Text

Forwards Proof & Review ABWR TS & Their Bases for Sections, 3.1,reactivity Control & 3.2,power Distribution
	ML20045G692
Person / Time
Site:	05200001
Issue date:	07/02/1993
From:	Crutchfield D; Office of Nuclear Reactor Regulation
To:	Marriott P; GENERAL ELECTRIC CO.
References
	NUDOCS 9307150054
	Download: ML20045G692 (78)
	v • d • e

,-

q MM MO gm co UNITED STATES

[ J N ' %,j NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 20555 0001 o

( %<

/

July 2,1993 Docket No.52-001 1

Mr. Patrick W. Marriott, Manager Licensing & Consulting Services GE Nuclear Energy 175 Curtner Avenue San Jose, California 95125 Dehr Mr. Marriott:

SUBJECT:

ADVANCED B01' LING WATER REACTOR (ABWR) TECHNICAL SPECIFICATIONS' MILESTONES Enclosed are the proof and review ABWR technical specifications and their bases for the following sections:

3.1 Reactivity Control 3.2 Power Distribution As we discussed at our management meeting on June 10, 1993, the Nuclear Regulatory Commission (NRC) staff will be regularly providing GE Nuclear Energy (GE), until August 31, selected sections of proof and review ABWR technical specifications. 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 l

anticipate that GE will interface very closely with the staff to resolve the I

majority of the 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 September 20, 1993, will be few.

l The electronic text of these sections is available on the NRC Technical Specifications Branch electronic bulletin boat d (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) =

l 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 techni-cal specifications in Wordperfect 5.1 format.

03 KiB FRE CMTER COPY 9307150054 930702 PDR ADOCK 05200001 A

PDR

i s

i 2-July 2, 1993 Mr. Patrick W. Marriott 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.

Sincerely, (Original signed by R. W. Borchardt for)

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:

Docket File PDST R/F DORS R/F OTSB R/F PDR Central Files TEMurley/FJMiraglia WTRussell JGPartlow DMCrutchfield BKGrimes CIGrimes BABoger JERichardson ACThadani FJCongel JSWermiel RBBarrett CEMcCracken RCJones CEBerlinger AEChaffee GHMarcus WBHardin, RES l

LCShao, RES JA0'Brien RWBorchardt JNWilson CPoslusny SNinh SMLMagruder GEGrant, 17G21 JEMoore, 15B18 RHLo PCHearn ACRS (11) (w/o enci)

FMReinhart PShea OFC:

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06 93 0FFICIAL RECORD COPY:

DOCUMENT NAME: TSP &R2.LTR i

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4 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 GE Nuclear Energy 12300 Twinbrook Parkway Suite 315 Rockville, Maryland 20852 Director, Criteria & Standards Division Office of Radiation Programs U. S. Environmental Protection Agency 401 M Street, S.W.

Washington, D.C.

20460 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.

1 Suite 800 '

Washington, D.C.

20004 Jay M. Gutierrez, Esq.

i Newman & Holtzinger, P.C.

1615 L Street, N.W.

Suite 1000 Washington, D.C.

20036 l

i SDM I

3.1.1 i

3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM) d LC0 3.1.1 SDM shall be:

i a.

2 0.38% ok/k, with the highest worth control rod or rod j

pair analytically determined; or b.

2 0.28% ok/k, with the highest worth control rod or rod pair determined by test.

i I

APPLICABILITY:

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

SDM not within limits A.1 Restore SDM to within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months in MODE 1 or 2.

limits.

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 not met.

C.

SDM not within limits C.1 Fully insert all I hour in MODE 3.

insertable control 4

rods.

D.

SDM not within limits D.1 Fully insert all I hour in MODE 4.

insertable control rods.

AND (continued)

ABWR TS 3.1-1 Rev.

O, 11/5/92

i SDM 3.1.1 ACTIONS l

CONDITION REQUIRED ACTION COMPLETION TIME D.

(continued) 0.2 Initiate action to I hour restore [ secondary i

cc.tainment] to CAABLE status.

AND D.3 Initiate action to I hour restore one standby gas treatment (SGT) subsystem to OPERABLE status.

AND D.4 Initiate action to I hour restore one isolation valve and associated instrumentation to OPERABLE status in each secondary containment penetration flow path not isolated.

E.

SDM not within limits E.1 Suspend CORE Immediately in MODE 5.

ALTERATIONS except for control rod insertion and fuel assembly removal.

AND (continued)

ABWR TS 3.1-2 Rev.

O, 11/5/92

SDM 3.1.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E.

(continued)

E.2 Initiate action to Immediately fully insert all insertable control rods in core cells containing one or more fuel assemblies.

AND j

E.3 Initiate action to I hour restore secondary containment to OPERABLE status.

AND E.4 Initiate action to I hour restore one SGT subsystem to OPERABLE status.

AND E.5 Initiate action to I hour restore one isolation valve and associated instrumentation to OPERABLE status in each secondary containment penetration flow path not isolated.

ABWR TS 3.1-3 Rev.

O, 11/5/92

SDM 3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1.1 Verify SDM is:

Priortoeacii in vessel fuel a.

2 0.38% Ak/k with the highest worth movement during control rod or control rod pair fuel loading analytically determined; or sequence b.

2 0.28% Ak/k with the highest worth AND

]

control rod or control rod pair determined by test.

Once within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after criticality following fuel movement or control rod replacement within the reactor pressure vessel i

ABWR TS 3.1-4 Rev.

O, 11/5/92

Reactivity Anomalies 3.1.2 3.1 REACTIVITY CONTROL SYSTEMS 3.1.2 Reactivity Anomalies LCO 3.1.2 The reactivity difference between the monitored core k,,, and 1

the predicted core k,,, shall be within 1% Ak/k.

j APPLICABILITY:

MODES 1 and 2.

ACTIONS i

CONDITION REQUIRED ACTION COMPLETION TIME A.

Core reactivity A.1 Restore core 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months difference not within reactivity difference limit.

to 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.

l l

ABWR TS 3.1-1 Rev.

O, 11/5/92

i i

Reactivity Anomalies 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l

SR 3.1.2.1 Verify core reactivity difference between Once within the monitored core k,n and the predicted 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after core k,n is within i 1% ok/k.

reaching equilibrium conditions l

following startup after fuel movement or control rod replacement l

within the I

reactor pressure vessel AND 1000 MWD /T thereafter l

ABWR TS 3.1-2 Rev.

O, 11/5/92

Control Rod OPERABILITY 3.1.3 3.1 REACTIVITY CONTROL SYSTEMS 3.1.3 Control Rod OPERABILITY LC0 3.1.3 Each control rod shall be OPERABLE.

APPLICABILITY:

MODES I and 2.

ACTIONS

_______________----------------------NOTE-------------------------------------

Separate Condition entry is allowed for each control rod.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One withdrawn control

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

rod stuck.

A stuck rod may be bypassed in the Rod Control and Information System (RC&lS) in accordance with SR 3.3.2.1.8 if required to allow jp continued operation.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months A.1 Disarm the associated control rod drive (CRD).

AND (continued) 1 l

ABWR TS 3.1-1 Rev.

O, 11/4/92 i

l Control Rod OPERABILITY 3.1.3 l

l ACTIONS 1

CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.2

NOTE---------

Not applicable when less than or equal to the low power setpoint (LPSP) of the RC&IS.

Perform SR 3.1.3.2 and SR 3.1.3.3 for each withdrawn 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE control rod.

AND A.3 Perform SR 3.1.1.1.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B.

Two or more withdrawn B.1 Disarm the associated I hour control rods stuck.

CRD.

AND B.2 Be in MODE 3.

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

I ABWR TS 3.1-2 Rev.

O, 11/4/92

Control Rod OPERABILITY 3.1.3 ACTIONS (continued) l CONDITION REQUIRED ACTION COMPLETION TIME C.

One or more control

NOTE---------

rods inoperable for Inoperable control l

I reasons other than rods may be bypassed Condition A or B.

in RC&IS in accordance with SR 3.3.2.1.8, if required, to allow insertion of inoperable control rod and continued I

operation.

l l

Inoperable Control l

Rods with failed l

Motor Drives can be l

fully inserted by l

individual scram.

C.1 Fully insert 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months inoperable control rod AND C.2 Disarm the associated 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months CRD.

D.

NOTE---------

D.1 Restore compliance 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Not applicable when with GWSR.

THERMAL POWER

> 10% RTP.

OB D.2 Restore control rod 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months One or more inoperable to OPERABLE status.

control rods not in compliance with Ganged Withdrawal Sequence (GWSR) and not separated by two or more OPERABLE control rods.

(continued) i ABWR TS 3.1-3 Rev.

O, 11/4/92

Control Rod OPERABILITY

~

3.1.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME l

E.

Required Action and F.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, C, D, or E not met.

OB Nine or more control rods inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.3.1 Determine the position of each control rod.

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

SR 3.1.3.2

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

Not required to be performed until 7 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RC&lS.

Insert each fully withdrawn control rod 7 days two notches.

l ABWR TS 3.1-4 Rev.

O, 11/4/92 l

\\

l l

w.-

-n p.

-e-u-

w+

Control Rod OPERABILITY 3.1.3 SURVEILLANCE RE0VIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.3.3

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

Not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RC&IS.

Insert each partially withdrawn control rod 31 days two notches.

SR 3.1.3.4 Verify each control rod scram time from In accordance fully withdrawn to 60% rod insertion is with 5 [ ] seconds.

SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 (continued)

SR 3.1.3.5 Verify each control rod does not go to the Once the first withdrawn overtravel position.

time the control rod is withdrawn to

" full out" position after the fuel movement within the RPV AND Prior to declaring l

control rod OPERABLE after work on control rod or CRD System that could affect coupling ABWR TS 3.1-5 Rev.

O, 11/4/92 1

1

Control Rod Scram Times 3.1.4 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Control Rod Scram Times LCO 3.1.4 a.

No more than [8] OPERABLE control rods shall be " slow,"

in accordance with Table 3.1.4-1; and b.

No more than 2 OPERABLE control rods that are " slow" shall occupy adjacent locations.

l APPLICABILITY:

MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Requirements of the A.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months LC0 not met.

SURVEILLANCE REQUIREMENTS

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

During single control rod pair scram time Surveillances, the control rod drive (CRD) pumps shall be isolated from the associated scram accumulator.

SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify each control rod scram time is Prior to within the limits of Table 3.1.4-1 with exceeding 2

reactor steam dome pressure 2 66.8 Kg/cm g 40% RTP after (950 psig).

fuel movement within the l

reactor l

pressure vessel AND (continued)

ABWR TS 3.1-1 Rev.

O, 11/6/92

l Control Rod Scram Times 3.1.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY-SR 3.1.4.1 (continued)

Prior to exceeding 40% RTP after each reactor shutdown j

2 120 days-SR 3..'. 4. 2 Verify, for a representative sample, each 120 days tested control rod scram time is within the cumulative limits of Table 3.1.4-1 with reactor steam operation in a

dome pressure 2 66.8Kg/cm g (950 psig).

MODE 1 SR 3.1.4.3 Verify each affected control rod scram time Prior to is within the limits of Table 3.1.4-1 with

' declaring any reactor steam dome pressure.

control rod OPERABLE after work on control rod or CRD System that could affect l

scram time SR 3.1.4.4 Verify each affected control rod scram time Prior to is within the limits of Table 3.1.4-1 with exceeding a

reactor steam dome pressure 2 66.8Kg/cm g 40% RTP after (950 psig).

Work on control rod or CRD System that could affect scram time 2

ABWR'TS 3.1-2 Rev.

0, 11/6/92 m-

~n y

-.----e zer-----

pyv..

.u-

--e

,i,--.

%-me-p-

9ap,.

<ey w

gr e-

Control Rod Scram Times 3.1.4 Table 3.1.4-1 Control Rod Scram Times

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

1.

OPERABLE control rods with scram times not within the limits of this Table are considered " slow."

2.

Control rods with scram times > [

] seconds to notch position [13] are inoperable, in accordance with SR 3.1.3.4, and are not considered " slow."

SCRAM TIMES (a)

(seconds)

REACTOR REACTOR REACTOR STEAM DOME STEAM DOME STEAM DOME PRESSURE (b)

PRESSURE (b)

NOTCH POSITION 0 psig

[950] psig

[1050] psig

[43]

(c)

[0.30]

[0.31]

[29]

(c)

[0.78]'

[0.84]

[13]

[

]

[1.40]

[1.53]

(a) Maximum scram time from fully withdrawn position, based on de-energization of scram pilot valve solenoids as time zero.

