OPERABLE status.

AND (continued)

(

ABWR TS 3.4-1 P&R, 07/22/93

6 RHR Shutdown Cooling System-Shutdown 3.4.7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.2 Verify an alternate I hour method of decay heat removal is available for each inoperable RHR shutdown cooling subsystem.

l E

A.3 Be in MODE 4.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months B.

No RHR shutdown B.1 Initiate action to Immediately l

~

cooling subsystem in restore one RHR operation.

shutdown cooling subsystem or one AND recirculation pump to operation.

No recirculation pump

{

in operation.

E i

B.2 Verify reactor I hour from coolant circulation discovery of no by an alternate reactor coolant method.

circulation i

M Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter AND B.3 Monitor reactor Once per hour coolant temperature and pressure.

J ABWR TS 3.4-2 P&R, 07/22/93 1

s e

RHR Shutdown Cooling System--Shutdown 3.4.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.7.1

------ ------- NOT E

= -----------

Not required to be met until 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after reactor steam dome pressure is < [the RHR cut in permissive pressure).

=- __-_ -== --------------------

Verify one RHR shutdown cooling subsystem 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or recirculation pump is operating.

e e

ABWR TS 3.4-3 P&R, 07/22/93

'a, O

RCS P/T Limits 3.4.8 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.8 RCS Pressure and Temperature (P/T) Limits LCO 3.4.8 RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculation pump starting temperature requirements shall be maintained within the limits specified in the PTLR.

APPLICABILITY:

At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

NOTE---------

A.1 Restore parameter (s) 30 minutes Required Action A.2 to within limits.

shall be completed if this Condition'is AND entered.

A.2 Determine RCS is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months acceptable for Requirements of the continued operation.

LC0 not met in MODES 1, 2, and 3.

I B.

Required Action and B.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A AND i

not met.

B.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)

ABWR TS 3.4-1 P&R, 07/22/93

RCS P/T Limits 3.4.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

NOTE---------

C.1 Initiate action to Immediately Required Action C.2 restore parameter (s) shall be completed if to within limits.

this Condition is entered.

AND

==-==__-

C.2 Determine RCS is Prior to Requirements of the acceptable for entering MODE 2 LCO not met in other operation.

or 3 than MODES 1, 2, and 3.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1

NOTE--------------------

Only required to be performed during RCS heatup and cooldown operations, and RCS inservice leak and hydrostatic testing.

Verify RCS pressure, RCS temperature, and 30 minutes RCS heatup and cooldown rates are within the limits specified in the PTLR.

SR 3.4.8.2 Verify RCS pressure and RCS temperature are Once within within the criticality limits specified in 15 minutes the PTLR.

prior to control rod withdrawal for the purpose of achieving criticality (continued)

ABWR TS 3.4-2 P&R, 07/22/93

4 e

RCS P/T Limits 3.4.8 l

SURVEILLANCE REQUIREMENTS (continued) l SURVEILLANCE FREQUENCY i

SR 3.4.8.3

NOTE==--- =-------------

Only required to be met in MODES 1, 2, 3, and 4 [with reactor steam dome pressure

_==- ________________. _====

1 L

Verify the difference between the bottom Once within head coolant temperature and the reactor 15 minutes l

pressure vessel (RPV) coolant temperature prior to each is within the limits specified in the PTLR.

startup of a recirculation pump SR 3.4.8.4

NOTE--------------

--

Only required to be met in MODES 1, 2, 3 and 4.

Verify the difference between the reactor Once within coolant temperature in the recirculation 15 minutes loop to be started and the RPV coolant prior to each temperature is within the limits specified startup of a in the PTLR.

recirculation pump SR 3.4.8.5

NOTE--------------------

Only required to be performed when tensioning the reactor vessel head bolting studs.

=- -=_________________

Verify reactor vessel flange and head 30 minutes flange temperatures are within the limits specified in the PTLR.

(continued) l

\\

ABWR TS 3.4-3 P&R, 07/22/93 l

l l

l

e RCS P/T Limits 3.4.8 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.8.6

NOTE--------------------

Not required to be performed until 30 minutes after RCS temperature s 80*F in MODE 4.

Verify reactor vessel flange and head 30 minutes flange temperatures are within the limits specified in the PTLR.

SR 3.4.8.7

NOTE--------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after RCS temperature s 100'F in MODE 4.

___________________________=

_==== _______

Verify reactor vessel flange and head 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months flange temperatures are within the limits specified in the PTLR.

b b

L b

ABWR TS 3.4-4 P&R, 07/22/93

e s

Reactor Steam Dome Pressure 3.4.9-

.i 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 Reactor Steam Dome Pressure LCO 3.4.9 The reactor steam dome pressure shall be s [1045] psig.

APPLICABILITY:

MODES I and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Reactor steam dome A.1 Restore reactor steam 15 minutes pressure not within dome pressure to limit.

within limit.

B.

Required Action and B.1 Be'in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met.

I SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 Verify reactor steam dome pressure is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s [1045] psig.

t ABWR TS 3.4-1 P&R, 07/22/93

a-o ECCS--Operating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)-

3.5.1 ECCS--Operating LC0 3.5.1 Each ECCS injection subsystem and the Automatic Depressurization System (ADS) function of [eight] safety /

relief valves shall be OPERABLE.

APPLICABILITY:

MODE 1, MODES 2 and 3, except ADS valves and_RCIC are not required to be OPERABLE with reactor steam dome pressure s [150] psig.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One low pressure ECCS A.1 Restore low pressure 7 days injection subsystem ECCS injection inoperable.

subsystem to OPERABLE status.

B.

One High Pressure Core B.1 Verify by Immediately Flooder (HPCF) System administrative means inoperable.

the other HPCF and RCIC System are OPERABLE when RCIC systems are OPERABLE.

6!!D 14 days B.2 Restore HPCF System to OPERABLE status.

(continued)

I l

ABWR TS 3.5-1 P&R, 07/22/93

4 4

ECCS--Operating 3.5.1 ACTIONS (continued)

CONDITION.

REQUIRED ACTION COMPLETION TIME C.

Two ECCS injection C.1 Restore one ECCS 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months subsystems (motor injection subsystem driven) inoperable.

to OPERABLE status.

D.

Required Action and D.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A, at Q J

B, or C not met.

D.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months E.

One ADS valve E.1 Restore ADS valve to 14 days inoperable.

OPERABLE status.

F.

One ADS valve F.1 Restore ADS valve to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable.

OPERABLE status.

AND QB One low pressuri ECCS F.2 Restore low pressure 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months injection subsystem ECCS injection inoperable.

subsystem to OPERABLE status.

G.

Two or more ADS valves G.I Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months inoperable.

AND QR (continued)

ABWR TS 3.5-2 P&R, 07/22/93 i

a ECCS--Operating 3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME G.

(continued)

G.2 Reduce reactor steam 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months dome pressure to Required Action and s [150] psig.

associated Completion Time cf Condition E or F not met.

A f

=

e e

I l

l 4

i P

i 1

)

ABWR TS 3.5-3 P&R, 07/22/93 i

l 1

4 4

ECCS-Operating 3.5.1 ACTIONS

~-

CONDITION REQUIRED ACTION COMPLETION TIME H.

HPCF and low pressure core spray (LPFL)

H.1 Enter LCO 3.0.3.

Imediately inoperable.

E i

Three or more ECCS injection subsystems inoperable.

E One or more ECCS injection subsystems and two or more ADS

~

valves inoperable.

t E

HPCF System and one or more ADS valves inoperable.

E

^

Two or more ECCS injection subsystems and one or more. ADS valves inoperable.

I. RCIC System I.1 Verify by Immediately inoperable.

administrative means HPCF is OPERABLE.

AND I.2 Restore RCIC System 14 days to OPERABLE status.

I l

ABWR TS 3.5-4 P&R, 07/22/93

e ECCS--Operating 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify, for each ECCJ.njection subsystem, 31 days the piping is filled with water from the pump discharge valve to the injection valve.

SR 3.5.1.2

== NOTE--------------------

=-

Low Pressure Flooder (LPFL) subsystems may be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than [the residual heat removal cut in permissive pressure] in MODE 3, if capable of being manually realigned and not otherwise inoperable.

Verify each ECCS injecti6n subsystem 31 days manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.5.1.3 Verify ADS [ Nitrogen gas supply system]

31 days pressure is 2 [150] psig.

(continued)

ABWR TS 3.5-5 P&R, 07/22/93 Om

A ECCS-Operating 3.5.1 j

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.4 Verify each ECCS pump develops the lInar.cordance specified flow rate [against a system head with the corresponding to the specified reactor Inservice pressure].

Testing 1

(SYSTEM HEAD Program or l

l.

CORRESPONDING l

TO A REACTOR l92 days l

I SYSTEM FLOW RATE PRESSURE OF1

_J LPFL 2 [4200] gpm 2 [225] psig I

HPCF 2 [3200] gpm 2 [1177] psig RCIC 2 [800] gpm 2 [1177] psig 1

SR 3.5.1.5

NOTE--------------------

Vessel injection / spray may be excluded.

Verify each ECCS injection / spray subsystem

[18] months actuates on an actual or simulated automatic initiation signal.

SR 3.5.1.6

NOTE--------------------

Valve actuation may be excluded.

_______________________.m__________________

Verify the ADS actuates on an actual or

[18] months simulated automatic initiation signal.

(continued)

ABWR TS 3.5-6 P&R, 07/22/93

i t

ECCS--Operating 3.5.1 SURVEiLLANCEREOUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.7

NOTE--------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam dome pressure is 2 [950] psig.

Verify each ADS valve opens when manually

[18] months on actuated.

a STAGGERED TEST BASIS for each valve solenoid s=e 5

P 4

I l

k ABWR TS 3.5-7 P&R, 07/22/93

-a s

ECCS--Shutdown 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS--Shutdown LCO 3.5.2 Two ECCS injection subsystems shall be OPERABLE.

APPLICABILITY:

MODE 4, MODE 5 except with the spent fuel pool gate removed and water level 2 7 m (23 ft) over the top of the reactor pressure vessel flange.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One required ECCS A.1 Restore required ECCS 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months injection subsystem injection / spray inoperable.

subsystem to OPERABLE 4

status.

B.

Required Action and B.1 Initiate action to Immediately associated Comp'ation suspend operations Time of Condition A with a potential for not met.

draining the reactor vessel (OPDRVs).

C.

Two required ECCS C.1 Initiate action to Immediately injection subsystems suspend OPDRVs.

inoperable.

AND j

l C.2 Restore one ECCS 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months injection subsystem to OPERABLE status.

(continued)

ABWR TS 3.5-I P&R, 07/22/93

d i

ECCS--Shutdown 3.5.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

Required Action C.2 D.1 Initiate action to Immediately and associated restore secondary Completion Time not containment to met.

OPERABLE status.

AND D.2 Initiate action to Immediately l

restore one standby gas treatment subsystem to OPERABLE status.

AND D.3 Initiate action to Immediately restore one isolation valve and associated instrumentation to OPERABLE status in each secondary containment penetration flow path not isolated.

t i

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify, for each required low pressure 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 5

Flooder (LPFL) subsystem, the suppression i

pool water level is 2 7 m (23 ft).

(continued) 1 ABWR TS 3.5-2 P&R, 07/22/93

I h

ECCS--Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.2 Verify, for each required High Pressure 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Core Flooder (HPCF) System, the:

a.

Suppression pool water level is 2 7 m (23 ft); or b.

Condensate storage tank water level is t

2[

3-i SR 3.5.2.3 Verify, for each required ECCS injection /

31 days i

spray subsystem, the piping is filled with t

water from the pump discharge valve to the injection valve.

1 SR 3.5.2.4

NOTE--------------------

Low Pressure Flooder (LPFL) subsystems may be considered OPERABLE during alignment and operation for decay heat removal, if capable of being manually realigned and not otherwise inoperable.

Verify each required ECCS injection 31 days subsystem manual, power operated, and i

automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

(continued) i i

ABWR TS 3.5-3 P&R, 07/22/93

d t

ECCS--Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.5 Verify each required ECCS pump develops the In accordance specified flow rate [against a system head with the 1

corresponding to the specified reactor Inservice pressure).

Testing Program or

[ SYSTEM HEAD 92 days CORRESPONDING TO A REACTOR SYSTEM FLOW RATE PRESSURE OF1 LPFL 2 [754] nd/ min 2 [281] kg/cm2 HPCF 2 [12.11] nd/ min 2 [7.03] kg/cm2 SR 3.5.2.6

NOTE--------------------

Vessel injection may be excluded.

Verify each required ECCS injection 18 months subsystem actuates on an actual or simulated automatic initiation signal.

k V

3 2

ABWR TS 3.5-4 P&R, 07/22/93

8 Responsibility 5.1 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.1.1 The [ Plant Superintendent] shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

The [ Plant Superintendent], or his designee, in accordance with approved administrative procedures, shall approve, prior to implementation, each proposed test or experiment and proposed changes and modifications to unit systems or equipment that affect nuclear safety.

5.1.2 The [ Shift Supervisor (SS)] shall be responsible for the control room command function. A management directive to this effect, signed by the [ highest level of corporate or site management]

shall be issued annually to all station personnel. During any absence of the [SS] from the control room while the unit is in MODE 1, 2, or 3, an individual with a valid Senior Reactor Operator (SRO) license shall be designated to assume the control room command function. During any absence of the [SS] from the control room while the unit is in MODE 4 or 5, an individual with a valid SR0 license or Reactor Operator license shall be designated to assume the control room command function.

I l

i 1

1 ABWR TS 5.0-1 P&R, 07/22/93

o a

Organization 5.2 l.

5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite and Offsite Oroanizations j

Onsite and offsite organizations shall be established for unit I

operation and corporate management, respectively. The onsite and I

offsite organizations shall include the positions for activities I

affecting safety of the nuclear power plant.

a.

Lines of authority, responsibility, and communication shall l

be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be~ documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel

~

positions, or in equivalent forms of documentation. These requirements shall be documented in the [SSAR];

b.

The [ Plant Superintendent] shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and.

maintenance of the plant; c.

The [a specified corporate executive position] shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety; and l

1 d.

The individuals who train the operating staff, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to I

ensure their independence from operating pressures.

5.2.2 Unit Staff The unit staff organization shall be as follows:

a.

Each on-duty shift shall be composed of at least the minimum I

shift crew composition shown in Table 5.2.2-1.

(continued)

ABWR TS 5.0-2 P&R, 07/22/93

.s Organization i

5.2 5.2 Organization l

5.2.2 Unit Staff (continued) b.

At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor.

In addition, while the unit is in MODE 1, 2, or 3, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room.

c.

A [ Health Physics Technician] shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to provide for unexpected absence,-

provided immediate action is taken to fill the required position.

d.

Either a licensed SR0 or licensed SR0 limited to fuel handling who has no concurrent responsibilities during this ~

operation shall be present during fuel handling and shall directly supervise all CORE ALTERATIONS.

e.

Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety related functions (e.g., licensed SR0s, licensed R0s, health physicists, auxiliary operators, and key maintenance personnel).

Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work an [8 or 12] hour day, nominal 40 hour4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months week, while the unit is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, or major plant modification, on a temporary basis the following guidelines shall be followed:

1.

An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight, excluding shift turnover time; 2.

An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7 day period, all excluding shift turnover time; (continued)

ABWR TS 5.0-3 P&R, 07/22/93

o 4

Organization 5.2 5.2 Organization 5.2.2 Unit Staff (continued) 3.

A break of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months should be allowed between work periods, including shift turnover time; 4.

Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

Any deviation from the above guidelines shall be authorized in advance by the [ Plant Superintendent] or his designee, in-accordance with approved administrative procedures, or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation.

Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the [ Plant Superintendent) or his designee to ensure that excessive hours have not been assigned.

Routine deviation from the above guidelines is not authorized.

DB The amount of overtime worked by unit staff members performing safety related functions shall be limited and controlled in accordance with the NRC Policy Statement on working hours (Generic Letter 82-12).

f.

The'[0perations Manager or Assistant Operations Manager) shall hold an SR0 license.

g.

The Shift Technical Advisor (STA) shall provide advisory technical support to the Shift Supervisor (SS) in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit.

i S

ABWR TS 5.0-4 P&R, 07/22/93 i

i

Organization 5.2 Table 5.2.2-1 (page 1 of 1)

Minimum Shift Crew Composition (a)

[ Single Unit Facility]

POSITION (b)

MINIMUM CREW NUMBER UNIT IN MODE 1, 2, OR 3 UNIT IN MODE 4 OR 5 SS 1

1 SR0 1

None R0 2

1 A0 (c) 2 1

STA 1

None (a) The shift crew composition may be one less than the minimum requirements ~

of Table 5.2.2-1 for not more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to accommodate unexpected absences of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 5.2.2-1.

This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

(b) Table Notation:

SS - [ Shift Supervisor] with a Senior Reactor Operator license; SRO - Individual with a Senior Reactor Operator license; RO - Individual with a Reactor Operator license; A0 - Auxiliary Operator; STA - Shift Technical Advisor.

(c) The STA position may be filled by an on-shift SS or SR0 provided'the individual meets the Commission Policy Statement on Engineering Expertise on Shift.

ABWR TS 5.0-5 P&R, 07/22/93

Organization 5.2 Table 5.2.2-1 (page 1 of 1)

Minimum Shift Crew Composition (a)

[Two Units With Two Control Rooms)

(Numbers for Each Unit)

POSITION (b)

MINIMUM CREW NUMBER UNIT IN MODE UNIT IN MODE 1, 2, OR 3; UNIT IN MODE 1, 2, OR 3; UNIT IN MODE OTHER UNIT 4 OR 5; OTHER OTHER UNIT IN 4 OR 5; OTHER IN MODE UNIT IN MODE MODE UNIT IN MODE 4 OR 5 OR 4 OR 5 OR 1, 2, OR 3 1, 2, OR 3 DEFUELED DEFUELED SS 1(d) 3(d) 1(d) 1(d)

SR0 1

None 1

None R0 2

1 2

1(8)

A0 STA(C) 2(d) 1 2

2 1

None 1

None (a) The shift crew composition may be one less than the minimum requirements of Table 5.2.2-1 for not more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 5.2.2-1.

This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

(b) Table Notation:

SS - [ Shift Supervisor) with a Senior Reactor Operator license; SRO - Individual with a Senior Reactor Operator license; RO - Individual with a Reactor Operator license; A0 - Auxiliary Operator; STA - Shift Technical Advisor.

(c)

The STA position may be filled by an on-shift SS or SR0 provided the individual meets the Commission Policy Statement on Engineering Expertise on Shift.

(d)

Individual may fill the same position on the other unit if licensed for both.

(e) One of the two required individuals may fill the same position on the other unit.

i l

ABWR TS 5.0-6 P&R, 07/22/93 i

Unit Staff Qualifications 5.3 5.0 ADMINISTRATIVE CONTROLS 5.3 Unit Staff Qualifications t

Reviewer's Note: Minimum qualifications for members of the unit staff shall be specified by use of an overall qualification statement referencing an ANSI Standard acceptable to the NRC staff or by specifying individual position qualifications. Generally, the first method is preferable; however, the second method is adaptable to those unit staffs requiring special qualification statements because of unique organizational structures.

5.3.1 Each member of the unit staff shall meet or exceed the minimum i

qualifications of [ Regulatory Guide 1.8, Revision 2,1987, or more recent revisions, or ANSI Standard acceptable to the NRC staff].

The staff not covered by [ Regulatory Guide 1.8] shall meet or exceed the minimum qualifications of [ Regulations, Regulatory Guides, or ANSI Standards acceptable to NRC staff].

In addition,'

the Shift Technical Advisor shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

6 l

i j

1 ABWR TS 5.0-7 P&R, 07/22/93 i

o e

Training 5.4 5.0 ADMINISTRATIVE CONTROLS 5.4 Training 5.4.1 A retraining and replacement training program for the unit staff shall be maintained under the direction of the [ position title]

and shall meet or exceed the requirements and recommendations of Section [

] of [an ANSI Standard acceptable to the NRC staff] and 10 CFR 55, and, for appropriate designated positions, shall include familiarization with relevant industry operational experience.

i l

c I

l ABWR TS 5.0-8 P&R, 07/22/93

s Reviews and Audits 5.5 5.0 ADMINISTRATIVE CONTROLS i

5.5 Reviews and Audits Reviewer's Note: The licensee shall describe the method (s) established to conduct independent reviews and audits. The methods may take a range of forms acceptable to the NRC. These methods may include creating an organizational unit or a standing or ad hoc committee, or assigning individuals capable of conducting these reviews and audits. When an individual performs a review j

function, a cross disciplinary review determination is necessary.

