ML20046C378
| ML20046C378 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 07/15/1993 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Marriott P GENERAL ELECTRIC CO. |
| References | |
| NUDOCS 9308100219 | |
| Download: ML20046C378 (99) | |
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NUCLEAR REGULATORY COMMISSION j[
f W ASHINGTJN. D.C. 20555-0001 o
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July 15,1993 g
Docket No.52-001 Mr. Patrick W. Marriott, Manager Licensing & Consulting Services GE Nuclear Energy 175 Curtner Avenue San Jose, California 95125
Dear Mr. Marriott:
ADVANCED BOILING WATER REACTOR (ABWR) TECHNICAL SPECIFICATIONS
SUBJECT:
Enclosed are the proof and review ABWR technical specifications and their bases for the following sections:
3.9 Refueling Operations 3.10 Special Operations 4.0 Design Features I
As we discussed at our management meeting on June 10, 1993, the Nuclear Regulatory Commission (NRC) staff will be providing GE Nuclear Energy (GE),
until August 31, selected sections of proof and review ABWR technical specifi-r These sections are based on the NRC staff review of the GE mark-up cations.
of the BWR-6 and BWR-4 Standard Technical Specifications; the sections, as As discussed, we anticipate that provided, are acceptable to the NRC staff.
GE will interfcce very closely with the staff to resolve any issues on these 31, 1993. Under this arrangement, we anticipate that sections prior to August formal comments to proof and review ABWR technical specifications made by September 20, 1993, will be few.
The electronic text of these sections is available on the NRC Technical Specifications Branch electronic bulletin board (OTSB-BBS) in Wordperfect 5.1 and the i
The data telephone number for the OTSB-BBS is (301) 504-1778, format.
system operator is Tom Dunning who is available for assistance at (301)
Also, in accordance with our agreements, GE will maintain these 504-1189.
sections in Wordperfect 5.1 format and will produce subsequent issues of the ABWR technical specifications in Wordperfect 5.1 format.
1 i
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PDR ADOCK 05200001 l
A PDR
Mr. Patrick W. Marriott July 15,1993 If you have any questions about technical specifications please contact Mark Reinhart with the Nuclear Reactor Regulation Technical Specifications Branch.
He may be reached at (301) 504-1185.
Sincerely, original signed by:
Jerry N. Wilson Dennis M. Crutchfield, Associate Director for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation
Enclosure:
As stated t
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OFFICIAL RECORD COPY:
1 Mr. Patrick W. Marriott Docket No.52-001 i
General Electric Company cc:
Mr. Robert Mitchell Mr. Joseph Quirk General Electric Company GE Nuclear Energy 175 Curtner Avenue General Electric Company San Jose, California 95125 175 Curtner Avenue, Mail Code 782 San Jose, California 95125 Mr. L. Gifford, Program Manager Regulatory Programs Mr. Raymond Ng GE Nuclear Energy 1776 Eye Street, N.W.
12300 Twinbrook Parkway Suite 300 Suite 315 Washington, D.C.
20006 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.
Suite 800 Washington, D.C.
20004 Jay M. Gutierrez, Esq.
Newman & Holtzinger, P.C.
1615 L Street, N.W.
Suite 1000 Washington, D.C.
20036 Mr. Steve Goldberg Budget Examiner 725 17th Street, N.W.
Room 8002 i
Washington, D.C.
20503 Mr. Frank A. Ross U.S. Department of Energy, NE-42 Office of LWR Safety and Technology 19901 Germantown Road Germantown, Maryland 20874
l 1
Refueling Equipment Interlocks 3.9.1 3.9 REFUELING OPERATIONS 3.9.1 Refueling Equipment Interlocks LCC 3.9.1 The refueling equipment interlocks shall be OPERABLE.
APPLICABILITY:
During in-vessel fuel movement with equipment associated with the interlocks.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more required A.1 Suspend in-vessel Immediately refueling equipment fuel movement with interlocks inoperable.
equipment associated with the inoperable interlock (s).
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.1.1 Perform CHANNEL FUNCTIONAL TEST on each of 7 days the following required refueling equipment interlock inputs:
a.
All-rods-in, b.
Refuel platform position, and c.
Refuel platform [ main] hoist, fuel
- loaded, t
ABWR TS 3.9-1 Rev.
O, 11/20/92
Refuel Position One-Rod-Out Interlock 3.9.2 3.9 REFUELING OPERATIONS 3.9.2 Refuel Position One-Rod-Out Interlock LCO 3.9.2 The refuel position one-rod-out interlock shall be OPERABLE.
APPLICABILITY:
MODE 5 with the reactor mode switch in the refuel position and any control rod withdrawn.
1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Refuel position one-A.I Suspend control rod Immediately rod-out interlock withdrawal.
AND A.2 Initiate action to Immediately fully insert all insertable control rods in core cells containing one or more fuel assemblies.
i 1
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.2.1 Verify reactor mode switch locked in refuel 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> position.
t (continued) 1 5
h t
ABWR TS 3.9-I Rev.
O, 11/20/92 O
Refuel Position One-Rod-Out Interlock 3.9.2 T
SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY I
NOTE--------------------
Not required to be performed until I hour after any control rod is withdrawn.
Perform CHANNEL FUNCTIONAL TEST.
7 days t
4 r
I t
t l
t i
t I
i ABWR TS 3.9-2 Rev.
O, 11/20/92
Control Rod OPERABILITY--Refueling 3.9.5 q
3.9 REFUELING OPERATIONS 3.9.5 Control Rod OPERABILITY--Refueling 3
LC0 3.9.5 Each withdrawn control rod shall be OPERABLE.
APPLICABILITY:
MODE 5.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more withdrawn A.1 Initiate action to Immediately control rods fully insert inoperable.
inoperable withdrawn control rods.
E SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.5.1 Insert each withdrawn control rod at least 7 days one step.
SR 3.9.5.2 Verify each withdrawn control rod scram 7 days accumulator pressure _is 2 SI units
[1520] psig.
[
E ABWR TS 3.9-1 Rev.
O, 11/20/92 1
Control Rod Position 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Control Rod Position LCO 3.9.3 All control rods shall be fully inserted.
APPLICABILITY:
When loading fuel assemblies into the core.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more control A.I Suspend loading fuel Immediately rods not fully assemblies into the inserted.
core.
SURVEILLANCE REQUIREMENTS j
SURVEILLANCE FREQUENCY I
SR 3.9.3.I Verify all control rods are fully inserted.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
[
ABWR TS 3.9-I Rev.
O, 11/20/92
Control Rod Position Indication 3.9.4 3.9 REFUELING OPERATIONS 3.9.4 Control Rod Position Indication LC0 3.9.4 One control rod " full-in" position indication channel for each control rod shall be OPERABLE.
l APPLICABILITY:
MODE 5.
ACTIONS
NOTE-------------------------------------
Separate Condition entry is L'; owed for each required channel.
CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more required A.1.1 Suspend in-vessel Immediately control rod position fuel movement.
indication channels inoperable.
AND A.I.2 Suspend control rod Immediately withdrawal.
AND I
A.I.3 Initiate action to Immediately fully insert all insertable control rods in core cells containing one or more fuel assemblies.
08-(continued) 5 1
i ABWR TS 3.9-I Rev.
O,11/20/92 i
1
Control Rod Position Indication 3.9.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
(continued)
A.2.1 Initiate action to Immediately fully insert the control rod -
associated with the inoperable position indicator.
AND A.2.2 Initiate action to _
Immediately disarm the associated fully inserted control rod drive.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.4.1 Verify the required channel has no " full-Each time the in" indication on each control rod that is control rod is not " full-in."
' withdrawn from the " full-in" position l
i i
. ABWR TS 3.9-2 Rev. O,11/20/92 I
Control Rod OPERABILITY--Refueling 3.9.5 3.9 REFUELING OPERATIONS 3.9.5 Control Rod OPERABILITY--Refueling LC0 3.9.5 Each withdrawn control rod shall be OPERABLE.
APPLICABILITY:
MODE 5.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more withdrawn A.1 Initiate action to Immediately control rods fully insert inoperable.
inoperable withdrawn control rods.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY j
SR 3.9.5.1 Insert each withdrawn control rod at least 7 days one step.
SR 3.9.5.2 Verify each withdrawn control rod scram 7 days accumulator pressure is 2 SI units
[1520) psig.
I ABWR TS 3.9-1 Rev.
O, 11/20/92
RPV Water Level --Irradiated Fuel 3.9.6 3.9 REFUELING OPERATIONS E
i 3.9.6 Reactor Pressure Vessel (RPV) Water Level --Irradiated Fuel LC0 3.9.6 RPV water level shall be 2 7.0 m (23 ft) above the top of the RPV flange.
APPLICABILITY:
During movement of irradiated fuel assemblies within the
[RPV),
During movement of new fuel assemblies or handling of control rods within the RPV, when irradiated fuel assemblies are seated within the [RPV).
j ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
RPV water level not A.I Suspend movement of Immediately within limit.
fuel assemblies and handling of control rods within the RPV.
t SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify RPV water level is 2 7.0 m (23 ft) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> above the top of the RPV flange.
r ABWR TS 3.9-1 Rev. O,11/20/92
[RPV) Water Level--New Fuel or Control Rods 3.9.7 REFUELING OPERATIONS (TO BE DELETED) 3.9 3.9.7
[ Reactor Pressure Vessel (RPV)] Water Level--New Fuel or Control Rods LCO 3.9.7
[RPV] water level shall be 2 [22 ft 8 inches] above the top of irradiated fuel assemblies seated within the [RPV).
APPLICABILITY:
During movement of new fuel assemblies or handling of control rods within the [RPV) when irradiated fuel assemblies are seated within the [RPV].
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
[RPV] water level not A.1 Suspend movement of Immediately within limit.
fuel assemblies and handling of control rods within the
[RPV).
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.7.1 Verify [RPV) water level is 2 [22 ft 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 8 inches] above the top of irradiated fuel assemblies seated within the [RPV).
ABWR TS 3.9-1 Rev.
O, 11/20/92
RHR--High Water Level 3.9.7 3.9 REFUELING OPERATIONS 3.9.7 Residual Heat Removal (RHR)--High Water Level LCO 3.9.8 One RHR shutdown cooling subsystem shall be OPERABLE and in operation.
NOTE----------------------------
The required RHR shutdown cooling subsystem may be removed from operation for up to [
] hours per [
] hour period.
APPLICABILITY:
MODE 5 with the water level 2 7.0 m (23 ft) above the top of the reactor pressure vessel (RPV) flange.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 1
A.
Required RHR shutdown A.1 Verify an alternate I hour cooling subsystem method of decay heat inoperable, removal is available.
AND Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter B.
Required Action and B.1 Suspend loading Immediately associated Completion irradiated fuel Time of Condition A assemblies into the not met.
RPV.
AND B.2 Initiate action to Immediately restore [ primary or secondary]
containment to OPERABLE status.
AND (continued)
ABWR TS 3.9-1 Rev.
O, 11/20/92
RHR--High Water Level
-3.9.7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B.
(continued)
B.3 Initiate action to Immediately restore one standby gas treatment subsystem to OPERABLE status.
AND B.4 Initiate action to Immediately restore one secondary containment isolation valve and associated instrumentation to OPERABLE status in each associated penetration flow path not isolated.
i C.
Required RHR shutdown C.1 Establish reactor 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from cooling subsystem not coolant circulation discovery of no in operation.
by an alternate reactor coolant method.
circulation AND C.2 Monitor reactor Once per hour coolant temperature.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.7.1 Verify one RHR shutdown cooling subsystem 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is operating.
ABWR TS 3.9-2 Rev.
O, 11/20/92
A RHR--Low Water Level 3.9.8-3.9 REFUELING OPERATIONS 3.9.8 Residual Heat Removal (RHR)--Low Water Level LC0 3.9.8 Two RHR shutdown cooling subsystems shall be OPERABLE, and one RHR shutdown cooling subsystem shall be in operation.
NOTE----------------------------
The required operating shutdown cooling subsystem may be removed from operation for up to [ ] hours per [ ] hour period.
APPLICABILITY:
MODE 5 with the water level < 7.0 m (23 ft) above the top of the reactor pressure vessel flange.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One or two RHR A.1 Verify an alternate I hour shutdown cooling method of decay heat subsystems inoperable.
removal is available AND for each inoperable RHR shutdown cooling Once per subsystem.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter B.
No RHR shutdown B.1 Establish reactor I hour from cooling subsystem in coolant circulation discovery of no operation.
by an alternate reactor coolant method.
circulation AND B.2 Monitor reactor Once per hour coolant temperature.
AND (continued)
ABWR TS 3.9-1 Rev.
O, 11/20/92
RHR--Low Water Level 3.9.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B.
(continued)
B.3 Initiate action to Immediately restore [ primary or secondary]
containment to OPERABLE status.
AND B.4 Initiate action to Immediately restore one standby gas treatment subsystem to OPERABLE status.
AND B.5 Initiate action to Immediately restore one secondary containment isolation valve and associated instrumentation to OPERABLE status in each associated penetration flow path not isolated.
9.
E3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.8.I Verify one RHR shutdown cooling subsystem 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is operating.
ABWR TS 3.9-2 Rev.