(b)

For immediate reactor steam dome pressures, the scram time criteria are determined by linear interpolation.

a (c)

For reactor steam dome pressure s 66.8 Kg/cm g (950 psig), only notch position (13] scram time limit applies.

2 ABWR TS 3.1-3 Rev.

O, 11/6/92 I

l

3.1 REACTIVITY CONTROL SYSTEMS 3.1.5 Control Rod Scram Accumulators I

LC0 3.1.5 Each control rod scram accumulator shall be OPERABLE.

l APPLICABILITY:

MODES 1 and 2.

l ACTIONS

...-----------.........---------.----NOTE-------------------------------------

l Separate Condition entry is allowed for each control rod scram accumulator.

i CONDITION REQUIRED ACTION COMPLETION TIME l

A.

One control rod scram A.1 Declare the B hours accumulator associated control inoperable.

rods inoperable.

(continued)

B.

Two or more control B.1 Declare the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months rod scram accumulators associated control inoperable, rod (s) inoperable.

C.

Required Action and C.1

NOTE---------

associated Completion Not applicable if all Time not met.

inoperable control rod. scram accumulators are associated with fully inserted control rods.

Place the reactor Immediately mode switch in the shutdown position.

1 Control Rod Scram Accumulators 3.1.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each control rod scram accumulator 7 days pressure is ;t SI units (1850 psig).

l ABWR TS 3.1-2 REV. 0 10/15/92 l

Rod Pattern Control 3.1.6 i

i 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 Rod Pattern Control i

LC0 3.1.6 OPERABLE control rods shall comply with the requirements of the [Gange Withdrawal Sequence Restrictions (GWSR)].

APPLICABILITY:

MODES 1 and 2 with THERMAL POWER 5 10% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more OPERABLE A.1

NOTE---------

control rods not in Affected control rods compliance with may be bypassed in

[GWSR).

Rod Control and Information System (RC&l3) in accordance with SR 3.3.2.1.8.

Move associated 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months control rod (s) to correct position.

08 A.2 Declare associated 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months control rod (s) inoperable.

(continued) l ABWR TS 3.1-1 Rev.

O, 11/6/92 l

Rod Pattern Control 3.1.6 l

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.

Nine or more OPERABLE B.1

NOTE---------

control rods not in Affected control rods compliance with may be bypassed in

[GWSR).

RC&lS in accordance with SR 3.3.2.1.8 for insertion only.

Suspend withdrawal of Immediately control rods.

AND B.2 Place the reactor 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months mode switch in the shutdown position.

i SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify all OPERABLE control rods comply 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with [GWSR].

l l

l 1

ABWR TS 3.1-2 Rev.

O, 11/6/92 I

. ~.-..-.. -

~-

SLC System-3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LC0 3.1.7 Two SLC subsystems shall be OPERABLE.

-i APPLICABILITY:

MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION

. COMPLETION-TIME-A.

Concentration of A.1 Restore concentration 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months boron in solution of boron in solution not within limits to within limits.

AND but > [ ).

10 days from-discovery of

~

failure to meet the LC0 B.

One SLC subsystem B.1 Restore SLC subsystem 7 days inoperable (for to OPERABLE status..

reasons other than AN.Q Condition A].

10 days from discovery of failure to meet the LCO C.

Two SLC subsystems C.1 Restore one SLC

8. hours l

inoperable (for subsystem to OPERABLE l

reasons other than status.

l Condition A].

l 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 not met.

i ABWR TS 3.1-1 Rev.

O, 11/4/92 L

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify available volume of sodium 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 2

pentaborate solution is [2 23.1 m (6103 gallons)].

SR 3.1.7.2 Verify temperature of sodium pentaborate 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months solution is within the limits of Figure 3.1.7-1.

SR 3.1.7.3 Verify temperature of pump suction piping 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is within the limits of [ Figure 3.1.7-1).

SR 3.1.7.4 Verify the concentration of boron in 31 days solution is within the limits of Figure 3.1.7-1.

AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after water or boron is added to solution AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after solution temperature is restored within the limits of Figure 3.1.7-1 (continued) 2 ABWR TS 3.1-2 Rev.

O, 11/4/92

9 1

SLC System 3.1.7

, SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.7.5 Verify each SLC subsystem manual, power 31 days operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position, or can be aligned to the correct position.

SR 3.1.7.6 Verify each pump develops a flow rate In accordance 2 [41.2] gpm at a discharge pressure with the l

2 [1300] psig.

Inservice Testing j

j Program or 92 days SR.3.1.7.7 Verify flow through one SLC subsystem from _

[18] months on l

pump into reactor pressure vessel, a STAGGERED l

TEST BASIS l

SR 3.1.7.8 Verify that simultaneous operation of both (18] months pumps develop a flow rate 6.301/s (100 gpm) at a pressure of SI units (1223 psig).

l SR 3.1.7.9 Verify all heat traced piping between

[18] months storage tank and pump suction is unblocked.

AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after solution temperature is restored within the limits of

[ Figure 3.1.7-1]

ABWR TS 3.1-3 Rev.

O, 11/4/92

(

4

..e,.

+-

4

--..,--rv._

.-,..,--.n

.-.--.-v,--

=

SLC System 3.1.7 SVRVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY (continued)

SR 3.1.7.10 Verify sodium pentaborate enrichment is Prior to 2 [60.0] atom percent B-10.

addition to SLC tank l

ABWR TS 3.1-4 Rev.

O, 11/4/92

SLC System 3.1.7 yj I/ // / / // //

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Figure 3.1.71 (page 1 of 1) j Sodiurn Pentaborate Solution Temperature / Concentration Recuirements ABWR TS 3.1-5 Rev.

O, 11/4/92

APLHGR 3.2.1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)

LC0 3.2.1 All APLHGRs shall be less than or equal to the limits specified in the COLR.

4 1

APPLICABILITY:

THERMAL POWER 2 25% RTP.

1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Any APLHGR not within A.1 Restore APLHGR(s) to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months i

limits.

within limits.

i B.

Required Action and B.1 Reduce THERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion to < 25% RTP.

Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify all APLHGRs are less than or equal Once within 4

to the limits specified in the COLR.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after 2 25% RTP AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter 4

ABWR TS 3.2-1 Rev.

O, 11/9/92

MCPR 3.2.2 3.2 POWER DISTRIBUTION LIMITS 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

LCO 3.2.2 All MCPRs shall be greater than or equal to the MCPR operating limits specified in the COLR.

APPLICABILITY:

THERMAL POWER 2 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Any MCPR not within A.1 Restore MCPR(s) to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months limits.

within limits.

B.

Required Action and B.1 Reduce THERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion to < 25% RTP.

Time not met.

l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify all MCPRs are greater than or equal Once within to the limits specified in the COLR.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after 2 25% RTP AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter l

ABWR TS 3.2-1 Rev.

O, 11/9/92

\\

LHGR (Optional)-

3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) (Optional)

LCO 3.2.3 All LHGRs shall be less than or equal to the limits specified in the COLR.

APPLICABILITY:

THERMAL POWER 2 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Any LHGR not within A.I Restore LHGR(s) to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months limits.

within limits.

B.

Required Action and B.I Reduce THERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion to < 25% RTP.

Time not met.

l SURVEILLANCE REQUIREMENTS l

[

SURVEILLANCE FREQUENCY l

SR 3.2.3.1 Verify all LHGRs are less than or equal to once within the limits specified in the COLR.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after 2 25% RTP AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter e-l l

ABWR TS 3.2-1 Rev.

O, 11/9/92

t l

Reactivity Control Systems Shutdown h1argin (SDht)

BASES SDh! requirements are specified to ensure:

(

The reactor can he made subcritical from all operating conditions and transients and Dmign Basis Events; The reactivity transients associated with postulated accident conditions are controllable within acceptable limits; and The reactor will he maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition.

These requirements are satisfied by the control rods, as described in GDC 26 (Reference B3.1.1-1), which can compensate for the reactivity effects of the fuel and water temperature changes experienced during all operating conditions.

APPLICABLE The control rod removal error during refueling accident analysis SAFETY ANALYSIS (Reference B3.1.1-2) assumes the core is subcritical with the highest worth control rod withdrawn. The analysis of this reactivity insertion event assumes the refueling interlocks are OPERABLE when the reactor is in the refueling mode of operation. These interlocks prevent the withdrawal of more than one control rod, or control rod pair, i

from the core during refueling. (Special consideration and j

requirements for multiple control rod withdrawal during refuelmg are l

casered in Special Operations LCO 3.10.6, "hf ultiple Control Rod Withdrawal-Refueling.") The analysis assumes this condition is acceptable since the core will be shut down with the highest worth control rod or rod pair withdrawn, if adequate SDh1 has been demonstrated.

APPLICABLE Prevention or mitigation of reactivity insertion events is necessary to SAFETY ANALYSIS limit energy deposition in the fuel to prevent (continued) significant fuel damage, which could result in undue release of 1

radioactivity (see Bases for LCO 3.1.7, " Standby Liquid Control (SLC) System"). Adequate SDh1 ensures inadvertent criticalities will j

not cause significant fuel damage.

SD51 satisfies Criterion 2 of the NRC Policy Statement.

LCO The specified SDh1 limit accounts for the uncertainty in the demonstration of SDh1 by testing. Separate SDh1 limits are provided i

for testing where the highest worth control rod or rod pair is determined analytically or by measurement. This is due to the reduced I

uncertainty in the SDh1 test when the highest worth control rod or rod

1 pair is determined by measurement. When SD51 is demonstrated by calculations not associated with a test, additional margin must be added to the specified SDA1 limit to account for uncertainties in the calculation. To ensure adequate SD51 during the design process, a design margin is included to account for uncertainties in the design calculations (Reference B3.1.1-3).

APPLICABILITY In h10 DES I and 2, SD51 must be provided because subcriticality with the highest worth control rod or rod pair withdrawn is assumed in the analysis (Reference B3.1.1-4). In $10 DES 3 and 4, SDh1 is required to ensure the reactor will be held subcritical with margin for a single withdrawn control rod or rod pair. SD51 is required in i

h10DE 5 to prevent an inadvertent criticality during the withdrawal of a single control rod from a core cell containing one or more fuel assemblies or of a control rod pair from loaded core cells during scram time testing.

ACTIONS i

A.1 With SDM not within the limits of the LCO in MODE 1 or 2, SDM must be restored within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Failure to meet the specified SDM may be caused by a control rod that cannot be inserted. The 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Completion

~

time is acceptable, considering that the reactor can still be shut down, assuming no additional failures of control rods to insert, and the low i

probability of an event occurring during this interval.

l B.1 If the SDM cannot be restored, 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 , to prevent the potential for further reductions in available SDM (e.g., additional stuck control rods). 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 operati'ig experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

C.1 With SDM not within limits in MODE 3, the operator must fully insert all insertable control rods within I hour. This action results in the least

reactive condition for the core. The allowed Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is acceptable, considering the reactor can still be shut down, assuming no failures of additional control rods to insert.

D.1, D.2, D.3, and D.4 l

With SDM not within limits in MODE 4, the operator must insert all insertable control rods in I hour. This action results in the least reactive condition for the core. The I hour Completion Time provides sufficient time to take corrective action and is acceptable, considering the reactor can still be shut down assuming no failures of additional control rods to insert. Actions must also be initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to provide means for control of potential radioactive releases. This includes ensuring secondary containment (LCO 3.6.4.1, " Secondary Containment") is OPERABLE; at least one Standby Gas Treatment (SGT) (LCO 3.6.4.3, " Standby Gas Treatment (SGT) System") subsystem is OPERABLE; and at least one j

secondary containment isolation valve (LCO 3.6.4.2, " Secondary Containment Isolation Valves (Slavs)") and associated instrumentation (LCO 3.3.6.2, " Secondary Containment Isolation Instrumentation") are OPERABLE in each associated penetration flow path not isolated. This may be performed as an administrative check, by examining logs or other information, to determine if the components are out of service for maintenance or other reasons. It is not necessary to perform the SRs needed to demonstrate the OPERABILITY of the components. If, L

however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, SRs may need to be performed to l

restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE.

E.1, E.2, E.3, E.4, and E.5 With SDM not within limits in MODE 5, the operator must immediately suspend CORE ALTERATIONS that could reduce SDM. The suspensions are on insertion of fuel in the core or the withdrawal of control rods.