If deemed necessary, such reviews shall be performed by the review personnel of the appropriate discipline.

Individual reviewers shall not review their own work.

Regardless of the method used, the licensee shall specify the functions, organizational arrangement, responsibilities, appropriate ANSI /ANS 3.1-1981 qualifications, and reporting requirements of each functional element or unit that contributes to these processes.

Reviews and audits of activities affecting plant safety have two distinct elements. The first element is the reviews performed by plant staff personnel to ensure that day to day activities are conducted in a safe manner.

These reviews are described in Section 5.5.1.

The second element, described in Section 5.5.2, is the [offsite] reviews and audits of unit activities and programs affecting nuclear safety that are performed independent of the plant staff. The [offsite] reviews and audits should provide integration of the reviews and audits into a cohesive program that provides senior level utility management with an assessment of facility operation and recommends actions to improve nuclear safety and plant reliability.

It should include an assessment

_of the effectiveness of reviews conducted according to Section 5.5.1.

5.5.1 Plant Reviews Reviewer's Note: The licensee shall describe provisions for plant reviews (organization, reporting, records) and the appropriate

_ ANSI /ANS Standard for personnel qualification.

5.5.1.1 Functions The (plant review method specified in Specification 5.5.1] shall, as a minimum, incorporate functions that:

a.

Advise the [ Plant Superintendent] on all matters related to nuclear safety; (continued)

ABWR TS 5.0-9 P&R, 07/22/93

s Reviews and Audits 5.5 5.5 Reviews and Audits 5.5.1.1 Functions (continued) b.

Recommend to the [ Plant Superintendent] approval or disapproval of items considered under Specifications 5.5.1.2.a through 5.5.1.2.e prior to their implementation, except as provided in Specification 5.7.1.3; c.

Determine whether each item considered under Specifications 5.5.1.2.a through 5.5.1.2.d constitutes an unreviewed safety question as defined in 10 CFR 50.59; and d.

Notify the [Vice President-Nuclear Operations] of any safety significant disagreement between the [ review organization or individual specified in Specification 5.5.1) and the [ Plant Superintendent) within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . However, the [ Plant Superintendent] shall have responsibility for resolution of such disagreements pursuant to Specification 5.1.1.

5.5.1.2 Responsibilities The [ plant review method specified in Specification 5.5.1] shall be used to conduct, as a minimum, reviews of the following:

a.

All proposed procedures required by Specification 5.7.1.1 and changes thereto; b.

All proposed programs required by Specification 5.7.2 and changes thereto; c.

All proposed changes and modifications to unit systems or equipment that affect nuclear safety; d.

All proposed tests and experiments that affect nuclear safety; and All-proposed changes to these Technical Specifications (TS),

e.

their Bases, and the Operating License.

P (continued)

ABWR TS 5.0-10 P&R, 07/22/93

l Retriews and Audits 5.5 l

5.5 Reviews and Audits (continued) 5.5.2 I0ffsitel Review and Audit Reviewer's Note: The licensee shall describe the provisions for reviews and audits independent of the plant's staff (organization, reporting, and records) and the appropriate ANSI /ANS Standards for personnel qualifications. These individuals may be located onsite or offsite provided organizational independence from plant staff is maintained. The [ technical] review responsibilities, Specification 5.5.2.4, shall include several individuals located onsite.

5.5.2.1 Functions The [offsite review and audit provisions specified in Specification 5.5.2] shall, as a minimum, incorporate the I

following functions that:

l a.

Advise the [Vice President-Nuclear Operations] on all matters related to nuclear safety; b.

Advise the management of the audited organization, and [its Corporate Management and Vice President-Nuclear Operations],

of the audit results as they relate to nuclear safety; c.

Recommend to the management of the audited organization, and its management, any corrective action to improve nuclear safety and plant operation; and l

d.

Notify the [Vice President-Nuclear Operations] of any safety significant disagreement between the [ review organization or i

individual specified in Specification 5.5.2] and the

[ organization or function being reviewed) within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

1 5.5.2.2

[0ffsite] Review Responsibilities 1

The [ review method specified in Specification 5.5.2] shall be responsible for the review of:

a.

The safety evaluations for changes to procedures, equipment, or systems, and tests or experiments completed under the provisions of 10 CFR 50.59, to verify that such actions do not constitute an unreviewed safety question as defined in 10 CFR 50.59; (continued)

ABWR TS 5.0-11 P&R, 07/22/93

s Reviews and Audits 5.5 5.5 Reviews and Audits 5.5.2.2

[0ffsite] Review Responsibilities (continued) b.

Proposed changes to procedures, equipment, or systems that involve an unreviewed safety question as defined in 10 CFR 50.59; c.

Proposed tests or experiments that involve an unreviewed safety question as defined in 10 CFR 50.59; d.

Proposed changes to TS and the Operating License; e.

Violations of codes, regulations, orders, license requirements, and internal procedures or instructions having nuclear safety significance; f.

All Licensee Event Reports required by 10 CFR 50.73; g.

Plant staff performance; h.

Indications of unanticipated deficiencies in any aspect of design or operation of structures, systems, or components that could affect nuclear safety; i.

Significant accidental, unplanned, or uncontrolled radioactive releases, including corrective action to prevent recurrence; j.

Significant operating abnormalities or deviations from normal and expected performance of equipment that affect nuclear safety; and k.

The performance of the corrective action system.

Reports or records of these reviews shall be forwarded to the

[Vice President-Nuclear Operations] within 30 days following completion of the review.

5.5.2.3 Audit Responsibilities The audit responsibilities shall encompass:

a.

The conformance of unit operation to provisions contained within the TS and applicable license conditions; b.

The training and qualifications of the unit staff; (continued)

ABWR TS 5.0-12 P&R, 07/22/93

t e

o Reviews and Audits 5.5 f

5.5 Reviews and Audits 5.5.2.3 Audit Responsibilities (continued) c.

The implementation of all programs required by Specification 5.7.2; d.

Actions taken to correct deficiencies occurring in equipment, structures, systems, components, or method of operation that affect nuclear safety; and e.

Other activities and documents as requested by the [Vice President-Nuclear Operations).

Reports or records of these audits shall be forwarded to the [Vice President-Nuclear Operations] within 30 days following completion of the review.

5.5.2.4

[ Technical] Review Responsibilities The [ technical] review responsibilities shall encompass:

a.

Plant operating characteristics, NRC issuances, industry advisories, l.icensee Event Reports, and other sources that may indicate areas for improving plant safety; b.

Plant operations, modifications, maintenance, and surveillance to verify independently that these activities are performed safely and correctly and that human errors are reduced as much as practical; Inte'rnal and external operational experience information c.

that may indicate areas for improving plant safety; and d.

Making detailed recommendations through the [Vice i

President-Nuclear Operations] for revising procedures, equipment modifications, or other means of improving nuclear j

safety and plant reliability.

5.5.3 Records Written records of reviews and audits shall be maintained. As a minimum these records shall include:

a.

Results of the activities conducted under the provisions of Section 5.5; (continued)

ABWR TS 5.0-13 P&R, 07/22/93

Reviews and Audits 5.5 5.5 Reviews and Audits 5.5.3 Records (continued) b.

Recommendations to the management of the organization being audited; c.

An assessment of the safety significance of the review or audit findings; d.

Recommended approval or disapproval of items considered under Specifications 5.5.1.2.a through 5.5.1.2.e; and e.

Determination whether each item considered under Specifications 5.5.1.2.a through 5.5.1.2.d constitutes an unreviewed safety question as defined in 10 CFR 50.59.

I P

i ABWR TS 5.0-14 P&R, 07/22/93

o TS Bases Control 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Technical Specifications (TS) Bases Control 5.6.1 Changes to the Bases of the TS shall be made under appropriate administrative controls and reviewed according to Specification 5.5.1.

5.6.2 Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:

a.

A change in the TS incorporated in the license; or b.

A change to the updated FSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.

5.6.3 The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

5.6.4 Proposed changes that meet the criteria of (a) or (b) above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71.

ABWR TS 5.0-15 P&R, 07/22/93

Procedures, Programs, and Manuals 5.7 5.0 ADMINISTRATIVE CONTROLS 5.7 Procedures, Programs, and Manuals 5.7.1 Procedures 5.7.1.1 Scope Written procedures shall be established, implemented, and maintained covering the following activities:

a.

The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978; b.

The emergency operating procedures required to implement the requirements of NUREG-0737 and to NUREG-0737, Supplement 1, as stated in [ Generic Letter 82-33];

c.

Security plan implementation; d.

Emergency plan implementation; Quality assurance for effluent and environmental monitoring; e.

f.

Fire Protection Program implementation; and 9

All programs specified in Specification 5.7.2.

5.7.1.2 Review and Approval Each procedure of Specification 5.7.1.1, and changes thereto, shall be reviewed in accordance with Specification 5.5.1, approved by the [ Plant Superintendent) or his designee in accordance with approved administrative procedures prior to implementation and reviewed periodically as set forth in administrative procedures.

5.7.1.3 Temporary Changes Temporary changes to procedures of Specification 5.7.1 may be made provided:

The intent of the existing procedure is not altered; a.

b.

The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator license on the unit affected; and (continued)

ABWR TS 5.0-16 P&R, 07/22/93

g

' Procedures, Programs, and Manuals

-5.7 5.7 Procedures,-Programs, and Manuals 5.7.1.3 Temporary Changes (continued) c.

The change is documented and reviewed in accordance with Specification 5.5.1 and approved by the [ Plant Superintendent] or his designee in accordance with approved administrative procedures within 14 days' of implementation.

5.7.2 Procrams and Manuals The following programs shall be established, implemented, and-j maintained.

5.7.2.1 Radiation Protection Program 1

Procedures for personnel radiation protection shall be prepared -

consistent with the requirements of 10 CFR 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure.

5.7.2.2 Process Control Program (PCP).

The PCP shall contain.the current formulas, sampling, analyses,

)

tests, and determinations to be made to ensure that processing and j

packaging of solid radioactive wastes will be accomplished to ensure compliance with 10 CFR 20, 10 CFR 61, and 10 CFR 71; state regulations; burial ground requirements; and other requirements governing the disposal of solid radioactive' waste'.

.l 1

Licensee ~ initiated changes to the PCP:

a.

Shall be documented and records of reviews performed shall be retained. This documentation shall' contain:

1.

sufficient information to support the change (s) and appropriate analyses or evaluations justifying the.

change (s),and 2.

a determination that the change (s) maintain the overall conformance of the solidified waste product to the 1

existing requirements of Federal, State, or other j

applicable regulations.

(continued) j ABWR TS 5.0-17 P&R, 07/22/93

a r

Procedures, Pr: grams, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.2 Process Control. Program (PCP)

(continued) b.

Shall be effective after review and acceptance by the

[ review method of Specification 5.5.1] and the approval of-the [ Plant Superintendent].

5.7.2.3 Offsite Dose Calculation Manual (0DCM)

The ODCM shall contain the methodology and parameters used a.

in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program; and b.

The ODCM shall also contain the Radioactive Effluent Controls and Radiological Environmental Monitoring programs required by Specification 5.7.2, and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release, reports required by Specification

[5.9.1.3] and Specification [5.9.1.4].

Licensee initiated changes to the ODCH:

Shall be documented and records of reviews performed shall a.

be retained by Specification 5.10.3.n.

This documentation shall contain:

1. ' sufficient information to support the change (s) together with the appropriate analyses or evaluations justifying the change (s), and 2.

a determination that the change (s) maintain the levels of radioactive effluent control required pursuant to 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations; b.

Shall become effective after review and acceptance by the

[ review method of Specification 5.5.1] and the approval of the [ Plant Superintendent]; and i

(continued)

ABWR TS 5.0-18 P&R, 07/22/93 i

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.3 Offsite Dose Calculation Manual (0DCM)

(continued)

{

c.

Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCH as a part of, or concurrent with, the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made.

Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.

5.7.2.4 Primary Coolant Sources Outside Containment This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. The systems include [the Low Pressure Core Flooder, High Pressure Core Flooder, Residual Heat Removal, Reactor Core Isolation Cooling, hydrogen recombiner, process sampling, and Standby Gas Treatment].

The program shall include the following:

Preventive maintenance and periodic visual inspection a.

requirements; and b.

Integrated leak test requirements for each system at refueling cycle intervals or less.

5.7.2.5 In Plant, Radiation Monitoring This program provides controls to ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:

a.

Training of personnel; b.

Procedures for monitoring; and c.

Provisions for maintenance of sampling and analysis equipment.

(continued)

ABWR TS 5.0-19 P&R, 07/22/93

o a

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2 Procrams and Manuals (continued) 5.7.2.6 Post Accident Sampling This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive gases, and particulates in plant gaseous effluents and containment atmosphere samples under accident conditions. The program shall include the following:

a.

Training of personnel; b.

Procedures for sampling ar.d :nalysis; and c.

Provisions for maintenance of sampling and analysis equipment.

~

5.7.2.7 Radioactive Effluent Controls Program This program conforms to 10 CFR 50.36a for the control of r :loactive effluents and for maintaining the doses to members of tai public from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (?)

shall be implemented by procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded.

The program shall include the following elements:

a.

Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM; b.

Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to 10 times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2401; Monitoring, sampling, and analysis of radioactive liquid and c.

gaseous effluents pursuant to 10 CFR 20.1302 and with the methodology and parameters in the ODCM; d.

Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, conforming to 10 CFR 50, Appendix I; (continued)

ABWR TS 5.0-20 P&R, 07/22/93 i

o Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.7 Radioactive Effluent Controls Program (continued) e.

Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days; f.

Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2 percent of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I; g.

Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas at or beyond the site boundary shall be limited to the following:

1.

for noble gases:

Less than or equal to a dose rate of 500 mrems/yr to the total body and less than or equal to a dose rate of 3000 mrems/yr to the skin, and 2.

for icdine-131, iodine-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days:

Less than or equal to a dose rate of 1500 mrems/yr to any onan, i

h.

Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; i.

Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives greater than 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and J.

Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

t 4

(continued)

ABWR TS 5.0-21 P&R, 07/22/93

a Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.8 Radiological Environmental Monitoring Program This program is for monitoring the radiation and radionuclides in the environs of the plant. The program shall provide representative measurements of radioactivity in the highest potential exposure pathways and verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall be contained in the ODCM, shall conform to the guidance of 10 CFR 50, Appendix I, and shall include the following:

a.

Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM; b.

A Land Use Census to ensure that changes in the use of areas-

'I at and beyond the site boundary are identified and that modifications to the monitoring program are made if required by the results of this census; and Participation in an Iriterlaboratory Comparison Program to c.

ensure that independent checks on the precision and accuracy t

of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

5.7.2.9 Component Cyclic or Transient Limit This program provides controls to track the FSAR, Section [

],

cyclic and transient occurrences to ensure that components are maintained within the design limits.

5.7.2.10 Inservice Inspection Program This program provides controls for inservice inspection of ASME Code Class 1, 2, and 3 components, including applicable supports.

The program shall include the following:

t I

(continued)

ABWR TS 5.0-22 P&R, 07/22/93

o e

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.10 Inservice Inspection Program (continued) a.

Provisions that inservice inspection of ASME Code Class 1, 2, and 3 components shall be performed in accordance with ASME Boiler and Pressure Vessel Code and Addenda,Section XI, as required by 10 CFR 50.55a; l

l b.

The provisions of SR 3.0.2 are applicable to the frequencies for performing inservice inspection activities; c.

An inservice inspection program for piping identified in NRC Generic Letter 88-01 in accordance with the NRC staff l

positions on schedule, methods, personnel, and sample expansion included in Generic Letter 88-01, or in accordance with alternate measures approved by the NRC staff; and d.

Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.

5.7.2.11 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports.

The program shall include the following:

Provisions that inservice testing of ASME Code Class 1, 2, a.

and 3 pumps, valves, and snubbers shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addende as required by 10 CFR 50.55a; b.

Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as follows:

(continued)

ABWR TS 5.0-23 P&R, 07/22/93

i Procedures, Programs, and ' Manuals 5.71 5.7 Procedures,: Programs, and Manuals 5.7.2.11 Inservice-Testing Program- (continued)

ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for Required Frequencies inservice testing for performing inservice activities testino activities i

Weekly At least once per 7 days i

Monthly At least once per 31 days-Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days

.t

~

Yearly or annually At least once per 366 days Biennially or every 2 years At least on e per 731 days t

c.

The provisions of SR 3.D.2 are applicable to the above required frequencies for performing inservice testing activities; d.

The provisions of SR 3.0.3 are applicable to inservice-testing activities; and e.

Nothing in the ASME Boiler and Pressure Vessel Code shall be.

construed to supersede the requirements of any TS.

5.7.2.12 Ventilation Filter Testing Program (VFTP) i A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in [ Regulatory Guide

],

and in accordance with Regulatory Guide 1.52, Revision 2; ASME N510-1989; and AG-1-1991.

i t

i i

(continued) l ABWR TS 5.0-24 P&R, 07/22/93 h

._1,.

m..

4 Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.12 Ventilation Filter Testing Program (VFTP)

(continued)

I a.

Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < [0.05]% when tested in I

accordance with Regulatory Guide 1.52, Revision 2, and ASME l

N510-1989 at the system flowrate specified below [110%):

j ESF Ventilation System Flowrate l

Control Room Habitability System I

Standby Gas Treatment System b.

Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass < [0.5]% when tested in accordance with Regulatory Guide 1.52, Revision 2, and ASME N510-1989 at the system flowrate specified below [i 10%):

l ESF Ventilation. System Flowrate Control Room Habitability System Standby Gas Treatment System c.

Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the value specified below when tested in accordance with [ ASTM D3803-1989] at a I

temperature of s [30*C] and greater than or equal to the relative humidity specified below:

ESF Ventilation System Penetration RH Control Room Habitability System Standby Gas Treatment System (continued) l l

ABWR TS 5.0-25 P&R, 07/22/93

Procedures,. Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 7

5.7.2.12 Ventilation. Filter Testing Program (VFTP)

(continued)

Reviewer's Note: Allowable penetration = [100% - methyl iodide efficiency for charcoal credited in staff safety evaluation]/

^

(safety factor).

Safety factor = [5] for systems with heaters.

d.

Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below when tested in accordance with Regulatory Guide 1.52, Revision 2, and ASME N510-1989 at the system flowrate specified below [f 10%]:

ESF Ventilation System Delta P Flowrate Control Room Habitability System Standby Gas Treatment System Demonstrate that the heaters for each of the ESF systems e.

dissipate the value specified below [i 10%] when tested in accordance with ASME N510-1989:

ESF Ventilation System Wattage i

Control Room Habitability System

' Standby Gas Treatment System The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

5.7.2.13 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the [ Waste Gas Holdup System), [the quantity of radioactivity contained in gas storage tanks or fed into the offgas treatment system, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks).

The gaseous radioactivity quantities shall be determined following the (continued)

ABWR TS 5.0-26 P&R, 07/22/93

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.13 Explosive Gas and Storage Tank Radioactivity Monitoring Program (continued) methodology in [ Branch Technical Position (BTP) ETSB 11-5,

" Postulated Radioactive Release due to Waste Gas System Leak or Failure"]. The liquid radwaste quantities shall be determined in accordance with [ Standard Review Plan, Section 15.7.3, " Postulated Radioactive Release due to Tank Failures").

The program shall include:

The limits for concentrations of hydrogen and oxygen in the a.

[ Waste Gas Holdup System] and a surveillance program to ensure the limits are maintained.

Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion);

b.