O, 11/20/92
Inservice Leak and Hydrostatic Testing Operation 3.10.1 l
i 3.10 SPECIAL OPERATIONS 3.10.1 Inservice Leak and Hydrostatic Testing Operation i
1 LC0 3.10.1 The average reactor coolant temperature specified in Table 1.1-1 for MODE 4 may be changed to "NA," and operation considered not to be in MODE 3; and the requirements of LC0 3.4.10 " Residual Heat Removal (RHR) Shutdown Cooling System--Cold Shutdown," may be suspended, to allow performance of an inservice leak or hydrostatic test i
provided the following MODE 3 LCOs are met:
a.
LC0 3.3.6.2, " Secondary Containment Isolation Instrumentation," [ Functions 1, 3, 4, and 5] of Table 3.3.6.2-1; b.
LCO 3.6.4.1, " Secondary Containment";
i c.
LCO 3.6.4.2, " Secondary Containment Isolation Valves (SCIVs)"; and d.
LC0 3.6.4.3, " Standby Gas Treatment (SGT) System."
APPLICABILITY:
MODE 4 with average reactor coolant temperature > [200]*F.
l t
l i
i i
i i
i f
3 l
t ABWR TS 3.10-1 Rev.
O, 11/20/92
Inservice Leak and Hydrostatic Testing Operation 3.10.1 ACTIONS
NOTE-------------------------------------
Separate Condition entry is allowed for each requirement of the LCO.
l i
j CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more of the A.1
NOTE---------
above requirements not Required Actions to
=
met.
be in MODE 4 include reducing average reactor coolant temperature to s [200]*F.
P 1
Enter the applicable Immediately Condition of the affected LCO.
08 F
A.2.1 Suspend activities Immediately that could increase the average reactor coolant temperature or pressure.
AND A.2.2 Reduce average 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> reactor coolant temperature to s [200]*F.
R ABWR TS 3.10-2 Rev.
O, 11/20/92
Inservice Leak and Hydrostatic Testing Operation 3,10.1
[
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l
r SR 3.10.1.1 Perform the applicable SRs for the required According to i
MODE 3 LCOs.
the applicable SRs a
i i
k i
P i
ABWR TS 3.10-3 Rev.
O, 11/20/92 t
Reactor Mode Switch Interlock Testing 3.10.2 3.10 SPECIAL OPERATIONS 3.10.2 Reactor Mode Switch Interlock Testing LC0 3.10.2 The reactor mode switch position specified in Table 1.1-1 (Section 1.1, Definitions) for MODES 3, 4, and 5 may be changed to include the run, startup/ hot standby, and refuel position, and operation considered not to be in MODE 1 or 2, to allow testing of instrumentation associated with the reactor mode switch interlock functions, provided:
a.
All control rods remain fully inserted in core cells containing one or more fuel assemblies; and b.
No CORE ALTERATIONS are in progress.
APPLICABILITY:
MODES 3 and 4 with the reactor mode switch in the run, startup/ hot standby, or refuel position, MODE 5 with the reactor mode switch in the run or startup/ hot standby position.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME
+
A.
One or more of the A.1 Suspend CORE Immediately above requirements not ALTERATIONS except met.
for control rod insertion.
AND A.2 Fully insert all I hour insertable control rods in core cells containing one or 4
more fuel assemblies.
NLD (continued) i i
1 ABWR TS 3.10-1 Rev.
O, 11/20/92 l
=
. Reactor Mode Switch Interlock Testing l
3.10.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
(continued)
A.3.1 Place the reactor 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> mode switch in the shutdown position.
OB A.3.2
NOTE---------
Only applicable in MODE 5.
Place the reactor 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> i
mode switch in the l
refuel position.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.2.1 Verify all control rods are fully inserted 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in core cells containing one or more fuel assemblies.
SR 3.10.2.2 Verify no CORE ALTERATIONS are in progress.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> h
ABWR TS 3.10-2 Rev.
O, 11/20/92
4 Control Rod Withdrawal--Hot Shutdown 3.10.3 3.10 SPECIAL OPERATI0i;S 3.10.3 Single Control Rod Withdrawal--Hot Shutdown t
LCO 3.10.3 The reactor mode switch position sp ecified in Table 1.1-1 for MODE 3 may be changed to include the refuel position, and operation considered not to be in MODE 2, to allow withdrawal of a single control rod or control rod pair, provided the following requirements are met:
a.
LC0 3.9.2, " Refuel Position One-Rod-Out Interlock";
b.
LCO 3.9.4, " Control Rod Position Indication",
c.
All other control rods are fully inserted; and i
d.
1.
LCO 3.3.1.1, " Reactor Protection System (RPS) i Instrumentation," MODE 5 requirements for Functions t
[1.a,1.b, Ic, 2.a, and 2.d,] of Table 3.3.1.1-1, and LC0 3.9.5, " Control Rod OPERABILITY--Refueling,"
08 2.
All other control rods in a five by five array centered on each control rod being withdrawn are disarmed, and LC0 3.1.1, " SHUTDOWN MARGIN (SDM)," MODE 5 i
requirements, except the single control rod or pair to be withdrawn may be assumed to be the highest worth control rod.
APPLICABILITY:
MODE 3 with the reactor mode switch in the refuel position.
ABWR TS 3.10-1 Rev.
O, 11/20/92
Control Rod Withdraeal--Hot Shutdown i
3.10.3 l
ACTIONS
NOTE-------------------------------------
Separate Condition entry is allowed for each requirement of the LCO.
4 CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more of the A.1
NOTES--------
above requirements not 1.
Required Actions met.
to fully insert all insertable control rods include placing i
the reactor mode switch in the shutdown position.
- 2. Only applicable if the requirement not met-is a required LCO.
Enter the applicable Immediately Condition of the affected LCO.
t 08 A.2.1 Initiate action to Immediately fully insert all insertable control rods.
AND A.2.2 Place the reactor 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> mode switch in the shutdown position.
t k
ABWR TS 3.10-2 Rev.
O, 11/20/92
Control Rod Withdrawal--Hot Shutdoen 3.10.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY i
SR 3.10.3.1 Perform the applicable SRs for the required According to LCOs.
the applicable SRs SR 3.10.3.2
NOTE--------------------
Not required to be met if SR 3.10.3.1 is satisfied for LC0 3.10.3.d.1 requirements.
Vt.rify all other control rods, other than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> the control rod being withdrawn, in a five by five array centered on each control rod being withdrawn, are disarmed.
SR 3.10.3.3 Verify all control rods are fully inserted.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ABWR TS 3.10-3 Rev.
O, 11/20/92
Control Rod Withdrawal--Cold Shutdown 3.10.4 3.10 SPECIAL OPERATIONS 3.10.4 Control Rod Withdrawal--Cold Shutdown LCO 3.10.4 The reactor mode switch position specified in Table 1.1-1 for MODE 4 may be changed to include the refuel position, I
and operation considered not to be in MODE 2, to allow withdrawal of a single control rod or control rod pair, and subsequent removal of the associated control rod drives (CRD) if desired, provided the following requirements are met:
a.
All other control rods are fully inserted; b.
1.
LC0 3.9.2, " Refuel Position One-Rod-Out Interlock,"
and LCO 3.9.4, " Control Rod Position Indication,"
08 2.
A control rod withdrawal block is inserted; and c.
1.
LC0 3.3.1.1, " Reactor Protection System (RPS)
Instrumentation," MODE 5 requirements for Functions
[1.a, 1.b, 1.c, 2.a, and 2.d.,]
of Table 3.3.1.1-1, LC0 3.3.1.2, " Reactor Protection System Trip Actuation MODE 5 Requirements and LC0 3.9.5, " Control Rod OPERABILITY--Refueling,"
OB 2.
All other control rods in a five by five array centered on the control rod being withdrawn are disarmed, and LC0 3.1.1, " SHUTDOWN MARGIN (SDM)," MODE 5 requirements, except the single control rod or control rod pair to be withdrawn may be assumed to be the highest worth control rod or control rod pair.
f' APPLICABILITY:
MODE 4 with the reactor mode switch in the refuel position.
ABWR TS 3.10-1 Rev.
O, 11/20/92
Control Rod Withdrawal--Cold Shutdown 3.10.4
}
ACTIONS
NOTE-------------------------------------
Separate Condition entry is allowed for each requirement of the LCO.
CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more of the A.1
NOTES----'----
above requirements not
- 1. Required Actions met with the affected to fully insert control rod (s) all insertable i
insertable.
control rods include placing the reactor mode switch in the shutdown position.
- 2. Only applicable if the requirement not met is a required LCO.
Enter the applicable Immediately Condition of the affected LCO.
og A.2.1 Initiate action to Immediately fully insert all insertable control rods.
AND A.2.2 Place the reactor I hour mode switch in the shutdown position.
(continued) i ABWR TS 3.10-2 Rev.
O, 11/20/92
+
Control Rod Withdrawal--Cold Shutdown 3.10.4 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME T
B.
One or more of the B.1 Suspend withdrawol of Immediately above requirements not the control rod and met with the affected removal of associated control rod not CRD.
insertable.
AND B.2.1 Initiate action to Immediately fully insert all control rods.
OR B.2.2 Initiate action to Immediately satisfy the requirements of this LCO.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQ0ENCY SR 3.10.4.1 Perform the applicable SRs for the required According to i
LCOs.
applicable SRs 4
NOTE--------------------
Not required to be met if SR 3.10.4.1 is satisfied for LCO 3.10.4.c.1 requirements.
Verify all other control rods a five 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by five array centered on eaEr ~.rtrol rod being withdrawn, are disarmeo (continued)
I ABWR TS 3.10-3 Rev.
O, 11/20/92 j
Control Rod Withdrewal--Cold Shutdown 3.10.4 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.10.4.3 Verify all other control rods are fully 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> inserted.
NOTE--------------------
Not required to be met if SR 3.10.4.1 is satisfied for LC0 3.19.4.b.1 requirements.
Verify a control rod withdrawal block is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> inserted.
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1 ABWR TS 3.10-4 Rev.
O, 11/20/92 l
1
CRD Removal--Refueling 3.10.5 3.10 SPECIAL OPERATIONS 3.10.5 Control Rod Drive (CRD) Removal--Refueling LC0 3.10.5 The requirements of LC0 3.3.1.1, " Reactor Protection System (RPS) Instrumentation"; LC0 3.3.8.2, " Reactor Protection i
System (RPS) Electric Power Monitoring"; LCO 3.9.1,
" Refueling Equipment Interlocks"; LCO 3.9.2, " Refueling Position One-Rod-Out Interlock"; LC0 3.9.4, " Control Rod Position Indication"; and LC0 3.9.5, " Control Rod OPERABILITY--Refueling," may be suspended in MODE 5 to allow the removal of a single CRD or CRD pair associated with a control rod (s) withdrawn from core cell (s) containing one or more fuel assemblies, provided the following requirements are met:
a.
All other control rods are fully inserted; b.
All other control rods in a five by five array centered i
on the control rod being removed are disarmed; c.
A control rod withdrawal block is inserted; i
d.
LC0 3.1.1, " SHUTDOWN MARGIN (SDM)," MODE 5 requirements, except the single control rod (or pair) to be withdrawn may be assumed to be the highest worth control rod pair; and e.
No other CORE ALTERATIONS are in progress.
APPLICABILITY:
MODE 5 with LC0 3.9.5 not met.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more of the A.1 Suspend removal of Immediately above requirements not the control rod and met.
associated CRD mechanism.
AND (continued)
ABWR TS 3.10-1 Rev.
O, 11/23/92 w
w
CRD Removal--Refueling 3.10.5' ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
(continued)
A.2.1 Initiate action to Immediately-fully inset all control rods.
i OR A.2.2 Initiate action to Immediately satisfy the requirements of this LCO.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.5.1 Verify all controls rods, other than the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> control rod withdrawn for the removal of the associated CRD, are fully inserted.
SR 3.10.5.2 Verify all control rods, other than the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> control rod withdrawn for the removal of the associated CRD, in a five by five array centered on each control rod withdrawn for the removal of the associated CRD, are disarmed.
SR 3.10.5.3 Verify a control rod withdrawal block is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> inserted.
(continued)
ABWR TS 3.10-2 Rev.
O, 11/23/92
CRD Removal--Refueling j
3.10.5 l
SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY j
SR 3.10.5.4 Perform SR 3.1.1.1.
According to SR 3.1.1.1 SR 3.10.5.5 Verify no CORE ALTERATIONS, other than the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> single control rod or control rod pair being removed, are in progress.
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i ABWR TS 3.10-3 Rev.
O, 11/23/92 l
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Multiple Control Rod Withdrawal--Refueling 3.10.6 3.10 SPECIAL OPERATIONS 3.10.6 Multiple Control Rod Withdrawal--Refueling i
LC0 3.10.6 The requirements of LC0 3.9.3, " Control Rod Position";
LCO 3.9.4, " Control Rod Position Indication"; and LC0 3.9.5,
" Control Rod OPERABILITY--Refueling," may be suspended, and the " full in" position indicators may be bypassed for any number of control rods in MODE 5, to allow withdrawal of these control rods, removal of associated control rod drives (CRDs), or both, provided the following requirements are met:
I a.
The four fuel assemblies are removed from the core cells associated with each control rod or CRD to be removed; b.
All other control rods in core cells containing one or more fuel assemblies are fully inserted; and c.
Fuel assemblies shall only be loaded in compliance with an approved spiral reload sequence.