Suspension of these activities shall not preclude completion of movement of a component to a safe condition. Inserting control rods or removing fuel from the core will reduce the total reactivity and are therefore excluded from the suspended actions.

Action must also be immediately initiated to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Action must continue until all insertable control rods in core cells containing one i

m

I l*

r or more fuel assemblies have been fully insened. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and therefore do not have to be inserted.

Action must also be initiated within I hour to provide means for control of potential radioactive releases. This includes ensuring secondary containment (LCO 3.6,4.1) is OPERABLE; at least one SGT subsystem (LCO 3.6.4.3) is OPERABLE; and at least one secondary containment isolation valve (LCO 3.6.4.2) and associated instrumentation (LCO 3.3.6.2) are OPERABLE in each associated penetration flow path not isolated. This may be performed as an administrative check, by examining logs or other information, to determine if the components are out of service for maintenance or other reasons. It is not necessary to perform the SRs needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, SRs may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE.

SR 3.1.1.1 l

Adequate SDM must be demonstrated to ensure the reactor can be made subcritical from any initial operating condition. Adequate SDM is demonstrated by testing before or during the first startup after fuel movement, control rod replacement, or shuffling within the reactor pressure vessel. Control rod replacement refers to the decoupling and removal of a control rod from a core location, and subsequent replacement l

with a new control rod or a control rod from another core location. Since I

core reactivity will vary during the cycle as a function of fuel depletion and poison burnup, the beginning of cycle (BOC) test must also account for changes in core reactivity during the cycle. Therefore, to obtain the i

SDM, the initial measured value must be increased by an adder, "R",

which is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated BOC core reactivity, if the value of R is negative (i.e., BOC is the most reactive point in the cycle), no correction to the BOC measured value is required l

(Reference B3.1.1-4).

l l

Tne SDM may be demonstrated during an in sequence control rod pair withdrawal, in which the highest worth control rod pair is analytically determined, or during local criticals, where the highest worth control rod

]

pair is determined by testmg. Local critical tests require the withdrawal of out of sequence control rods. This testing would therefore require bypassing of the Rod Worth Minimizer to allow the out of sequence withdrawal, and therefore additional requirements must be met (see LCO 3.10.7, " Control Rod Testing-Operating").

The Frequency of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after reaching criticality is allowed to provide a reasonable amount of time to perform the required calculations and appropriate verification.

During MODE 5, adequate SDM is also required to ensure the reactor does not reach criticality during control rod withdrawals. An evaluation of each in vessel fuel movement during fuel loading (including shuffling fuel l

within the core) shall be performed to ensure adequate SDM is maintained i

during refueling. This evaluation ensures the intermediate loading patterns are bounded by the safety analyses for the final core loading pattern. For example, bounding analyses that demonstrate adequate SDM for the most reactive configurations during the refueling may be performed to demonstrate acceptability of the entire fuel movement sequence. For the SDM demonstrations that rely solely on calculation, additional margin (0.10% Dk/k) must be added to the SDM limit of 0.28% Dk/k to account for uncertainties in the calculation. Spiral offload or reload sequences inherently satisfy the SR, provided the fuel assemblies are reloaded in the same configuration analyzed for the new cycle. Removing fuel ftom the l

core will always result in an increase in SDM.

REFERENCES l

10 CFR 50, Appendix A, GDC 26.

ABWR SSAR, Section 15.4.1.

ABWR SSAR, Section 4.3.2.4.1.

NDE 24011 P-A-9, "GE Standard Application for Reactor Fuel,"

Section 3.2.4.1 Sept.1988.

b Reactivity Control Systerns Reactivity Anomalies BASES BACKGROUND in accordance with GDC 26, GDC 28, and GDC 29 (Reference B3.1.2-1), reactivity shall be controllable such aat suberiticality is maintained under cold conditions and acceptable fuel design limits are not exceeded during normal operation and anticipated operational occurrences. Reactivity anomaly is used as a measure of the predicted versus measured core reactivity during power operation. The continual confirmation of core reactivity is necessary to ensure that the Design Basis Accident (DBA) and transient safety analyses remain valid. A large reactivity anomaly could be the result of unanticipated changes in fuel reactivity, control rod worth, or operation at conditions not consistent with those assumed in the predictions of core reactivity, and could potentially result in a loss of SDhl or violation of acceptable fuel design limits.

Comparing predicted versus measured core reactivity validates the nuclear methods used in the safety analysis and supports the SDh!

demonstrations (LCO 3.1.1, "SIIUTDOWN h!ARGIN (SDht)") in ensuring the reactor can be brought safely to co!d, suberitical conditions.

f When the reactor core is critical or in normal power operation, a reactivity balance exists and the net reactivity is zero. A comparison of predicted and measured reactivity is convenient under such a balance, since parameters are being maintained relatively stable under steady l

state power conditions. The positive reactivity inherent in the core design is balanced by the negative reactivity of the control components, thermal feedback, neutron leakage, and materials in the core that absorb neutrons, such as burnable absorbers, producing zero net reactivity.

In order to achieve the required fuel cycle energ3 output, the uranium l

enrichment in the new fuel loading and the fuel loaded in the previous cycles provide excess positive reactivity beyond that required to sustain steady state operation at the beginning of cycle (BOC). When the reactor is critical at RTP and operating moderator temperature, the excess positive reactivity is compensated by burnable absorbers (if any), control rods, and whatever neutron poisons (mainly xenon and samarium) are present in the fuel.

I l

f l

l

l.

The predicted core reactivity, as represented by k effective (kJ, is calculated by a 3D core simulator code as a function of cycle exposure.

This calculation is performed for projected operating states and conditions throughout the cycle. The monitored klis calculated by the i

core monitoring system for actual plant conditions and is then compared to the predicted value for the cycle exposure.

Accurate prediction of core reactivity is either an explicit or implicit assumption in the accident analysis evaluations. Every accident evaluation (Reference B3.1.2-2) is, therefore, dependent upon accurate evaluation of core reactivity. In particular, SD51 and reactivity transients, such as control rod withdrawal accidents, are very sensitive to accurate prediction of core reactivity. These accident analysis evaluations rely on computer codes that have been qualified against available test data, operating plant data, and analytical benchmarks.

Monitoring reactisity anomaly provides additional assurance that the nuclear methods provide an accurate representation of the core t

reactivity.

The comparison between measured and predicted initial core reactivity l

provides a normalization for the calculational models used to predict core reactivity. If the measured and predicted k/ for identical core conditions at BOC do not reasonably agree, then the assumptions used in the reload cycle design analysis or the calculation models used to predict k/ may not be accurate. If reasonable agreement between measured and predicted core reactivity exists at BOC, then the prediction may be normalized to the measured value. Thereafter, any significant deviations in the measured k/ from the predicted k. that develop during fuel depletion may be an indication that the assumptions of the DBA and transient analyses are no longer valid, or that an unexpected change in core conditions has occurred.

Reactivity anomalies satisfy Criterion 2 of the NRC Policy Statement.

The reactivity anomaly limit is established to ensure plant operation is maintained within the assumptions of the safety analyses. Large differences between monitored and predicted core reactivity may indicate that the assumptions of the DBA and transient analyses are no longer valid, or that the uncertainties in the Nuclear Design Methodology are larger than expected. A limit on the difference between the monitored core k/ and the predicted core k/ of 1% Dk/k has been established based on engineering judgment. A > 1%

r deviation in reactivity from that predicted is larger than expected for normal operation and should therefore he evaluated.

APPLICABILITY In 510DE 1, most of the control rods are withdrawn and steady state operation is typically achieved. Under these conditions, the comparison between predicted and monitored core reactivity provides an effective measure of the reactivity anomaly. In h! ODE 2, control rods are typically being withdrawn during a startup. In h! ODES 3 and 4, all control rods are fully inserted, and, therefore, the reactor is in the least reactive state, where monitoring core reactivity is not necessary. In 510DE 5, fuel loading results in a continually changing core reactivity. SDh! requirements (LCO 3.1.1) ensure that fuel movements are performed within the bounds of the safety analysis, and an SDh1 demonstration is required during the first startup following operations that could have altered core reactivity (e.g., fuel movement, control rod replacement, control rod shuffling). The SDh!

test, required by LCO 3.1.1, provides a direct comparison of the predicted and monitored core reactivity at cold conditions; therefore, reactivity anomaly is not required during these conditions.

A.1 l

l Should an anomaly develop between measured and predicted core reactivity, the core reactivity difference must be restored to within the limit to ensure continued operation is within the core design assumptions.

Restoration to within the limit could be performed by an evaluation of the i

core design and safety analysis to determine the reason for the anomaly.

This evaluation normally reviews the core conditions to determine their consistency with input to design calculations. Measured core and process I

parameters are also normally evaluated to determine that they are within the bounds of the safety analysis, and safety analysis calculational models may be reviewed to verify that they are adequate for representation of the core conditions. The required Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is based on the low probability of a DBA during this period, and allows suf0cient time to assess the physical condition of the reactor and complete the l

evaluation of the core design and safety analysis.

i I

l i

I..

B.1 If the core reactivity cannot be restored to within the 1% Dk/k limit, 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 l

an orderly manner and without challenging plant systems.

l l

SR 3.1.2.1 Verifying the reactivity difference between the morJtored and predicted core k,'is within the limits of the LCO provides further assurance that plant operation is maintained within the assumptions of the DBA and transient analyses. The Core Monitoring System calculates the core k.' for the reactor conditions obtained from plant instrumentation. A comparison of the monitored core k,' to the predicted core k,' at the same cycle exposure is used to calculate the reactivity difference. The comparison is required when the core reactivity has potentially changed by a signincant amount. This may occur following a refueling in which new fuel l

assemblies are loaded, fuel assemblies are shufded within the core, or control rods are replaced or shuf0ed. Control rod replacement refers to the decoupling and removal of a control rod from a core location, and subsequent replacement with a new control rod or a control rod from p-y

another core location. Also, core reactivity changes during the cycle. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval after reaching equilibrium conditions following a startup is based on the need for equilibrium xenon concentrations in the core, l

such that an accurate comparison between the monitored and predicted core k,ff values can be made. For the purposes of this SR, the reactor is assumed to be at equilibrium conditions when steady state operations (no control rod movement or core flow changes) at h 75% RTP have been obtained. The 1000 MWD /T Frequency was developed, considering the relatively slow change in core reaa.vity with exposure and operating experience related to variations in core reactivity.

10 CFR 50, Appendix A, GDC 26, GDC 28, and GDC 29.

ABWR SSAR, Chapter 15, i

l t

I l

l l

l I

l l.

Reactivity Control Systems l

Control Rod OPERABILITY BASES i

i BACKGROUND Control rods are components of the Control Rod Drive (CRD) System, which is the primary reactivity control system for the reactor. In conjunction with the Reactor Protection System, the CRD System I

provides the means for the reliable control of reactivity changes to ensure that under conditions of normal operation, including anticipated operational occurrences, specified acceptable fuel design limits are not exceeded. In addition, the control rods provide the i

capability to hold the reactor core subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by j

u malfunction in the CRD System. The CRD System is designed to satisfy the requirements of GDC 26, GDC 27, GDC 28, and GDC 29, 1

(Reference B3.1.3-1).

l The CRD System consists of 205 fine motion control rod drive 1

(FMCRD) mechanisms and 103 hydraulic control unit (IICU)

~

assemblies. The FMRCD is an electro-hydraulic actuated mechanism that provides normal positioning of the control rods using an electric motor, and scram insertion of the control rods using hydraulic power.

i The hydraulic power for scrum is provided by high pressure water J

stored in the individual IICU accumulators, each of which supplies sufficient volume to scram two F14RCDs. Normal control rod positioning is performed using a Lall nut and rotating ball-screw arrangement driven by an electric stepping motor. A hollow piston, which is coupled at the upper end to the control rod, rmts on the ball-nut. The hall-nut inserts the hollow piston and comiected control rod into the core or withdraws them depending on the direction of

)

rotation of the stepping motor. An electromechanical brake mechanism engages the motor drive shaft when the motor is deenergized to prevent inadvertent withdrawal of the control rod, but does not restrict scram insertion.

This Specification, along with LCO 3.1.4, " Control Rod Scram Times," and LCO 3.1.5, " Control Rod Scram Accumulators," ensure that the performance of the control rods in the event of a Design Basis Accident (DBA) or transient meets the assumptions used in the safety 1

analyses of Reference B3.1.3-2, Reference B3.1.3-2, Reference B3.1.3-3, and Reference B3.1.3-5.