A surveillance program to ensure that the quantity of radioactivity contained in [each gas storage tank and fed into the offgas treatment system) is less than the amount that would result in a whole body exposure of 2 0.5 rem to any individual in an unrestricted area, in the event of [an uncontrolled release of the tanks' contents); and A surveillance program to ensure that the quantity of c.

radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overficws and surrounding area drains connected to the

[ Liquid Radwaste Treatment System] is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

(continued)

ABWR TS 5.0-27 P&R, 07/22/93

j a

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2 Proarams and Manuals (continued) 5.7.2.14 Diesel Fuel Oil Testing Program t

A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established.

The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a.

Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1.

an API gravity or an absolute specific gravity within

limits, 2.

a flash point and kinematic viscosity within limits for ASTM 2D fuel oil, 3.

a clear and bright appearance with proper color; b.

Other properties for ASTM 2D fuel oil are within limits within 30 days following sampling and addition to storage tanks; and Total particulate concentration of the fuel oil is s 10 mg/l c.

when tested every 31 days in accordance with ASTM D-2276, Method A-2 or A-3.

5.7.2.15 Fire Protection Program This program provides controls to ensure that appropriate fire protection measures are maintained to protect the plant from fire and to ensure the capability to achieve and maintain safe shutdown in the event of a fire is maintained.

ABWR TS 5.0-28 P&R, 07/22/93

a SFDP 5.8 5.0 ADMINISTRATIVE CONTROLS 5.8 Safety Function Determination Program (SFDP) 5.8.1 This program ensures loss of safety function is detected and appropriate actions taken. Upon failure to meet two or more LCOs at the same time, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate limitations and remedial or compensatory actions may be identified to be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6.

5.8.2 The SFDP shall contain the following:

a.

Provisions for cross division checks to ensure a loss of the~

capability to perform the safety function assumed in the accident analysis does not go undetected; I

b.

Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists; c.

Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and d.

Other appropriate limitations and remedial or compensatory actions.

i 5.8.3 A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a.

A required system redundant to system (s) supported by the l

inoperable support system is also inoperable (Case A); or b.

A required system redundant to system (s) in turn supported by the inoperable supported system is also inoperable (Case B); or l

[

(continued) l l

ABWR TS 5.0-29 P&R, 07/22/93

SFDP 5.8 5.8 SFDP 5.8.3 (continued) c.

A required system redundant to support system (s) for the supported systems (a) and (b) above is also inoperable (Case C).

Generic Example:

Division A Division 8 System i System i

+-Case C 4

1 System 11

(Support System System 11 j

1 Inoperable) 1 System iii System iii +-Case A I

1 System iv System iv

+-Case 8

~

l 5.8.4 The SFDP identifies where a loss of safety function exists.

If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

ABWR TS 5.0-30 P&R, 07/22/93 1

l l

i o

Reporting Requirements 5.9 5.0 ADMINISTRATIVE CONTROLS 5.9 Reporting Requirements 5.9.1 Routine Reports The following reports shall be submitted in accordance with 10 CFR 50.4.

5.9.1.1 Startup Report A summary report of plant startup and power escalation testing shall be submitted following:

T a.

Receipt of an Operating License; b.

Amendment to the license involving a planned increase in power level; Installation of fuel that has a different design or has been c.

manufactured by a different fuel supplier; and d.

Modifications that may have significantly altered the t

nuclear, thermal, or hydraulic performance of the unit.

The initial Startup Report shall address each of the startup tests identified in FSAR, Chapter [14), and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any correctlye actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report. Subsequent Startup Reports shall address startup tests that are necessary to demonstrate the acceptability of changes and modifications.

Startup Reports shall be submitted within 90 days following completion of the Startup Test Program; 90 days following resumption or commencement of commercial power operation; or 9 months following initial criticality, whichever is earliest.

If the Startup Report does not cover all three events (i.e., initial criticality, completion of Startup Test Program, and resumption or commencement of commercial operation), supplementary reports shall be submitted at least every 3 months until all three events have been completed.

(continued) 1 ABWR TS 5.0-31 P&R, 07/22/93 l

.e Reporting Requirements 5.9 5.9 Reporting Requirements 5.9.1 Routine Reports (continued) 5.9.1.2 Annual Reports

NOTE-------------------------------

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station.

Annual Reports covering the activities of the unit as described below for the previous calendar year shall be submitted by March 31 of each year.

[The initial report shall be submitted by March 31 of the year following initial criticality.]

Reports required on an annual basis include:

t a.

Occupational Radiation Exposure Report A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) for whom monitoring was required, receiving an annual deep dose equivalent > 100 mrem and the associated collective deep dose equivalent (reported in person-rem) according to work and job functions (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance [ describe maintenance), waste processing, and refueling). This tabulation supplements the reqpirements of 10 CFR 20.2206. The dose assignments to various duty functions may be estimated based on pocket dosimeter, thermoluminescent dosimeter (TLD), or film badge measurements.

Small exposures totalling < 20% of the individual total dose need not be accounted for.

In the aggregate, at least 80% of the total deep dose equivalent received from external sources should be assigned to.

specific major work functions;- and

[b.

Any other unit unique reports required on an annual basis.]

(continued)

ABWR TS 5.0-32 P&R, 07/22/93

=

6 Reporting Requirements 5.9 5.9 Reporting Requirements 5.9.1 Routine Reoorts (continued) 5.9.1.3 Annual Radiological Environmental Operating Report

NOTE-------------------------------

=

A single submittal may be made for a multiple unit station.

The submittal should combine sections common to all units at the station.

The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with-the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements [in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979).

[The report shall identify the TLD results that represent collocated dosimeters in relation to the NRC TLD program and the exposure period associated with each result.]

In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.

The missing data shall be submitted in a supplementary report as soon as possible.

(continued)

ABWR TS 5.0-33 P&R, 07/22/93

)

e e

Reporting Roquirements 5.9 5.9 Reporting Requirements 5.9.1 Routine Reoorts (continued) 5.9.1.4 Radioactive Effluent Release Report

~

=_---NOTE-

______________________=

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

The Radioactive Effluent Release Report covering the operation of the unit during the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.

5.9.1.5 Monthly Operating Reports Routine reports of operating statistics and shutdown experience [,

including documentation of all challenges to the safety / relief valves ] shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

5.9.1.6 CORE OPEftATING LIMITS REPORT (COLR) a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

The individual specifications that address core operating limits must be referenced here.

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

(continued)

ABWR TS 5.0-34 P&R, 07/22/93

, Q

'9 e

Reporting Requirements-5.9 5.9 Reporting Requirements 5.9.1.6 CORE OPERATING LIMITS REPORT (COLR)

(continued)

Identify the Tcpical Report (s) by number, title, date, and NRC staff approval document, or identify the staff Safety Evaluation Report for a plant specific methodology by NRC letter and date.

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the-NRC.

5.9.1.7 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

The RCS pressure and temperature limits, including heatup and cooldown rates, criticality, and hydrostatic and leak test limits, shall be established and documented in the PTLR.

[Theindividual Specifications that' address the reactor vessel pressure and temperature limits and the heatup and cooldown rates may be referenced.] The analytical methods used to determine the pressure and temperature limits including the heatup and cooldown rates shall be those previously reviewed and approved by the NRC in [ Topical Report (s), number, title, date, and NRC staff approval document, or staff safety evaluation report for a plant specific methodology by NRC letter and date). The reactor vessel pressure and temperature limits, including those for heatup and cooldown rates, shall be determined so that all applicable limits (e.g.,

heatup limits, cooldown limits, and inservice leak and hydrostatic testing limits) of the analysis are met. The PTLR, including revisions or supplements thereto, shall be provided upon issuance for each reactor vessel fluency period.

(continued)

ABWR TS 5.0-35 P&R, 07/22/93

e o

Reporting Requirements 5.9 5.9 Reporting Requirements (continued) 5.9.2 Special Reports Special Reports may be required covering inspection, test, and maintenance activities. These special reports are determined on an individual basis for each unit, and their preparation and submittal are designated in the Technical Specifications.

Special Reports shall be submitted in accordance with 10 CFR 50.4 within the time period specified for each report.

The following Special Reports shall be submitted:

a.

In the event an ECCS is actuated and injects water into the RCS in MODE 1, 2, or 3, a Special Report shall be prepared j-and submitted within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to -

l date.

The current value of the usage factor for each affected safety injection nozzle shall be provided in this l

Special Report whenever its value exceeds 0.70.

b.

If an individual emergency diesel generator (EDG) experiences four or more valid failures in the last 25 demands, these failures and any nonvalid failures experienced by that EDG in that time period shall be reported within 30 days.

Reports on EDG failures shall include the information recommended in Regulatory Guide 1.9, Revision 3, Regulatory Position C.5, or existing Regulatory

)

Guide 1.108 reporting requirement.

1 1

c.

When a Special Report is required by Condition B or G of j

LCO 3.3.[3.1], " Post Accident Monitoring (PAM) i Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

l (continued)

ABWR TS 5.0-36 P&R, 07/22/93

Record Retentien 5.10 5.0 ADMINISTRATIVE CONTROLS 5.10 Record Retention 5.10.1 The following records shall be retained for at least 3 years:

All License Event Reports required by 10 CFR 50.73; a.

b.

Records of changes made to the procedures required by Specification 5.7.1.1; and c.

Records of radioactive shipments.

5.10.2 The following records shall be retained for at least 5 years:

a.

Records and logs of unit operation covering time intervals -

at each power level; b.

Records and logs of principal maintenance activities-inspections, repair, and replacement of principal items of equipment related to nuclear safety; c.

Records of surveillance activities, inspections, and calibrations required by the Technical Specifications (TS) [and the Fire Protection Program];

d.

Records of sealed source and fission detector leak tests and results; and l

e.

Records of annual physical inventory of all sealed source material of record.

t 5.10.3 The following records shall be retained for the duration of the l

unit Operating License:

a.

Records and drawing changes reflecting unit design modifications made to systems and equipment described in the FSAR; b.

Records of new and irradiated fuel inventory, fuel transfers, and assembly burnup histories; Records of radiation exposure for all individuals entering c.

radiation control areas; (continued)

ABWR TS 5.0-37 P&R, 07/22/93 i

i e

o t

Record Rstention 5.10

- 7 5.10 Record Retention 5.10.3 (continued) t d.

Records of gaseous and liquid radioactive material released to the environs-e.

Records of transient or operational cycles for those unit components identified in [FSAR, Section X];

f.

Records of reactor tests and experiments; g.

Records of training and qualification for members of the unit staff; h.

Records of inservice inspections performed pursuant to the TS; i.

Records of quality assurance activities required by the Operational Quality Assurance (QA) Manual [not listed in Specification 5.10.1 and which are classified as permanent records by applicable regulations, codes, and standards];

J.

Records of reviews performed for changes made to procedures, equipment, or reviews of tests and experiments pursuant to 10 CFR 50.59; k.

Records of the reviews and audits required by Specification 5.5.1 and Specification 5.5.2; 1.

Records of the service lives of all hydraulic and mechanical snubbers required by [ document where snubber requirements relocated to], including the date at which the service life commences, and associated installation and maintenance records; i

m.

Records of analyses required by the Radiological Environmental Monitoring Program that would permit evaluation of the accuracy of the analysis at a later date (these records should include procedures effective at specified times and QA records showing that these procedures were followed);

I t

Records of reviews performed for changes made to the Offsite n.

Dose Calculation Manual and the Process Control Program; and ABWR TS 5.0-38 P&R, 07/22/93

s High Radiation Area

~

5.11 5.0 ADMINISTRATIVE CONTROLS 5.11 High Radiation Area l

5.11 Hiah Radiation Areas 1

As provided in paragraph 20.1601 (c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601 (a) and (b) of 10 CFR Part 20:

I i

5.11.1 Hiah Radiation Areas with Dose Rates not Exceedino 1.0 rem / hour:*

i Each entryway to such an area shall be barricaded and a.

conspicuously posted as a high radiation area. Such barricades may be breached only during periods of entry or exit.

b.

Access to, and activities in, each such area shall be controlled by means of a Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area (s) and other appropriate radiation protection equipment and measures.

Individuals qualified in radiation protection procedures c.

(e.g., health physics technicians) and personnel i

continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while 3

performing their assigned duties'provided that they are following plant radiation protection procedures for entry to. exit from, and work in such areas.

d.

Each individual (whether alone or in a group) entering such an area shall possess:

1.

a radiation monitoring device that continuously i

displays radiation dose rates in the area (" radiation monitoring and indicating device"), or 2.

a radiation monitoring device that continuously integrates the radiation dose rates in the area and i

alarms when the device's dose alarm setpoint is reached

(" alarming dosimeter"), with an appropriate alarm setpoint, or (continued)

ABWR TS 5.0-39 P&R, 07/22/93 l

e High Radiaticn Area 5.11 5.11 High Radiation Area 5.11.1 (continued) 3.

a radiation monitoring device that continuously transmits dose rate and cumulative dose to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or 4.

a self-reading dosimeter and, (a) be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual work site, qualified in radiation protection procedures, equipped with a radiation monitoring and indicating device who is responsible for controlling personnel radiation exposure within the area, or (b) be under the surveillance, as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area.

Entry into such areas shall be made only after dose rates in e.

the area have been determined and entry personnel are knowledgeable of them.

5.11.2 Hioh Radiation Areas with Dose Rates Greater than 1.0 rem / hour.*

but less than 500 rads / hour:**

Each entryway to such an area shall be conspicuously posted a.

as a high radiation area and shall be provided with a locked door or gate that prevents unauthorized entry, and in addition:

1.

all such door and gate keys shall be maintained under the administrative control of the shift foreman or the 4

health physics supervisor on duty.

2.

doors and gates shall remain locked except during periods of personnel entry or exit.

(continued)

ABWR TS 5.0-40 P&R, 07/22/93

o 4

High Radiation Area 5.11 5.11 High Radiation Areas 9

5.11.2 (continued) b.

Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area (s) and other appropriate radiation protection equipment and measures.

Individuals qualified in radiation protection procedures may c.

be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are following plant radiation protection procedures for entry to, exit from, and work in such areas.

d.

Each individual entering such an area shall possess:

1.

an alarming dosimeter with an appropriate alarm setpoint, or i

2.

a radiation monitoring device that continuously transmits dose rate and cumulative dose to a remote receiver monitored. by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area, or 3.

a self-reading dosimeter and, (a) be under the surveillance, as specified in the RWP or equivalent, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring and indicating device who is responsible for controlling personnel exposure within the area, or (b) be under the surveillance, as specified in the RWP or equivalent, by means of closed circuit' television, of personnel qualified.in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area.

Entry into such areas shall be made only after dose rates in e.

the area have been determined and entry personnel are knowledgeable of them.

(continued)

ABWR TS 5.0-41 P&R, 07/22/93 l

I High Radiation Area 5.11 5.11 High Radiation Areas i

l l

5.11.2 (continued)

I f.

Such individual areas that are within a larger area that is controlled as a high radiation area, where no enclosure i

exists for the purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate, but shall be barricaded and conspicuously posted as a high radiation area, and a conspicuous, clearly visible flashing light shall be activated at the area as a warning device.

l At 30 centimeters (12 inches) from the radiation source or from any surface penetrated by the radiation.

- At I meter from the radiation source or from any surface penetrated by the radiation.

D h

i b

i k

ABWR TS 5.0-42 P&R, 07/22/93

4 Recirculation Loops Operating B 3.4.1

+

B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1 Reactor Internal Pumps (RIPS) Operating BASES BACKGROUND The Reactor Coolant Recirculation System is designed to provide a forced coolant flow through the core to remove heat from the fuel. The forced coolant flow removes more heat from the fuel than would be possible with just natural circulation. The forced flow, therefore, allows operation at significantly higher power than would otherwise be 1

possible.

The recirculation system also controls reactivity over a wide span of reactor power by varying the recirculation flow rate to control the void content of the moderator. The Reactor Coolant Recirculation System consists of ten recirculation pumps internal to the reactor-vessel. These reactor internal pumps (RIPS) directly provide the driving flow of water through the reactor vessel.

Each RIP contains a wet motor, an adjustable speed drive (ASD) to control pump speed, an external heat exchanger to cool the pump motor, and associated instrumentation. The RIP motors, which protrude from the bottom of the reactor vessel into the lower drywell, are part of the reactor coolant pressure boundary. The RIP impellers are reactor vessel internals.

The recirculated coolant consists of saturated water from the steam separators and dryers that has been subcooled by incoming feedwater. This water passes down the annulus between the reactor vessel wall and the core shroud and becomes the suction flow for the RIPS. This flow enters the ten RIPS at suction inlets located equidistant around the plate (or pump deck) forming the bottom of the annulus area.

The total flow then passes through the RIP impeller into the area below the core (lower plenum), gaining sufficient head in the process to drive the required flow upward through the core.

The subcooled water enters the bottom of the fuel channels and contacts the fuel cladding, where heat is transferred to the coolant. As it rises, the coolant begins to boil, creating steam voids within the fuel channel that continue until the coolant exits the core. Because of reduced moderation, the steam voiding introduces negative reactivity that must be compensated for to maintain or to increase reactor power. The recirculation flow control allows (continued)

ABWR TS B 3.4-1 P&R, 07/22/93

Recirculation Loops Operating B 3.4.1 BASES BACKGROUND operators to increase recirculation flow and sweep some of (continued) the voids from the fuel channel, overcoming the negative reactivity void effect.

Thus, the reason for having variable recirculation flow is to compensate for reactivity effects of boiling over a wide range of power generation (i.e., 55 to 100% RTP) without having to move control rods and disturb desirable flux patterns.

Each RIP is manually started from the control room.

The ASDs provide regulation of individual AIP speed and, therefore, flow. The flow through each RIP can be manually or automatically controlled.

APPLICABLE The operation of the Reactor Coolant Recirculation System

~

SAFETY ANALYSES (with at least nine RIPS in operation) is an initial condition assumed in the design basis loss of coolant accident (LOCA) (Ref. B 3.4-1).

During a LOCA, the operating RIPS are all assumed to trip at time zero due to a coincident loss of offsite power.

The subsequent core flow coast down will be immediate and rapid because of the relatively low inertia of the pumps and motors. However, the RIPS are assumed to have sufficient flow coastdown characteristics to maintain fuel thermal margins during abnormal operational transients (Ref. 3.4.1-2), which are analyzed in Chapter 15 of the SSAR.

A plant specific LOCA analysis may be performed assuming only nine operating RIPS. This analysis shall demonstrate that, in the event of a LOCA, the Emergency Core Cooling System response will provide adequate core cooling, provided the APLHGR requirements are modified accordingly (Ref. B 3.41-3).

The transient analyses of Chapter 15 of the SSAR may also be performed for RIPS in operation (Ref. 3.4.1-3) to demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR requirements are modified. During operation with only nine RIPS, modification to the Reactor Protection System average power range monitor (APRM) instrument setpoints is also required to account for the different relationships between recirculation drive flow and reactor core flow.

The APLHGR (continued) 4 ABWR TS B 3.4-2 P&R, 07/22/93

~.

-f 4

Recirculation Loops Operating-B 3.4.1 BASES APPLICABLE and MCPR.setpoints for nine RIPS in operation are to be SAFETY ANALYSES specified in the COLR. The APRM flow biased simulated (continued)

. thermal' power setpoint is in LCO 3.3.1.I, " Reactor Protection System (RPS) Instrumentation."

RIPS operating satisfies Criterion 2 of the NRC Policy i

Statement.

LCO At least nine RIPS are required to be in' operation to ensure:

that during a postulated LOCA or transient the assumptions of the associated analyses are satisfied.

[With only [nine]

RIPS in operation, modifications to the required APLHGR limits (LC0 3.2.I, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), MCPR limits (LCO 3.2.2, " MINIMUM CRITICAL ~ ~

POWER RATIO (MCPR)"), and APRM Flow Biased Simulated Thermal Power-High setpoint (LCO 3.3.I.1) may be applied to allow contin'ed. operation consistent with the assumptions of u

Reference B 3.4.1-I.]

1 i

APPLICABILITY In MODES I and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there.is considerable energy in the reactor core and the limiting design basis transients'and accidents are assumed to occur.-

In MODES 3, 4,.and 5, the consequences of an accident are reduced and the flow and coastdown characteristics of the

~'

RIPS are not important.

l ACTIONS A.]