APPLICABILITY:
MODE 5 with LCO 3.9.3, LCO 3.9.4, or LC0 3.9.5 not met.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more of the A.1 Suspend withdrawal of Immediately above requirements not control rods and met.
removal of associated i
CRDs.
I AND (continued)
ABWR TS 3.10-1 Rev.
O, 11/23/92
Multiple Control Rod Withdrawal-Refueling-3.10.6 i
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
(continued)
A.2.1 Initiate action to Immediately I
fully insert all control rods in core cells containii,g one or more fuel assemblies.
E A.2.2 Initiate action to Immediately satisfy the requirements of this LCO.
t SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY-SR 3.10.6.1 Verify the four fuel assemblies are removed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from core cells associated with each control rod 'or CRD removed.
SR 3.10.6.2 Verify all other control rods in core cells 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> containing one or more fuel assemblies are fully inserted.
NOTE--------------------
Only required to be met during fuel loading.
i-Verify fuel assemblies being loaded are in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> compliance with an approved [ spiral] reload sequence.
]
)
ABWR TS 3.10-2 Rev.
O,-11/23/92
i Control Rod Testing--Operating 3.10.7 3.10 SPECIAL OPERATIONS l
3.10.7 Control Rod Testing--Operating LC0 3.10.7 The requirements of LC0 3.1.6, " Rod Pattern' Control," may be suspended and control rods bypassed in the Rod Action Control Cabinet as allowed by SR 3.3.2.1.6, to allow performance of SDM demonstrations, control rod scram time testing, control rod friction testing, and the Startup Test Program, provided conformance to the approved control rod sequence for the specified test is verified by a second licensed operator or other qualified member of the technical staff.
i APPLICABILITY:
MODES 1 and 2 with LCO 3.1.6 not met.
i ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l
A.
Requirements of the A.1 Suspend performance Immediatoly LC0 not met.
of the test and exception to LCO 3.1.6.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY' SR 3.10.7.1 Verify movement of control rods is in During control compliance with the approved control rod rod movement sequence for the specified test by a second licensed operator or other qualified member of the technical staff.
i i
ABWR TS 3.10-1 Rev.
O, 11/23/92 t
SDM Test--Refueling 3.10.8 1
3.10 SPECIAL OPERATIONS 3.10.8 SHUTDOWN MARGIN (SDM) Test-Refueling LC0 3.10.8 The reactor mode switch position specified in Table 1.1-1 for MODE 5 may be changed to include the startup/ hot standby i
position, and operation considered not to be in MODE 2, to allow SDM testing, provided the following requirements are met:
a.
1.
LC0 3.3.2.1, " Control Rod Block Instrumentation,"
MODE 2 requirements for Function 1.b of Table 3.3.2.1-1, OR 2.
Conformance to the approved cor al rod sequence for the SDM test is verified by a.;cond licensed operator or other qualified member of the technical staff; b.
Each withdrawn control rod shall be coupled to the associated CRD; c.
All control rod withdrawals [during out of sequence control rod moves] shall be made in notch out mode; and d.
No other CORE ALTERATIONS are in progress.
APPLICABILITY:
MODE 5 with the reactor mode switch in startup/ hot standby position.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more of the A.1 Place the reactor Immediately above requirements not mode switch in the met.
shutdown or refuel position.
ABWR TS 3.10-1 Rev.
O, 11/23/92
SDM Test--Refueling 3.10.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.8.1
NOTE--------------------
Not required to be met if SR 3.10.8.2 satisfied.
Perform the applicable SRs for LCO 3.3.2.1, According to Function 1.b.
the applicable i
SRs SR 3.10.8.2
NOTE--------------------
Not requit ed to be met if SR 3.10.8.1 satisfied.
Verify movement of control rods is in During control compliance with the approved control rod rod movement sequence for the SDM test by a second licensed operator or other qualified member of the technical staff.
SR 3.10.8.3 Verify no other CORE ALTERATIONS are in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> progress.
(continued) 9 ABWR TS 3.10-2 Rev.
O,11/23/92
f 1
SDM Test--Refueling 3.10.8 SURVEILLANCE REQUIREMENTS (continued) l SURVEILLANCE FREQUENCY SR 3.10.8.4 Verify each withdrawn control rod does not Fach time the I
go to the withdrawn overtravel position.
vattrol rod is withdrawn to
" full out" position AND l
Prior to satisfying l
LC0 3.10.8.b requirement i
after work on control rod or CRD System that could affect coupling i
i 5
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ABWR TS 3.10-3 Rev.
O, 11/23/92
~
Reactor Internal Pumps--Testing 3.10.9 i
3.10 SPECIAL OPERATIONS 3.10.9 Reactor Internal Pumps--Testing LCO 3.10.9 The requirements of LC0 3.4.1, " Recirculation loops Operating," may be suspended for s 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to allow a.
PHYSICS TESTS, provided THERMAL POWER is s [5]% RTP; and b.
Performance of the Startup Test Program.
APPLICABILITY:
MODES 1 and 2 with less than five Reactor Internal Pumps in operation.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Requirements of A.I Insert all insertable I hour LC0 3.4.1 not met for control rods.
> 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B.
Requirements of the B.1 Place the reactor Immediately LC0 not met for mode switch in the reasons other than shutdown position.
Condition A.
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5 ABWR TS 3.10-1 Rev.
O,11/23/92
Reactor Internal Pumps-Testing 3.10.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.9.1 Verify LC0 3.4.1 requirements suspended for I hour s 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SR 3.10.9.2 Verify THERMAL POWER is s 5% RTP during_
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> PHYSICS TESTS.
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ABWR TS 3.10-2 Rev.
O, 11/23/92
Training Startups 3.10.10 3.10 SPECIAL OPERATIONS 3.10.10 Training Startups LC0 3.10.10 The low pressure coolant injection (LPCI) OPERABILITY requirements specified in LC0 3.5.1, "ECCS--Operating," may be changed to allow one residual heat removal subsystem to be aligned in the shutdown cooling mode for training startups, provided the following requirements are met:
a.
All OPERABLE intermediate range monitor (IRM) channels are s [25/40] divisions of full scale on Range 7; and b.
Average reactor coolant temperature is < 200*F.
APPLICABILITY:
MODE 2 with one LPCI subsystem suction valve closed.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME P
A.
One or more of the A.1 Place the reactor Immediately above requirements not mode switch in the met.
shutdown position.
P I
l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.10.1 Verify all OPERABLE IRM channels are I hour s [25/40] divisions of full scale on Range 7.
1 SR 3.10.10.2 Verify average reactor coolant temperature I hour is < 200*F.
ABWR TS 3.10-1 Rev.
O, 11/23/92 k
i Design Features 4.0 4.0 DESIGN FEATURES i
4.1 Site i
4.1.1 Site and Exclusion Area Boundaries The site and exclusion area boundaries [shall be as described or as shown in Figure 4.1-1].
4.1.2 Low Population Zone (LPZ)
The LPZ [shall be as described or as shown in Figure 4.1-2].
4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain [800] fuel assemblies.
Each assembly shall consist of a matrix of zirconium alloy fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U0 ) as fuel material [, and water rods).
Limited 2
substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used.
Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
t 4.2.2 Control Rod Assemblies The reactor core shall contain [193] cruciform shaped control rod-assemblies.
The control material shall be [ boron carbide, hafnium metal] as approved by the NRC.
i l
(continued)
ABWR TS 4.0-1 Rev.
O,11/23/92
Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:
a.
Fuel assemblies having a maximum [k-infinity of [1.31] in the normal reactor core configuration at cold conditions]
[ average U-235 enrichment of [4.5] weight percent];
b.
kiNc# s 0.95 if fully flooded with unborated water, which 1udes an allowance for uncertainties as described in
[Section 9.1 of the FSAR];
[c. A nominal fuel assembly center to center storage spacing of [7] inches within rows and [12.25] inches between rows in the [ low density storage racks] in the upper containment pool; and]
[d.
A nominal fuel assembly center to center storage spacing of [6.26] inches, within a neutron poison material between storage spaces, in the [high density storage racks] in the spent fuel storage pool and in the upper containment pool.]
4.3.1.2 The new fuel storage racks are designed and shall be maintained with:
a.
Fuel assemblies having a maximum [k-infinity of [1.31] in the normal reactor core configuration at cold conditions]
[ average U-235 enrichmant of [4.5] weight percent];
k,,,ludes an allowance for uncertainties as described ins 0.95 if fully flood b.
inc
[Section 9.1 of the FSAR];
k,,,llowance for uncertainties as described ins 0.98 if moderated by aqueo c.
an a
[Section 9.1 of the FSAR]; and d.
A nominal [6.26] inch center to center distance between fuel assemblies placed. in storage racks.
(continued)
ABWR TS 4.0-2 Rev.
O, 11/23/92
Design Features 4.0 i
4.0 DESIGN FEATURES 1
4.3 fuel Storage (continued) 4.3.2 Drainaae The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation [202 ft 5.25 inches].
l 4.3.3 Capacity 4.3.3.1 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than [2324] fuel assemblies.
4.3.3.2 No more than [800) fuel assemblies may be stored in the upper containment pool, s
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ABWR TS 4.0-3 Rev.
O, 11/23/92
?
Design Features 4.0 t
b This figure shall consist of [a map of] the site area and provide, as a minimum, the information described in Section
[2.1.2] of the FSAR relating to [the map].
S r
Figure 4.1-1 (page 1 of 1)
Site and Exclusion Area Boundaries 1
i 1
ABWR TS 4.0-4 Rev.
O, 11/23/92 l
~
Design Features 4.0 i
i i
t b
This figure shall consist of [a map of] the site area showing the LPZ boundary.
Features such as towns, roads, and recreational areas shall be indicated in sufficient detail to allow identification of significant shifts in l
population distribution within the LPZ.
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t i
Figure 4.1-2 (page 1 of 1)
Low Population Zone I
ABWR TS 4.0-5 Rev.
O, 11/23/92 i
\\
)
B 3.9 Refueling Operations Refueling Equipment Interlocks BASES BACKGROUND Refueling equipment interlocks restrict the operation of the refueling equipment or the withdrawal of control rods to reinforce unit procedures in preventing the reactor from achieving criticality during refueling. The refueling interlock circuitry senses the conditions of the refueling equipment and the control rods. Depending on the sensed conditions, interlocks are actuated to prevent the operation of the refueling equipment or the withdrawal of control rods.
GDC 26 of 10 CFR 50, Appendix A, requires that one of the two required independent reactivity control systems be capable of holding the reactor core suberitical under cold conditions (Ref. B3.9.1-1). The control rods, when fully inserted, serve as the system capable of maintaining the reactor suberitical in cold conditions during all fuel movement activities and accidents.
Two channels of instrumentation are provided to sense the position of the refueling platform, the loading of the refueling platform main hoist, and the full insertion of all control rods. With the reactor mode switch in the shutdown or refueling position, the indicated conditions are combined in logic circuits to determine if all restrictions on refueling equipment operations and control rod insertion are satisfied.
A control rod not at its full-in position interrupts power to the refueling equipment and prevents operating the equipment over the reactor core when loaded with a fuel assembly. Conversely, the refueling equipment located over the core and loaded with fuel inserts a control rod withdrawal block in the Rod Control and Information System (RC&lS)to prevent withdrawing a control rod.
The refueling platform has two mechanical switches that open before the platform and the fuel grapple are physically located over the reactor vessel. The main hoist has two switches that open when the hoist is loaded with fuel. The refueling interlocks use these indications i
to prevent operation of the refueling equipment with fuel loaded over the core whenever any ccmtrol rod is withdrawn, or to prevent control rod withdrawal wheneter fuel loaded refueling equipment is over the core (Ref. B3.9.1-2).
The hoist switches open at a load lighter than the weight of a single fuel assembly in water.
The refueling interlocks are explicitly assumed in the SSAR analysis of
the control rod removal error during refueling (Ref. B3.9.1-3). This analysis evaluates the consequences of c(mtrol rod withdrawal during refueling. A prompt reactivity excursion during refueling could potentially result in fuel failure with subsequent release of radioactive material to the environment.
Criticality and, therefore, subsequent prompt reactivity excursions are prevented during the insertion of fuel, provided all control rods are fully inserted during the fuel insertion. The refueling interlocks accomplish this by preventing loading fuel into the core with any control rod withdrawn, or by preventing withdrawal of a rod from the core during fuel loading.
The refueling platform location switches activate at a point outside of the reactor core, such that, considering switch hysteresis and maximum platform momentum toward the core at the time of power loss with a fuel assembly loaded and a control rod withdrawn, the fuel is not over the core.
Refueling equipment interlocks satisfy Criterion 3 of the NRC Policy Statement.
LCO To prevent criticality during refueling, the refueling interlocks ensure that fuel assemblies are not loaded with any control rod withdrawn.
To prevent these conditions from developing, the all-rods-in, the refueling platform position, and the refueling platform main hoist fuel loaded inputs are required to be OPERABLE. These inputs are combined in logic circuits that provide refueling equipment or control rod blocks to prevent operations that could result in criticality during refueling operations.
APPLICABILITY In MODE 5, a prompt reactivity excursion could cause fuel damage and subsequent release of radioactive material to the environment.
The refueling equipment interlocks protect against prompt reactivity excursions during MODE 5. The interlocks are only required to be OPERABLE during in-vessel fuel movement with refueling equipment associated with the interlocks.