W 4

l APPLICABLE The analytical methods and assumptions used in the evaluations SAFETY ANALYSES involving control rods are presented in Reference B3.1.3-2, Reference B3.1.3-3, Reference B3.1.3-4, and Reference B3.1.3-5. The control rods provide the primary means for rapid reactivity control (reactor scram), for maintaining the reactor suberitical, and for limiting the potential effects of reactivity insertion events caused by malfunctions in the CRD System.

l The capability of inserting the control rods ensures that the assumptions for scram reactivity in the DBA and transient analyses are not violated. Since the SDh1 ensures the reactor will be subcritical with the highest worth control rod pair withdrawn (assumed single failure of an IICU), the additional failure of a second control rod to insert could invalidate the demonstrated SDh1 and potentially limit the ability of the CRD System to hold the reactor suberitical. Therefore, the requirement that all control rods he OPERABLE ensures the CRD System can perform its intended function.

The control rods also protect the fuel from damage that could result in release of radioactivity. The limits protected are the htCPR Safety Limit (SL) (see Bases for LCO 3.2.2, "h11Nih1Uh1 CRITICAL POWER RATIO ($1CPR)"), the 1% cladding plastic strain fuel design limit (see Bases for LCO 3.2.1, " AVERAGE PLANAR LINEAR IIEAT GENERATION RATE (APLGIIR)," and LCO 3.2.3,

" LINEAR llEAT GENERATION RATE (LIIGR)"), and the fuel damage limit (see Bases for LCO 3.1.6, " Rod Pattern Control")

during reactivity insertion events.

The negative reactivity insertion (scram) provided by the CRD System provides the analytical hasis for determination of plant thermal limits and provides protection against fuel damage limits during a Rod Withdrawl Error (RWE) event. Bases for LCO 3.1.4, LCO 3.1.5, and LCO 3.1.6 discuss in more detail how the SLs are protected by the CRD System.

Control rod OPERABILITY satisfies Criterion 3 of the NRC Policy l

Statement.

OPERABILITY of an individual control rod is based on a combination of factors, primarily the scram insertion times, the control rod coupling integrity, and the ability to determine the control rod position. Accumulator OPERABILITY is addressed by LCO 3.1.5. Although not all control rods are required to be

i w

OPERABLE to satisfy the intended reactivity control requirements, strict control over the number and distribution ofinoperable control rods is required to satisfy the assumptions of the DBA and transient analyses.

In MODES I and 2, the control rods are assumed to function during a DBA or transient and are therefore required to be OPERABLE in these MODES. In MODES 3 and 4, control rods are only allowed to be withdrawn under Special Operations LCO 3.10.3, " Control Rod Withdrawal-Ilot Shutdown," and LCO 3.10.4, " Control Rod Withdrawal-Cold Shutdown," which provide adequate requirements for control rod OPERABILITY during these conditions. Control rod requirements in MODE 5 are located in LCO 3.9.5, " Control Rod OPERABILITY-Refueling."

The ACTIONS table is modified by a Note indicating that a separate Condition entry is allowed for each ctmtrol rod. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable control rod. Complying with the Required Actions may allow for continued operation, and subsequent inoperable control rods are governed by subsequent Condition entry and application of associated Required Actions.

A.1, A.2, and A.3 j

A control rod is considered stuck if it will not insert by either FMCRD j

drive motor torque or scram pressure. The failure of a control rod to insert during SR 3.2.3.2 or SR 3.2.3.3 alone, however, does not necessarily mean that the control rod is stuck, since failure of the motor drive would also result in a failure of these tests. Verincation of a stuck rod can be make by attempting to withdraw the rod. If the motor is working and the rod is actually stuck, the traveling nut will back down from the bottom of the drive and a rod separation alarm and rod block will result (see LCO 3.3.2). Conversely, if the motor drive is known to be failed, the rod is not necessarily inoperable since it is probably still capable of scram. However, at the next required performance of SR 3.1.3.2 or 3.1.3.3, there would be no way of verifying insertability, I

except by scram. In this case, an individual scram should be attempted. If the rod scrams, the rod is not stuck but should be consider inoperable and bypassed in RC&lS since it cannot be withdrawn and a separation situation will exist until the motor is repaired and the traveling nut is i

n

run-in to the full position. If the rod fails to insen by individual scram, it should be considered stuck and the appropriate ACTIONS taken. The failure of a control rod pair to in:,ert is assumed in the design basis transient and accident analyses and therefore, with one withdrawn control rod stuck, some time is allowed to make the control rod insertable, With a fully inserted control rod stuck, no actions are required as long as the l

control rod remains fully inserted. The Required Actions are modified by a Note that allows a stuck control rod to be bypassed in the Rod Control Information System (RC&lS) to allow continued operation. [SR 3.3.2.1.6]

provides additional requirements when control rods are bypassed in RC&lS to ensure compliance with the RWE analysis. With one withdrawn control rod stuck, the control rod must be disarmed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The motor drive may be disarmed by placing the rod in RD&lS bypass or by manually disconnecting its power supply. The allowed Completion Time i

of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is acceptable, considering the reactor can still be shut down, assuming no additional control rods fail to insert, and provides a reasonable amount of time to perform the Required Action in an orderly manner. Isolating the control rod from scram prevents damage to the CRD and surrounding fuel assemblies should a scram occur. The control rod can be isolated from scram by isolating its associate hydraulic control unit.

Two CRDs sharing an HCU can be individually isolated from scram Monitoring of the insertion capability of withdrawn control rods must also be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . SR 3.1.3.2 and SR 3.1.3.3 perform periodic tests of the control rod insertion capability of withdrawn control j

rods. Testing withdrawn control rods ensures that a generic problem does not exist. The allowed Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provides a l

reasonable time to test the control rods, considering the potential for a need to reduce power to perform the tests. Required Action A.2 is modified by a Note that states the requirement is not applicable when below the actual low power setpoint (LPSP) of the RC&lS, since the notch insertions may not be compatible with the requirements of rod pattern control (LCO 3.1.6) and the RC&lS (LCO 3.3.2).

To allow continued operation with a withdrawn control rod stuck, an evaluation of adequate SDM is also required within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Should a DBA or transient require a shutdown, to preserve the single failure criterion an additional control rod would have to be assumed to have failed to insert when required. Therefore, the original SDM demonstration may not be valid. The SDM must therefore be evaluated (by measurement

w or analysis) with the stuck control rod at its stuck position and the highest worth OPERABLE control rod assumed to be fully withdrawn.

With a single control rod stuck in a withdrawn position, the remaining OPERABLE control rods are capable of providing the required scram and shutdown reactivity. Failure to reach hiODE 4 is only likely if an additional control rod adjacent to the stuck control rod also fails to insert I

during a required scram. Even with the postulated additional single failure.

of an adjacent control rod to insert, sufficient reactivity control remains to j

reach and maintain hiODE 3 conditions (Reference B3.1.3-6). Required action A.2 performs a step test on each remaining withdrawn control rod to ensure that no additional control rods are stuck. Therefore,72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed to perform the analysis or test in Required Action A.3.

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B.1 and B.2 With two or more withdrawn control rods stuck, the stuck control rods should be isolated from scram pressure within I hour and the plant brought to hiODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The occurrence of more than one control rod stuck at a withdrawn position increases the probability that the reactor cannot be shut down if required. Insertion of all insertable control rods eliminates the possibility of an additional failure of a control rod to insert. 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 hiODE 3 from full power conditions in an orderly manner and without challenging plant systems.

C.1 and C.2 With one or more control rods inoperable for reasons other than being stuck in the withdrawn position, operation may continue, provided the control rods are fully inserted within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and disarmed (however, they j

do not need to be isolated from scram) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected. The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations. The control rods can be disarmed by disconnecting power to the motor drive or by placing the rod in RC&lS INOP Bypass. Required Action C.1 is modified by a Note that allows control rods to be bypassed in the RC&lS if required to allow insertion of the inoperable control rods and continued operation. Also, as noted, control rods declared inoperable with a failed motor drive can only be inserted by scram. Control rods with failed motor drives are not inoperable for this reason alone, but must be considered so upon failure of l

6.'

SR 3.1.3.2 or SR 3.1.3.3, or when not in compliance with GWSR (see LCO 3.1.6) This does not conflict with SR 3.0.1 since the ability to move the control rod via the FMCRD, as discussed in the bases for SR 3.1.3.2 and SR 3.1.3.3, is required to prove that the rod is not stuck. Likewise, loss of position indication, assuming no rod movement, would not result in control rod (s) inoperability until failure of SR 3.1.3.1. SR 3.3.2.1.6 provides additional requirements when the control rods are bypassed to ensure compliance with the RWE analysis.

The allowed Completion Times are reasonable, considering the small number of allowed inoperable control rods, and provide time to insert and disarm the control rods in an orderly manner and without challenging plant systems.

I D.1 and D.2 Out of sequence control rods may increase the potenti d reactivity worth of a control rod, or gang of control rods, during a RWE and therefore the distribution of inoperable control rods must be controllec. r sow f

10% RTP, the generic ganged withdrawal sequence restrictions (GWSR)

(which is equivalent to previous banked position withdrawal sequence, l

BPWS) analysis (Reference B3.1.3-6) requires inserted control rods not in compliance with GWS to be separated by at least two OPERABLE control j

rods in all directions, including the diagonal. Therefore, if one or more inoperable control rods are not in compliance with GWS and not separated l

by at least two OPERABLE control rods, action must be taken to restore l

compliance with GWSR or restore the control rods to OPERABLE status.

A Note has been added to the Condition to clarify that the Condition is not applicable when > 10% RTP since the GWSR is not required to be l

followed under these conditions, as described in the Bases for LCO 3.1.6.

E.1 If any Required Action and associated Completion Time of Condition A, C, or D are not met or nine or more inoperable control rods exist, 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 MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This ensures all insertable control rods are inserted and places the reactor in a condition that does not require the active function (i.e.,

scram) of the control rods. De number of control rods permitted to be inoperable when operating above 10% RTP could be more than the value specified, but the occurrence of a large number of inoperable control rods i

o could be indicative of a generic problem, and investigation and resolution of the potential problem should be undenaken. 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.

SR 3.1.3.1 The position of each control rod must be determined, to ensure adequate information on control rod position is available to the operator for determining CRD OPERABILITY and controlling rod patterns. Control l

rod position may be determined by the use of OPERABLE position indicators, by moving control rods to a position with an OPERABLE indicator, or by the use of other appropriate methods. The 24 hout Frequency of this SR is based on operating experience related to expected changes in control rod position and the availability of control rod position indications in the control room.

SR 3.1.3.2 and SR 3.1.3.3 Control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at two notches and observing that the control rod moves. The control rod may then be returned to its original position. This ensures the control rod is not stuck and is free to insert on a scram signal. These Surveillances are not required when below the actual l-LPSP of the RC&ls since the notch insertions may not be compatible with l

the requirements of rod pattern control (LCO 3.1.6) and the RC&lS l

(LCO 3.3.2.1). The 7 day Frequency of SR 3.1.3.2 is based on operating experience related to the changes in CRD performance and the ease of performing notch testing for fully withdrawn control rods. Partially withdrawn control rods are tested at a 31 day Frequency, based on the potential power reduction required to allow the control rod movement, and l

considering the large testing sample of SR 3.1.3.2. Furthermore, the 31 day Frequency takes into account operating experience related to changes in CRD performance. At any time, if a control rod is immovable, a determination of that control rod's appropriate action must be taken.

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i SR 3.1.3.4 Verifying the scram time for each control rod to 60% rod insertion position is s [ ] seconds provides reasonable assurance that the control rod will insert when required during a DBA or transient, thereby completing its shutdown function. This SR is performed in conjunction with the control rod scram time testing of SR 3.1.4.1, SR 3.14.2, SR 3.1.4.3, and SR 3.1.4.4. The LOGIC SYSTEM FUNC7iONAL TEST in LCO 3.3.1.1 and LCO 3.3.1.2 overlap this Surveillance to provide complete testing of the assumed safety function. The associated Frequencies are acceptable, considering the more frequent testing performed to demonstrate other aspects of control rod OPERABILITY and operating experience, which shows scram times do not significantly change over an operating cycle.