With the requirements of the LC0 not met, at ~1 east nine RIPS must be restored to operation within.24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . RIP is considered not in operation when the pump is idle. -Should a l

LOCA occur with less than nine RIPS in operation, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses.. Therefore, only a-limited time is allowed to restore a sufficient number of the inoperable RIPS to operating status.

1 (continued)

ABWR TS B 3.4-3 P&R, 07/22/93

. t Recirculation Loops Operating B 3.4.1 BASES (continued)

ACTIONS Alternatively, if the requirements for nine RIPS in the (continued) operation part of the LCO are applied to operating limits and RPS setpoints, operation with only nine RIPS would satisfy the requirements of the LCO and the initial conditions of the accident sequence.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is based on the low probability of an accident occurring during this time period, on a reasonable time to complete the Required Action, and on frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected.

ACTIONS 1L1 With less than half of the required RIPS (four or fewer) in operation, or the Required Action and associated Completion Time of Condition A not met, the unit is required to be brought to a MODE in which the LCO does not apply.

To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

In this condition, the RIPS are not required to be operating because of the reduced severity of DBAs and minimal dependence on the RIPS flow and coastdown characteristics.

The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This SR ensures that the number of RIPS in operation is consistent with the assumptions of the applicable DBA and transient analyses.

This surveillance is required to be performed once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Operating experience with previous BWR designs has demonstrated that a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months frequency for this type of surveillance is adequate.

(continued)

ABWR TS B 3.4-4 P&R, 07/22/93

m Recirculation Loops Operating B 3.4.1 BASES (continued)

REFERENCES 1.

ABWR SSAR, Section 6.3.3.

2.

ABWR SSAR, Chapter 15.

3.

Plant specific analysis for nine RIPS operating.

9 9

e l

l l

l ABWR TS B 3.4-5 P&R, 07/22/93 l

e S/RVs B 3.4.2 P 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.2 Safety / Relief Valves (S/RVs)

BASES BACKGROUND The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Ref. B 3.4.2-1) requires the Reactor Pressure Vessel be protected from overpressure during upset conditions by self actuated safety valves. As part of the nuclear pressure relief system, the size and number of safety / relief valves (S/RVs) are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary (RCPB).

~

The S/RVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell.

Each S/RV discharges steam through a discharge line to a point below the minimum water level in the suppression pool.

The S/RVs can actuate by either of two modes:

the safety mode or the relief mode.

In the safety mode (or spring mode of operation), the direct action of the steam pressure in the main steam lines will act against a spring loaded disk that will pop open when the valve inlet pressure exceeds the spring force.

In the relief mode (or power actuated mode of I

operation), a pneumatic piston or cylinder and mechanical i

linkage assembly are used to open the valve by overcoming i

the spring force, even with the valve inlet pressure equal i

to O psig. The pneumatic operator is arranged so that its malfunction will not prevent the valve disk from lifting if steam inlet pressure reaches the spring lift set pressures.

In the relief mode, valves may be opened manually or automatically at the selected preset pressure.

Eight of the S/RVs that provide the relief function are part of the Automatic Depressurization System specified in LC0 3.5.1, "ECCS-Opera ting. " The instrumentation for the ADS function is discussed in LC6 3.3.5.1, " Emergency Core Cooling Systems (ECCS) Instrumentation."

l (continued)

ABWR TS B 3.4-1 P&R, 07/22/93

S/RVs B 3.4.2 P

BASES (continued)

APPLICABLE The overpressure protection system must accommodate the SAFETY ANALYSES most severe pressure transient.

Evaluations have determined that the most severe transient is the closure of all main steam isolation valves (MSIVs) followed by reactor scram on high neutron flux (i.e., failure of the direct scram i

associated with MSIV position) (Ref. B 3.4.2-2).

For the purpose of the analyses, [ eleven) of the S/RVs are assumed to operate in the safety mode and no credit is taken for the relief mode of operation.

The analysis results demonstrate that the design S/RV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel 2

design pref)sure, i.e.,110% x 87.9 Kg/cm (1250 psig) -

96.7 Kg/cm (137S psig) (8618 KPa).

[ Twelve] S/RVs are required to be OPERABLE in the safety made to meet single failure considerations. This acceptancelimitof96.7Kg/cm[C0helpstoensurethatthe.

~

(137S psig) (9488 KPa) is met during the design basis event.

Reference B 3.4.2-3 discusses additional events that are expected to actuate the S/RVs.

From an overpressure standpoint, these events are bounded by the MSIV closure with flux scram event described above.

S/RVs satisfy Criterion 3 of the NRC Policy Statement.

LC0 The safety function of [ twelve] S/RVs is required to be OPERABLE in the safety mode.

The requirements of this LC0 are applicable only to the capability of the 5 e

mechanically open to relieve excess pressure. /RVs to In l

Reference B 3.4.2-1, an evaluation was performed to i

establish the parametric relationship between the peak i

vessel pressure and the number of OPERABLE S/RVs. The i

results show that with a minimum of [ eleven] S/RVs in the safety mode OPERABLE, the ASME Code limit of 96.7 Kg/cm2 (1375 psig) (9488 KPa) is not exceeded.

[Tweleve] S/Rvs are i

required to be OPERABLE in the safety mode to meet single failure considerations.

The S/RV setpoints are established to ensure the ASME Code limit on peak reactor pressure is satisfied. The ASME Code r

specifications require the lowest safety valve be set at or below vessel design pressure, i.e., 87.9 KG/cm (1250 psig)

(8618 KPa) and the highest safety valve be set so the total accumulated pressure does not exceed 110% of the design (continued)

ABWR TS B 3.4-2 P&R, 07/22/93

r 4

S/RVs l

B 3.4.2 I

t BASES (continued)

LCO pressure for conditions. The transient evaluations in (continued)

Reference B 3.4.2-3 are based on these setpoints, but also l

include the additional uncertainties of i 1% of the nominal l

setpoint to account for potential setpoint drift to provide an added degree of conservatism.

Operation with fewer valves OPERABLE than specified, or with i-I setpoints outside the ASME limits, could result in a more I

severe reactor response to a transient than predicted, i

possibly resulting in the ASME Code limit on reactor pressure being exceeded.

l APPLICABILITY In MODES 1, 2, and 3, the specified number of S/RVs must be.

OPERABLE since there may be considerable energy in the reactor core and the limiting design basis transients are assumed to occur. The S/RVs may be required to provide pressure relief to discharge energy from the core until such time that the Residual Heat Removal (RHR) System is capable of dissipating the heat.

In MODE 4, decay heat is low enough for the RHR System to i

provide adequate cooling, and reactor pressure is low enough that the overpressure limit cannot be approached by assumed operational transients or accidents.

In MODE 5, the reactor vessel head is unbolted or removed and the reactor is at atmospheric pressure. The S/RV function is not needed l

during these conditions.

ACTIONS Ad With the safety function of one [ required] S/RV inoperable, the remaining OPERABLE S/RVs are capable ot' providing the necessary overpressure protection.

[Because of additional design margin, the ASME Code limits for the RCPB can also be satisfied with two S/RVs inoperable.] However, the overall reliability of the pressure relief system is reduced because additional failures in the remaining OPERABLE S/RVs could result in failure to adequately relieve pressure during a limiting event.

For this reason, continued operation is permitted for a limited time only.

l (continued)

ABWR TS B 3.4-3 P&R, 07/22/93

a S/RVs B 3.4.2 BASES ACTIONS The 14 day Completion Time to restore the inoperable (continued) required S/RVs to OPERABLE status is based on the relief capability of the remaining S/RVs, the low probability of an event requiring S/RV actuation, and a reasonable time to complete the Required Action.

B.1 and 8.2 With less than the minimum number of required S/RVs OPERABLE, a transient may result in the violation of the ASME Code limit on reactor pressure.

[If the inoperable required S/RV cannot be restored to OPERABLE status within.

the associated Completion Time of Required Action A.1] or if

[two] or more [ required] S/RVs are inoperable, the plant must be brought to a MODE in which the LC0 does not apply.

To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.2 1 1

REQUIREMENTS This Surveillance demonstrates that the [ required] S/RVs will open at the pressures assumed in the safety analysis of Reference B 3.4.2-2.

The demonstration of the S/RV safety function lift settings must be performed during shutdown, since this is a bench test [, and in accordance with the Inservice Testing Program).

The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.

The S/RV setpoint is i [3]% for OPERABILITY; however, the valves are reset to 1% during the Surveillance to allow for drift.

The [18 month] Frequency was selected because this Surveillance must be performed during shutdown conditions and is based on the time between refuelings.

t (continued)

ABWR TS B 3.4-4 P&R, 07/22/93

e o

1 S/RVs B 3.4.2 BASES SURVEILLANCE SR 3.4.2.2 REQUIREMENTS A manual actuation of each [ required] S/RV is performed to verify that, mechanically, the valve is functioning properly and no blockage exists in the valve discharge line.

This can be demonstrated by the response of the turbine control valves or bypass valves, by a change in the measured steam flow, or any other method suitable to verify steam flow.

Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve.

Sufficient time is therefore allowed after the required pressure is achieved to perform this test. Adequatepregsureatwhich this test is to be performed is [66.8] Kg/cm (950 psig)

(the pressure recommended by the valve manufacturer).

Plant startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam dome pressure 2

is 2 ([66.8] Kg/cm (950 psig).

The 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions for testing and provides a reasonable time to complete the SR.

If the valve j

fails to actuate due only to the failure of the solenoid but is capable of opening on overpressure, the safety function l

of the S/RV is considered OPERABLE.

The [18] month on a STAGGERED TEST BASIS Frequency ensures that each solenoid for each S/RV is alternately tested. The Frequency is consistent with SR 3.4.4.1 to ensure that the S/RVs are manually actuated following removal for refurbishment or lift setpoint testing.

REFERENCES 1.

ASME, Boiler and Pressure Vessel Code,Section III.

2.

ABWR SSAR, Section 5.2.2.

3.

ABWR SSAR, Section 15.

ABWR TS B 3.4-5 P&R, 07/22/93

'e b

RCS Operational LEAKAGE B 3.4.3 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.3 RCS Operational LEAKAGE BASES BACKGROUND The RCS includes systems and components that contain or transport the coolant to or from the reactor core.

The pressure containing components of the RCS and the portions of connecting systems out to and including the isolation valves define the reactor coolant pressure boundary (RCPB).

The joints of the RCPB components are welded or bolted.

During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration. ~

Limits on RCS operational LEAKAGE are required to ensure appropriate action is taken before the integrity of the RCPB is impaired. This LC0 specifies the types and limits of LEAKAGE.

This protects the RCS pressure boundary described in 10 CFR 50.2, 10 CFR 50.55a(c), and GDC 55 of 10 CFR 50, Appendix A (Refs. B 3.4.3-1, B 3.4.3-2, and B 3.4.3-3).

The safety significance of leaks from the RCPB varies widely depending on the source, rate, and duration.

Therefore, detection of LEAKAGE in the primary containment is necessary. Methods for quickly separating the identified i

LEAKAGE from the unidentified LEAKAGE are necessary to provide the operators quantitative information to permit them to take corrective action should a leak occur detrimental to the safety of the facility or the public.

A limited amount of leakage inside primary containment is expected from auxiliary systems that cannot be made 100%

l leaktight.

Leakage from these systems should be detected and isolated from the primary containment atmosphere, if i

possible, so as not to mask RCS operational LEAKAGE detection.

This LC0 deals with protection of the RCPB from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded. The consequences of violating this LC0 include the possibility of a loss of coolant accident.

(continued)

ABWR TS B 3.4-1 P&R, 07/22/93 l

e RCS Operational LEAKAGE B 3.4.3 f

l BASES (continued)

APPLICABLE The allowable RCS operational LEAKAGE limits are based on SAFETY ANALYSES the predicted and observed leakage in operating plants. The normally expected background LEAKAGE due to equipment design and the detection capability of the instrumentation for determining system LEAKAGE were also considered.

Tha evidence from experiments suggests, for LE#GGE even greater than the specified unidentified LEAKAGE limits, the probability is small that the imoerfection or crack associated with such LEAKAGE would grow rapidly.

The unidentified LEAKAGE flow limit allows time for corrective action before the RCPB could be significantly 3

compromised. The 1.14 m /hr (5 gpm) limit is a small fraction of the calculated flow from a critical crack in the l-primary system piping. Crack behavior from experimental programs (Refs. B 3.4.3-4 and B 3.4.3-5) shows leak rates of tens of thousands liters per second (hundreds of gallons per minute) will precede crack instability (Ref. B 3.4.3-6).

No applicable safety analysis assumes the total LEAKAGE limit.

The total LEAKAGE limit considers RCS inventory makeup capability and drywell floor sump capacity.

1 RCS operational LEAKAGE satisfies Criterion 2 of the NRC Policy Statement.

l LC0 RCS operational LEAKAGE shall be limited to:

I a.

Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material degradation.

LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE.

Violation of this LC0 could result in continued degradation of the RCPB.

LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

i (continued)

ABWR TS B 3.4-2 P&R, 07/22/93

s RCS Operational LEAKAGE B 3.4.3 BASES (continued)

LCO b.

Unidentified LEAKAGE (continued) 3 Unidentified LEAKAGE of 1.14 m /hr (5 grm) is allowed as a reasonable minimum detectable amount that the drywell air monitoring, drywell sump level monitoring, and drywell air cooler condensate flow rate monitoring equipment can detect within a reasonable time period.

Violation of this LCO could result in continued degradation of the RCPB.

c.

Total LEAKAGE The total LEAKAGE limit is based on a reasonable minimum detectable amount.

The limit also accounts for LEAKAGE from known sources (identified LEAKAGE).

Violation of this LC0 indicates an unexpected amount

~

of LEAKAGE and, therefore, could indicate new or additional degradation in an RCPB component or system.

APPLICABILITY In MODES 1, 2, and 3, the RCS operational LEAKAGE LCO applies because the potential for RCPB LEAKAGE is greatest when the reactor is pressurized.

In MODES 4 and 5, RCS operational LEAKAGE limits are not required since the reactor is not pressurized and stresses in the RCPB materials and potential for LEAKAGE are reduced.

ACTIONS A_1 With RCS unidentified or total LEAKAGE greater than the limits, actions must be taken to reduce the leak.

Because the LEAKAGE limits are conservatively below the LEAKAGE that would constitute a critical crack size, 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is allowed to reduce the LEAKAGE rates before the reactor must be shut down.

If an unidentified LEAKAGE has been identified and quantified, it may be reclassified and considered as identified LEAKAGE. However, the total LEAKAGE limit would remain unchanged. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time is needed to properly verify the source before the reactor must be shut down.

(continued)

ABWR TS B 3.4-3 P&R, 07/22/93

c,,

RCS Operational LEAKAGE B 3.4.3 BASES (continued)

ACTIONS B.1 and B.2 (continued)

If any Required Action and associated Completion Time of Condition A or B is not met or if pressure boundary LEAKAGE l

exists, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.3.1 REQUIREMENTS The RCS LEAKAGE is monitored by a variety of instruments designed to provide alarms when LEAKAGE is indicated and to quantify the various types of LEAKAGE.

Leakage detection instrumentation is discussed in more detail in the Bases for LCO 3.4.5, "RCS Leakage Detection Instrumentation." Sump level and flow rate are typically monitored to determine actual LEAKAGE rates. However, any method may be used to quantify LEAKAGE within the guidelines of Reference B 3.4.3-7.

In conjunction with alarms and other administrative controls, an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Frequency for this Surveillance is appropriate for identifying changes in LEAKAGE and for tracking required trends (Ref. B 3.4.3-8).

REFERENCES 1.

10 CFR 50.2.

2.

10 CFR 50.55a(c).

3.

10 CFR 50, Appendix A, GDC 55.

4.

GEAP-5620, April 1968.

5.

NUREG-76/067, October 1975.

(continued)

ABWR TS B 3.4-4 P&R, 07/22/93

o a

RCS Operational LEAKAGE B 3.4.3 BASES (continued)

REFERENCES 6.

ABWR SSAR, Section 5.2.5.

(continued) 7.

Regulatory Guide 1.45.

8.

Generic Letter 88-01, Supplement 1.

l 1

1 ABWR TS B 3.4-5 P&R, 07/22/93

a a

RCS PIV Leakage B 3.4.4 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.4 RCS Pressure Isolation Valve (PIV) Leakage BASES BACKGROUND RCS PIVs are defined as any two normally closed valves in series within the reactor coolant pressure boundary (RCPB).

The function of RCS PIVs is to separate the high pressure RCS from an attached low pressure system.

This protects the RCS pressure boundary described in 10 CFR 50.2, 10 CFR 50.55a(c), and GDC 55 of 10 CFR 50, Appendix A (Refs. B 3.4.4-1, B 3.4.4-2, and B 3.4.4-3).

PIVs are designed to meet the requirements of Reference B 3.4.4-4.

During their lives, these valves can produce varying amounts of reactor coolant leakage through either normal operational wear or mechanical deterioration.

i i

The RCS PIV LCO allows RCS high pressure operation when Teakage through these valves exists in amounts that do not 1

compromise safety. The PlV leakage limit applies to each individual valve.

Leakage through these valves is not included in any allowable LEAKAGE specified in LC0 3.4.5, i

"RCS Operational LEAKAGE."

Although this specification provides a limit on allowable PIV leakage rate, its main purpose is to prevent overpressure failure of the low pressure portions of connecting systems. The leakage limit is an indication that the PlVs between the RCS and the connecting systems are degraded or degrading.

PIV leakage could lead to overpressure of the low pressure piping or components.

Failure consequences could be a loss of coolant accident i

(LOCA) outside of containment, an unanalyzed accident which could degrade the ability for low pressure injection.

j A study (Ref. B 3.4.4-5) evaluated various PIV configurations to determine the probability of intersystem LOCAs. This study concluded that periodic leakage testing of the PIVs can substantially reduce intersystem LOCA probability.

PIVs are provided to isolate the RCS from the following typically connected systems:

(continued)

ABWR TS B 3.4-1 P&R, 07/22/93

t r

o.

RCS PIV Leakage' B 3.4.4 BASES BACKGROUND a.

Residual Heat Removal (RHR) System; (continued) b.

Low Pressure Core Flooder System; c.

High Pressure Core Flooder System; and i

d.

Reactor Core Isolation Cooling System.

The PIVs are listed in Reference B 3.4.4-6.

APPLICABLE Reference B 3.4.4-5 evaluated various PIV configurations, SAFETY ANALYSES leakage testing of the valves, and operational changes to determine the effect on the probability of intersystem LOCAs. This study concluded that periodic leakage testing of the PIVs can substantially reduce the probability of an intersystem LOCA.

PIV leakage is not considered in any Design Basis Accident analyses.

This Specification provides for monitoring the condition of the RCPB to detect PIV degradation that has the potential to cause a LOCA outside of containment.

RCS PIV leakage satisfies Criterion 2 of the NRC Policy Statement.

LCO RCS PIV leake te is leakage-into closed systems connected to the RCS.

Ist,3ation valve leakage is usually on the order of drops per minute.

Leakage that increases significantly suggests that something is operationally wrong and corrective action must be taken.

Violation of this LCO t

could result in continued degradation of a PIV, which could lead to overpressurization of a low pressure system and the loss of the integrity of a fission product barrier.

3 The LC0 PIV leakage limit is 114 m /hr (0.5 gpm) per 2.54 cm (noming1 inch) of valve size with a maximum limit of 1.14 m /hr (5 gpm) (Ref. B 3.4.4-7).

Reference B 3.4.4-7 permits leakage testing at a lower pressure differential than between the specified maximum RCS pressure and the normal pressure of the connected system during RCS operation (the maximum pressure differential).

The observed rate may be adjusted to the maximum pressure (continued)

ABWR TS B 3.4-2 P&R, 07/22/93

o a

RCS PIV Leakage B 3.4.4 BASES LCO differential by assuming leakage is directly proportional to (continued) the pressure differential to the one-half power.

APPLICABILITY In MODES 1, 2, and 3, this LCO applies because the PIV leakage potential is greatest when the RCS is pressurized.

In MODE 3, valves in the RHR flowpath are not required to r

meet the requirements of this LCO when in the RHR mode of operation.