In MODES 1,2,3, and 4, the reactor pressure vessel head is on, and no fuel loading activities are possible. Therefore, the refueling interlocks are not required to be OPLRABLE in these MODES.
ACTIONS A.1 With one or more of the required refueling equipment interlocks
inoperable, the unit must be placed in a condition in which the LCO does not apply. In-vessel fuel movement with the affected refwling equipment must be immediately suspended. *Ihis action ensures that operations are not performed with equipment that would potentially not be blocked from unacceptable operations (e.g., loading fuel into a cell with a control rod withdrawn). Suspension of in-vessel fuel movement shall not preclude completion of movement of a component to a safe position.
SR 3.9.1.1 Performance of a CHANNEL FUNCTIONAL TEST demonstrates each required refueling equipment interlock will function properly when a simulated or actual signal indicative of a required condition is injected into the logic. The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping, or total channel steps so that the entire channel is tested.
The 7 day Frequency is based on engineering judgment and is considered adequate in view of other indications of refueling interlocks and their associated input status that are available to unit operations personnel.
10 CFR 50, Appendix A, GDC 26.
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B 3.9 Refueling Operations Refuel Position One-Rod-Out Interlock BASES BACKGROUND The refuel position one-rod-out interlock restricts the movement of control rods to reinforce unit procedures that prevent the reactor from becoming critical during refueling operations. During refueling operations, no more than one control rod is permitted to be withdrawn. To ensure one-rod-out interlock, RC&IS GANG / SINGLE selection switch must be in " Single" mode. Otherwise, it is possible to withdraw the two rods associated with the same IICU while in the REFUEL MODE.
GDC 26 of 10 CFR 50, Appendix A, requires that one of the two required independent reactivity control systems be capable of holding the reactor core subcritical under cold conditions (Ref. B3.9.2-1). The control rods serve as the system capable of maintaining the reactor suberitical in cold conditions.
The refuel position one-rod-out interlock prevents the selection cf a second control rod for movement when any other control rod is not fully inserted (Ref. B3.9.2-2). It is a logic circuit that has redundant channels. It uses the all-rods-in signal (from the control rod full-in position indicators discussed in LCO 3.9.4, " Control Rod Position Indication") and a rod selection signal (from the Rod Control and Information System).
This Specification ensures that the performance of the refuel position one-rod-out interlock in the event of a Design Basis Accident meets the assumptions used in the safety analysis of Reference B3.9.2-3.
The refuel position one-rod-out interlock is explicitly assumed in the SSAR analysis of the control rod removal error during refueling (Ref. B3.9.2-3). This analysis evaluates the consequences of control rod withdrawal during refueling. A prompt reactivity excursion during refueling could potentially result in fuel failure with subsequent release of radioactive material to the emironment.
The refuel position one-rod-out interlock and adequate SDM (LCO 3.1.1) prevent criticality by preventing withdrawal of more than one c<mtrol rod. With one control rod withdrawn, the core will remain suberitical, thereby preventing any prompt critical excursion.
The refuel position one-rod-out interlock satisfies Criterion 3 of the NRC Policy Statement.
LCO To prevent criticality during MODE 5, the refuel position one-rod-out i
interlock ensures no more than one control rod may be withdrawn.
Both channels of the refuel position one-rod-out interlock are required to be OPERABLE.
APPLICABILITY In MODE 5, with the reactor mode switch in.be refuel position and the RC&lS GANG / SINGLE selection switch in " Single" mode, the OPERABLE refuel position one-rod-out interlock provides protection against prompt reactivity excursions.
In MODES 1,2,3, and 4, the refuel position one-rod-out interlock is not required to be OPERABLE and is bypassed. In MODES 1 and 2, the Reactor Protection System (LCO 3.3.1.1) and the control rods (LCO 3.1.2) provide mitigation of potential reactivity excursions. In MODES 3 and 4, with the reactor mode switch in the shutdown position, a control rod block (LCO 3.3.2.1) ensures all control rods 4
are inserted, thereby preventing criticality during shutdown conditions.
ACTIONS A.1 and.A.2 With one or both channels of the refuel position one-rod-out interlock inoperable, the refueling inte 'ocks may not be capable of preventing more than one control rod from being withdrawn. This condition may lead to criticality.
Control rod withdrawal must be immeNely suspended, and action must be immediately initiated to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Action must continue until all such control rods are fully inserted. 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.
SR 3.9.2.1 Proper functioning of the refueling position one-rod-out interlock requires the reactor mode switch to be in Refuel and the RC&lS GANG / SINGLE selection switch in " Single" mode. During control rod withdrawal in -
MODE 5, improper positioning of the reactor mode switch and the RC&lS GANG / SINGLE selection switch could, in some instances, allow improper bypassing of required interlocks. Therefore, this Surveillance i
imposes an additional level of assurance that the refueling position j
one-rod-out interlock will be OPERABLE when required. By " locking" j
the reactor mode switch in the proper position and placing the RC&lS i
1
=
GANG / SINGLE selection switch in " Single" mode, an additional administrative control is in place to preclude operator errors from resulting in unanalyzed operation.
The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient in view of other administrative controls utilized during refueling operations to ensure safe operation.
Performance of a CHANNEL FUNCTIONAL TEST on each channel demonstrates the associated refuel position one-rod-out interlock will function properly when a simulated or actual signal indicative of a
[
required condition is injected into the logic. The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping, or total channel steps so that the entire channel is tested. The 7 day Frequency is considered adequate because of demonstrated circuit reliability, procedural controls on control rod withdrawals, and visual and audible indications available in the ccmtrol room to alert the operator of control rods not fully inserted. To perform the required testing, the applicable condition must be entered (i.e., a control rod must be withdrawn from its full-in position). Therefore, SR 3.9.2.2 has been modified by a Note that states the CHANNEL FUNCTIONAL TEST is not required to be performed until I hour after any control rod is i
withdrawn.
10 CFR 50, Appendix A, GDC 26.
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B 3.9 Refueling Operations Control Rod Position BASES BACKGROUND Control rods provide the capability to maintain the reactor suberitical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the Control Rod Drive System. During refueling, rr9tement of control rods is limited by the refueling interlocks (LCO 3.9.1 and LCO 3.9.2) or the control rod block with the reactor mode switch in the shutdown position (LCO 3.3.2.1).
GDC 26 of 10 CFR 50, Appendix A, requires that one of the two required independent reactivity control systems be capable of holding the reactor core subcritical under cold conditions (Ref. B3.9.3-1). The control rods serve as the system capable of maintaining the reactor suberitical in cold conditions.
The refueling interlocks and the RC&IS GANG / SINGLE selection switch allow a single control rod to be withdrawn at any time unless fuel is being loaded into the core. To preclude loading fuel assemblies into the core with a control rod withdrawn, all control rods must be fully inserted. This prevents the reactor from achieving criticality during refueling operations.
Prevention and mitigation of prompt reactivity excursions during refueling are provided by the refueling interlocks (LCO 3.9.1 and LCO 3.9.2), the SDM (LCO 3.1.1), the startup range monitor neutron flux scram (LCO 3.3.1.1), the average power range monitor neutron flux scram (LCO 3.3.1.1), and the c<mtrol rod block instrumentation (LCO 3.3.2.1).
The safety analysis of the control rod removal error during refueling in the SSAR (Ref. B3.9.3-2) assumes the functioning of the refueling interictks and adequate SDM. Additionally, prior to fuel reload, all control rods must be fully inserted to ensure that an inadvertent criticality does not occur.
Control rod position satisfies Criterion 3 of the NRC Policy Statement.
LCO All control rods must be fully inserted during applicable refueling conditions to prevent an inadvertent criticality during refueling.
APPLICABILITY During MODE 5, loading fuel into a core cell with the control rod withdrawn may result in inadvertent criticality. Therefore, the c(mtrol rod must be inserted before loading fuel into a core cell. All control rods must be laserted before loading fuel to ensure that a fuel loading f
1 I
error does not result in loading fuel into a core cell with the control rod withdrawn.
In MODES 1,2,3, and 4, the reactor pressure vessel head is on, and no fuel loading activities are possible. Therefore, this Specification is not applicable in these MODES.
ACTIONS A
I
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With all control rods not fully inserted during the applicable conditions,-
an inadvertent criticality could occur that is not analyzed in the SSAR. All fuel loading operations must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position.
SR 3.9.3.1 During refueling, to ensure that the reactor remains suberitical, all control rods must be fully inserted prior to and during fuel loading. Periodic j
checks of the control rod position ensure this condition is maintained.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency takes into consideration the procedural controls on control rod movement during refueling as well as the redundant functions of the refueling interlocks.
10 CFR 50, Appendix A, GDC 26.
B 3.9 Refueling Operations Control Rod Position Indication BASES BACKGROUND The full-in position indication channel for each control rod provides information necessary to the refueling interlocks to prevent inadvertent criticalitics during refueling operations. During refueling, the refueling interlocks (LCO 3.9.1 and LCO 3.9.2) use the full-in position indication channel to limit the operation of the refueling equipment and the movement of the control rods. The absence of the full-in position indication channel signal for any control rod removes the all-rods-in permissive for the refueling equipment interlocks and prevents fuel loading. Also, this condition causes the refuel position one-rod-out interlock to not allow the withdrawal of any other control rod.
GDC 26 of 10 CFR 50, Appendix A, requires that one of the two required independent reactivity c(mtrol systems be capable of holding the reactor core suberitical under cold conditions (Ref. B3.9.4-1). The control rods serve as the system capable of maintaining the reactor suberitical in cold conditions.
Prevention and mitigation of prompt reactivity excursions during refueling are provided by the refueling interlocks (LCO 3.9.1 and LCO 3.9.2), the SDM (LCO 3.1.1), the startup range monitor neutron flux scram (LCO 3.3.1.1), the average power range monitor neutron flux scram (LCO 3.3.1.1), and the control rod block instrumentation (LCO 3.3.2.1).
The safety analysis for the control rod removal error during refueling (Ref. B3.9.4-2) assumes the functioning of the refueling interk>cks and adequate SDM. The full-in position indication channel is required to be OPERABLE so that the refueling interlocks can ensure that fuel cannot be loaded with any control rod withdrami and that no more than one control rod can be withdrawn at a time.
Control rod position indication satisfies Criterion 3 of the NRC Policy Statement.
LCO One of the two control rod full-in position indication channels must be OPERABLE to provide the required inputs to the refueling interlocks.
A channel is OPERABLE ifit provides correct position indication to the refueling interlock logic.
APPLICABILITY During MODE 5, the control rods must have OPERA 3LE full-in position indication channels to ensure the applicable refueling
interlocks will be OPERABLE.
In MODES 1 and 2, requirements for control rod position are specified in LCO 3.1.3, " Control Rod OPERABILITY." In MODES 3 and 4, with the reactor mode switch in the shutdown position, a control rod block (LCO 3.3.2.1) ensures all control rods are inserted, thereby preventing criticality during shutdown conditions.
ACTIONS A Note has been provided to modify the ACTIONS related to control rod position indication channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent trains, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. Ilowever, the Required Actions for control rods with inoperable position indication channels provide appropriate compensatory measures. As such, this Note has been provided, which allows separate Condition entry for each control rod with inoperable position indication channels.
A.1.1, A.1.2, A.1.3, A.2.1, and A.2.2 With required full-in position indication channels inoperable for one or more control rods, compensating actions must be taken to protect against potential reactivity excursions from fuel assembly insertions or control rod withdrawals. This may be accomplished by immediately suspending in-vessel fuel movement and control rod withdrawal, and immediately initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Actions must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted. Suspension of in-vessel fuel movements and control rod withdrawal shall not preclude moving a component to a safe position.
Alternatively, actions may be immediately initiated to fully insert the control rod (s) associated with the inoperable full-in position indicators (s) and to disarm the drive (s) to ensure that the control rod is not withdrawn. Actions must continue until all associated control rods are fully inserted and drives are disarmed.
Under these conditions, an inoperable full-in channel may be bypassed to allow refueling operations to proceed. An alternate method must be i
used to ensure the control rod is fully inserted (e.g., use the "00" notch position indication). Another option is to bypass Synchro A (which is the current position probe) and use Synchro B instead. If the readings of the two Synchros do not agree, the condition will be alarmed to the operator to Initiate bypass of Synchro A and to use Synchro B. If all fuel is removed from a core cell, the full-in position indications may be bypassed, since the control rod may be withdrawn and the position indication is not required to be OPERABLE.
SR 3.9.4.1 The full-in position indication channels provide input to the one-rod-out interlock and other refueling interlocks that require an all-rods-in permissive. The interlocks are activated when the full-in position indication for any control rod is not present, since this indicates that all rods are not fully inserted. Therefore, testing of the full-in position indication channels is performed to ensure that when a control rod is withdrawn, the full-in position indication is not present. Performing the SR each time a control rod is withdrawn is considered adequate because of the procedural controls on control rod withdrawals and the visual and audible indications available in the control room to alert the operator to control rods not fully inserted.
10 CFR 50, Appendix A, GDC 26.
)
i B 3.9 Refueling Operations Control Rod OPERABILITY-Rtfueling BASFS BACKGROUND Control rods are components of the Control Rod Drive (CRD) System, the primary reactivity control system for the reactor. In conjunction with the Reactor Protection System, the CRD System provides the means for the reliable control of reactivity changes during refueling operation. In addition, the control rods provide the capability to maintain the reactor suberitical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the CRD System.