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SR 3.1.3.5 Coupling veri 6 cation is performed to ensure the control rod is connected to the CRDM and will perform its intended function when necessary. The Surveillance requires verifying that a control rod does not go to the withdrawn overtravel position when it is fully withdrawn. The overtravel position feature provides a positive check on the coupling integrity, since only an uncoupled hollow piston can reach the overtravel position. The veri 0 cation is required to be performed once the first time a control rod is withdrawn to the " full out" position after fuel movement within the RPV or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling. This includes control rods inserted one notch and then returned to the " full out" position during the performance of SR 3.1.3.2. This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved and operating experience related to uncoupling

events, i

REFERENCES l

10 CFR 50, Appendix A, GDC 26, GDC 27, GDC 28, and GDC 29.

ABWR SSAR, Section 4.6.2.

ABWR SS AR, Section 5.2.2.

ABWR SSAR, Section 15.4.1.

ABWR SSAR, Section 15.4.9.

NEDO-21231, " Banked Position Withdrawal Sequence," Section 7.2, January 1977.

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l Reactivity Control Systems Control Rod Scram Times l

BASES 1

BACKGROUND The scram function of the Control Rod Drive (CRD) System controls reactivity changes during abnormal operational transients to ensure that specified acceptable fuel design limits are not exceeded (Reference B3.1.4-1). The control rods are scrammed by positive means, using hydraulic pressure exerted on the CRD piston.

A single hydraulic control unit (IICU) powers the scram action of two fine motion control rod drives (FMCRDs). When a scram signal is initiated, control air is vented from the scram valve in each IICU, allowing it to open by spring action. Iligh pressure nitrogen then raises the piston within the IICU accumulator and forces the displaced water through the scram piping to the connected FMCRDs. Inside each FMCRD, the high pressure water lifts the hollow piston off the ball-nut and drives the control rod into the core. A buffer assembly j

stops the hollow piston at the end of its stroke. Departure from the ball-nut releases spring-loaded latches in the hollow piston that engage slots in the guide tube. These latches support the control rod in the inserted position. The control rod cannot he withdrawn until the i

hall-nut is driven up and engaged with the hollow piston Stationary fingers on the ball-nut then cam the latches out of the slots and hold them in the retracted position. A scram action is complete when every FMCRD has reached their fully inserted position.

APPLICABLE The analytical methods and assumptions used in evaluating the control SAFETY ANALYSES rod scram function are presented in Reference B3.1.4-2, Reference B3.1.4 3, and Reference B3.1.4-4. The Design Basis Accident (DBA) and transient analyses assume that all of the control rods scram at a specified I

insertion rate. The resulting negative scram reactivity forms the basis for the determination of plant thermal limits (e.g., the MCPR). Other distributions of scram times (e.g., several control rods scramming slower than the average time, with several control rods scramming faster than the average time) can also provide sufficient scram reactivity. Surveillance of each individual control rodensure the j

scram reactivity assumed in the DBA and transient analyses can be met.

The scram function of the CRD System protects the MCPR Safety l

l

POWER RATIO (MCPR)"), and the 1% cladding plastic strain fuel design limit (see Bases for LCO 3.2.1, " AVERAGE PLANAR

]

LINEAR IIEAT GENERATION RATE (APLIIGR)," and LCO 3.2.3,

" LINEAR IIEAT GENERATION RATE (LIIGR)"), which ensure j

l that no fuel damage will occur if these limits are not exceeded. Above 66.8 Kg/cm 2G (950 psig), the scram function is designed to insert i

negative reactivity at a rate fast enough to prevent the actual MCPR from becoming less than the MCPR SL during the analyzed limiting

)

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power transient. Below 66.8 Kg/cm 2G (950 psig), the scram function is assumed to perform during the Rod Withdrawal Error (RWE) event drop accident (Reference B3.1.4-4) and, therefore, also provides protection against violating fuel damage limits during reactivity insertion accidents (see Bases for LCO 3.1.6, " Rod Pattern Control").

For the reactor vessel overpressure protection analysis, the scram function, along with the safety / relief valves, ensure that the peak vessel pressure is maintained within the applicable ASME Code limits.

Control rod scram times satisfy Criterion 3 of the NRC Policy 1

Statement.

LCO The scram times specified in Table 31.4-1 (in the accompanying LCO) are required to ensure that the scram reactivity assumed in the DBA and transient analysis is met. To account for single failure and " slow" scramming control rods, the scram times specified in Table 3.1.4-1 are faster than those assumed in the design basis analysis. The scram times have a margin to allow up to 8.0% of the control rods to have l

l scram times that exceed the specified limits (i.e., " slow" control rods) assuming a single stuck control rod (as allowed by LCO 3.1.3,

" Control Rod OPERABILITY") and an idi!!cnal control rod failing to scram per the single failure criterion. the scram times are specified l

as a function of reactor steam dome pressure to account for the l

pressure dependence of the scram times. The scram times are specified relative to measurements based on reed switch positions, which provide the control rod position indication. The reed switch closes

(" pickup") when the hollow piston passes a specific location and then f

opens (" dropout") as the index tube travels upward Verification of the specified scram times in Table 3.1.4-1 is accomplished through measurement of the " dropout" times.

To ensure that local scram reactivity rates are maintained within acceptable limits, no more than two of the allowed " slow" control i

i

7 rods may occupy adjacent locations.

Table 3.1.4-1 is modified by two Notes, which state control rods with scram times not within the limits of the Table are considered " slow" and that control rods with scrum times > [ ] seconds to 60%

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insertion are considered inoperable as required by SR 3.1.3.4.

t APPLICABILITY In MODES I and 2, a scram is assumed to function during transients and accidents analyzed for these plant conditions. These events are assumed to occur during startup and power operation; therefore, the scram function of the control rods is required during these MODES.

j In MODES 3 and 4, the control rods are only allowed to be withdrawn under Special Operations LCO 3.10.3, " Control Rod Withdrawal-Ilot Shutdown," and LCO 3.10.4, " Control Rod Withdrawal-Cold Shutdown," which provide adequate requirements for control rod scram capability during these conditions. Scram requirements in MODE 5 are contained in LCO 3.9.5, " Control Rod OPERABILITY-Refueling."

l ACTIONS A.1 When the requirements of this LCO are not met, the plant must be l

brought to a MODE in which the LCO does not apply. To achieve this l

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 . The l

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 l

experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE The four SRs of this LCO are modified by a Note stating that during a REQUIREMENTS single control rod scram time surveillance, the CRD pumps shall be j

l isolated from the associated scram accumulator. With the CRD pump isolated (i.e., charging valve closed), the influence of the CRD pump head does not affect the single control rod scram times. During a full core l

scram, the CRD pump head would be seen by all control rods and would have a negligible effect on the scram insertion times.

SR 3.1.4.1 l

The scram reactivity used in DBA and transient analyses is based on assumed control rod scram time. Measurement of the scram times with reactor steam dome pressure 2 66.8 Kg/cm 'G (950 psig) demonstrates acceptable scram times for the transie-nts analyzed in Reference B3.1.4-2 and Reference B3.1.4-3.

Scram insertion times increase with increasing reactor pressure because of l

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the competing effects of reactor steam dome pressure and stored accumulator energy. Therefore, demonstration of adequate scram times at reactor steam dome pressure greater than 66.8 Kg/cm G (950 psig) ensures that the scram times will be within the speciDed limits at higher pressures.

Limits are specified as a function of reactor pressure to account for the sensitivity of the scram insertion times with pressure and to allow a range of pressures over which scram time testing can be performed. To ensure scram time testing is performed within a reasonable time following a refueling or after a shutdown 2: 120 days, all c(mtrol rods are required to be tested before exceeding 40% RTP following a shutdown. This Frequency is acceptable, considering the additional surveillances performed for control rod OPERABILITY, the frequent verification of adequate accumulator pressure, and the required testing of control rods affected by work on control rods or the CRD System.

SR 3.1.4.2 Additional testing of a sample of control rods is required to verify the continued performance of the scram function during the cycle. A representative sample contains at least 10% of the control rods, with no more than 20% of the control rods in the sample " slow." If more than 20% of the sample is declared to be " slow" per the criteria in Table 3.1.4-1, additional control rods are tested until this 20% criterion is satisfied, or Required Action A.1 must be taken. For planned testing, the control rods selected for the sample should be different for each test. Data from inadvertent scrams should be used whenever possible to avoid unnecessary testing at power, even if the control rods with data were previously tested in a sample. The 120 day Frequency is based on operating experience that has shown control rod scram times do not significantly change over an operating cycle. This Frequency is also reasonable, based on the additional Surveillances done on the CRDs at more frequent intervals in accordance with LCO 3.1.3 and LCO 3.1.5,

" Control Rod Scram Accumulators."

SR 3.1.4.3 When work that could affect the scram insertion time is performed on a control rod or the CRD System, testing must be done to demonstrate that each affected control rod retains adequate scram performance over the range of applicable reactor pressures from zero to the maximum permissible pressure. The scram testing must be performed once before

i declaring the control rod OPERABLE. The required scram time testing must demonstrate that the affected control rod is still within the limits of Table 3.1.41, for startup conditions.

Specific examples of work that could affect the scram times include (but are not limited to) the following: removal of any CRD for maintenance or modification; replacement of a control rod; and maintenance or modification of a scram solenoid pilot valve, scram valve, accumulator isolation valve, or check valves in the piping required for scram.

The Frequency of once prior to declaring the affected control rod OPERABLE is acceptable because of the capability of testing the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY.

SR 3.1.4.4 l

When work that could affect the scram insertion time is performed on a control rod or CRD System, testing must be done to demonstrate each affected control rod is still within the limits of Table 3.1.4-1 with the reactor steam dome pressure 2 66.8 Kg/cm G (950 psig). Where work has been performed at high reactor pressure, the requirements of SR 3.1.4.3 and SR 3.1.4.4 will be satisfied with one test. For a control rod affected by work performed while shut down, however, a zero pressure and a high pressure test may be required. This testing ensures that the control rod scram performance is acceptable for operating reactor l

pressure conditions prior to withdrawing the control rod for continued operation. Alternatively, a test during hydrostatic pressure testing could also satisfy both criteria.

The Frequency of once prior to exceeding 40% RTP is acceptable because

(

of the capability of testing the control rod at the different conditions and the more frequent surveillances on other aspects of control rod OPERABILITY.

REFERENCES 10 CFR 50, Appendix A, GDC 10.

ABWR SS AR, Section 4.6.2.

ABWR SS AR, Section 5.2.2.

FABWR SSAR, Section 15.4.1.

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Reactivity Control Systems Control Rod Scram Accumulators BASES BACKGROUND The control rod scram accumulators are part of the Control Rod Drive (CRD) System and are provided to ensure that the control rods scram under varying reactor conditions. The control rod scram accumulators store sufficient energy to fully insert two control rods at any reactor vessel pressure. The accumulator is a hydraulic cylinder with a free floating piston The pl.cton separates the water used to scram the control rods from the nitrogen, which provides the required energy. The scram accumulators are necessary to scram the control rods within the required insertion times of LCO 3.1.4, " Control Rod Scram Times."

APPLICABLE The analytical methods and assumptions used in evaluating the control SAFETY ANALYSES rod scram function are presented in Reference B3.1.5-1, Reference B3.1.5-2, Reference B3.1.5-3, and Reference B3.1.5-4. The Design Basis Accident (DBA) and transient analyses assume that all of l

the control rods scram at a specified insertion rate. OPERABILITY of l

each individual control rod scram accumulator, along with LCO 3.1.3,

" Control Rod OPERABILITY," and LCO 3.1.4, ensures that the scram reactivity assumed in the DBA and transient analyses can he j

met. The existence of an inoperable accumulator may invalidate prior scram time measurements for the associated control rod.

The scram function of the CRD System, and, therefore, the OPERABILITY of the accumulators, protects the MCPR Safety Limit (see Bases for LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)") and the 1% cladding plastic strain fuel design limit (see Bases for LCO 3.2.1, " AVERAGE PLANAR LINEAR IIEAT l

GENERATION RATE (APLIIGR)," and LCO 3.2.3, " LINEAR IIEAT GENERATION RATE (LIIGR)"), which ensure that no fuel damage will occur if these limits are not exceeded (see Bases for LCO 3.1.4). Also, the scram function at low reactor vessel presure l

(i.e., startup conditions) provides protection against violating fuel design limits during reactivity insertion accidents (see Bases for LCO 3.1.6, " Rod Pattern Control").

Control rod scram accumulators satisfy Criterion 3 of the NRC Policy Statement.

LCO The OPERABILITY of the control rod scram accumulators is

required to ensure that adequate scram insertion capability exists when needed over the entire range of reactor pressurts. The OPERABILITY of the scram accumulators is based on maintaining adequate accumulator pressure.