In MODES 4 and 5, leakage limits are not provided because the lower reactor coolant pressure results in a reduced potential for leakage and for a LOCA outside the containment. Accordingly, the potential for the consequences of reactor coolant leakage is far lower during these MODES.

ACTIONS The ACTIONS are modified by two Notes.

Note I has been provided to modify the ACTIONS related to RCS PIV flow paths.

Section 1.3, Completion Times, specifies once a Condition has been entered, subsequent trains, subsystems, components or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition.

Section 1.3 also specifies Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for the Condition of RCS PIV leakage limits exceeded provide appropriate compensatory measures for separate affected RCS PIV flow paths.

As such, a Note has been provided that allows separate Condition entry for each affected RCS PIV flow path. Note 2 requires an evaluation of affected systems if a PIV is inoperable.

The leakage may have affected system OPERABILITY, or isolation of a leaking flow path with an alternate valve may have degraded the ability of the interconnected system to perform its safety function. As a result, the applicable Conditions and Required Actions for systems made inoperable by PIVs must be entered. This ensures appropriate remedial actions are i

taken, if necessary, for the affected systems.

(continued)

ABWR TS B 3.4-3 P&R, 07/22/93

o RCS PIV Leakage B 3.4.4 BASES (continued)

ACTIONS A.1 and A.2 (continued)

If leakage from one or more RCS PIVs is not within limit, the flow path must be isolated by at least one closed manual, deactivated, automatic, or check valve within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Required Action A.I and Required Action A.2 are modified by a Note stating that the valves used for isolation must meet the same leakage requirements as the PIVs and must be on the RCPB [or the high' pressure portion of the system.]

Four hours provides time to reduce leakage in excess of the allowable limit and to isolate the flow path if leakage cannot be reduced while corrective actions to reseat the leaking PIVs are taken. The 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allows time for these

~

actions and restricts the time of operation with leaking valves.

Required Action A.2 specifies that the double isolation barrier of two valves be restored by closing another valve qualified for isolation or restoring one leaking PIV.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time after exceeding the limit considers the time required to complete the Required Action, the low probability of a second valve failing during this time period, and the low probability of a pressure boundary rupture of the low pressure ECCS piping when overpressurized to reactor pressure (Ref. B 3.4.4-8).

BiandB.2 If leakage cannot be reduced or the system isolated, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to j

MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This action may reduce the leakage and also reduces the potential for a LOCA outside the containment. The Completion Times are reasonable, based on operating experience, to achieve the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

(continued)

ABWR TS B 3.4-4 P&R, 07/22/93

O J

RCS PIV Leakage B 3.4.4 i

BASES

)

SURVEILLANCE SR 3.4.4.1 REQUIREMENTS Performance of leakage testing on each RCS PIV is required to verify that leakage is below the specified limit and }/hr o

identify each leaking valve. The leakage limit of 114 m (0.5 gpm) per 2.54 cm (nominal inch) of valve diameter up to 1.14 m /hr (5 gpm) maximum applies to each valve.

Leakage

^

testing requires a stable pressure condition.

For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves.

If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement.

In this situation, the protection provided by i

redundant valves would be lost.

The 18 month Frequency required by the Inservice Testing Program is within the ASME Code,Section XI, Frequency requirement and is based on the need to perform this Surveillance under the conditions that apply during an outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

Therefore, this SR is modified by a Note that states the leakage Surveillance is not required to be performed in MODE 3.

Entry into MODE 3 is permitted for leakage testing at high differential pressures with stable conditions not possible in the lower MODES.

REFERENCES 1.

10 CFR 50.2.

2.

10 CFR 50.55a(c).

s 3.

10 CFR 50, Appendix A, GDC 55.

4.

ASME, Boiler and Pressure Vessel Code,Section XI, Subsection IWV.

5.

NUREG-0677, May 1980.

6.

ABWR SSAR, Section 3.9.

(continued)

ABWR TS B 3.4-5 P&R, 07/22/93

RCS PIV Leakage-B 3.4.4 BASES REFERENCES 7.

ASME, Boiler and Pressure Vessel Code,Section XI, (continued)

IWV-3423(e).

8.

NEDC-31339, November 1986.

4 7

e t

t ABWR TS B 3.4-6 P&R, 07/22/93

2 RCS Leakage Detection Instrumentation B 3.4.5 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.5 RCS Leakage Detection Instrumentation BASES BACKGROUND GDC 30 of 10 CFR 50, Appendix A (Ref.1), requires means for detecting and, to the extent practical, identifying the location of the source of RCS LEAKAGE.

Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.

Limits on LEAKAGE from the reactor coolant pressure boundary (RCPB) are required so that appropriate action can be taken before the integrity of the RCPB is impaired (Ref. 2).

Leakage detection systems for the RCS are provided to alert the operators when leakage rates above normal background levels are detected and also to supply quantitative measurement of rates. The Bases for LC0 3.4.5, "RCS Operational LEAKAGE," discuss the limits on RCS LEAKAGE rates.

Systems for separating the LEAKAGE of an identified source from an unidentified source are necessary to provide prompt and quantitative information to the operators to permit them to take immediate corrective action.

LEAKAGE from the RCPB inside the drywell is detected by at least one of two or three independently monitored variables, sugh as sump level changes and drywell gaseous and particulate radioactivity levels. The primary means of quantifying LEAKAGE in the drywell is the drywell floor drain sump monitoring system.

The drywell floor drain sump monitoring system monitors the LEAKAGE collected in the floor drain sump. This unidentified LEAKAGE consists of LEAKAGE from control rod drives, valve flanges or packings, floor drains, the Closed Cooling Water System, and drywell air cooling unit condensate drains, and any LEAKAGE not collected in the drywell equipment drain sump. The drywell floor drain sump has transmitters that supply level indications in the main control room.

The floor. drain sump level indicators have switches that start and stop the sump pumps when required. A timer starts each time the sump is pumped down to the low level setpoint.

(continued)

ABWR TS B 3.4-1 P&R, 07/22/93

RCS Leakage Detection Instrumentation B 3.4.5 BASES BACKGROUND If the sump fills to the high level setpoint before the (continued) timer ends, an alarm sounds in the control room, indicating a LEAKAGE rate into the sump in excess of a preset limit. A second timer starts when the sump ptmps start on high level.

Should this timer run out before the sump level reaches the low level setpoint, an alarm is sounded in the control room indicating a LEAKAGE rate into the sump in excess of a preset limit. A flow indicator in the discharge line of the drywell floor drain sump pumps provides flow indication in the control room.

The drywell air monitoring systems continuously monitor the drywell atmosphere for airborne particulate and gaseous radioactivity. A sudden increase of radioactivity, which may be attributed to RCPB steam or reactor water LEAKAGE, is annunciated in the control room. The drywell atmosphere particulate and gaseous radioactivity monitoring systems are not capable of quantifying leakage rates, but are sensitive enough to indicate increased LEAKAGE rates of I gpm within I hour.

Larger changes in LEAKAGE rates are detected in proportionally shorter times (Ref. 3).

Condensate from four of the six drywell coolers is routed to i

the drywell floor drain sump and is monitored by a flow transmitter that provides indication and alarms in the control room. This drywell air cooler condensate flow rate monitoring system serves as an added indicator, but not

___ quantifier, of RCS unidentified LEAKAGE.

t APPLICABLE A threat of significant compromise to the RCPB exists if the i

SAFETY ANALYSES barrier contains a crack that is large enough to propagate rapidly.

LEAKAGE rate limits are set low enough to detect the LEAKAGE emitted from a single crack in the RCPB (Refs. 4 and 5).

Each of the leakage detection systems inside the drywell is designed with the capability of detecting LEAKAGE i

less than the established LEAKAGE rate limits and providing appropriate alarm of excess LEAKAGE in the control room.

i A control room alarm allows the operators to evaluate the significance of the indicated LEAKAGE and, if necessary, shut down the reactor for further investigation and corrective action. The allowed LEAKAGE rates are well below the rates predicted for critical crack sizes (Ref. 6).

(continued)

ABWR TS B 3.4-2 P&R, 07/22/93

r~

)

RCS Leakage Detection Instrumentation B 3.4.5 BASES APPLICABLE Therefore, these actions provide adequate response before a SAFETY ANALYSES significant break in the RCPB can occur.

(continued)

RCS leakage detection instrumentation satisfies Criterion 1 of the NRC Policy Statement.

LC0 The drywell floor drain sump monitoring system is required to quantify the unidentified LEAKAGE from the RCS. Thus, for the system to be considered OPERABLE, either the flow monitoring or the sump level monitoring portion of the system must be OPERABLE. The other monitoring systems provide early alarms to the operators so closer examination of other detection systems will be made to determine the extent of any corrective action that may be required. With~

the leakage detection systems inoperable, monitoring for LEAKAGE in the RCPB is degraded.

1 APPLICABILITY In MODES 1, 2, and 3, leakage detection systems are required to be OPERABLE to support LC0 3.4.5.

This Applicability is consistent with that for LC0 3.4.5.

ACTIONS A_d With the drywell floor drain sump monitoring system inoperable, no other form of sampling can provide the equivalent information to quantify leakage. However, the drywell atmospheric activity monitor [and the drywell air cooler condensate flow rate monitor] will provide i

indications of changes in leakage.

With the drywell floor drain sump monitoring system inoperable, but with RCS unidentified and total LEAKAGE being determined every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (SR 3.4.5.1), operation may i

continue for 30 days.

The 30 day Completion Time of Required Action A.1 is acceptable, based on operating experience, considering the multiple forms of leakage detection that are still available. Required Action A.1 is modified by a Note that states that the provisions of LC0 3.0.4 are not applicable. As a result, a MODE change is allowed when the drywell floor drain sump monitoring system (continued)

ABWR TS B 3.4-3 P&R, 07/22/93

o i

s.

t

\\

j RCS Leakage Detection Instrumentation B 3.4.5 BASES ACTIONS Ad (continued) is inoperable. This allowance is provided because other instrumentation is available to monitor RCS leakage.

B.1 and B.2 With both gaseous and particulate drywell atmospheric monitoring channels inoperable, grab samples of the containment atmosphere shall be taken and analyzed to provide periodic leakage information.

[Provided a sample is obtained and analyzed every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , the plant may be operated for up to 30 days to allow restoration of at least one of the required monitors.]

[Provided a sample is obtained and analyzed every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , the plant may continue, operation since at least one other form of drywell leakage detection (i.e., air cooler condensate flow rate monitor) is avail able. ]

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval provides periodic information that is adequate to detect LEAP',GE. The 30 day Completion Time for restoration recognizes siat at least one other form of leakage detection is avetlable.

The Required Actions are modified by a Note that states that the provisions of LC0 3.0.4 are not applicable. As a result, a MODE change is allowed when both the gaseous and particulate primary containment atmospheric monitoring channels are inoperable.

This allowance is provided because other instrumentation is available to monitor RCS leakage.

L1 With the required drywell air cooler condensate flow rate monitoring system inoperable, SR 3.4.7.1 is performed every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to provide periodic information of activity in the drywell at a more frequent interval than the routine Frequency of SR 3.4.7.1.

The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months interval provides periodic information that is adequate to detect LEAKAGE and recognizes that other forms of leakage detection are available. However, this Required Action is modified by a Note that allows this action to be not applicable if the i~

required drywell atmospheric monitoring system is inoperable. Consistent with SR 3.0.1, Surveillances are not l

_ required to be performed on inoperable equipment.

(continued)

ABWR TS B 3.4-4 P&R, 07/22/93 L

1

c 1

=4 RCS Lemkage Detection Instrumentation B 3.4.5 BASES ACTIONS

- D.1 and D.2

~~

(continued)

With both the gaseous and particulate drywell atmospheric monitor channels and the drywell air cooler condensate flow rate monitor inoperable, the only means of detecting LEAKAGE is the drywell floor drain sump monitor. This Condition does not provide the required diverse means of leakage detection.

The Required Action is to restore either of the inoperable monitors to OPERABLE status within 30 days to regain the intended leakage detection diversity. The 30 day Completion Time ensures that the plant will not be operated in a degraded configuration for a lengthy time period. The Required Actions are modified by a Note that states that the provisions of LCO 3.0.4 are not applicable.

As a result, a MODE change is allowed when both the gaseous and particulate primary containment atmospheric monitoring channels and air cooler condensate flow rate are inoperable. This allowance is provided because other instrumentation is available to

__ monitor RCS leakage.

E.1 and E.2 If any Required Action of Condition A, B, [C, or D] cannot be met within the associated Completion Time, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions in an orderly manner and without challenging plant systems.

L.1 With all required monitors inoperable, no required automatic means of monitoring LEAKAGE are available, and immediate plant shutdown in accordance with LCO 3.0.3 is required.

SURVEILLANCE SR 3.4.7.1 REQUIREMENTS This SR requires the performance of a CHANNEL CHECK of the required drywell atmospheric monitoring system. The check gives reasonable confidence that the channel is operating (continued)

ABWR TS B 3.4-5 P&R, 07/22/93

r s

RCS Leakage Detection Instrumentation B 3.4.5 BASES SURVEILLANCE SR 3.4.7.1 (continued)

REQUIREMENTS properly. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is based on instrument reliability and is reasonable for detecting off normal conditions.

SR 3.4.7.2 This SR requires the performance of a CHANNEL FUNCTIONAL TEST of the required RCS leakage detection instrumentation.

The test ensures that the monitors can perform their function in the desired manner. The test also verifies the alarm setpoint and relative accuracy of the instrument string.

The Frequency of 31 days considers instrument reliability, and operating experience has shown it proper

~

for detecting degradation.

SR 3.4.7.3 This SR requires the performance of a CHANNEL CALIBRATION of the required RCS leakage detection instrumentation channels.

The calibration verifies the accuracy of the instrument string, including the instruments located inside containment. The Frequency of [18] months is a typical refueling cycle and considers channel reliability.

Operating experience has proven this Frequency is acceptable.

REFERENCES 1.

10 CFR 50, Appendix A, GDC 30.

2.

Regulatory Guide 1.45, May 1973.

3.

FSAR, Section [5.2.5.2].

4.

GEAP-5620, April 1968.

5.

NUREG-75/067, October 1975.

6.

FSAR, Section [5.2.5.5.3].

ABWR TS B 3.4-6 P&R, 07/22/93

,4 s

-s RCS Specific Activity B 3.4.6 8 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.6 RCS Specific Activity BASES BACKGROUND During circulation, the reactor coolant acquires radioactive materials due to release of fission products from fuel leaks into the coolant and activation of corrosion products in the reactor coolant. These radioactive materials in the coolant can plate out in the RCS, and, at times, an accumulation will break away to spike the normal level of radioactivity.

The release of coolant during a Design Basis Accident (DBA) could send radioactive materials into the environment.

Limits on the maximum allowable level of radioactivity in the reactor coolant are established to ensure, in the event.

l of a release of any radioactive material to the environment during a DBA, radiation doses are maintained within the limits of 10 CFR 100 (Ref. B 3.4.6-1).

This LC0 contains both' iodine and gross specific activity limits.

The iodine isotopic activities per gram of reactor coolant are expressed in terms of a DOSE EQUIVALENT I-131.

Total specific reactor coolant activity is limited on the basis of the weighted average beta and gamma energy levels in the coolant. The allowable levels are intended to limit i

the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months radiation dose to an individual at the site boundary to a small fraction of the 10 CFR 100 limit.

APPLICABLE Analytical methods and assumptions involving radioactive SAFETY ANALYSES material in the primary coolant are presented in the SSAR (Ref. B3.4.6-2).

The specific activity in the reactor coolant (the source term) is an initial condition for evaluation of the consequences of an accident due to a main steam line break (MSLB) outside containment.

No fuel damage is postulated in the MSLB accident, and the release of

)

radioactive material to the environment is assumed to end when the main steam isolation valves (MSIVs) close completely.

This MSLB release forms the basis for determining offsite i

doses (Ref. B 3.4.6-2).

The limits on the specific activity j

of the primary coolant ensure that for a MSLB with an assumed pre-accident iodine spike corresponding to the (continued)

ABWR TS B 3.4-1 P&R, 07/22/93

r e

RCS Specific Activity B 3.4.6 BASES APPLICABLE maximum iodine concentration, the calculated doses will not SAFETY ANALYSES exceed the guideline values of 10 CFR 100 and for a MSLB (continued) with an assumed iodine concentration corresponding to the equilibrium value for continued full power cperation the 2-hour thyroid and whole body doses at the site boundary, will J

l not exceed 10% of the dose guidelines of 10 CFR 100.

The limits on specific activity are values from a parametric i

evaluation of typical site locations. These limits are conservative because the evaluation considered more l

restrictive parameters than for a specific site, such as the l

location of the site boundary and the meteorological conditions of the site.

RCS specific activity satisfies Criterion 2 of the NRC Policy Statement.

LC0 The specific iodine activity is limited to s 7400 g/gm (0.2 pCi/gm) DOSE EQUIVALENT I-131, and the gross specific activity is limited to 3.7E + 6/E q/gm (100/E pCi/gm).

These limits ensure the source term assumed in the safety analysis for the MSLB is not exceeded, so any release of radioactivity to the environment during an MSLB is less than a small fraction of the 10 CFR 100 limits.

APPLICABILITY In MODE 1, and MODES 2 and 3 with any main steam line not isolated, limits on the primary coolant radioactivity are applicable since there is an escape path for release of radioactive material from the primary coolant to the environment in the event of an MSLB outside of primary containment.

In MODES 2 and 3 with the MSIVs closed, such limits do not apply since an escape path does not exist.

In MODES 4 and 5, no limits are required since the reactor is not pressurized and the potential for leakage is reduced.

(continued)

ABWR TS B 3.4-2 P&R, 07/22/93

a b

M RCS Specific Activity B 3.4.6 BASES ACTIONS A.1 and A.2 When the reactor coolant specific activity exceeds the LCO DOSE EQUIVALENT I-131 limit, but is s 148,000 g/gm (4.0 pCi/gm), samples must be analyzed for DOSE EQUIVALENT I-131 at least once every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

In addition, the specific activity must be restored to the LCO limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

The Completion Time of once every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is based on the time needed to take and analyze a sample.

The 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Completion Time to restore the activity level provides a reasonable time for temporary coolant activity increases (iodine spikes or crud bursts) to be cleaned up with the normal processing systems.

~

B.l. B.2.1. B.2.2.1. and B.2.2.2 If the DOSE EQUIVALENT I-131 cannot be restored to s 7400 i

q.gm (0.2 pCi/gm) within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , or if at any time it is

> 148,000 q/gm (4.0 yCi/gm), it must be determined at least i

every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and all the main steam lines must be isolated j

within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Isolating the main steam lines precludes i

the possibility of releasing radioactive material to the environment in an amount that is more than a small fraction of the requirements of 10 CFR 100 during a postulated MSLB accident.

Alternately, the plant can be brought to MODE 3 within

12. hours and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

This option is provided for those instances when isolation of main steam lines is not desired (e.g., due to the decay heat loads).

In MODE 4, the requirements of the LC0 are no longer applicable.

1 The Completion Time of once every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is the time needed to take and analyze a sample. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Completion Time is reasonable, based on operating experience, to isolate the main steam lines in an orderly manner and without challenging plant systems. Also, the allowed Completion Times for Required Actions B.2.2.1 and B.2.2.2 for bringing the plant to MODES 3 and 4 are reasonable, based on i

operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

(continued)

ABWR TS B 3.4-3 P&R, 07/22/93

~

r s

RCS Specific Activity B 3.4.6 BASES ACTIONS C.I. C.2.1. and C.2.2 (continued)

When the reactor coolant specific activity is > 3.7E + 6E q/gm (100/E pci/gm), all main steam lines must be isolated within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Closing the MSIVs eliminates the potential radioactivity release path to the environment during an MSLB event.

Alternately, the plant can be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

This option is provided for those instances when isolation of main steam lines is not desired (e.g., due to the decay heat loads).

In MODE 4, the requirements of the LC0 are no longer applicable.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Completion Time is reasonable, based on

~

operating experience, to isolate the main steam lines without challenging plant systems. Also, the allowed Completion Times for Required Actions C.2.1 and C.2.2 are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This SR requires performing a gamma isotopic analysis as a measure of the gross specific activity of the reactor coolant at least once per 7 days. While basically a quantitative measure of radionuclides with half lives longer than 15 minutes, excluding iodines, this measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken.