The CRD system also includes the Fine Motion Control Rod Drives (BICRDs) and the CRD system instrumentation with which the RC&lS directly interfaces. The BICRDs can be inserted either hydraulically or electrically. In response to a scram signal, the BICRD is inserted hydraulically via the stored energy in the scram accumulators. A redundant signalis also given to insert the B!CRD electrically via its motor drive. This diversity provides a high degree of assurance of rod insertion on demand.
GDC 26 of 10 CFR 50, Appendix A, requires that one of the two required independent reactivity control systems be capable of holding the reactor core subcritical under cold conditions (Ref. B3.9.5-1). The CRD System is the system capable of maintaining the reactor i
suberitical in cold conditions.
Prevention and mitigation of prompt reactivity excursions during refueling are provided by refueling interlocks (LCO 3.9.1 and LCO 3.9.2), the SDM (LCO 3.1.1), the startup range monitor neutron flux scram (LCO 3.3.1.1), the average power range monitor neutron Dux scram (LCO 3.3.1.1), and the control rod block instrumentation (LCO 3.3.2.1).
The safety analysis for the control rod removal error during refueling (Ref. B3.9.5-2) evaluates the consequences of control rod withdrawal during refueling. A prompt reactivity excursion during refueling could potentially result in fuel failure with subsequent release of radioactive material to the environment. Control rod scram provides backup protection should a prompt reactivity excursion occur.
Control rod OPERABILITY during refueling satisfies Criterion 3 of j
the NRC Policy Statement.
LCO Each withdrawn control rod must be OPERABLE. The withdrawn.
l i
t control rod is considered OPERABLE if the scram accumulator pressure is > 107 Kg/cm 2
g (1520 psig) and the control rod is capable of being automatically inserted upon receipt of a i
scram signal. Inserted control rods have already completed their reactivity control function.
APPLICABILITY During MODE 5, withdrawn c(mtrol rods must be OPERABLE to ensure that in a scram the control rods willinsert and provide the required negative reactivity to maintain the reactor subcritical.
For MODES I and 2, control rod requirements are found in LCO 3.1.2, " Reactivity Anomalies," LCO 3.1.3, " Control ~ Rod OPERABILITY," LCO 3.1.4, " Control Rod Scram Times," and LCO 3.1.5, " Control Rod Scram Accumulators." During MODES 3 and 4, control rods are only allowed to be withdrawn under LCO 3.10.3, " Control Rod Withdrawal IIot Shutdown," and LCO 3.10.4, " Control Rod Withdrawal Cold Shutdown," in the Special Operations section. These provide adequate requirements for control rod OPERABILITY during these conditions.
ACTIONS A.1 With one or more withdrawn control rods inoperable, action must be immediately initiated to fully insert the inoperable control rod (s). Inserting the control rod (s) ensures that the shutdown and scram capabilities are not adversely affected. Actions must continue until the inoperable control rod (s) is fully inserted.
SR 3.9.5.1 and SR 3.9.5.2 During MODE 5, the OPERABILITY of control rods is primarily required to ensure that a withdrawn control rod will automatically insert if a signal requiring a reactor shutdown occurs. Because no explicit analysis exists for automatic shutdowa during refueling, the shutdown function is satisfied if the withdrawn control rod is capable of automatic insertion and the associated CRD scram accumulator pressure is 2107 Kg/cm g (1520 psig).
He 7 day Frequency takes into consideration equipment reliability, procedural controls over the scram accumulators, and control room alarms and indicating lights that indicate low accumulator charge pressures.
10 CFR 50, Appendix A, GDC 26.
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B 3.9 Refueling Operations Reactor Pressure Vessel (RPV) Water Level BASES BACKGROUND The movement of fuel assemblies or handling of control rods within the RPV requires a minimum water level of 7.0 m (23 ft) above the top of the RPV flange. Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling l
accident (Refs. B3.9.6-1 and B3.9.6-2). Sufficient iodine activity would be retained to limit offsite doses from the accident to < 25% of 10 CFR 100 limits, as provided by the guidance of Reference B3.9.6-3.
During movement of fuel assemblies or handling of control rods, the water level in the RPV and the spent fuel pool is an initial condition design parameter in the analysis of a fuel handling accident in containment postulated by Regulatory Guide 1.25 (Ref. B3.9.6-1). A minimum water level of 7.0 m (23 ft) allows a decontamination factor of 100 (Ref. B3.9.6-4) to be used in the accident analysis for lodine.
This relates to the assumption that 99% of the totallodine released from the pellet to cladding gap of all the dropped fuel assembly rods is retained by the refueling cavity water. The fuel pellet to cladding gap is assumed to contain 10% of the total fuel rod iodine Inventory (Ref. B3.9.6-1).
Analysis of the fuel handling accident inside containment is decribed in Reference B3.9.6-2. With a minimum water level of 7.0 m (23 ft) and a minimum decay time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to fuel handling, the analysis and test programs demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water, and that offsite doses are maintained within allowable limits (Ref. B3.9.6-5).
RPV water level satisfies Criterion 2 of the NRC Policy Statement.
LCO A minimum water level of 7.0 m (23 ft above the top of the RPV flange is required to ensure that the radiological consequences of a postulated fuel handling accident are within acceptable limits, as provided by the guidance of Reference B3.9.6-3.
APPLICABILITY LCO 3.9.6 is applicable when moving fuel assemblies or handling control rods (i.e., movement with other than the normal control rod drive) within the RPV. The LCO minimizes the possibility of a fuel handling accident in containment that is beyond the assumptions of the safety analysis. If irradiated fuel is not present within the RPV,
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there can be no significant radioactivity release as a result of a postulated fuel handling accident. Requirements for fuel handling accidents in the spent fuel storage pool are covered by LCO 3.7.6,
" Fuel Pool Water I2 vel."
ACTIONS A.1 If the water level is < 7.0 m (23 ft) above the top of the RPV flange, all operations involving movement of fuel assemblies and handling of control rods within the RPV shall be suspended immediately to ensure that a fuel handling accident cannot occur. The suspension of fuel movement and I
control rod handling shall not preclude completion of movement of a component to a safe position.
Verification of a minimum water level of 7.0 m (23 ft) above the top of the RPV flange ensures that the design basis for the postulated fuel handling accident analysis during refueling operations is met. Water at the required level limits the consequences of damaged fuel rods, which are postulated to result from a fuel handling accident in containment (Ref. B3.9.6-2).
The Frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is based on engineering judgment and is considered adequate in view of the large volume of water and the normal procedural controls on valve positions, which make significant unplanned level changes unlikely.
B3.9.6.1 Regulatory Guide 1.25, March 23,1972.
B3.9.6.2 ABWR SSAR, Section 15.7.4.
B3.9.6.3 NUREG-0800, Section 15.7.4.
B3.9.6.4 NUREG-0831, Supplement 6, Section [16.4.2].
B3.9.6.5 10 CFR 100.11.
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B 3.9 Refueling Operations ResidualIIcat Removal (RilR)-Iligh Water Level BASES BACKGROUND The purpose of the RIIR System in MODE 5 is to remove decay heat and sensible heat from the reactor coolant, as required by GDC 34.
Each of the three shutdown cooling loops of the RIIR System can provide the required decay heat removal. Each loop consists of one motor driven pump, a heat exchanger, and associated piping and valves. Each loop has a dedicated suction nonle from the reactor vessel. Each pump discharges the reactor coolant, after it has been -
cooled by circulation through the respective heat exchangers, to the reactor via feedwater line B for subsystem A, and via the individual RIIR inlet nonles for subsystems B and C. The RIIR heat exchangers transfer heat to the Reactor Building Cooling Water (RCW) system (LCO 3.7.2). The RIIR shutdown cooling mode is manually controlled.
In addition to the RIIR subsystems,'the volume of water above the reactor preisure vessel (RPV) flange provides a heat sink for decay heat removal.
With the unit in MODE 5, the RIIR System is not required to mitigate any esents or accidents evaluated in the safety analyses. The RIIR System is required for removing decay heat to maintain the temperature of the reactor coolant.
Although the RIIR System does not meet a specific criterion of the NRC Policy Statement, it was identified in the NRC Policy Statement as an important contributor to risk reduction. Therefore, the RIIR System is retained as a Specification.
LCO Only one RIIR shutdown cooling subsystem is required to he OPERABLE in MODE 5 with the water level 2 7.0 m (23 ft) above the RPV flange. Only one subsystem is required because the volume of water above the RPV flange provides backup decay heat removal capability.
An OPERABLE RIIR shutdown cooling subsystem consists of an RIIR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path.
Additionally, each RIIR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. Operation (either continuous or intermittent) of one subsystem can maintain and reduce
the reactor coolant temperature as required. Ilowever, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required. A Note is provided to allow a [ ] hour exception to shut down the operating subsystem every [ ] hours.
APPLICABILITY One RIIR shutdown cooling subsystem must he OPERABLE in MODE 5, with the water level 2: 7.0 m (23 ft) above the top of the RPV flange, to provide decay heat removal. RIIR System requirements in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS); Section 3.5, Emergency Core Cooling Systems (ECCS); and Section 3.6, Containment Systems. RIIR System requirements in MODE 5, with the water level < 7.0 m (23 ft) above the RPV flange, are given in LCO 3.9.8, " Residual IIent Removal (RIIR) IAw Water Level."
ACTIONS A.1 With no RHR shutdown cooling subsystem OPERABLE, an alternate method of decay heat removal must be established within I hour. In this condition, the volume of water above the RPV flange provides adequate capability to remove decay heat from the reactor core. However, the overall reliability is reduced because loss of water level could result in reduced decay heat removal capability. The I hour Completion Time is based on the decay heat removal function and the probability of a loss of the available decay heat removal capabilities. Furthermore, verification of the functional availability of these alternate method (s) must be reconfirmed every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter. This will ensure continued heat removal capability.
Alternate decay heat removal methods are available to the operators for review and preplanning in the unit's Operating Procedures. For example, in addition to the three RHR shutdown cooling loops, this may include the use of the Reactor Water Cleanup System, operating with the regenerative heat exchanger bypassed. The method used to remove the decay heat should be the most prudent choice based on unit conditions.
B.1, B.2, B.3, and B.4 If no RHR shutdown cooling subsystem is OPERABLE and an alternate method of decay heat removal is not available in accordance with Required Action A.1, actions shall be taken immediately to suspend operations involving an increase in reactor decay heat load by suspending the loading of irradiated fuel assemblies into the RPV.
Additional actions are required to minimize any potential fission product release to the environment. His includes initiating immediate action to restore the following to OPERABLE status: secondary containment, one standby gas treatment subsystem, and one secondary containment isolation valve and associated instrumentation in each associated penetration not isolated. This may be performed as an administrative check, by examining logs or other information to determine whether the components are out of service for maintenance or other reasons. It does not mean to perform the -
surveillances 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, a surveillance may need to be performed to restore the component to OPERABLE status.
Actions must continue until all required components are OPERABIE.,
C.1 and C.2 If no RHR Shutdown Cooling System is in operation, an alternate method of coolant circulation is required to be established within I hour. The Completion Time is modified such that I hour is applicable separately for each occurrence involving a loss of coolant circulation.
During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR Shutdown Cooling System), the reactor coolant temperature must be periodically monitored to ensure proper functioning of the alternate method. The once per hour Completion Time is deemed appropriate.
SR 3.9.7.1 This Surveillance demonstrates that the RHR subsystem is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability.
The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient in view of other visual and -
audible indications available to the operator for monitoring the RHR subsystem in the control room.
REFERENCES None.
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I B 3.9 Refueling Operations Residual IIcat Removal (RIIR)-Low Water Level BASES BACKGROUND The purpose of the RIIR System in MODE S is to remove decay heat and sensible heat from the reactor coolant, as required by GDC 34.
Each of the three shutdown cooling loops of the RIIR System can f
provide the required decay heat removal. Each loop consists of one i
motor driven pump, a heat exchanger, and associated piping and valves. Each loop has a dedicated suction nozzle from the reactor vessel. Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor via feedwater line B for subsystem A, and via the individual RIIR inlet nozzles for subsystems B and C. The RIIR heat exchangers transfer heat to the Reactor Building Cooling Water (RCW) system (LCO 3.7.2). The RilR shutdown cooling mode is manually controlled.
With the unit in MODE 5, the RIIR System is not required to mitigate any events or accidents evaluated in the safety analyses. The RIIR System is required for removing decay heat to maintain the temperature of the reactor coolant.
Although the RIIR System does not meet a specific criterion of the NRC Policy Statement, it was identified in the NRC Policy Statement as an important contributor to risk reduction. Therefore, the RIIR System is retained as a Specification.
LCO In MODE S with the water level < 7.0 m (23 ft) above the reactor pressure vessel (RPV) flange two RIIR shutdown cooling subsystems must be OPERABLE.
An OPERABLE RIIR shutdown cooling subsystem consists of an RIIR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path.
Additionally, each RIIR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. Operation (either continuous or intermittent) of one subsystem can maintain and reduce i
the reactor coolant temperature as required. Ilowever, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, continuous operation is required. A Note is provided to allow a [ ] hour exception to shut down the operating subsystem every [ ] hours.
o APPLICABILITY Two RIIR shutdown cooling subsystems are required to be OPERABLE in MODE 5, with the water level < 7.0 m (23 ft) above the top of the RPV flange, to provide decay heat removal. RHR System requirements in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS); Section 3.5, Emergency Core Cooling Systems (ECCS); and Section 3.6, Containment Systems.