APPLICAlllLITY In MODES I and 2, the scram function is required for mitigation of DBAs and transients and, therefore, the scram accumulators must he OPERABLE to support the scram function. in MODES 3 and 4, control rods are only allowed to be withdrawn under Special Operations LCO 3.10.3, " Control Rod Withdrawal-Ilot Shutdown,"

and LCO 3.10.4, " Control Rod Withdrawal-Cold Shutdown," which provide adequate requirements for control rod scram accumulator OPERABILITY under these conditions. Requirements for scram accumulators in MODE 5 are contained in LCO 3.9.5, " Control Rod OPERABILITY-Refueling."

ACTIONS The ACTIONS table is modified by a Note indicating that a separate Condition entry is allowed for each control rod. This is acceptable since the Required Actions for each Condition provide appropriate compensatory action for each inoperable control rod. Complying with the Required Actions may allow for continued operation and subsequent inoperable control rods governed by subsequent Condition entry and application of associated Required Actions.

A.I and A.2 With one or more control rod scram accumulators inoperable the scram function could become severely degraded because the accumulators are the primary source of scram force for the control rods at all reactor pressures.

Therefore, it must be verified immediately that all control rods associated with inoperable scram accumulators are fully inserted. The associated control rods must also be declared inoperable within I hour. The allowed Completion Time of I hour is reasonable for Required Action A.2, considering the low probability of a DBA or transient occurring during the time the accumulator is inoperable. Additionally, an automatic reactor scram function is provided on sensed low pressure in the CRD charging water header (see LCO 3.3.1.1, "RPS Instrumentation *'). This anticipatory reactor trip protects against the possibility of significant pressure degradation (and thus reduced scram force) concurrently in multiple control rod scram accumulators due to a transient in the CRD hydraulic system.

B.1 The reactor mode switch must be immediately placed in the shutdown position if any Required Action and associated Completion Time cannot be l

met. This ensures that all insertable control rods are inserted and that the reactor is in a condition that does not require the active function (i.e.,

scram) of the control rods. This Required Action is modified by a Note stating that the Required Action is not applicable if all control rods associated with the inoperable scram accumulators are fully inserted, since the function of the control rods has been performed.

SURVEILLANCE SR 3.1.5.1 REQUIREMENTS SR 3.1.5.1 requires that the accumulator pressure be checked every 7 days to ensure adequate accumulator pressure exists to provide sufficient scram force. The primary indicator of accumulator OPERABILITY is the accumulator pressure. A minimum accumulator pressure is specified, below which the capability of the accumulator to perform its intended

f. <ction becomes degraded and the accumulator is considered inoperable.

The minimum accumulator pressure of 130.1 Kg/cm g(1850 psig) is well below the expected pressure of 151.2 Kg/cm g(2150 psig)

(Reference B3.1.5-2). Declaring the accumulator inoperable when the minimum pressure is not maintained ensures that significant degradation in scram times does not occur. The 7 day Frequency has been shown to be j

acceptable through operating experience and takes into account other indications available in the control room.

1 REFERENCES NEDE-24011-P-A, " General Electric Standard Application Fuel,"

September,1988.

ABWR SSAR, Section 4.6.1.

ABWR SSAR, Section 5.2.2.

APWR SSAR, Section 15.4.1.

W Reactivity Control Systems Rod Pattern Control BASES BACKGROUND Control rod patterns during startup conditions are controlled by the operator and the rod worth minimizer (RWM) (LCO 3.3.2.1,

" Control Rod Block Instrumentation"), so that only specified control rod sequences and relative positions are allowed over the operating range of all control rods inserted to 10% RTP. The sequences effectively limit the potential amount of reactivity addition that could occur during a control rod withdrawal, specifically the rod withdrawal error (RWE) event.

APPLICABLE The analytical methods and assumptions used in evaluating the RWE SAFETY ANALYSES are summarized in Reference B3.1.6-1 and Reference B3.1.6-2. RWE analyses assume that the reactor operator follows prescribed withdrawal sequences. These seque% brine the potential initial conditions for the RWE anal.v6. The RW \\1 (LCO 3.3.2.1) provides backup to operator control of tiie withdrawal sequences to ensure that the initial conditions of the RWE analysis are not violated.

Prevention or mitigation of positive reactivity insertion events is necessary to limit the energy deposition in the fuel, thereby preventing significant fuel damage, which could result in undue release of radioactivity (Reference B3.1.6-4). Since the failure consequences for UO have been shown to be insignificant below fuel energy depositions 2

of 300 cal /gm, the fuel damage limit of 280 cal /gm provides a margin of safety from significant core damage, which would result in release of radioactivity (Reference B3.1.6-2). Generic analysis of the GWSR (BPWS, see Reference B3.1.6-3) has demonstrated tha. the 280 cal /gm fuel damage limit will not he violated during a postulated reactivity transient while following the GWSR mode of operation.

Control rod patterns analyzed in Reference B3.1.6-1 and Reference B3.1.6-2 follow the GWSR which is the same as the banked position withdrawal sequence (BPWS) described in i

Reference B3.1.6-3. The GWSR is applicable from the condition of all control rods fully inserted to 10% RTP. For the GWSR, the control rods are required to be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions. The banked positions are defined to minimize the maximum l

incremental control rod worths without being overly restrictive during i

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l normal plant operation. The generic BPWS analysis (Reference B3.1.6-3) also evaluated the effect of fully inserted, l

inoperable control rods not in compliance with the sequence, to allow a limited number (i.e., eight) and distribution of fully inserted, i

inoperable control rods.

Rod pattern control satisfies the requirements of Criterion 3 of the NRC Policy Statement.

LCO Compliance with the prescribed control rod sequences minimizes the

')

l potential consequences of a RWE by limiting the initial conditions to thost consistent with the GWSR. This LCO only applies to i

OPERABLE control rods. For inoperable control rods required to be j

j inserted, separate requirements are specified in LCO 3.1.3, " Control Rod OPERABILITY," consistent with the allowances for inoperable j

control rods in the GWSR.

1 i

APPLICABILITY Compliance with GWSR is required in' MODES I and 2, when TIIERMAL POWER is s 10% RTP. When TIIERMAL POWER is

> 10% RTP, there is no credible control rod configuration that results in a control rod worth that could exceed the 280 cal /gm fuel damage limit during a RWE. In MODES 3,4, and 5, since only a total of ane control rod or control rod pair can be withdrawn from I

I core cells containing fuel assemblies, adequate SDM ensures that the I

reactor will remain subcritical.

I A.1 and A.2 With one er more OPERABLE control rods not in compliance with the prescribed control rod sequence, action may be taken to either correct the control rod pattern or declare the associated control rods inoperable within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Noncompliance with the prescribed sequence may be the result of failed synchros, drifting from a control rod drive purge water transient, l

leaking scram valves, or a power reduction to s 10% RTP before establishing the correct control rod pattern. The number of OPERABLE control rods not in compliance with the prescribed sequence is limited to eight to prevent the operator from attempting to correct a control rod pattern that significantly deviates from the prescribed sequence. When the control rod pattern is not in compliance with the prescribed sequence, all control rod movement should be stopped except for moves needed to correct the control rod pattern, or scram if warranted.

Required Action A.1 is modified by a Note, which allows control rods to be bypassed in RC&lS to allow the affected control rods to be returned to

s s

their correct position. This ensures that the control rods will be moved to the correct position. A control rod not in compliance with the prescribed sequence is not considered inoperable except as required by Required Action A.2. OPERABILITY of control rods is determined by compliance with LCO 3.1.3; LCO 3.1.4, " Control Rod Scram Times"; and LCO 3.1.5, " Control Rod Scram Accumulators." The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is reasonable, considering the restrictions on the number of allowed out of sequence control rods and the low probability of a RWE occurring during the time the control rods are out of sequence.

B.1 and B.2 If nine or more OPERABLE control rods are out of sequence, the control rod pattern significantly deviates from the prescribed sequence. Control rod withdrawal should be suspended immediately to prevent the potential for further deviation from the prescribed :;equence. Control rod insertion to correct control rods withdrawn beyond their allowed position is allowed since, in general, insertion of control rods has less impact on control rod worth than withdrawals have. Required Action B.1 is modified by a Note that allows the affected control rods to be bypassed in RC&lS in accordance with SR 3.3.2.1.8 to allow insertion only.

l With nine or more OPERABLE control rods not in compliance with GWSR, the reactor mode switch must be placed in the shutdown position within I hour. With the reactor mode switch in shutdown, the reactor is j

shut down, and therefore does not meet the applicability requirements of l

this LCO. The allowed Completion Time of I hour is reasonable to allow insertion of control rods to restore compliance, and is appropriate relative to the low probability of a RWE occurring with the control rods out of sequence.

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v SURVEILLANCE SR 3.I.6.1 REQUIREMENTS The control rod pattern is verified to be in compliance with the GWSR at a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency, ensuring the assumptions of the RWE analyses are met. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency of this Surveillance was developed considering that the primary check of the control rod pattern compliance with the GWSR is performed by the RWM (LCO 3.3.2.1). The RWM provides control rod blocks to enforce the required control rod sequence and is required to be OPERABLE when operating at s 10% RTP.

l 1

l NEDE-24011-P-A-9-US, " General Electric Standard Application for Reactor Fuel-Supplement for United States," September,1988.

ABWR SS AR, Section 15.4.1.

NUREG-0800, " Standard Review Plan," Section 15.4.1, "UncontrolledControl Rod Assembly Withdrawal from a Suberitical or Low Power Startup Conditon," Revision 2, July 1981.

10 CFR 100.11, " Determination of Exclusion Area Low Population l

Zone and Population Center Distance."

NEDO-21231, " Banked Position Withdrawal Sequence," January 1977.

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I Reactivity Control Systems Rod Pattern Control BASES BACKGROUND Control rod patterns during startup conditions are controlled by the operator and the rod worth minimizer (RWM) (LCO 3.3.2.1, l'

" Control Rod Block Instrumentation"), so that only specified control rod sequences and relative positions are allowed over the operating range of all control rods inserted to 10% RTP. The sequences effectively limit the potential amount of reactivity addition that could occur during a control rod withdrawal, specifically the rod withdrawal error (RWE) event.

APPLICABLE The analytical methods and assumptions used in evaluating the RWE SAFETY ANALYSES are summarized in Reference B3.1.6-1 and Reference B3.1.6-2. RWE

(

analyses assume that the reactor operator follows prescribed l

withdrawal sequences. These sequences define the potential initial l

conditions for the RWE analysis. The RWM (LCO 3.3.2.1) provides l

backup to operator control of the withdrawal sequences to ensure that the initial conditions of the RWE analysis are not violated.

l Prevention or mitigation of positive reactivity insertion events is necessary to limit the energy deposition in the fuel, thereby preventing significant fuel damage, which could result in undue release of radioactivity (Reference B3.1.6-4). Since the failure consequences for UO have been shown to be insignificant below fuel energy depositions 2

of 300 cal /gm, the fuel damage limit of 280 cal /gm provides a margin of safety from significant core damage, which would result in release of radioactivity (Reference B3.1.6-2). Generic analysis of the GWSR (BPWS, see Reference B3.1.6-3) has demonstrated that the 280 cal /gm fuel damage limit will not be violated during a postulated reactivity transient while following the GWSR mode of operation.

Control rod patterns analyzed in Reference B3.1.6-1 and Reference B3.1.6-2 follow the GWSR which is the same as the banked position withdrawal sequence (BPWS) described in Reference B3.1.6-3. The GWSR is applicable from the condition of all control rods fully inserted to 10% RTP. For the GWSR, the control rods are required to be moved in groups, with all control rods assigned to a specific group required to be within r,pecified banked positions. The banked positions are defined to minimize the maximum incremental control rod worths without being oserly restrictive during

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normal plant operation. The generic BPWS analysis (Reference B3.1.6-3) also evaluated the effect of fully inserted, inoperable control rods not in compliance with the sequence, to allow a limited number (i.e., eight) and distribution of fully inserted, inoperable control rods.

Rod pattern control satisfies the requirements of Criterion 3 of the NRC Policy Statement.

LCO Compliance with the prescribed control rod sequences minimizes the potential consequences of a RWE by limiting the initial conditions to those consistent with the GWSR. This LCO only applies to OPERABLE control nods. For inoperable control rods required to be inserted, separate requirements are specified in LCO 3.1.3, " Control Rod OPERABILITY," consistent with the allowances for inoperable j

control rods in the GWSR.