This Surveillance provides an indication of any increase in gross specific activity.

Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LC0 limit under normal operating conditions.

The 7 day Frequency considers the unlikelihood of a gross fuel failure during this short time frame.

(continued)

ABWR TS B 3.4-4 P&R, 07/22/93 1

o i

RCS Specific Activity B 3.4.6 BASES SURVEILLANCE SR 3.4.6.2 REQUIREMENTS (continued)

This Surveillance is performed to ensure iodine remains within limit during normal operation. The 31 day Frequency is adequate to trend changes in the iodine activity level considering gross specific activity is monitored every 7 days.

This SR is modified by a Note that requires this Surveillance to be performed' only in MODE 1 because the level of fission products generated in other MODES is much less.

i SR 3.4.6.3 A radiochemical analysis for E determination is required every 184 days with the plant operating in MODE 1 with i

equilibrium conditions.

The E determination directly relates to the LCO and is required to verify plant operation within the gross specific activity LCO limit. The analysis for E is a nieasurement of the average energies per disintegration for isotopes with half lives longer than 15 minutes, excluding iodines.

Operating experience has shown that E does not change rapidly and the Frequency of 184 days recognizes this.

This SR has been modified by a Note that states that sampling is required to be performed within 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for at least 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . This ensures the i

radioactive materials are at equilibrium so the analysis for E is representative and not skewed by a crud burst or other similar abnormal event.

REFERENCES 1.

10 CFR 100.11, 1973.

2.

ABWR SSAR, Chapter 15.

)

i ABWR TS B 3.4-5 P&R, 07/22/93

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RHR. Shutdown Cooling System-Hot Shutdown j

B 3.4.7 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System-Hot Shutdown i

i BASES BACKGROUND Irradiated fuel in the shutdown reactor core generates heat during the decay of fission products and increases the temperature of the reactor coolant.

This decay heat must be removed to reduce the temperature of the reactor coolant to

$ 93*C (200*F). This decay heat removal is in preparation for performing refueling or maintenance operations, or for keeping the reactor in the Hot Shutdown condition.

The three redundant, manually controlled shutdown cooling subsystems of the RHR System provide decay heat removal.

Each loop consists of a motor driven pump, a heat exchanger, i

and associated piping and valves.

Each loop has its own dedicated suction from the RPV.

Each pump discharges the reactor coolant, after circulation through the respective heat exchanger, to the reactor via feedwater line "A" for RHR loop "A" and via the RHR low pressure flooder spargers for loops "B" and "C".

The RHR heat exchangers transfer heat to the Reactor Building Cooling Water System (LC0 3.7.2).

APPLICABLE Decay heat remor=1 by the RHR System in the shutdown cooling SAFETY ANALYSES mode is not required for mitigation of any event or accident evaluated in the safety analyses.

Decay heat removal is, however, an important safety function that must be l

accomplished or core damage could result. Although the RHR Shutdown Cooling System does not meet a specific criterion of the NRC Policy Statement, it was identified in the NRC Policy Statement as a significant contributor to risk reduction. Therefore, the RHR Shutdown Cooling System is retained as a Technical Specification.

LCO Two RHR shutdown cooling subsystems are required to be OPERABLE, and one shutdown cooling subsystem must be in i

operation. An OPERABLE RHR shutdown cooling subsystem i

consists of one OPERABLE RHR pump, a heat exchanger, a::d the (continued)

ABWR TS B 3.4-1 P&R, 07/22/93 l

l

RHR Shutdown Cooling System-Hot Shutdown

[

B 3.4.7 BASES LCO associated piping and valves.

Each shutdown cooling (continued) subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat.

In MODE 3, [one] RHR shutdown cooling subsystem can provide the required cooling, but

[two] subsystems are required to be OPERABLE to provide redundancy. Operation of [one] subsystem can maintain or reduce the reactor coolant temperature as required.

However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly r

continuous operation is required.

Note 1 permits two RHR shutdown cooling subsystems to be i

shut down for a period of E hours in an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period, provided [one) subsystem 1.s OPERABLE.

Note 2 allows one RHR shutdown cooling subsystem to be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for performance of surveillance tests provided one of the remaining RHR shutdown cooling subsystem is OPERABLE.

These tests may be on the affected RHR System or on some other plant system or component that necessitates placing the RHR system in an inoperable status during the performance. This is permitted because the core heat generation can be low enough and the heatup rate slow enough to allow some changes to the RHR subsystems or other operations requiring RHR flow interruption and loss of redundancy.

APPLICABILITY In MODES I and 2, and in MODE 3 with reactor steam dome pressure above the RHR cut in permissive pressure, this LCO is not applicable. Operation of the RHR System in the shutdown cooling mode is not allowed above this pressure because the RCS pressure may exceed the design pressure of the shutdown cooling piping. Decay heat removal at reactor pressures above the RHR cut in permissive pressure is typically accomplished by condensing the steam in the main condenser. Additionally, in MODE 2 below this pressure, the OPERABILITY requirements for the Emergency Core Cooling Systems (ECCS) (LCO 3.5.1, "ECCS-Operating") do not allow placing the low pressure RHR shutdown cooling subsystem into operation.

In MODE 3 with reactor steam dome pressure below the RHR cut in permissive pressure (i.e., the actual pressure at which the interlock resets) the RHR System may be operated in the (continued)

ABWR TS B 3.4-2 P&R, 07/22/93

'e t

RHR Shutdown Cooling System-Hot Shutdown B 3.4.7 BASES APPLICABILITY shutdown cooling mode to remove decay heat to reduce or (continued) maintain coolant temperature.

The requirements for decay heat removal in MODES 4 and 5 are discussed in LCO 3.4.8, " Residual Heat Removal (RHR)

Shutdown Cooling System-Cold Shutdown"; LC0 3.9.7,

" Residual Heat Removal (RHR)-High Water Level"; and LC0 3.9.8, " Residual Heat Removal (RHR)-Low Water Level."

ACTIONS A.I. A.2. and A.3 With one or both required RHR shutdown cooling subsystems inoperable for decay heat removal, the inoperable required _

subsystem (s) must be restored to OPERABLE status without del ay.

In this condition, the remaining OPERABLE required subsystem can provide the necessary decay heat removal.

The overall reliability is reduced, however, because a single failure in the OPERABLE required subsystem could result in reduced RHR shutdown c60 ling capability.

With one of the two required RHR shutdown cooling subsystems inoperable, the remaining required subsystem is capable of providing the required decay heat removal.

However, the overall reliability is reduced. Therefore an alternate required method of decay heat removal must be provided (such as the third RHR Shutdown Cooling Subsystem). With both RHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability.

This re-establishes backup decay heat removal capabilities, similar to the requirements of the LCO. The I hour Completion Time is based on the decay heat removal function and the probability of a loss of the available decay heat removal capabilities.

The required cooling capacity of the alternate method should be ensured by verifyinC (by calculation or demonstration)

)

its capability to maintain or reduce temperature.

Decay i

heat removal by ambient losses can be considered as contributing to the alternate method capability. Alternate methods that can be used include (but are not limited to)

[the third RHR Shutdown Cooling Subsystem), the Spent Fuel Pool Cooling System, or the Reactor Water Cleanup System.

(continued)

ABWR TS B 3.4-3 P&R, 07/22/93

RHR Shutdown Cooling System-Hot Shutdown B 3.4.7 BASES ACTIONS A.1. A.2. and A.3 (continued)

However, due to the potentially reduced reliability of the alternate methods of decay heat removal, it is also required to reduce the reactor coolant temperature to the point where HODE 4 is entered.

B.1. B.2. and 8.3 With no RHR shutdown cooling subsystem in operation, except as is permitted by the LCO Note, reactor coolant circulation by the RHR shutdown cooling subsystem or [enough RIPS so that at least [4] are operating] must be restored without del ay.

Until RHR operation is re-established, an alternate method of reactor coolant circulation must be placed into service.

3 This will provitie the necessary circulation for monitoring coolant temperature and pressure. The I hour Completion Time is based on the coolant circulation function and is modified such that the I hour is applicable separately for each occurrence involving a loss of coolant circulation.

Furthermore, verification of the functioning of the alternate method must be reconfirmed every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

This will provide assurance of continued temperature monitoring capability.

During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR Shutdown Cooling System), the reactor coolant temperature and pressure must be periodically monitored to ensure proper function of the alternate method. The once per hour Completion Time is deemed appropriate.

SURVEILLANCE SR 3.4.7.1 REQUIREMENTS This Surveillance verifies that [one] RHR shutdown cooling subsystem is in operation and circulating reactor coolant.

The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient in view (continued)

ABWR TS B 3.4-4 P&R, 07/22/93

a:

16 l

RHR Shutdown Cooling System--Hot Shutdown B 3.4.7 i

BASES SURVEILLANCE SR 3.4.7.1 (continued)'

REQUIREMENTS of other visual and audible indications available to the operator for monitoring the RHR subsystem in the control

-j room.

This ' Surveillance is modified by a Note allowing sufficient' l

time to align the RHR System for shutdown cooling operation after clearing the pressure interlock that isolates the i

system. The Note takes exception to the. requirements of the Surveillance being met-(i.e., forced coolant circulation is not required for this initial 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period), which also allows entry into the Applicability of this Specification in j

accordance with SR 3.0.4 since the Surveillance will not be "not met" at the time of entry into the Applicability.

i REFERENCES None.

i l

i

-)

-j ABWR TS B 3.4-5 P&R, 07/22/93 j

e.

s RCS P/T Limits B 3.4.8 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.8 RCS Pressure and Temperature (P/T) Limits BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

The PTLR contains P/T limit curves for heatup, cooldown, and inservice leak and hydrostatic testing, and data for the maximum rate of change of reactor coolant temperature. The.

heatup curve provides limits for both heatup and criticality.

Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

The LC0 establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the

~

component of most concern in regard to brittle failure.

Therefore, the LCO limits apply mainly to the vessel.

10 CFR 50, Appendix G (Ref. B 3.4.8-1), requires the establishment of P/T limits for material fracture toughness requirements of the RCPB materials.

Reference B 3.4.8-1 requires an adequate margin to brittle failure during normal 4

operation, anticipated operational occurrences, and system hydrostatic tests.

It mandates the use of the American Society of Mechanical Engineers (ASME) Code,Section III, Appendix G (Ref. B 3.4.8-2).

The actual shift in the RT of the vessel material will be gg established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. B 3.4.8-3) and 10 CFR 50, Appendix H (Ref. B 3.4.8-4).

The operating P/T limit curves will be j

(continued)

ABWR TS B 3.4-1 P&R, 07/22/93

s RCS P/T Limits B 3.4.8 BASES BACKGROUND adjusted, as necessary, based on the evaluation findings and (continued) the recommendations of Reference B 3.4.8-5.

The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive regions.

The heatup curve represents a different set of restrictions than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed.

The.

thermal gradient reversal alters the location of the tensile stress between the outer and inner walls.

The criticality limits include the Reference 3.4.8-1 requirement that they be at least 22*C (40*F) above the heatup curve or the cooldown curve and not lower than the minimum permissible temperature for the inservice leak and hydrostatic testing.

The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident.

In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components. The ASME Code,Section XI, Appendix E (Ref. B 3.4.8-6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.

APPLICABLE The P/T limits are not derived from Design Basis Accident SAFETY ANALYSES (DBA) analyses. They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, a condition that is unanalyzed.

Reference B 3.4.8-7 establishes the methodology for determining the P/T limits.

(continued)

ABWR TS B 3.4-2 P&R, 07/22/93

)

s e

RCS P/T Limits B 3.4.8-BASES APPLICABLE Since the P/T limits are not derived from any DBA, there are SAFETY ANALYSES no acceptance limits related to the P/T limits.

Rather, the (continued)

P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.

RCS P/T limits satisfy Criterion 2 of the NRC Policy Statement.

l LCO The elements of this LC0 are:

j a.

RCS pressure, temperature, and heatup or cooldown rate are within the limits specified in the PTLR.

b.

RCS pressure and temperature are within the criticality limits specified in the PTLP.

The reactor vessel flange and the head flange c.

temperatures are within the limits of the PTLR when reactor vessel head bolting studs are tensioned.

These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide margin to nonductile failure.

The rate of change of temperature limits control the thermal gradient through the vessel wall and are used as inputs for -

calculating the heatup, cooldown, and inservice leak and

~

hydrostatic testing P/T limit curves. Thus, the LCO for the J

rate of change of temperature restricts stresses caused by i

thermal gradients and also ensures the validity of the P/T limit curves.

Violation of the limits places the reactor vessel outside of 1

the bounds of the stress analyses and can increase stresses in other RCS components. The consequences depend on several factors, as follows:

i a.

The severity of the departure from the allowable operating pressure temperature regime or the severity of the rate of change of temperature; b.

The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronounced); and (continued)

ABWR TS B 3.4-3 P&R, 07/22/93

e' a

RCS P/T Limits B 3.4.8 BASES LCO c.

The existence, sizes, and orientations of flaws in (continued) the vessel material.

APPLICABILITY The potential for violating a P/T limit exists at all times.

For example, P/T limit violations could result from ambient temperature conditions that result in the reactor vessel metal temperature being less than the minimum allowed s

temperature for boltup. Therefore, this LCO is applicable even when fuel is not loaded in the core.

1 ACTIONS The Actions designated by this Specification are based on the premise that a violation of the limits occurred during normal plant maneuvering.

Severe violations caused by abnormal transients, which may be accompanied by equipment failures, may also require additional actions based on emergency operating procedures.

A.1 and A.2 Operation outside the P/T limits while in MODE I, 2, or 3 roust be corrected so that the RCPB is returned to a condition that has been verified by stress analyses.

The 30 minute Completion Time reflects the urgency of restoring the parameters to within the analyzed range. Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner.

Besides restoring operation within limits, an evaluation is required to determine if RCS operation can continue.

The evaluation must verify the RCPB integrity remains acceptable i

and must be completed if continued operation is desired.

Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components.

ASME Code,Section XI, Appendix E (Ref. B 3.4.8-6), may be used to support the evaluation.

However, its use is restricted to evaluation of the vessel beltline.

(continued)

ABWR TS B 3.4-4 P&R, 07/22/93

't "0

RCS P/T Limits B 3.4.8 BASES ACTIONS A.1 and A.2 (continued)

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is reasonable to accomplish the evaluation of a mild violation. More severe violations may require special, event specific stress analyses or inspections. A favorable evaluation must be completed if continued operation is desired.

Condition A is modified by a Note requiring Required Action A.2 be completed whenever the Condition is entered.

The Note emphasizes the need to perform the evaluation of i

the effects of the excursion outside the allowable limits.

Restoration alone per Required Action A.1 is insufficient because higher than analyzed stresses may have occurred and i

may have affected the RCPB integrity.

1 B.1 and B.2 If a Required Action and associated Completion Time of Condition A are not met, the plant must be brought to a lower MODE because either the RCS remained in an unacceptable P/T region for an extended period of increased stress, or a sufficiently severe event caused entry into an unacceptable region.

Either possibility indicates a need l

for more careful examination of the event, best accomplished with the RCS at reduced pressure and temperature. With the reduced pressure and temperature conditions, the possibility of propagation of undetected flaws is decreased.

Pressure and temperature are reduced by bringing the plant to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

1 Operation outside the P/T limits in other than MODES 1, 2, and 3 (including defueled conditions) must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses. The Required Action must be initiated without delay and continued until the limits are restored.

(continued)

ABWR TS B 3.4-5 P&R, 07/22/93 l

o RCS P/T Limits B 3.4.8 BASES ACTIONS C.1 and C.2 (continued)

Besides restoring the P/T limit parameters to within limits, an evaluation is required to determine if RCS operation is allowed. This evaluation must verify that the RCPB integrity is acceptable and must be completed before approaching criticality or heating up to > 93*C (200*F).

Several methods may be used, including comparison with pre-analyzed transients, new analyses, or inspection of the components. ASME Section-XI, Appendix E (Ref. B 3.4.8-6),

may be used to support the evaluation; however, its use is restricted to evaluation of the beltline.

SURVEILLANCE SR 3.4.8.1

~

REQUIREMENTS Verification that operation is within PTLR limits is i

required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes.

This Frequency is considered reasonable in view of the control room indication available to monitor RCS status. Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits assessment and correction of minor deviations.

Surveillance for heatup, cooldown, or inservice leakage and hydrostatic testing may be discontinued when the criterfa given in the relevant plant procedure for ending the activity are satisfied.

This SR has been modified by a Note that requires this Surveillance to be performed only during system heatup and couldown operations and inservice leakage and hydrostatic testing.

SR 3.4.8.2 A separate limit is used when the reactor is approaching criticality.

Consequently, the RCS pressure and temperature must be verified within the appropriate limits before j

withdrawing control rods that will make the reactor critical.

j i

(continued)

ABWR TS B 3.4-6 P&R, 07/22/93 i

o e

RCS P/T Limits B 3.4.8 BASES SURVEILLANCE SR 3.4.8.2 REQUIREMENTS (continued)

Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal.

SR 3.4.8.3 and SR 3.4.8.4 and SR 3.4.8.5 Limits on the reactor vessel flange and head flange temperatures are generally bounded by the other P/T limits during system heatup and cooldown. However, operations approaching MODE 4 from MODE 5 and in MODE 4 with RCS temperature less than or equal to certain specified values require assurance that these temperatures meet the LCO limits.

The flange temperatures must be verified to be above the limits 30 minutes befo're and while tensioning the vessel head bolting studs to ensure that once the head is tensioned the limits are satisfied. When in MODE 4 with RCS temperature 5 27'C (80*F), 30 minute checks of the flange temperatures are required because of the reduced margin to the limits. When in MODE 4 with RCS temperature $ 38'C (100*F), monitoring of the flange temperature is required every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to ensure the temperatures are within the limits specified in the PTLR.

The 30 minute Frequency reflects the urgency of maintaining the temperatures within limits, and also limits the time that the temperature limits could be exceeded. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is reasonable based on the rate of temperature change possible at these temperatures.

REFERENCES 1.

10 CFR 50, Appendix G.

2.

ASME, Boiler and Pressure Vessel Code,Section III, Appendix G.

3.

ASTM E 185-82, July 1982.

4.

10 CFR 50, Appendix H.

(continued)

ABWR TS B 3.4-7 P&R, 07/22/93

s RCS P/T Limits B 3.4.8 BASES REFERENCES 5.

Regulatory Guide 1.99, Revision 2, May 1988.

(continued) 6.

ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.

f 7.

NED0-21778-A, December 1978.

l O

f e

I ABWR TS B 3.4-8 P&R, 07/22/93

e a

Reactor Stearr Dome Pressure B 3.4.9 8 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.9 Reactor Steam Dome Pressure BASES BACKGROUND The reactor steam dome pressure is an assumed initial condition of Design Basis Accidents (DBAs) and transients and is also an assumed value in the determination of compliance with reactor pressure vessel overpressure protection criteria.

t i

2 APPLICABLE The reactor steam dome pressure of s 73.1 Kg/cm (1040 psig).

SAFETY ANALYSES (7170 KPa) is an initial condition of the vessel overpressure protection analysis of -

Reference B 3.4.9-1.

This analysis assumes an initial maximum reactor steam dome pressure and evaluates the response of the pressure relief system, primarily the safety / relief valves, during the limiting pressurization transient.

The determination of compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure ensures that the assumptions of the overpressure protection analysis are conserved. Reference B.3.4.9-2 also assumes an initial reactor steam dome pressure for the analysis of DBAs and transients used to determine the limits for fuel cladding integrity MCPR (see Bases for.

LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)") and 1% cladding plastic strain (see Bases for LC0 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)").

Reactor steam dome pressure satisfies the requirements of Criterion 2 of the NRC Policy Statement.

LC0 Thespecifiedreactorsteamdomepressurelimitofs73.1 i

Kg/cm (1040 psig) (7170 KPa) ensures the plant is operated within the assumptions of the transient analyses. Operation (continued)

ABVR TS B 3.4-1 P&R, 07/22/93

s.

3 i

Reactor Steam Dome Pressure B 3.4.9 BASES above the limit may result in a transient response more severe than analyzed.