RIIR System requirements in MODE 5, with the water level h 7.0 m (23 ft)above the RPV flange, are given in LCO 3.9.7, " Residual IIcat Removal (RIIR) IIigh Water Level."
ACTIONS A.1 With one of the two required RHR shutdown cooling subsystems inoperable, the remaining subsystem is capable of providing the required decay heat removal. However, the overall reliability is reduced. Therefore an alternate method of decay heat removal must be provided (such as the third RHR shutdown cooling subsystem). With both required RHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability. This re-establishes backup decay heat removal capabilities, similar to the requirements of the LCO. The I hour Completion Time is based on the decay heat removal function and the probability of a loss of the available decay heat removal capabilities. Furthermore, verincation of the functional availability of these alternate method (s) must be reconfirmed every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.
This will ensure continued heat removal capability.
Alternate decay heat removal methods are available to the operators for review and preplanning in the unit's Operating Procedures. For example, in addition to the third RHR shutdown cooling loop, this may include the use of the Reactor Water Cleanup System, operating with the regenerative heat exchanger bypassed. The method used to remove decay heat should be the most prudent choice based on unit conditions.
B.1, B.2, B.3, and B.4 If no RHR shutdown cooling subsystem is in operation, an alternate method of coolant circulation is required to be established within I hour.
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The Completion Time is modified such that the I hour is applicable separately for each occurrence involving a loss of coolant circulation.
During the period when the reactor coolant is being circulated by an
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alternate method (other than by the required RHR Shutdown Cooling System), the reactor coolant temperature and level must be periodically i
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monitored to ensure proper function of the alternate method. The once per hour Completion Time is deemed appropriate.
j If at least one RHR subsystem is not restored to OPERABLE status i
immediately, additional actions are required to minimize any potential fission product release to the environment. 'Iliis includes initiating immediate action to restore the following to OPERABLE status: secondary containment, one standby gas treatment subsystem, and one secondary containment isolation valve and associated instrumentation in each associated penetration not isolated. This may be performed as an administrative check, by examining logi er other information to determine whether the components are out of service for maintenance or other reasons. It is not necessary to perform the surveillances 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, the surveillance may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE.
SR 3.9.8.I This Surveillance demonstrates that one RHR subsystem is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability.
The Frequency of 12 hourr ', st#icient in view of other visual and audible indications available a t*m operator for monitoring the RHR subsystem in the control room.
REFERENCES None.
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Special Operations Inservice Leak and Hydrostatic Testing Operation BASES The purpose of this Special Operations LC0 is to allow certain reactor coolant pressure tests to be performed in MODE 4 when the metallurgical characteristics of the reactor pressure vessel (RPV) require the pressure testing at temperatures > 93*C (200*F) (normally corresponding to MODE 3).
Inservice testing and system leakage pressure tests required by Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Ref. B3.10.1-1) are performed prior to the reactor going critical after a refueling outage. Reactor Internal l
pump operation and a water solid RPV (except for an air bubble for pressure control) are used to achieve the necessary temperatures and pressures required for these tests. The minimum temperatures (at the required pressures) allowed for these tests are determined from the RPV pressure and temperature (P/T) limits required by LCO 3.4.9, " Reactor Coolant System (RCS) Pressure and Temperature (P/T) Limits." These limits are conservatively based on the fracture toughness of the reactor vessel, taking into account anticipated vessel neutron fluence.
With increased reactor vessel fluence over time, the minimum allowable vessel temperature increases at a given pressure. Periodic updates to the RPV P/T limit curves are performed as necessary, based on the results of analyses of irradiated surveillance specimens removed from the vessel. Hydrostatic and leak testing will eventually be required with minimum reactor coolant temperatures > 93*C (200*F).
The hydrostatic test, generally performed every ten
]
years, requires increasing pressure to 110% of operating pressure 72.1 Kg/cm 1
i (1025 psig) or 79.3 Kg/cm g (1128 psig), and because of the expected increase in reactor vessel fluence,
o the minimum allowable vessel temperature according to LCO 3.4.9 is increased i
as shown in the PTLR (Ref. B3.10.3). This increase to 110% of operating pressure does not exceed the Safety Limit of 1375 psig.
Allowing the reactor to be considered in MODE 4 during hydrostatic or leak testing, when the reactor coolant temperature is >93*C (200*F), effectively provides an exception to MODE 3 requirements, including OPERABILITY of primary containment and the full complement of redundant Emergency Core Cooling Systems (ECCS). Since the hydrostatic or leak tests are performed water solid, at low decay heat values, and near MODE 4 conditions, the stored energy in the reactor core will be very low. Under these conditions, the potential for failed fuel and a subsequent increase in coolant activity above the limits
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of LC0 3.4.6, " Reactor Coolant System (RCS) Specific Activity," are minimized. In addition, the secondary containment will be OPERABLE, in accordance with this Specic1 Operations LCO, and will be capable of handling any airborne radioactivity or steam leaks that could occur during the performance of hydrostatic or leak testing. The consequences of a steam leak under pressure testing conditions, with secondary containment OPERABLE, will be conservatively bounded by the consequences of the accident postulated main steam line break outside o' primary containment analysis described in Reference B3.10.1-2. Therefore, requiring the secondary containment to be OPERABLE will conservatively ensure that any potential airborne radiation from steam leaks will be filtered through the Standby Gas Treatment System, thereby limiting radiation releases to the envir'onment.
In the event of a large primary system leak, the reactor vessel would rapidly depressurize, allowing the low pressure core cooling systems to operate. The capability of the low pressure flooder subsystems, as required in MODE 4 by LCO 3.5.2, "ECCS-Shutdown," would be more than adequate to keep the core flooded under this low decay heat load condition. Small system leaks would be detected by leakage inspections before significant inventory loss occurred.
For the purposes of this test, the protection provided by normally required MODE 4 applicable LCOs, in addition to
the secondary containment requirements required to be met by this Special Operations LCO, will ensure acceptable consequences during normal hydrostatic test conditions l
and during postulated accident conditions.
As described in LC0 3.0.7, compliance with Special Operations LCOs is optional, and therefore, no criteria of the NRC Policy Statement apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.
1 As described in LCO 3.0.7, compliance with this Special Operations LCO is optional. Operation at reactor coolant temperatures >93*C (200*F), can be in accordance with.
Table 1.1-1 for MODE 3 operation without meeting this Special Operations LC0 or its ACTIONS. This option may be required due to P/T limits, however, which require testing at temperatures > 93*C (200*F), while the ASME inservice test itself requires the safety / relief valves to be gagged, preventing their OPERABILITY.
If it is desired to perform these tests while complying with this Special Operations LCO, then the MODE 4 applicable LCOs and specified MODE 3 LCOs must be met.
This Special Operations LCO allows changing Table 1.1-1 temperature limits for MODE 4 to "NA" and suspending the requirements of LC0 3.4.8, " Residual Heat Removal (RHR)
Shutdown Cooling System-Cold Shutdown." The additional requirements for secondary containment LCOs to be met will provide sufficient protection for operations at reactor coolant temperatures > ?"C (200*F) for the purposes of performing either an inservice leak or i
hydrostatic test.
This LC0 allows primary containment to be open for frequent unobstructed access to perform inspections, and i
for outage activities on various systems to continue I
consistent with the MODE 4 applicable requirements that.
are in effect immediately prior to and immediately after this operation.
The MODE 4 requirements may only be modified for the
]
performance of inservice leak or hydrostatic tests so that these operations can be considered as in MODE 4,
]
even though the reactor coolant temperature is > 93*C l
~
=
+
\\
k (200*F). The additional requirement for secondary containment OPERABILITY according to the imposed MODE 3 requirements provides conservatism in the response of the unit to any event that may occur. Operations in all other MODES are unaffected by this LCO.
A Note has been provided to modify the ACTIONS related to inservice leak and hydrostatic testing operation.
Section 1.3, Completion Times, specifies once a Condition has been entered, subsequent trains, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits, will not result in separate entry into the Condition.
Section-1.3 also specifies Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for each requirement of the LCO not met provide appropriate compensatory measures for separate requirements that are not met. As such, a Note has been provided that allows separate Condition entry for each requirement of the LCO.
A.1 If an LCO specified in LCO 3.10.1 is not met, the ACTIONS applicable to the stated requirements shall be entered immediately and complied with.
Required Aczion A.1 has been modified by a Note that clarifies the intent of another LCO's Required Action to be in MODE 4 includes reducing the average reactor coolant temperature to s 93*C (200*F).
A.2.1 and A.2.2 Required Actions A.2.1 and A.2.2 are alternate Required Actions that can be taken instead of Required Action A.1 to restore compliance with the normal MODE 4 requirements, and thereby exit this Special Operations LCO's Applicability. Activities that could further increase reactor coolant temperature or pressure are suspended immediately, in accordance with Required Action A.2.1, and the reactor coolant temperature is reduced to establish normal MODE 4 requirements. The allowed Completion Time of l
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for Required Action A.2.2 is based on engineeringjudgment and provides sufficient time to reduce the average reactor coolant temperature from the highest expected value to s 93*C (200*F) with normal cooldown procedures. The Completion Time is also consistent with the time provided in LCO 3.0.3 for reaching MODE 4 from MODE 3.
SR 3.10.1.1 The LCOs made applicable are required to have their Surveillances met to
establish that this LCO is being met. A r.'scussion of the applicable SRs is provided in their respective Bases.
B3.10.1 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section XI.
B3.10.2 ABWR SSAR, Section 15.1.
B3.10.3 Appendix G of 10 CFR 50.
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5 b
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l 1
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l 1
a
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Special Operations Single Control Rod Withdrawal-IIot Shutdown BASES The purpose of this MODE 3 Special Operations LCO is to permit the withdrawal of a single control rod, or control rod pair, for testing while in hot shutdown, by impasing certain restrictions. In MODE 3, the reactor mode switch is in the shutdown position, and all control rods are inserted and blocked from withdrawal. Many systems and functions are not required in these conditions, due to other installed interk>cks that are actuated when the reactor mode switch is in the shutdown position. Ilowever, circumstances will arise while in j
MODE 3 that present the need to withdraw a single control rod, or control rod pair, for various tests (e.g., friction tests, scram timing, j
and coupling integrity checks). These single control rod, or control rod pair, withdrawals are normally accomplished by selecting the refuel
-i position for the reactor mode switch. A control rod pair (those i
associated by a shared CRD hydraulic control unit) may be withdrawn by utilizing the Rod Test Switch which " gangs" the two rods together for rod position and control purposes. This Special Operations LCO
{
provides the appropriate additional controls to allow a single control i
rod, or control rod pair, withdrawal in MODE 3.
With the reactor mode switch in the refuel position, the analyses for j
control rod withdrawal during refueling are applicable and, provided l
the assumptions of these analyses are satisfied in MODE 3, these '
analyses will bound the consequences of an accident. Explicit safety analyses in the SSAR (Ref. B3.10.3-1) demonstrate that the functioning of the refueling interh>cks and adequate SDM will preclude unacceptable reactivity excursions.
Refueling interlocks restrict the movement of control rods to reinforce operational procedures that prevent the reactor from becoming critical. These interlocks prevent the withdrawal of more than one control rod. Under these conditions, since only one control rod can he withdrawn, the core will always be shut down even with the highest j
worth control rod pair withdrawn if adequate SDM exists.
.j COPY MISSING - INSERT Q ???
The control rod scram function provides backup protection to normal I
refueling procedures and the refueling interlocks, which prevent inadvertent criticalities during refueling.
-i
Alternate backup protection can be obtained by ensuring that a five by five array of c<mtrol rods, centered on the withdraim ccmtrol rod, are inserted and incapable of withdrawal.
As described in LCO 3.0.7, compliance with Special Operations LCOs j
is optional, and therefore, no criteria of the NRC Policy Statement apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A
)
discussion of the criteria satisfied for the other LCOs is provided in j
their respective Bases.
As described in LCO 3.0.7, compliance with this Special Operations LCO is optional. Operation in MODE 3 with the reactor mode switch in the refuel position can be performed in accordance with other Special Operations LCOs (i.e., LCO 3.10.2, " Reactor Mode Switch Interlock Testing," and LCO 3.10.4, " Control Rod Withdrawal-Cold Shutdown") without meeting this Special Operations LCO or its ACTIONS. Ilowever, if a single control rod, or control rod pair, withdrawal is desired in MODE 3, controls consistent with those required during refueling must be implemented and this Special Operations LCO applied. The refueling interk>cks of LCO 3.9.2,
" Refuel Position One Rod Out Interlock," required by this Special Operations LCO, will ensure that only one control rod, or control rod pair, can be withdrawn.
To hack up the refueling interlocks (LCO 3.9.2), the ability to scram the withdrawn control rod (s) in the event of an inadvertent criticality is provided by this Special Operations LCO's requirements in item d.1. Alternately, provided a sufficient number of control rods in the vicinity of the withdrawn control rod (s) are known to be inserted and incapable of withdrawal, the possibility of criticality on withdrawal of these control rod (s) is sufficiently precluded, so as not to require the scram capability of the withdrawn control rod (s).