APPLICABILITY Compliance with GWSR is required in h10 DES I and 2, when TIIERh!AL POWER is s 107, RTP. When TIIERh!AL POWER is

> 10*/o RTP, there is no credible control rod configuration that results in a control rod worth that could exceed the 280 cal /gm fuel damage limit during a RWE. In h! ODES 3,4, and 5, since only a total of one control rod or control rod pair can be withdrawn from core cells containing fuel assemblies, adequate SDh1 ensures that the l

reactor will remain suberitical.

1 A.I and A.2 With one or more OPERABLE control rods not in compliance with the prescribed control rod sequence, action may be taken to either correct the control rod pattern or declare the associated control rods inoperable within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Noncompliance with the prescribed sequence may be the result of failed synchros, drifting from a control rod drive purge water transient, leaking scram valves, or a power reduction to s 10% RTP before establishing the correct control rod pattern. The number of OPERABLE control rods not in compliance with the prescribed sequence is limited to eight to prevent the operator from attempting to correct a control rod pattern that significantly deviates from the prescribed sequence. When the control rod pattern is not in compliance with the prescribed sequence, all i

control rod movement should be stopped except for moves needed to correct the control rod pattern, or scram if warranted.

Required Action A.1 is modified by a Note, which allows ccmtrol rods to be bypassed in RC&IS to allow the affected control rods to be returned to

i their correct position. This ensures that the control rods will be moved to the correct position. A control rod not in compliance with the prescribed sequence is not considered inoperable except as required by Required Action A.2. OPERABILITY of control rods is determined by compliance with LCO 3.1.3; LCO 3.1.4, " Control Rod Scram Times"; and LCO 3.1.5, " Control Rod Scram Accumulators." The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is reasonable, considering the restrictions on the number of allowed out of sequence control rods and the low probability of a RWE occurring during the time the control rods are out of sequence.

B.1 and B.2 If nine or more OPERABLE control rods are out of sequence, the control rod pattern signi6cantly deviates from the prescribed sequence. Control rod withdrawal should be suspended immediately to prevent the potential for further deviation from the prescribed sequence. Control rod insertion to correct control rods withdrawn beyond their allowed position is allowed l

since, in general, insertion of control rods has less impact on control rod worth than withdrawals have. Required Action B.1 is modi 6ed by a Note that allows the affected control rods to be bypassed in RC&lS in accordance with SR 3.3.2.1.8 to allow insertion only.

With nine or more OPERABLE control rods not in compliance with l

GWSR, the reactor mode switch must be placed in the shutdown position within I hour. With the reactor mode switch in shutdown, the reactor is shut down, and therefore does not meet the applicability requirements of this LCO. The allowed Completion Time of I hour is reasonable to allow insertion of control rods to restore compliance, and is appropriate relative to the low probability of a RWE occurring with the control rods out of

sequence, l

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1 -

SURVEILLANCE SR 3.1.6.1 REQUIREMENTS The control rod pattern is verified to be in compliance with the GWSR at a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency, ensuring the assumptions of the RWE analyses are met. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency of this Surveillance was developed considering that the primary check of the control rod pattern compliance with the GWSR is performed by the RWM (LCO 3.3.2.1). The RWM provides control rod blocks to enforce the required control rod sequence and is required to be OPERABLE when operating at s 10% RTP.

NEDE-240ll-P-A-9-US, " General Electric Standard Application for Reactor Fuel-Supplement for United States," September,1988.

l ABWR SSAR, Section 15.4.1.

NUREG-0800, " Standard Review Plan," Section 15.4.1, "UncontrolledControl Rod Assembly Withdrawal from a Subcritical or Low Power Startup Conditon," Revision 2, July 1981.

10 CFR 100.11, " Determination of Exclusion Area Low Population l

Zone and Population Center Distance.'

NEDO-21231, " Banked Position Withdrawal Sequence," January 1977.

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Reactivity Control Systems Standby Liquid Control (SLC) System BASES BACKGROUND The SLC System is designed to provide the capability of bringing the reactor, at any time in a fuel cycle, from full power and minimum control rod inventory (which is at the peak of the xenon transient) to a subcritical condition with the reactor in the most reactive xenon free state without taking credit for control rod movement. The SLC l

l System satisfies the requirements of 10 CFR 50.62 l

l (Reference B3.1.7-1) on anticipated transient without scram (ATWS).

The SLC System consists of a boron solution storage tank, two positive displacement pumps, two motor operated injection valves, which are provided in parallel for redundancy, and associated piping and valves used to transfer horated water from the storage tank to the reactor pressure vessel (RPV). The horated solution is discharged through the "B" Iligh Pressure Core Flooder (IIPCF) subsystem.

APPLICABLE The SLC System is automatically initiated. The SLC System is used l

SAFETY ANALYSES in the event that not enough control rods can be inserted to accomplish shutdown and cooldown in the normal manner. The SLC System injects horated water into the reactor core to compensate for all of the various reactivity effects that could occur during plant operation. To meet this objective, it is necessary to inject a quantity of boron that produces a concentration of 850 ppm of natural boron in the reactor core at 21*C (70*F). To allow for potential leakage and imperfect mixing in the reactor system, an additional amount of boron equal to 25% of the amount cited above is added (Reference B3.1.7-2).

The temperature versus concentration limits in Ilgure 3.1.7-1 (in the accompanying LCO) are calculated such that the required concentration is achieved accounting for dilution in the RPV with normal water level and including the water volume in the residual heat removal shutdown cooling piping. This quantity of borated solution is the amount that is above the pump suction shutoff levelin the boron solution storage tank. No credit is taken for the portion of the tank volume that cannot be injected.

The SLC System satisfies the requirements of the NRC Policy Statement because operating experience and probabilistic risk assessment have generally shown it to be important to public health and safety.

s-l LCO The OPERABILITY of the SLC System provides backup capability for reactivity control, independent of normal reactivity control provisions provided by the control rods. The OPERABILITY of the SLC System is based on the conditions of the horated solution in the l

I storage tank and the availability of a flow path to the RPV, including the OPERABILITY of the pumps and valves. Because the minimum required boron solution concentration is the same for both ATWS mitigation and cold shutdown (unlike some previous reactor designs) then if the boron solution concentration is less than the required limit, both SLC subsystems shall be declared inoperable. Two SLC subsystems are required to be OPERABLE, each containing an OPERABLE pump, a motor operated injection valve, and associated piping, valves, and instruments and controls to ensure nn OPERABLE flow path.

APPLICABILITY In MODES 1 and 2, shutdown capability is required. In MODES 3 and 4, control rods are only allowed to be withdrawn under Special l

Operations LCO 3.10.3, " Control Rod Withdrawal-llot Shutdown,"

I and LCO 3.10.4, " Control Rod Withdrawal-Cold Shutdown," which provide adequate controls to ensure the reactor remains subcritical. In MODE 5, only a single control rod or control rod pair can be j

withdrawn from a core cell containing fuel assemblies. Demonstration i

of adequate SDM (LCO 3.1.1, "SIIUTDOWN MARGIN (SDM)")

ensures that the reactor will not become criilcal. Tbrefore, the SLC System is not required to be OPERABLE during thu e 2nditions, when only a single control rod or control rod pair can be withdrawn.

ACTIONS A.1 If one SLC System subsystem is inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this condition, j

the remaining OPERABLE subsystem is adequate to perform the shutdown function. However, the overall reliability is reduced because a single failure in the remaining OPERABLE subsystem could result in reduced SLC System shutdown capability. The 7 day Completion Time is based on the availability of an OPERABLE subsystem capable of performing the intended SLC System function and the low probability of a Design Basis l

Accident (DBA) or severe transient occurring concurrent with the failure of the Control Rod Drive System to shut down the plant. The maximum Completion Time of 10 days is allowed for this LCO in the event of l

multiple Condition entry.

B.1 If both SLC subsystems are inoperable, at least one subsystem must be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is considered acceptable, given the low probability of a DBA or transient occurring concurrect with the failure of the control rods to shut down the reactor.

C.1 If any Required Action and associated Completion Time is 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 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.

SURVEILLANCE SR 3.1.7.1, SR 3.1.7.2, and SR 3.1.7.3 REQUIREMENTS SR 3.1.7.1 through SR 3.1.7.3 are 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Surveillances, verifying certain characteristics of the SLC System (e.g., the volume and temperature of the borated solution in the storage tank), thereby ensuring the SLC System OPERABILITY without disturbing normal plant operation. These Surveillances ensure the proper borated solution and temperature, including the temperature of the pump suction piping, are-maintained. Maintaining a minimum specified borated solution temperature is important in ensuring that the boron remains in solution and does not precipitate out in the storage tank or in the pump suction piping. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency of these SRs is based on operating experience that has shown there are relatively slow variations in the measured parameters of volume and temperature.

SR 3.1.7.4 This Surveillance requires an examination of the sodium pentaborate solution by using chemical analysis to ensure the proper concentration of boron exists in the storage tank. SR 3.1.7.4 must be performed anytime boron or water is added to the storage tank solution to establish that the boron solution concentration is within the specified limits. This Surveillance must be performed anytime the temperature is restored to l

l

4 within the limits of Figure 3.1.7-1, to ensure no significant boron precipitation occurred. The 31 day Frequency of this Surveillance is appropriate because of the relatively slow variation of boron concentration between surveillances.

SR 3.1.7.5 SR 3.1.7.5 verifies each valve in the system is in its correct position.

Verifying the correct alignment for manual, power operated, and automatic valves in the SLC System flow path ensures that the proper flow paths will exist for system operation. This Surveillance does not apply to valves that are locked, sealed, or otherwise secured in position, j

since they were verified to be in the correct position prior to locking, sealing, or securing. This verification of valve alignment does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct positions. The 31 day Frequency is based on engineering judgment and is consistent with the procedural controls governing valve operation that ensure correct valve positions.

1 SR 3.1.7.6 Demonstrating each SLC System pump develops a flow rate 211.4 m i

i s

1 1

/h (50 gpm) at a discharge pressure 2 86.0 Kg/cm j

l 2

g (1223 psig) ensures that pump performance has not degraded during the fuel cycle. This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reducticn, cooldown of the moderator, and xenon decay. This test confirms one point on the pump design curve, and is indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance.

The Frequency of this Surveillance is in accordance with the Inservice Testing Program or 92 days, I

i l

l SR 3.1.7.7 and SR 3.1.7.8 l

These Surveillances ensure that there is a functioning flow path from the

boron solution storage tank to the RPV. The pump and injection valve tested should be alternated such that both complete flow paths are tested every 36 months, at alternating 18 month intervals. The Surveillance may be performed in separate steps to prevent injecting boron into the RPV.

An acceptable method for verifying flow from the pump to the RPV is to pump demineralized water from a test tank through one SLC subsystem and into the RPV. 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 these components usually pass the Surveillance test when performed at the 18 month Frequency; therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

Demonstrating that all heat traced piping between the boron solution storage tank and the suction inlet to the injection pumps is unblocked ensures that there is a functioning flow path for injecting the sodium l

pentaborate solution. An acceptable method for verifying that the suction piping is unblocked is to pump from the storage tank to the test tank. The 18 month Frequency is acceptable since there is a low probability that the subject piping will be blocked due to precipitation of the boron from solution in the heat traced piping. This is especially true'in light of the daily temperature verification of this piping required by SR 3.1.7.3.

However, if, in performing SR 3.1.7.8, it is determined that the temperature of this piping has fallen below the specified minimum, this j

Surveillance must be performed once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the piping l

temperature is restored within the limits of Figure 3.1.7-1.

10 CFR 50.62.

ABWR SSAR, Section 9.3.5.3.

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J Power Distribution Limits B 3.2.1 Average Planar Linear llent Generation Rate (APLIIGR)

BASES BACKGROUND The APLilGR is a measure of the average LilGR of all the fuel rods in a fuel assembly at any axiallocation. Limib on the APLIIGR are specified to ensure that the fuel design limits identified in Reference B3.2.2-1 are not exceeded during anticipated operational occurrences (AOOs) and that the peak cladding temperature (PCT) during the postulated design basis loss of coolant accident (LOCA) does not exceed the limits specified in 10 CFR 50.46.