APPLICABILITY In MODES 1 and 2, the reactor steam dome pressure is (continued) required to be less than or equal to the 1;mit.

In these MODES, the reactor may be generating significant steam, and the DBAs and transients are bounding.

In MODES 3, 4, and 5, the limit is not applicable because the reactor is shut down.

In these MODES, the reactor pressure is well below the required limit, and no anticipated events will challenge the overpressure limits.

ACTIONS M

With the reactor steam dome pressure greater than the limit, prompt action should be taken to reduce pressure to below the limit and return the reactor to operation within the bounds of the analyses. The 15 minute Completion Time is reasonable considering the importance of maintaining the pressure within limits. This Completion Time also ensures that the probability of an accident while pressure is greater than the limit is minimal.

If the operator is unable to restore the reactor steam dome pressure to below the limit, then the reactor should be brought to MODE 3 to be.within the assumptions of the transient analyses.

M If the reactor steam dome pressure cannot be restored to within the limit within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply.

To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed' Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

(continued)

ABWR TS B 3.4-2 P&R, 07/22/93

s se Reactor Steam Dome Pressure B 3.4.9 BASES SURVEILLANCE SR 3.4.9.1 REQUIREMENTS Verification that reactor steam dome pressure is s 73.1 Kg/cm2 (1040 psig) (7170 KPa) ensures that the initial conditions of the DBAs and transients are met. Operating SURVEILLANCE SR 3.4.9.1 (continued) experience has shown the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency to be sufficient for identifying trends and verifying operation within safety analyses assumptions.

REFERENCES 1.

ABWR SSAR, Section 5.2.2.

2.

ABWR SSAR, Chapter 15.

9

)

(continued) j ABWR TS B 3.4-3 P&R, 07/22/93

\\

l

ECCS-Operating B 3.5.1 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.1 ECCS-Operating BASES BACKGROUND The ECCS is designed, in conjunction with the primary and secondary containment, to limit the release of radioactive materials to the environment following a loss of coolant accident (LOCA). The ECCS directs water to both inside and outside the core shroud to cool the core during a LOCA.

The ECCS network is composed of the High Pressure Core Flooder (HPCF) System, the Reactor Core Isolation Cooling (RCIC)

System, and the low pressure Core Flooder (LPFL) mode of the Residual Heat Removal (RHR) System.

The ECCS also consists _

of the Automatic Depressurization System (ADS). The suppression pool provides the required source of water for the ECCS. Although no credit is taken in the safety analyses for the condensate storage tank (CST), it is capable of providing a source of water for both the RCIC system and the two HPCF subsystems.

On receipt of an initiation signal, ECCS pumps automatically start; simultaneously the. system aligns, and the pumps inject water, taken either from the CST or suppression pool, into the Reactor Coolant System RCS) as RCS pressure is overcome by the discharge pressur(e of the ECCS pumps.

Although the system is initiated, ADS action is delayed, to allow time for conformation of the initiating signal. The discharge pressure of the HPCF pumps almost immediately exceeds that of the RCS, and the pumps inject coolant into the flooding sparger above the core. Once the steam driven RCIC turbine has accelerated, the RCIC pumps discharge pressure also quickly exceeds that of the RCS and the pump injects coolant into the reactor pressure vessel (RPV) via one of the feedwater lines.

If the break is small, RCIC or either of the HPCF pumps will maintain coolant inventory, as well as vessel level, while the RCS is still pressurized.

If RCIC and HPCF fail, they are backed up by ADS in combination with the LPFL.

In this event, the ADS timed sequence would be allowed to time out and open the selected safety / relief valves (S/RVs), depressurizing the RCS and allowing the LPFL to overcome RCS pressure and inject coolant into the vessel.

If the break is large, RCS pressure initially drops rapidly, and the HPCF and LPFL subsystems cool the core.

(continued)

BWR/6 STS B 3.5-1 P&R, 07/22/93

s ECCS-Operating 8 3.5.1 BASES Water from the break returns to the suppression pool where it is used again and again.

Water in the suppression pool is circulated through a heat exchanger cooled by the Reactor Building Cooling Water (RCW) System.

Depending on the location and size of the break, portions of the ECCS may be ineffective; however, the overall design is effective in cooling the core regardless of the size or location of the piping break. Apart from its ECCS function, the RCIC System is also designed to operate either automatically or manually following reactor pressure vessel (RPV) isolation accompanied by a loss of coolant flow from the feedwater system to provide adequate core cooling and control of RPV water level. Under these conditions, the HPCF and RCIC systems perform similar functions.

The RCIC System design requirements ensure that the criteria of Reference 10CFR50,.

Appendix A, GDC 33 are satisfied.

All ECCS subsystems are designed to ensure that no single active component failure will prevent automatic initiation and successful operation of the minimum required ECCS subsystems.

ECCS injection systems are arranged in three separate divisions each comprised of a high pressure and low pressure subsystem.

ECCS Division 1 consists of the RCIC system and LPFL-A.

ECCS Division 2 consists of HPCF-B and LPFL-B, ECCS Division 3 consists of HPCF-C and LPFL-C, LPFL is an independent operating mode of the RHR System.

There are three LPFL subsystems.

Each LPFL subsystem (Ref. 83.5.1) consists of a motor driven pump, a heat exchanger, piping, and valves to transfer water from the suppression pool to the RPV.

Each LPFL subsystem has its own suction and discharge piping. Each LPFL subsystem takes suction from the suppression pool.

LPFL subsystems B and C have dedicated discharge nozzles to the RPV that connect to l

flooding sparges in the vessel annulus area outside the core shroud.

LPFL subsystem A discharges to one of the main l

feedwater injection lines and thus also supplies coolant to the vessel annulus area outside the core shroud via the feed water sparger. The LPFL subsystems are designed to provide l

core cooling at low RPV pressure.

Upon receipt of an initiation signal, each LPFL pump is automatically started approximately 10 seconds after electrical power is available.

When the RPV pressure drops sufficiently, LPFL flow to the RPV begins.

RHR System valves in the LPFL flew path are automatically positioned to ensure the proper flow (continued)

BWR/6 STS B 3.5-2 P&R, 07/22/93

i

^

i ECCS-Operating B 3.5.1 BASES path for water from the suppression pool to inject into the RPV. A discharge test line is provided to route water from i

and to the suppression pool to allow testing of each LPFL pump without injecting water into the RPV.

BACKGROUND The HPCS System (Ref. 3) consists of a single motor driven pump, a spray sparger above the core, and piping and valves to transfer water from the suction source to the sparger.

Suction piping is provided from the CST and the suppression pool.

Pump suction is normally aligned to the CST source to minimize injection of suppression pool water into the RPV.

However, if the CST water supply is low or the suppression pool level is high, an automatic transfer to the suppression, pool water source ensures a water supply for continuous operation of the HPCS System.

The HPCS System is designed to provide core cpoling over a wide range of RPV pressures (0 to 82.75 Kg/cm vessel to suction source). Upon receipt of an initiation signal, the HPCF pumps automatically start-(when AC power is avail'able) and valves in the flow path begin to open.

Since the HPCF System is designed to operate over the full range of expected RPV pressures, HPCF flow begins as soon as the necessary valves are open. A full flow test line is provided to route water from and to the CST to allow testing of the HPCF System during normal operation without injecting water into the RPV.

The RCIC System (Ref. 83.5.1-1) consists of a steam driven turbine pump unit, piping, and valves to provide steam to the turbine, as well as piping and valves to transfer water from the suction source to the core via the feedwater system line. Suction piping is provided from the condensate storage tank (CST) and the suppression pool. Pump suction is normally aligned to the CST to minimize injection of suppression pool water into the RPV. However, if the CST water supply is low, or the suppression pool level is high, an automatic transfer to the suppression pool water source ensures a water supply for continuous operation of the RCIC System.

The steam supply to the turbine is piped from main steam line B, upstream of the inboard main steam line isolation valve.

The RCIC System is designed to provide corp cooling for a wide range of reactor pressures, 11.6Kg/cm g(165 psig) to 2

81.2 Kg/cm g(1155 psig). Upon receipt of an initiation (continued)

BWR/6 STS B 3.5-3 P&R, 07/22/93

A 1

ECCS--Operating B 3.5.1 BASES signal, the RCIC turbine accelerates to a specified speed.

As the RCIC flow increases, the turbine control valve is automatically adjusted to maintain design flow.

Exhaust steam from the RCIC turbine is discharged to the suppression pool.

A full flow test line is provided to route v'+er from and to the suppression pool to allow testing of the RCIC System during normal operation without injecting water into the RPV.

The ECCS pumps are provided with minimum flow bypass lines, which discharge to the suppression pool. The valves in these lines automatically open to prevent pump damage due to overheating when other discharge line valves are closed or RPV pressure is greater than the LPFL pump discharge pressures following system initiation.

To ensure rapid delivery of water to the RPV and to minimize water hammer effects, the ECCS discharge line " keep fill " systems are designed to maintain all pump discharge lines filled with water.

The ADS (Ref. B3.5.1-1) consists of 8 of the 20 S/RVs.

It is designed to provide depressurization of the primary system during a small break LOCA if RCIC and HPCF fail or l

are unable to maintain required water level in the RPV. ADS operation reduces the RPV pressure to within the operating pressure range of the low pressure ECCS subsystems (LPFL),

so that these subsystems can provide core cooling.

Each ADS valve is supplied with pneumatic power from either its own dedicated accumulator located in the drywell, or from the atmospheric control system (ACS) directly when pneumatic power from the accumulators is not needed.

The ACS also l

supplies the nitrogen (at pressure) necessary to assure the ADS accumulators remain charged for use in emergency actuation.

(continued)

BWR/6 STS B 3.5-4 P&R, 07/22/93

A ECCS-Operating B 3.5.1 BASES (continued)

The ECCS performance is evaluated for the entire spectrum of break sizes for a postulated LOCA. The accidents for which ECCS operation is required are presented in References 83.5.1-2, B3.5.1-3, and B3.5.1-4.

The required analyses and assumptions are defined in 10 CFR 50 (Ref. 83.5.1-5), and the results of these analyses are described in Reference 9 B3.5.1-6.

This LC0 helps to ensure that the following acceptance criteria for the ECCS, established by 10 CFR 50.46 (Ref. B3.5.1-7), will be met following a LOCA assuming the worst case single active component failure in the ECCS:

a.

Maximum fuel element cladding temperature is s 1204*C (2200*F);

b.

Maximum cladding oxidation is s 0.17 times tnc total cladding thickness before oxidation; c.

Maximum hydrogen generation from zirconium water reaction is s 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react; d.

The core is maintained in a coolable geometry; and e.

Adequate long term cooling capability is maintained.

~

The limiting single failures are discussed in Reference B3.5.1-6.

For any LOCA, failure of ECCS subsystems in Division (HPCF-B and LPFL-B) or Division 3 (HPCF-C and LPFL-C) due to failure of its associated diesel generator is the most severe failure. One ADS valve failure i

is analyzed as a limiting single failure for events requiring ADS operation, however the above single failure of a diesel generator and associated motor driven ECCS injection subsystems in the division, is a more limiting failure. The remaining OPERABLE ECCS subsystems provide the capability to adequately cool the core and prevent excessive i

fuel damage. An additional function of the RCIC system is l

to respond to transient events by providing make-up coolant to the reactor vessel, including as an AC independent source during station blackout.

i (continued)

BWR/6 STS B 3.5-5 P&R, 07/22/93

A ECCS-Operating B 3.5.1 BASES (continued)

In order to provide increased margin to ECCS acceptance criteria (i.e., 10 CFR 50.46), the ECCS as designed to the more stringent goal of no core uncovery for any postulated DBA or transient event, even given the most limiting single failure.

This design philosophy resulted in substantially improved ECCS performance such that, when analyzed consistent with typical licensing basis methodologies, (i.e., assuming only the traditional limiting single failure), there was considerable margin relative to existing regulatory requirements. The magnitude of such margin suggested that the ECCS would still be able to perform its intended safety function, even under various situations with some equipment initially out of service or unavailable due to multiple postulated failures.

Therefore, further ECCS analyses were performed (see Reference B3.5.1-8) in an attempt to identify the minimum amount of ECCS equipment that must operate such that the plant could still meet the 10 CFR 50.46 acceptance criteria listed above.

Analyses were performed for a set of identified limiting scenarios, assuming the unavailability (or failure) of multiple ECCS subsystems, and using the same calculational methods as were used for the traditional design basis analyses. The results of these analyses demonstrated that

" success" (i.e., no violation of the above stated 50.46 limits) was achieved under various postulated accident scenarios provided at least one motor driven ECCS injection subsystem was capable of successfully injecting water into the RPV.

For any such scenarios also requiring

~

depressurization, " success" was achieved with the actuation of at least five SR/Vs in the ADS mode (in conjunction with successful vessel injection from the one required ECCS subsystem).

The ECCS satisfy Criterion 3 of the NRC Policy Statement.

LCO Each ECCS injection subsystem and seven ADS valves are required to be OPERABLE.

The ECCS injection subsystems are defined as the three LPFL subsystems, the RCIC System, and the HPCF subsystems.

The motor driven ECCS injection subsystems are defined as the two HPCF subsystems and the three LPFL subsystems.

(continued)

BWR/6 STS B 3.5-6 P&R, 07/22/93

+,

a ECCS-Operating B 3.5.1 BASES i

With less than the required number of ECCS subsystems OPERABLE during a limiting design basis LOCA concurrent with the worst case single failure, the margins to the limits specified in 10 CFR 50.46 (Ref. 10) could potentially be exceeded. All ECCS subsystems must therefore be OPERABLE to satisfy the single failure criterion required by 10 CFR 50.46 (Ref. 10). The ECCS is supported by other systems that provide automatic ECCS initiation signals (LC0 3.3.5.1, " Emergency Core Cooling System (ECCS) l Instrumentation"), service water to cool rooms containing l

ECCS equipment (LC0 3.7.1, "[ Standby Service Water (SSW)]

l System and [ Ultimate Heat Sink (UHS)]," and LC0 3.7.2, "High l

Pressure Core Spray (HPCS) Service Water System (SWS)"),

l electrical power (LCO 3.8.1, "AC Sources-Operating," and

~

LC0 3.8.4, "DC Sources-Operating").

A LPCI subsystem may be considered OPERABLE during alignment and operation for decay heat removal when below the actual RHR cut in permissive pressure in MODE 3, if capable of being manually realigned (remote or local) to the LPFL mode i

and not otherwise inoperable. At these low pressures and I

decay heat levels, a reduced complement of ECCS subsystems can provide the required core cooling, thereby allowing i

operation of an RHR shutdown cooling loop when necessary.

l APPLICABILITY All ECCS subsystems are required to be OPERABLE during MODES 1, 2, and 3 when there is considerable energy in the 1

reactor core and core cooling would be required to prevent fuel damage in the event of a break in the primary system l

piping.

InMODES2and3,theRCICsystemignotrequired to be OPERABLE when pressure is s 10.55Kg/cm g(150 psig) l since other ECCS subsystem can provide sufficient flow to the vessel.

In MODES 2 and 3, the ADS requiredwhenpressureiss3.515Kg/cm[unctionisnot g (50 psig) because the low pressure ECCS subsystems (LPFL) are capable of providing flow into the RPV below this pressure.

ECCS requirements for MODES 4 and 5 are specified in LCO 3.5.2, "ECCS-Shutdown. "

l l

(continued)

BWR/6 STS B 3.5-7 P&R, 07/22/93

m

^

ECCS--Operating B 3.5.1 BASES ACTIONS 611 If any one motor driven ECCS injection subsystem (i.e., HPCP or LPFL) is inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days.

In this Condition, the remaining OPERABLE subsystems provide more than adequate core cooling during a LOCA. However, overall ECCS reliability is reduced; and a single failure impacting one or more of the remaining OPERABLE subsystems concurrent with a LOCA would result in degraded ECCS performance and reduced margins to 10 CFR 50.46 acceptable criteria.

Nevertheless, even given the worse case single failure concurrent with a LOCA initiated from this condition, there will always be at least one ECCS subsystem available to

~

inject water into the RPV. Additional analyses of limiting design basis scenarios demonstrate that in such cases 10 CFR 50.46 acceptance criteria will still be met.

Furthermore, results of PRA sensitivity studies performed (see Reference B3.5.1-9) show that this situation is acceptable from an overall plant risk perspective. The 7 day Completion Time is thus based on the overall redundancy provided by the ECCS and its continued ability to perform its intended safety function, while assuring a return to full ECCS capability in a reasonable time so as to not significantly impact overall ECCS reliability.

9 (continued)

BWR/6 STS B 3.5-8 P&R, 07/22/93

A q

ECCS-Operating B 3.5.1 BASES Ed If the RCIC System is inoperable during MODE 1, or p0 DES 2 or 3 with reactor steam dome pressure > 10.55 Kg/cm g (150 psig), it must be restored to OPERABLE status within 14 days.

In this Condition, loss of the RCIC System will not affect the overall plant capability to provide makeup inventory at high RPV pressure since the two HPCF subsystems would still be available to provide make-up to the reactor during a loss of coolant accident (LOCA) in which the RPV remained at high pressure. However, for transients and certain and certain abnormal events with no LOCA, RCIC (as opposed to HPCF) is the preferred source of makeup coolant because its relatively small capacity and automatic flow

~

control capability allows for easier control of RPV water

~

level.

Furthermore, with its steam driven turbine, the RCIC system provides the only source of reactor coolant make-up during a complete loss of AC power event. Therefore, only a limited time is allowed to restore the inoperable RCIC to OPERABLE status. The 14 day Completion Time is based on PRA sensitivity studies (Reference B3.5.1-9) that show the unavailability of RCIC to be a greater contributor to

~

overall plant risk than the unavailability of any other ECCS subsystem. Thus, a more restrictive completion time is imposed relative to Condition A.

Additional a 14 day Completion Time for RCIC has been found to be acceptable through operating experience.

~

C.1 With any two ECCS injection subsystems inoperable, at least one ECCS injection subsystem must be restored to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

In this Condition, the remaining OPERABLE subsystems provide adequate core cooling during a LOCA. However, overall ECCS reliability is reduced in this Condition because a single failure in one of the remaining OPERABLE subsystems concurrent with a design basis LOCA may result in the ECCS not being able to perform its intended safety function (depending on the size and location of the postulated break). Since the ECCS availability is reduced relative to Conditions A or B, a more restrictive Completion Time is imposed. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is based on the overall redundancy provided by the ECCS and its continued ability to perform the intended safety function while assuring a return towards full ECCS capability in a (continued)

BWR/6 STS B 3.5-9 P&R, 07/22/93

- s ECCS-Operating.

B 3.5.1 BASES reasonable time so as to not significantly impact overall ECCS reliability.

D_d With any three ECCS injection subsystem inoperable, at least one ECCS injection subsystem must be restored to OPERABLE-status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

In this condition, the remaining OPERABLE subsystems provide adequate core ~ cooling during a LOCA. However, overall ECCS reliability is reduced, because a single failure in one of the remaining OPERABLE subsystems concurrent with LOCA may result in the ECCS not being able to perform its intended safety function' i

(depending on the size and. location of the postulated i

break). Since the ECCS availability is reduced relative to_

Condition C, a more restrictive Completion Time is -imposed.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is based on~ the overall (but_~

.j reduced) redundancy provided by the ECCS and its continued ability to perform the intended safety function, while assuring a_ return towards full ECCS capability in a-reasonable time so as to not significantly impact overall ECCS reliability.

E.1 and E.2 i

If any Required Action and associated Completion Time of Condition A, B, C, or D are not met, the plant must~be brought to a MODE in which the LCO does not apply.

To_

achieve this status, the. plant must be brought to at least MODE 3 within 12' hours and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The allowed Completion Times are reasonable,. based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without-i challenging plant systems.

f.d

- 1 When four or more ECCS subsystems are inoperable, as-stated in Condition F, the plant is in a condition outside~

of the accident analyses. Therefore, LCO 3.0.3 must be entered immediately.

(continued) l BWR/6 STS B 3.5-10 P&R, 07/22/93 i

c ECCS-Operating B 3.5.1 i

BASES L1 P

With one required ADS valve inoperable,'an ADS valve must be restored to OPERABLE status within 14 days. The LC0 i

requires seven ADS valves to be OPERABLE to provide the ADS function.