Control rod withdrawals are adequately controlled in MODES 1,2, and 5 by existing LCOs. In MODES 3 and 4, control rod withdrawal is only allowed if performed in accordance with this Special Operations LCO or Special Operations LCO 3.10.4, and if limited to one control rod, or control rod pair. This allowance is only provided with the rouctor mode switch in the refuel position. For these conditions, the one-rod-out interlock (LCO 3.9.2), control rod position indication (LCO 3.9.4, " Control Rod Position Indication") full
^
insertion requirements for all other control rods and scram functions (LCO 3.3.1.1, " Reaction Protection System (RPS) Instrumentation,"
and LCO 3.9.5, " Control Rod OPERAllILITY-Refueling"), or the added administrative control in Item d.2 of this Special Operations LCO, minimizes potential reactivity excursions.
A Note has been provided to modify the ACTIONS related to a single or dual control rod withdrawal while in MODE 3. Section 1.3, Completion Times, specifies once a Condition has been entered, subsequent trains, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. Ilowever, the Required Actions for each requirement of the LCO not met provide appropriate compensatory measures for separate requirements that are not met. As such, a Note has been provided that allows separate Condition entry for each requirement of the LCO.
A.1 If one or more of the requirements specified in this Special Operations LCO are not met, the ACTIONS applicable to the stated requirements of the affected LCOs are immediately entered as directed by Required Action A.I. This Required Action has been modified by a Note that clarifies the intent of any other LCO's Required Actions, in accordance with the other applicable LCOs, to insert all control rods and to also require exiting this Special Operations Applicability LCO by returning the reactor mode switch to the shutdown position. A second Note has been added, which clarifies that this Required Action is only applicable if the requirements not met are for an affected LCO.
A.2.1 and A.2.2 Required Actions A.2.1 and A.2.2 are alternative Required Actions that can be taken instead of Required Action A.1 to restore compliance with the normal MODE 3 requirements, thereby exiting this Special Operations LCO's Applicability. Actions must be initiated immediately to insert all insertable control rods. Actions must continue until all such control rods are fully inserted. Placing the reactor mode switch in the shutdown position will ensure that all inserted rods remain inserted and restore operation in accordance with Table 1.1-L The allowed Completion Time of I hour to place the resctor mode switch in the shutdown position
provides sufficient time to normally insert the control rods.
SR 3.10.3.1, SR 3.10.3.2, and SR 3.10.3.3 The other LCOs made applicable in this Special Operations LCO are required to have their Surveillances met to establish that this Special Operations LCO is being met. If the local array of control rods is inserted and disarmed while the scram function for the withdrawn rod (s) is not available, periodic verification in accordance with SR 3.10.3.2 is required to preclude the possibility of criticality. SR 3.10.3.2 has been modified by a Note, which clarifies that this SR is not required to be met if I
SR 3.10.3.1 is satisfied for LCO 3.10.3.d.1 requirements, since l
SR 3.10.3.2 demonstrates that the alternative LCO 3.10.3.d.2 requirements are satisfied. Also, SR 3.10.3.3 verifies that all control rods other than the control rod (s) being withdrawn are fully inserted. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is acceptable because of the administrative controls on control rod withdrawals, the protection afforded by the LCOs involved, and hardware interlocks that preclude additional control rod withdrawals.
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Special Operations Single Control Rod Withdrawal-Cold Shutdown BASES The purpose of this 510DE 4 Special Operations LCO is to permit the withdrawal of a single control rod, or ccmtrol rod pair, for testing or maintenance, while in cold shutdown, by imposing certain restrictions.
In 510DE 4, the reactor mode switch is in the shutdown position, and all control rods are inserted and blocked from withdrawal. Afany systems and functions are not required in these conditions, due to the installed interlocks associated with the reactor mode switch in the shutdown position. Circumstances will arise while in A10DE 4, however, that present the need to withdraw a single control rod, or control rod pair, for various tests (e.g., friction tests, scram time testing, and coupling integrity checks). Certain situations may also require the removal of the associated control rod drives (CRD). These single or dual c<mtrol rod withdrawals and possible subsequent removals are normally accomplished by selecting the refuel position for the reactor mode switch. A control rod pair (those associated by a-single CRD hydraulic control unit) may be withdrawn by utilizing the Rod Test Switch, which " gangs" the two rods together for rod position and control purposes.
With the reactor mode switch in the refuel position, the analyses for control rod withdrawal during refueling are applicable and, provided the assumptions of these analyses are satisfied in h10DE 4, these analyses will bound the consequences of an accident. Explicit safety analyses in the SSAR (Ref. B3.10.4-1) demonstrate that the functioning of the refueling interlocks and adequate SDh1 will preclude unacceptable reactivity excursions.
Refueling interlocks restrict the movement of control rods to reinforce operational procedures that prevent the reactor from beco:ning critical. Those interlocks prevent the withdrawal of more than one control rod, or control rod pair. Under these conditions, the core will always he shut down even with the highest worth control rod pair withdrawn if adequate SDh1 exists.
TEXT 511SSING -INSERT Q ????
The control rod scram function provides backup protection to normal refueling procedures and the refueling interlocks, which prevent inadvertent criticalities during refueling. Alternate backup protection
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t i
P can he obtained by ensuring that a five by five array of control rods, e
centered on the withdrawn control rod (s), are inserted and incapable of withdrawal. This alternate backup protection is required when removing the CRD hecause this removal renders the withdrawn j
control rod (s) incapable of being scrammed.
As described in LCO 3.0.7, compliance with Special Operations 1,COs '
i i
is optional, and therefore, no criteria of the NRC Policy Statement apply. Special Operations LCOs provide flexibility to perform certain j
operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in j
their respective Bases.
As described in LCO 3.0.7, compliance with this Special Operations
.l LCO is optional. Operation in MODE 4 with the reactor mode switch
[
in the refuel position can be performed in accordance with other LCOs (i.e., Special Operations LCO 3.10.2, " Reactor Mode Switch Interlock Testing," and LCO 3.10.3, " Control Rod Withdrawal-Hot Shutdown") without meeting this Special Operations LCO or its l
ACTIONS. If a single control rod, or control rod pair, withd.awal is desired in MODE 4, ccmtrols consistent with those required during refueling must he implemented and this Special Operations LCO applied.
l The refueling interlocks of LCO 3.9.2, " Refuel Position One-Rod-Out Interlock," required by this Special Operations LCO 3.10.4 will ensure that only one control rod, or control rod pair, can he withdrawn. At the time CRD removal begins, the disconnection of the position indication prohe will cause LCO 3.9.4, " Control Rod Position -
Indication," and therefore, LCO 3.9.2 to fail to be met.-At this time, a control rod withdrawal block will be inserted to ensure tliat no additional control rods can be withdrawn and that compliance with this Special Operations LCO is maintained.
To hack up the refueling interlocks (LCO 3.9.2) or the control rod withdrawal block, the ability to scram the withdrawn control rod (s) in I
the event of an inadvertent criticality is provided by this Special Operations LCO's requirements in Item c.1. Alternatively, when the scram function is not OPERABLE, or the CRD is to be removed, a sufficient number of rods in the vicinity of the withdrawn control rod (s) are required to he inserted and made incapable of withdrawal.
This precludes the possibility of criticality upon withdrawal of this l
4
Control rod withdrawals are adequately controlled in MODES 1,2, and 5 by existing LCOs In MODES 3 and 4, control rod withdrawal is only allowed if performed in accordance with Special Operations LCO 3.10.3, or this Special Operations LCO, and if limited to one control rod, or control rod pair. This allowance is only provided with the reactor mode switch in the refuel position.
During these conditions, the full insertion requirements for all other control rods, the one-rod-out interlock (LCO 3.9.2), ccmtrol rod position indication (LCO 3.9.4), and scram functions (LCO 3.3.1.1,
" Reactor Protection System (RPS) Instrumentation," and LCO 3.9.5,
" Control Rod OPERABILITY-Refueling"), LCO 3.3.1.2, " Reactor Protection System (RPS) Trip Actuation," or the added administrative controls in Item b.2 and Item c.2 of this Special Operations LCO, provide mitigation of potential reactivity excursions.
A Note has been provided to modify the ACTIONS related to a single or dual control rod withdrawal while in MODE 3. Section 1.3, Completion Times, specifies once a Condition has been entered, subsequent trains, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Secthn 1.3 also specifies Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. Ilowever, the Required Actions for each requirement of the LCO not met provide appropriate compensatory measures for separate requirements that are not met. As such, a Note has been provided that allows separate Condition entry for each requirement of the LCO.
A.1, A.2.1, and A.2.2 If one or more of the requirements of this Special Operations LCO are not met with the affected control rod insertable, these Required Actions restore operation consistent with normal MODE 4 conditions (i.e., all rods inserted) or with the exceptions allowed in this Special Operations LCO.
Required Action A.1 has been modified by a Note that clarifies that the intent of any other LCO Required Actions, in accordance with the other applicable LCOs, to insert all control rods includes exiting this Special Operations Applicability LCO by returning the reactor mode switch to the shutdown position. A second Note has been added to Required Action A.1 to clarify that this Required Action is only applicable if the requirements
not met are for an affected LCO.
Required Actions A.2.1 and A.2.2 are specified, based on the assumption of the control rod (s) being withdrawn. If a control rod is still insertable, actions must be immediately initiated to fully insert all insertable control rods and within I hour place the reactor mode switch in the shutdown position. Action must continue until all such control rods are fully inserted. The allowed Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for placing the reactor mode switch in the shutdown position provides sufficient time to normally insert the control rods.
B.1, B.2.1, and B.2.2 If one or more of the requirements of this Special Operations LCO are not met with the affected control rod (s) not insertable, withdrawal of the control rod and removal of the associated CRD must immediately be suspended. If the CRD has been removed, such that the control rod is not insertable, the Required Actions require the most expeditious action be taken to either initiate action to restore the CRD and insert its control rod, or restore compliance with this Special Operations LCO.
SR 3.10.4.1, SR 3.10.4.2, SR 3.10.4.3, and SR 3.10.4.4 The other LCOs made applicable by this Special Operations LCO are required to have their associated Surveillances met to establish that this Special Operations LCO is being met. If the local array of control rods is inserted and disarmed while the scram function for the withdrawn rod is not available, periodic verification is required to ensure that the possibility of criticality remains precluded. Also, all the control rods are verified to be inserted, as well as the control rod withdrawal block. Verification that all the other control rods are fully inserted is required to meet the SDM requirements. Verification that a control rod withdrawal block has been inserted ensures that no other control rods can be inadvertently withdrawn under conditions when position indication instrumentation is inoperable for the affected control rod. The 74 hour8.564815e-4 days <br />0.0206 hours <br />1.223545e-4 weeks <br />2.8157e-5 months <br /> Frequency is acceptable because of the administrative controb antrol rod withdrawals, the protection afforded by the LCOs involved, and hardware interlocks to preclude an additional control rod withdrawal.
SR 3.10.4.2 and SR 3.10.4.4 have been modified by Notes, which clarify that these SRs are not required to be met if the alternative requirements demonstrated by SR 3.10.4.1 are satisfied.
i
Special Operations Control Rod Drive (CRD) Removal-Refueling BASES The purpose of this MODE 5 Special Operations LCO is to permit the removal of a CRD during refueling operations by imposing certain administrative controls. Refueling interlocks restrict the movement of control rods and the operation of the refueling equipment to reinforce operational procedures that prevent the reactor from becoming critical during refueling operations. During refueling operations, no more than one control rod, or control rod pair, is permitted to be withdrawn from a core cell containing one or more fuel assemblies.
The refueling interlocks use the " full in" position indicators to determine the position of all control rods. If the " full in" position I
signal is not present for every control rod, then the all rods in permissive for the refueling equipment interlocks is not present and fuel loading is prevented. Also, the refuel position one-rod-out -
I interlock will not allow the withdrawal of a second control rod. A control rod drive pair (those associated by a shared CRD hydraulic control unit) may be removed under the control of the one-rod-out interlock by utilizing the rod test switch. This switch allows the CRD pair to be treated as one CRD for purposes of the one-rod-out interlock.
The control rod scram function provides backup protection to normai refueling procedures as do the refueling interlocks described above, which prevent inadvertent criticalities during refueling. The requirement for this function to be OPERABLE precludes the possibility of removing the CRD once a control rod is withdrawn from a core cell containing one or more fuel assemblies. This Special Operations LCO provides controls sufficient to ensure the possibility of an inadvertent criticality is precluded, while allowing a single CRD, or control rod pair, to be removed from core cell (s) containing one or f
more fuel assemblies. The removal of the CRD involve disconnecting the position indication probe, which causes noncompliance with LCO 3.9.4, " Control Rod Position Indication," and, therefore, LCO 3.9.I, " Refueling Equipment Interlocks," and LCO 3.9.2,
" Refueling Position One-Rod-Out Interlock." The CRD removal also requires isolation of the CRD from the CRD Ilydraulic System, thereby causing inoperability of the control rod (LCO 3.9.5, " Control
Rod Ol'ERABILITY-Refueling").
'l With the reactor mode switch in the refuel position, the analyses for control rod withdrawal during refueling are applicable and, provided the assumptions of these analyses are satisfied, these analyses will bound the consequences of accidents. Explicit safety analyses in the SSAR (Ref. B3.10.5-1) demonstrate that the proper operation of the refueling interlocks and adequate SD51 will preclude unacceptable reactivity excursions.
TEXT 511SSING - INSERT Q ?????