APPLICABLE The analytical methods and assumptions used in evaluating the SAFTTY ANALYSES fuel design limits are presented in the SSAR, Chapter 4, and in Reference B3.2.2-1. The analytical methods and assumptions used in j

evaluating Design Basis Accidents (DBAs), anticipated operational transients, and normal operations that determine APLIIGR limits are presented in SSAR, Chapters 4,6, and 15, and in Reference B3.2.2-1.

Fuel design evaluations are performed to demonstrate that the 17e limit on the fuel cladding plastic strain and other fuel design limits described in Reference B3.2.2-1 are not exceeded during AOOs for operation with LIIGR up to the operating limit LIIGR. APLIIGR limits are equivalent to the LIIGR limit for each fuel rod divided by the local peaking factor of the fuel assembly. APLIIGR limits are developed as a function of exposure and the various operating core flow and power states to ensure adherence to fuel design limits during the limiting AOOs. Flow dependent APLIIGR limits are determined using the three dimensional BWR simulator code (Reference B3.2.1-2) to analyze slow flow runout transients. The flow dependent multiplier, MAPFAC,, is dependent on the maximum core flow runout capability.

The maximum runout flow is dependent on the existing setting of the core now limiter in the Recirculation Flow Control System.

APPLICABLE Based on analyses of limiting plant transients (other than core now SAFETY ANALYSES increases) over a range of power and flow conditions, power dependent multipliers, MAPFAC,, are also generated. Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which turbine stop valve closure and turbine control valve fast closure scram signals are bypassed, both high and low core flow MAPFAC, limits are provided for operation at power levels between 257c RTP and the previously mentioned bypass power I

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level. The exposure dependent APLIIGR limits are reduced by h1APFAC, and h1APFAC, at various operating conditions to ensure that all fuel design criteria are met for normal operation and AOOs.

A complete discussion of the analysis code is provided in Reference B3.2.1-3.

LOCA analyses are then performed to ensure that the above determined APLIIGR limits are adequate to meet the PCT and maximum oxidation limits of 10 CFR 50.46. The analysis is performed using calculational models that are consistent with the requirements of 10 CFR 50, Appendix K. A complete discussion of the analysis code is provided in Reference B3.2.2-1. The PCT following a postulated LOCA is a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is not strongly influenced by the rod to rod power distribution within an assembly. The APLIIGR limits specified are equivalent to the LIIGR of the highest powered fuel rod assumed in the LOCA analysis divided by its local peaking factor. A conservative multiplier is applied to the LIIGR assumed in the LOCA analysis to account for the uncertainty associated with the measurement of the APLIIGR.

l The APLIIGR satisfies Criterion 2 of the NRC Policy Statement.

l LCO The APLilGR limits specified in the COLR are the resuh of fuel l

design, DBA, and transient analyses. The limit is determined by multiplying the smaller of the AIAPFAC, and 51APFAC, factors times the exposure dependent APLIIGR Ilmits.

APPLICABILITY The APLHGR limits are primarily derhed from fuel design evaluations and LOCA and transient analyses that are assumed to occur at high power levels. Design calculations (Reference B3.2.1-3) and operating experience have shown that as power is reduced, the margin to the required APLIIGR limits increases. This trend j

continues down to the power range of 5% to 15% RTP when entry into 510DE 2 occurs. When in h10DE 2, the Startup Range Neutron hfonitor (SRN51) scram function provides prompt scram initiation l

during any significant transient, thereby effectively removing any i

APLilGR limit compliance concern in AIODE 2. Therefore, at TilERA1AL POWER levels s 25% RTP, the reactor operates with substantial margin to the APLIIGR limits; thus, this LCO is not i

required.

1

I ACTIONS A.1 If any APLHGR exceeds the required limits, an assumption regarding an initial condition of the DBA and transient analyses may not be met.

Therefore, prompt action is taken to restore the APLHGR(s) to within the required limits such that the plant will be operating within analyzed conditions and within the design limits of the fuel rods. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months l

Completion Time is sufficient to restore the APLHGR(s) to within its i

limits and is acceptable based on the low probability of a transient or DBA j

occurring simultaneously with the APLHGR out of specification.

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l B.1 If the APLHGR cannot be restored to within its required limits within the l

associated Completion Time, the plant must be brought to a hiODE or other specified condition in which the LCO does not apply. To achieve this status, THERhiAL POWER must be reduced to < 25% RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The allowed Completion Time is reasonable, based on operating experience, to reduce THERhfAL POWER to < 25% RTP in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.1.1 REQUIRE 51ENTS APLHGRs are required to be initially calculated within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERhiAL POWER is 2 25% RTP and then every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter.

They are compared to the specified limits in the COLR to ensure that the j

reactor is operating within the assumptions of the safety analysis. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution under normal conditions.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after THERMAL POWER 2 25% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels.

NEDO-240ll-P-A, " General Electric Standard Application for Reactor Fuel," September,1988.

NEDO-301300-A, " Steady State Nuclear Methods," May,1985.

NEDO-24154, " Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors," October,1978.

.r Power Distribution Limits B 3.2.2 i

Minimum Critical Power Ratio (MCPR)

BASES BACKGROUND The MCPR is the ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The MCPR Safety Limit (SL) is set such that 99.9'7o of the fuel rods avoid boiling transition if the limit is not violated (refer to the Bases for SL 2.1.2). The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (AOOs). Although fuel damage does not necessarily occur if a fuel rod actually experiences boiling transition (Reference B3.2.2-1), the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion.

The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations have been developed to predict critical bundle power (i.e., the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling). Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensurint, that fuel failures due to inadequate cooling do not occur.

l APPLICABLE The analytical methods and assumptions used in evaluating the SAFETY ANALYSES AOOs to <stablish the operating limit MCPR are presented in the SSAR, Chapters 4,6, and 15, and Reference B3.2.2-2. To ensure that the MCPR SL is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR). The types of transients evaluated are loss of feedwater flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest change in CPR (DCPR). When the largest DCPR is added to the MCPR SL, the required operating limit MCPR is obtained.

APPLICABLE The MCPR operating limits derived from the transient analysis are SAFETY ANALYSES dependent on the operating core flow and power state (MCPR, and MCPR,, respectively) to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency. Flow

t 1

dependent MCPR limits are determined by steady state thermal hydraulic methods with key physics response inputs benchmarked using the three dimensional BWR simulator code (Reference B3.2.2-3) to analyze slow flow runout transients. The operating limit is dependent on the maximum core flow limiter setting in the Recirculation Flow Control System.

Power dependent MCPR limits (MCPR,) are determined by the one dimensional transient code (Reference B3.2.2-4) for the anticipated 1

transients that are significantly affected by power. Due to the i

sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine f

control valve fast closure scram trips are bypassed, high and low flow MCPR limits are provided for operating between 25% RTP and m

the previously mentioned bypass power level.

The MCPR satisfies Criterion 2 of the NRC Policy Statement.

l The MCPR operating limits specified in the COLR are the result of the Design Basis Accident (DBA) and transient analysis. The MCPR operating limits are determined by the larger of the MCPR, and MCPR, limits.

APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 257c RTP, the reactor is operating at minimum reactor internal pump speed and the moderator void ratio is small. Surveillance of thermal limits below 25% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient r,ccurs.

Statistical analyses documented in Reference B3.2.2-5 indicate that the l

nominal value of the initial MCPR expected at 25% RTP is > 3.5.

I Studies of the variation of limiting transient behavior have been performed over the range of power and flow conditions. These studies encompass the range of key actual plant parameter values important to typically limiting transients. The results of these studies demonstrate that a margin is expected between performance and the MCPR requirements, and that margins increase as power is reduced to 25% RTP. This trend is expected to continue to the 5% to 15% power range when entry into MODE 2 occurs. When in MODE 2, the Startup Range Neutron Monitor (SRNM) provides rapid scram initiation for any significant power increase transient, which I

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l effectively eliminates any MCPR compliance concern. Therefore, at TIIERMAL POWER levels < 25% RTP, the reactor is operating with substantial margin to the MCPR limits and this LCO is not required.

ACTIONS A.1 If any MCPR is outside the required limits, an assumption regarding an initial condition of the design basis transient analyses may not be met.

Therefore, prompt action should be taken to restore the MCPR(s) to within the required limits such that the plant remains operating within -

analyzed conditions. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is normally sufEcient to restore the MCPR(s) to within its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the MCPR out of specification.

B.1 If the MCPR cannot be restored to within the required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 25% RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.2.1 l

REQUIREMENTS l

The MCPR is required to be initially calculated within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after l

THERMAL POWER is 2 25% RTP and then every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter.

It is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution during normal operation.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after THERMAL POWER reaches 2 25% RTP is acceptable given the large inherent margin to operating limits at low power levels.

NUREG-0562, June,1979.

NEDO-240ll-P-A, " General Electric Standard Application for Reactor Fuel", September,1988 NEDO-30131-A, " Steady State Nuclear Methods", May,1985.

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NEDO-24154, "Qualif'ication of the One-Dimensional Core Transient Model for Boiling Water Reactors", October,1978."

"BWR/6 General Rod Withdrawal Error Analysis", Appendix 15B, General Electric Standard Safety Analysis Report (GESSAR-II).

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Power Distribution Limits Linear IIcal Generation Rate (LIIGR) (Optional)

BASES BACKGROUND The LIIGR is a measure of the heat generation rate of a fuel rod in a fuel assembly at any axial location. Limits on the LIIGR are specified to ensure that fuel design limits are not exceeded anywhere in the core during normal operation, inclu< ling anticipated operational occurrences (AOOs). Exceeding the LIIGR limit could potentially l

result in fuel damage and subsequent release of radioactive materials.

Fuel design limits are specified to ensure that fuel system damage, fuel rod failure or inability to cool the fuel does not occur during the anticipated operating conditions identified in Reference B3.2.3-1.

APPLICABLE The analytical methods and assumptions used in evaluating the fuel SAFETY ANALYSES system design are presented in Reference B3.2.3-1 and Reference B3.2.3-2. The fuel assembly is designed to ensure (in conjunction with the core nuclear and thermal hydraulic design, plant equipment, instrumentation, and protection system) that fuel damage will not result in the release of radioactive materials in excess of the guidelines of 10 CFR, Parts 20,50, and 100. The mechanisms that could cause fuel damage during operational transients and that are considered in fuel evaluations are:

Rupture of the fuel rod cladding caused by strain from the relative expansion of the 00, pellet; and Severe overheating of the fuel rod cladding caused by inadequate cooling.

A value of 1% plastic strain of the Zircaloy cladding has been defined as the limit below which fuel damage caused by overstraining of the fuel cladding is not expected to occur (Reference B3.2.3-3). The MCPR Safety Limit ensures that fuel damage caused by severe overheating of the fuel rod cladding is avoided.

j Fuel design evaluations have been performed and demonstrate that the 1% fuel cladding plastic strain design limit is not exceeded during continuous operation with LIIGRs up to the operating limit specified in the COLR.

APPLICABLE The analysis also includes allowances for short term transient SAFETY ANALYSES operation above the operating limit to account for AOOs, plus an allowance for densification power spiking.

The LIIGR satisfies Criterion 2 of the NRC Policy Statement.

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9 LCO The LIIGR is a basic assumption in the fuel design analysis. The fuel has been designed to operate at rated core power with sufficient design margin to the LIIGR calculated to cause a 1% fuel cladding plastic strain. The operating limit to accomplish this objective is specified in the COLR.

j APPLICABILITY The LilGR limits are derived from fuel design analysis that is limiting at high power level conditions. At core thermal power levels i

< 25% RTP, the reactor is operating with a substantial margin to the LIIGR limits and, therefore, the Specification is only required when the reactor is operating at h 25% RTP.

ACTIONS A.1 If any LHGR exceeds its required limit, an assumption regarding an initial condition of the fuel design analysis is not met. Therefore, prompt action j

should be taken to restore the LHGR(s) to within its required limits such that the plant is operating within analyzed conditions. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is normally sufficient to restore the LHGR(s) to within its limits and is acceptable based on the low probability of a transient or Design Basis Accident occurring simultaneously with the LHGR out of specification.

B.1 If the LHGR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 25% RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.3.1 REQUIREMENTS The LHGR is required to be initially calculated within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is 2 25% RTP and then every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, it is compared with the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on both engineering judgment and recognition l

of the slowness of changes in power distribution under normal conditions.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after THERMAL POWER 2 25% RTP is i

~

s achieved is acceptable given the large inherent margin to operating limits at lower power levels.

[Non GE Fuel Analysis).

ABWR SSAR Chapter 4.

NUREG-0800,Section II A.2(g), Revision 2, July 1981.

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