Reference B3.5.1-6 contains.the'.results of the t

traditional design basis analysis that evaluated the effect of one ADS valve (out of eight total) being out of service.

The results of this analysis were bounded by those of other more limiting ECCS single failure scenarios. Thus, additional analyses were performed for identified limiting events assuming the unavailability of multiple ADS valves.

i Per these analyses, operation of only five ADS valves will provide the required depressurization. However, even in this condition with six ADS valves remaining OPERABLE, overall reliability of the ADS is reduced and there is a-marked reduction in depressurization capability.

Therefore, operation is only allowed for a limited time. The 14 day l

Completion Time is based on~ the overall redundancy and capacity of the ADS system and its continued ability to perform its intended safety function, while assuming a return toward full ADS capability in a reasonable time so as to not significantly impact overall ADS or ECCS reliability.

This conditions has been modified _by Note that allows concurrent existence.with conditions A, B, C, or D, Concurrent existence is justified by the Additional Eccs analyses that were performed and greatly simplifies the necessary Required Actions.'

H.1 and H.2 If the Required Action and associated Completion Time of-Condition G is not met or if two or more required ADS valves

~

are inoperable, the plant must be brought to a condition in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and 2

reactor steam dome pressure reduced to 5; 3.5 Kg/cm g(50 psig) within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in-an orderly manner and without challenging plant systems.

(continued)

BWR/6 STS B 3.5-11 P&R, 07/22/93

s ECCS-Operating B 3.5.1 BASES SR 3.5.1.1 The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge lines of the HPCF subsystems, RCIC system, and LPCI subsystems full of water ensures that the system will perform properly, injecting their full capacity into the RCS upon demand. This will also prevent a water hammer following an ECCS initiation signal. One acceptable method of ensuring the line are full is to vent at the high points.

The 31 day Frequency is based on operating experience, on the procedural controls governing system operation, and on the gradual nature of void buildup in the ECCS piping.

SR 3.5.1.2 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves potentially capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

The 31 day Frequency of this SR was derived from the Inservice Testing Program requirements for performing valve testing at least once every 92 days.

The Frequency of 31 days is further justified because the valves are operated under procedural control and because improper valve alignment would only affect a single subsystem. This Frequency has been shown to be acceptable through operating experience.

(continued)

BWR/6 STS B 3.5-12 P&R, 07/22/93

e ECCS--Operating B 3.5.1 BASES i

This SR is modified by a Note that allows a LPFL subsystem to be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the RHR cut in permissive pressure in MODE 3, if capable of being manually realigned (remote or local) to the LPFL mode and not otherwise inoperable. This allows operation in the RHR shutdown cooling mode during MODE 3 if necessary.

SR 3.5.1.3 Verification every 31 days that ADS air receiver pressure is 2

2 11.3 Kg/cm g (161 psig) assures adequate air pressure for reliable ADS operation. The accumulator on each ADS valve provides pneumatic pressure for valve actuation. The

~

designed pneumatic supply pressure requirements for the accumulator are such that, following a failure of the pneumatic supply to the accumulator, at least one valve actuation can occur with the drywell at design pressure, or five valve actuations 'can occur with the drywell at atmospheric pressure (Ref B3.5.1-10). The ECCS safety analysis assumes only one actuation to achieve the depressurization required for operation of the low pressure ECCS. This minimum required pressure of 11.3 Kg/cm 9 (161 psig) is provided by the Atmospheric Control System (ACS).

The 31 day Frequency takes into consideration administrative controi over operation of the ACS and alarms for low pneumatic pressure.

SR 3.5.1.4. SR 3.5.1.5 and SR 3.5.1.6 i

The performance requirements of the ECCS pumps are determined through application of the 10 CFR 50, Appendix K, criteria (Ref. B3.5.1-5).

These periodic Surveillances are performed (in accordance with the ASME Code,Section XI, requirements for the ECCS pumps) to verify that the ECCS l

pumps will develop the flow rates required by the respective analyses. The ECCS pump flow rates ensure that adequate core cooling is provided to satisfy the acceptance criteria of 10 CFR 50.46 (Ref. B3.5.1-7).

The RCIC pump flow rates also ensure that the system can maintain reactor coolant inventory during pressurized conditions with the RPV isolated.

(continued)

BWR/6 STS B 3.5-13 P&R, 07/22/93 P

~

n ECCS--Operating B 3.5.1 i

BASES The pump flow rates are verified against a system head that is equivalent to the RPV pressure expected during a LOCA.

The total system pump outlet pressure is adequate to overcome the elevation head pressure between the pump suction and the vessel discharge, the piping friction losses, and RPV pressure present during LOCAs. These values may be established during pre-operational testing.

The RCIC pump flow rates ensure that the system can maintain reactor coolant inventory during pressurized conditions with the RPV isolated. The flow tests for the RCIC System are performed at two different pressure ranges such that system capability to provide rated flow is tested both at the higher and lower operating ranges of the system.

Since the required reactor steam dome pressure must be available to parfu.m SR 3.5.1.5 and SR 3.5.1.6 sufficient time is allowed af'.er edequate pressure is achieved to perform these SRs.

Reactor startup is allowed prior to performing the low pressure Surveillance because the reactor pressure is low and the time to satisfactorily perform the Surveillance is short. The reactor pressure is allowed to be increased to normal operating pressure since it is assumed that the low pressure test has been satisfactorily completed and there is no indication or reason to believe thtt RCIC is inoperable.

Therefore, these SRs are modified by !?otes that state the Surveillances are not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the specified reactor steam dome pressure is reached.

A 92 day Frequency for SR 3.5.1.4 and SR 3.5.1.5 is consistent with the Inservice Testing Program requirements.

The 18 month Frequency for SR 3.5.1.6 is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned 4

transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle.

Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.5.1.7 The ECCS subsystems are required to actuate automatically to perform their design functions.

This Surveillance test verifies that, with a required system initiation signal (continued) t i

BWR/6 STS B 3.5-14 P&R, 07/22/93

~

r ECCS-Operating B 3.5.1 BASES (actual or simulated), the automatic initiation logic ol HPCF, RCIC, and LPFL will cause the systems or subsystems to operate as designed, including actuation of the system throughout its emergency operating sequence, automatic pump startup, and actuation of all automatic valves to their required positions. This Surveillance also ensures that the HPCF and RCIC Systems will automatically restart on an RPV low water level (Levels 1.5 and 2, respectively) signal received subsequent to an RPV high water level (Level 8) trip and that the suction is automatically transferred from the CST to the suppression pool. The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlaps this Surveillance to provide complete testing of the assumed safety function.

The 18 month Frequency is based on the need to perform this ~

Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which i

is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

This SR is modified by a Note that excludes vessel injection during the Surveillance.

Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.

i l

SR 3.5.1.8 i

The ADS designated S/RVs are required to actuate automatically upon receipt of specific initiation signals.

A system functional test is performed to demonstrate that the mechanical portions of the ADS function (i.e.,

solenoids) operate as designed when initiated either by an actual or simulated initiation signal, causing proper actuation of all the required components.

SR 3.5.1.9 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function.

'4 (continued)

BWR/6 STS B 3.5-15 P&R, 07/22/93

ECCS-Operating B 3.5.1 BASES The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

This SR is modified by a Note that excludes valve actuation.

This prevents an RPV pressure blowdown.

SR

3. 5. L1 A manual actuation of each ADS valve is performed to verify that the valve and solenoids are functioning properly and that no blockage exists in the S/RV discharge lines. This is demonstrated by the' response of the turbine control or bypass valve, by h change in the measured steam flow, or by any other method suitable te verify steam flow.

Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Sufficient time is i

therefore allowed, after the required pressure is achieved, to perform this test. Adequatepressureatwhichthistest is to be performed is 66.8 Kg/cm g (950 psig) (the pressure recommended by the valve manufacturer).

Reactor startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam dome pressure is ;t 66.8 2

Kg/cm g (950 psig). SR 3.5.1.8 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed i.

safety function.

The Frequency of 18 months on a STAGGERED TEST BASIS ensures l

that both solenoids for each ADS valve are alternately l

tested. The Frequency is based on the need to perform this l

Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

I l

1 (continued)

BWR/6 STS B 3.5-16 P&R, 07/22/93 l

v-e-

ECCS--Operating B 3.5.1 BASES Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

B 3.5.1.1, ABWR SSAR. Section 6.3.2.

B 3.5.1-2, ABWR SSAR. Section 15.6.4.

B 3.5.1-3, ABWP SSAR. Section 15.6.5.

B 3.5.1-4, ABWR SSAR. Section 15.6.6.

B 3.5.1-5, 10 CFR 50. Accendix K.

B 3.5.1-6, ABWR SSAR. Section 6.3.3.

B 3.5.1-7, 10 CFR 50.46.

B 3.5.1-8,

???

B 3.5.1-9, PRA Studies B 3.5.1-10, ABWR SSAR. Seciton 7.3.1.1.1.2.

mm L

BWR/6 STS B 3.5-17 P&R, 07/22/93

\\>

ECCS-Shutdown B 3.5.2 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.2 ECCS-Shutdown BASES BACKGROUND A description of the High Pressure Core Flooder (HPCF) and the Low Pressure Flooder (LPFL) subsystems of the Residual Heat Removal (RHR) System is provided in the Bases for LCO 3.5.I, "ECCS--Operating." The Reactor Core Isolation Cooling (RCIC) system steam driven turbines can not operate with the reactor shutdown and so are not available.

~

APPLICABLE ECCS performance is evaluated for the entire spectrum of SAFETY ANALYSES break siles for a postulated loss of coolant accident (LOCA).

The long term cooling analysis following a design basis LOCA (Ref.-B3.5.2-1) demonstrates that only one motor driven ECCS injection subsystem is required, post LOCA, to maintain the peak cladding temperature below the allowable limit. To provide redundancy, a minimum of two ECCS 1

subsystems are required to be OPERABLE in MODES 4 and 5.

Two OPERABLE ECCS injection subsystems also ensure adequate inventory makeup in the reactor pressure vessel (RPV) ir,the event of an inadvertent vessel draindown.

The ECCS satisfy Criterion 3 of the NRC Policy Statement.

i LCO Two ECCS injection subsystems are required to be OPERABLE.

)

The ECCS injection subsystems are defined as the three LPFL and the two HPCF subsystems.

Each LPFL subsystem consists of one motor driven pump, piping, and valves to transfer water from the suppression pool to the RPV.

Each HPCF subsystem consists of one motor driven pump, piping, and valves to transfer water from the suppression pool or condensate storage tank (CST) to the RPV.

3 Any LPFL subsystem may be aligned for the shutdown cooling mode of the decay heat removal system in MODE 4 or 5 and considered OPERABLE for the ECCS function, if it can be

)

manually realigned (remote or local) to the LPFL mode and is not otherwise inoperable. Because of low pressure and low (continued)

BWR/6 STS B 3.5-1 P&R, 07/22/93

,V y

ECCS-Shutdown B 3.5.2 BASES OPERABLE status is based on engineering judgment that considered the availability of or.e subsystem and the low probability of a vessel draindown event.

With the inoperable subsystem not restored to OPERABLE status within the required Completion Time, action must be initiated immediately to suspend operations with a potential for draining the reactor vessel (0PDRVs) to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

C.1. C.2. D.l. D.2. and D.3 If both of the required ECCS injection subsystems are inoperable, all coolant inventory makeup capability may be unavailable. Therefore, actions must be initiated immediately to suspend OPDRVs in order to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until OPDRVs are suspended. One ECCS injection subsystem must also be restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

If at least one ECCS injection subsystem is not restored to OPERABLE status within the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time, additional actions are required to minimize any potential fission product release to the environment.

This includes initiating immediate action to restore the following to OPERABLE status: secondary containment, one standby gas treatment subsystem, and one isolation valve and associated instrumentation in each secondary containment penetration i

flow path not isolated. This may be performed by an administrative check, by examining logs or other information, to determine if the components are out of service for maintenance or other reasons. Verification does not require performing the Surveillances needed to demonstrate OPERABILITY of the components.

If, however, any required component is inoperable, then it must be restored to OPERABLE status.

In this case, the Surveillances may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE.

(continued)

BWR/5 STS B 3.5-3 P&R, 07/22/93

r-ECCS--Shutdown B 3.5.2 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.2 ECCS--Shutdown BASES BACKGROUND A description of the High Pressure Core Flooder (HPCF) and the Low Pressure Flooder (LPFL) subsystems of the Residual Heat Removal (RHR) System is provided in the Bases for LCO 3.5.1, "ECCS--Operating." The Reactor Core Isolation Cooling (RCIC) system steam driven turbines can not operate with the reactor shutdown and so are not available.

~

APPLICABLE ECCS performance is evaluated for the entire spectrum of SAFETY ANALYSES break sizes for a postulated loss of coolant accident (LOCA). The long term cooling analysis following a design basis LOCA (Ref.-B3.5.2-1) demonstrates that only one motor driven ECCS injection subsystem is required, post LOCA, to maintain the peak cladding temperature below the allowable limit. To provide redundancy, a minimum of two ECCS subsystems are required to be OPERABLE in MODES 4 and 5.

Two OPERABLE ECCS injection subsystems also ensure adequate inventory makeup in the reactor pressure vessel (RPV) in the event of an inadvertent vessel draindown.

The ECCS satisfy Criterion 3 of the NRC Policy Statement.

LCO Two ECCS injection subsystems are required to be OPERABLE.

The ECCS injection subsystems are defined as the three LPFL and the two HPCF subsystems.

Each LPFL subsystem consists of one motor driven pump, piping, and valves to transfer water from the suppression pool to the RPV.

Each HPCF subsystem consists of one motor driven pump, piping, and v:lves tc transfer water from the suppression pool or condensate storage tank (CST) to the RPV.

Any LPFL subsystem may be aligned for the shutdown cooling mode of the decay heat removal system in MODE 4 or 5 and considered OPERABLE for the ECCS function, if it can be m9nually realigned (remote or local) to the LPFL mode and is not otherwise iroperable.

Because of low pressure and low (continued)

BWR/6 STS B 3.5-1 P&R, 07/22/93

ECCS-Shutdown B 3.5.2 BASES temperature conditions in MODES 4 and 5, sufficient time will be available to manually align and initiate LPFL subsystem operation to provide core cooling prior to postulated fuel uncovery.

APPLICABILITY OPERABILITY of the ECCS injection subsystems is required in MODES 4 and 5 to ensure adequate coolant inventory and sufficient heat removal capability for the irradiated fuel in the core in case of an inadvertent draindown of the vessel. Requirements for ECCS OPERABILITY during MODES I, 2, and 3 are discussed in the Applicability section of the Bases for LCO 3.5.I.

ECCS subsystems are not required to be OPERABLE during MODE 5 with the spent fuel pool gate removed, and the water level maintained at ;t 7 m (23 ft) above the RPV flange.

This provides sufficient coolant inventory to allow operator action to terminate the inventory loss prior to fuel uncovery in case of an inadvertent draindown.

The Automatic Depressurization System is not required to be OPERABLE during MODES 4 and 5 because the RPV pressure is

< 3.5 Kg/cm' g(50 psig), and the LPFL and HPCF subsystems can provide core cooling without any depressurization of the primary system.

Because the Reactor Core Isolation Cooling (RCIC) system requires steam to operate, it is not required to be OPERABLE during MODES 4 and 5.

1 ACTIONS A.1 and B.1 If any one required ECCS injection subsystem is inoperable, i

the required inoperable ECCS injection subsystem must be restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

In this Condition, the remaining OPERABLE subsystem can provide sufficient RPV flooding capability to recover from an inadvertent vessel draindown. However, overall system reliability is reduced because a single failure in the remaining OPERABLE subsystem concurrent with a vessel draindown could result in the ECCS not being able to perform its intended function.

The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time for restoring the required ECCS injection / spray subsystem to (continued)

BWR/6 STS B 3.5-2 P&R, 07/22/93

,-J

- w ECCS-Shutdown B 3.5.2 BASES i

OPERABLE status is based on engineering judgment that considered the availability of one subsystem and the low probability of a vessel draindown event.

With the inoperable subsystem not restored to OPERABLE status within the required Completion Time, action must be initiated immediately to suspend operations with a potential for draining the reactor vessel (OPDRVs) to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

C.I. C.2. D.l. D.2. and 0.3 If both of the required ECCS injection subsystems are

~

inoperable, all coolant inventory makeup capability may be unavailable. Therefore, actions must be initiated immediately to suspend OPDRVs in order to minimize the probability of a vessel draindown and the subsequent potential for fission product release.

Actions must.

continue until OPDRVs are susper.ded. One ECCS injection subsystem must also be restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

If at least one ECCS injection subsystem is not restored to OPERABLE status within the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time, additional actions are required to minimize any potential fission product release to the environment. This includes initiating immediate action to restore the following to OPERABLE status:

secondary containment,_one standby gas treatment subsystem, and one isolation valve and associated instrumentation in each secondary containment penetration flow path not isolated. This may be performed by an administrative check, by examining logs or other information, to determine if the components are out of service for maintenance or other reasons. Verification does not require performing the Surveillances needed to demonstrate OPERABILITY of the components.

If..however, any required component is inoperable, then it must be restored to OPERABLE status.

In this case, the Surveillances may

]

need to be performed to restore the component to OPERABLE status. Actions must continue until all required components i

are OPERABLE.

i i

(continued)

BWR/6 STS B 3.5-3 P&R, 07/22/93 i

V, n

ECCS-Shutdown B 3.5.2 BASES The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time to restore at least one ECCS injection subsystem to OPERABLE status ensures that prompt action will be taken to provide the required cooling f

capacity or to initiate actions to place the plant in a condition that minimizes any potential fission product release to the environment.

i SR 3.5.2.1 and SR 3.5.2.2 The minimum water level of 7 m (23 ft) required for the suppression pool is periodically verified to ensure that the suppression pool will provide adequate net positive suction head (NPSH) for the ECCS pumps, recirculation volume, and vortex prevention. With the suppression pool water level I

less than the required limit, all ECCS injection subsystems are inoperable.

When the suppression pool level is < 7 m (23 ft), the HPCF is considered OPERABLE' only if it can take suction from the CST and the CST water level is sufficient to provide the required NPSH for the HPCF pump. Therefore, a verification that either the suppression pool water level is 2 7 m (23 ft) or the HPCF System is aligned to take suction from the CST and the CST contains 2 liters ([

] gallons) of water, equivalent to [

]m, ensures that the HPCF System can supply makeup water to the RPV.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency of these SRs was developed considering operating experience related to suppression pool and CST water level var *ations and instrument drift during the applicable MODES.

Furthermore, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications in the control room, including alarms, to alert the operator to an abnormal suppression pool or CST water level condition.

SR 3.5.2.3. SR 3.5.2.5. and SR 3.5.2.6 The Bases provided for SR 3.5.1.1, SR 3.5.1.4, and SR 3.5.1.7 are applicable to SR 3.5.2.3, SR 3.5.2.5, and SR 3.5.2.6, respectively.

i (continued) i BWR/6 STS B 3.5-4 P&R, 07/22/93

M, 9 ; ;w ECCS-Shutdown B 3.5.2 i

BASES SR 3.5.2.4 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otharwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. Ynis SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. The 31 day Frequency is appropriate because the valves are operated under procedural control and the probability of their being mispositioned during this time period is low.

In MODES 4 and 5, the RHR System may operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor.

Therefore, RHR valves that are required for LPFL subsystem operation may be aligned for the shutdown cooling mode. Therefore, this SR is modified by a Note that allows one LPFL subsystem of the RHR System to be considered OPERABLE for the ECCS function if all the required valves in the LPFL flow path can be manually realigned (remote or local) to allow injection into the RPV and the system is not otfierwise inoperable. This will ensure adequate core cooling if an inadvertent vessel draindown should occur.

REFERENCES APPLICABLE SAFETY ANALYSES ABWR SSAR, Section 6.3.

SURVEILLANCE REQUIREMENTS REFERENCES k

BWR/6 STS B 3.5-5 P&R, 07/22/93

ML20056D479

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