Refueling interlocks restrict the movement of control rods and the operation of the refueling equipment to reinforce operational 3
procedures that prevent the reactor from becoming critical. These interlocks prevent the withdrawal of more than one control rod, or control rod pair. Under these ctmditions, the core will always he shut down even with the highest worth control rod pair withdrawn if adequate SD51 exists. By requiring all other control rods to be inserted and a control rod withdrawal block initiated, the function of the inoperable one-rod-out interlock (LCO 3.9.2) is adequately maintained. This Spec!al Operations LCO requirement to suspend all CORE ALTERATIONS adequately compensates for the inoperable all rods in permissive for the refueling equipment interlocks (LCO 3.9.1).
The control rod scram function provides backup protection to normal refueling procedures and the refueling interlocks, which prevent inadvertent criticalities during refueling. Since the scram function and refueling interlocks may be suspended, alternate backup protection required by this Special Operations LCO is obtained by ensuring that a five by five array of control rods, centered on the withdrawn control rod, are inserted and are incapable of being withdrawn (by insertion of a control rod block).
As described in LCO 3.0.7, compliance with Special Operations LCOs is optional, and therefore, no criteria of the NRC Policy Statement apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.
As described in LCO 3.0.7, compliance with this Special Operations LCO is optional. Operation in 510DE 5 with any of the following LCOs-LCO 3.3.1.1, " Reactor Protection System (RPS)
Instrumentation," LCO 3.3.8.2, " Reactor Protection System (RPS)
Electric Power hionitoring," LCO 3.9.1, LCO 3.9.2, LCO 3.9.4, or LCO 3.9.5-not met can he performed in accordance with the Required Actions of these LCOs without meeting this Special Operations LCO or its ACTIONS. Ilowever, if a single CRD or CRD drive pair removal from a core cell containing one or more fuel assemblies is desired in h10DE 5, controls consistent with those required by LCO 3.3.1.1, LCO 3.3.8.2, LCO 3.9.1, LCO 3.9.2, LCO 3.9.4, and LCO 3.9.5 must be implemented and this Special Operations LCO applied.
By requiring all other control rods to be inserted and a control rod withdrawal block initiated, the function of the inoperable one-rod-out interlock (LCO 3.9.2) is adequately maintained. This Special Operations LCO requirement to suspend all CORE ALTERATIONS adequately compensates for the inoperable all rods in permissive for the refueling equipment interlocks (LCO 3.9.1). Ensuring that the five by five array of control rods, centered on each withdrawn control rod, are inserted and incapable of withdrawal adequately satisfies the backup protection that LCO 3.3.1.1 and LCO 3.9.2 would have otherwise provided.
The exception granted in this Special Operations LCO to assume that the withdrawn c(mtrol rod, or control rod pair, is the highest worth control rod pair to satisfy LCO 3.1.1, "SIIUTDOWN AIARGIN (SD51)," and the inability to withdraw another control rod during this operation without additional SD51 demonstrations, is conservative (i.e., the withdrawn control rod pair may not he the highest worth control rod pair).
Operation in h10DE 5 is controlled by existing LCOs. The allowance to comply with this Special Operations LCO in lieu of the ACTIONS of LCO 3.3.1.1, LCO 3.3.8.2, LCO 3.9.1, LCO 3.9.2, LCO 3.9.4, and LCO 3.9.5 is appropriately controlled with the additional administrative controls required by this Special Operations LCO, which reduces the potential for reactivity excursions.
A.1, A.2.1, and A.2.2 j
If one or more of the requirements of this Special Operations LCO are not met, the immediate implementation of these Required Actions restores cperation consistent with the normal requirements for failure to meet LCO 3.3.1.1. LCO 3.9.1, LCO 3.9.2, LCO 3.9.4, and LCO 3.9.5 (i.e.,
all control rods inserted) or with the allowances of this Special Operations
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LCO. The Completion Times for Required Action A.1, Required Action A.2.1, and Required Action A.2.2 are intended to require these Required Actions be implemented in a very short time and carried through in an expeditious manner to either initiate action to restore the CRD and insert its control rod, or initiate action to restore compliance with this Special Operations LCO. Actions must continue until either Required Action A.2.1 or Required Action A.2.2 is satisfied.
SR 3.10.5.1, SR 3.10.5.2, SR 3.10.5.3, SR 3.10.5.4, and SR 3.10.5.5 Verification that all the control rods, other than the control rod withdrawn for the removal of the associated CRD, are fully inserted is required to ensure the SDM is within limits. Verification that the local five by five array of control rods other than the control rod withdrawn for the removal of the associated CRD, is inserted and disarmed, while the scram function for the withdrawn rod is not available, is required to ensure that the possibility of criticality remains precluded. Verification that a control rod withdrawal block has been inserted ensures that no other control rods can be inadvertently withdrawn under conditions when position indication instrumentation is inoperable for the withdrawn control rod. The Surveillance for LCO 3.1.1, which is made applicable by this Special Operations LCO, is required in order to establish that this Special Operations LCO is being met. Verification that no other CORE ALTERATIONS are being made is required to ensure the assumptions of the safety analysis are satisfied.
Periodic verification of the administrative controls established by this Special Operations LCO is prudent to preclude the possibility of an inadvertent criticality. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is acceptable, given the administrative controls on control rod removal and hardware interlocks to block an additional control rod withdrawal.
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Special Operations A!ultiple Control Rod Withdrawal-Refueling BASES The purpose of this MODE 5 Special Operations LCO is to permit multiple control rod withdrawal during refueling by imposing certain administrative controls.
Refueling interlocks restrict the movement of control rods and the operation of the refueling equipment to reinforce operational procedures that prevent the reactor from becoming critical during refueling operations. During refueling operations, no more than one control rod, or control rod pair, is permitted to be withdrawn from a core cell containing one or more fuel assemblies. When all four fuel assemblies are removed from a cell, the control rods may be withdrawn with no restrictions. Any number of control rods may be withdrawn and removed from the reactor vesselif their cells contain no fuel.
The refueling interlocks use the " full in" position indicators to determine the position of all control rods. If the " full in" position signal is not present for every control rod, then the all rods in permissive for the refueling equipment interlocks is not present and fuel loading is prevented. Also, the refuel position one-rod-out interlock will not allow the withdrawal of additional control rods.
To allow more than one control rod pair to be withdrawn during refueling, these interlocks must be defeated. This Special Operations LCO establishes the necessary administrative controls to allow bypass of the "fullin" position indicators.
Explicit safety analyses in the SSAR (Ref. B3.10.6-1) demonstrate that the functioning of the refueling interlocks and adequate SDM will prevent unacceptable reactivity excursions during refueling. To allow multiple control rod withdrawals (e.g. more than one control rod or control rod pair), control rod removals, associated control rod drive (CRD) removal, or any combination of these, the " full in" position indication is allowed to be bypassed for each withdrawn control rod if all fuel has been removed from the cell. With no fuel assemblies in the core cell, the associated control rod has no reactivity control function and is not required to remain inserted. Prior to reloading fuel into the ce!!, however, the associated control rod must be inserted to ensure that an inadvertent criticality does not occur, as evaluated in the i
Reference B3.10.6-1 analysis.
As described in LCO 3.0.7, compliance with Special Operations LCOs is optional, and therefore, no criteria of the NRC Policy Statement j
apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in j
their respective Bases.
As described in LCO 3.0.7, compliance with this Special Operations LCO is optional. Operation in MODE 5 with LCO 3.9.3, " Control Rod Position," LCO 3.9.4, " Control Rod Position Indication," or LCO 3.9.5, " Control Rod OPERABILITY-Refueling," not met, can be performed in accordance with the Required Actions of these LCOs without meeting this Special Operations LCO or its ACTIONS. If multiple control rod withdrawal or removal, or CRD removal is desired, all four fuel assemblics are required to be removed from the associated cells. Prior to entering this LCO, any fuel remaining in a cell whose control rod was previously removed under the provisions of another LCO must be removed.
When loading fuel into the core with multiple control rods withdrawn, special spiral reload sequences are used to ensure that reactivity additions are minimized. Otherwise, all control rods must be fully I
inserted before loading fuel.
Operation in MODE 5 is controlled by existing LCOs. The exceptions from other LCO requirements (e.g., the ACTIONS of LCO 3.9.3, LCO 3.9.4 or LCO 3.9.5) allowed by this Special Operations LCO are appropriately controlled by requiring all fuel to be removed from cells whose " full in" indicators are allowed to be bypassed.
A.1, A.2.1, and A.2.2 If one or more of the requirements of this Special Operations LCO are not met, the immediate implementation of these Required Actions restores operation consistent with the normal requirements for refueling (i.e., all control rods inserted in core cells containing one or more fuel assemblies) or with the exceptions granted by this Special Operations LCO. The Completion Times for Required Action A.1, Required Action A.2.1, and Required Action A.2.2 are intended to require that these Required Actions be implemented in a very short time and carried through in an expeditious manner to either initiate action to restore the affected CRDs and insert their control rods, or initiate action to restore compliance with this Special Operations LCO.
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SR 3.10.6.1, SR 3.10.6.2, and SR 3.10.6.3 Periodic verification of the administrative controls established by this Special Operations LCO is prudent to preclude the possibility of an inadvertent criticality. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is acceptable, given the -
administrative controls on fuel assembly and control rod removal, and takes into account other indications of control rod status available in the control room.
B3.10.6.1 ABWR SSAR, Section 15.4.1 l
Special Operations Control Rod Testing-Operating BASES The purpose of this Special Operations LCO is to permit control rod testing, while in MODES 1 and 2, by imposing certain administrative controls. Control rod patterns during startup conditions are controlled by the operator and the rod worth minimizer (RWM) (LCO 3.3.2.1,
" Control Rod lilock Instrumentation"), such that only the specified control rod sequences and relative positions required by LCO 3.1.6,
" Rod Pattern Control," are allowed over the operating range from all control rods inserted to the low power setpoint (LPSP) of the RWM.
The sequences effectively limit the potential amount and rate of reactivity increase that could occur during a rod withdrawal error (RWE). During these conditions, control rod testing is sometimes required that may result in control rod patterns not in compliance with the prescribed sequences of LCO 3.1.6. These tests may include SDM demonstrations, control rod scram time testing, control rod friction testing, and testing performed during the Startup Test Program. This Special Operations LCO provides the necessary exceptions to the requirements of LCO 3.1.6 and provides additional administrative controls to allow the deviations in such tests from the prescribed sequences in LCO 3.1.6.
The analytical methods and assumptions used in evaluating the RWE are summarized in References B3.10.7-1 and B3.10.7-2. RWE analyses assume the reactor operator follows prescribed withdrawal sequences.
These sequences define the potential initial conditions for the RWE analyses. The RWE provides backup to operator control of the withdrawal sequences to ensure that the initial conditions of the RWE analyses are not siolated. For special sequences developed for control rod testing, the initial control rod patterns assumed in the safety analyses of References B3.10.7-1, B3.10.7-2,3, and 4 may not be preserved. Therefore, special RWE analyses are required to demonstrate that these special sequences will not result in unacceptable consequences, should a RWE occur during the testing.
These analyses, performed in accordance with an NRC approved methodology, are dependent on the specific test being performed.
As described in LCO 3.0.7, compliance with Special Operations LCOs is optional, and therefore, no criteria of the NRC Policy Statement i
4 apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.
As described in LCO 3.0.7, compliance with this Special Operations LCO is optional. Control rod testing may be performed in compliance with the prescribed sequences of LCO 3.1.6, and during these tests, no exceptions to the requirements of LCO 3.1.6 are necessary. For testing performed with a sequence not in compliance with LCO 3.1.6, I
the requirements of LCO 3.1.6 may be suspended, provided additional administrative controls are placed on the test to ensure that the assumptions of the special safety analysis for the test sequence remain valid. When deviating from the prescribed sequences of LCO 3.1.6, individual control rods must be bypassed in the Rod Action Control Cabinet (RACC). Assurance that the test sequence is followed can be provided by a second licensed operator or other qualified member of the technical staff verifying conformance to the approved test sequence. These controls are consistent with those normally applied to operation in the startup range as defined in SR 3.3.2.1.6, when it is necessary to deviate from the prescribed sequence (e.g., an inoperable control rod that must be fully inserted).
Control rod testing, while in MODES 1 and 2 with TIIERMAL POWER greater than the LPSP of the RWM,is adequately controlled by the existing LCOs on power distribution limits and control rod block instrumentation. Control rod movement during these conditions is not restricted to prescribed sequences and can be performed within the constraints of LCO 3.2.1, " AVERAGE PLANAR LINEAR IIEAT GENERATION RATE (APLIIGR)," LCO 3.2.2, " MINIMUM l
CRITICAL POWER RATIO (MCPR)," LCO 3.2.3, " LINEAR IIEAT GENERATION RATE (LIIGR)," and LCO 3.3.2.1, With TIIERMAL POWER less than or equal to the LPSP of the RWM, the provisions of this Special Operations LCO are necessary to perform special tests that are not in conformance with the prescribed control rod sequences of LCO 3.1.6. While in MODES 3 and 4, control rod withdrawal is
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only allowed if performed in accordance with Special Operations LCO 3.10.3, " Control Rod Withdrawal-Ilot Shutdown" or Special j
Operations LCO 3.10.4, " Control Rod Withdrawal-Cold Shutdown,"
l which provide adequate controls to ensure that the assumptions of the 1
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safety analyses of Reference B3.10.7-1 and B3.10.7-2 are satisfied.'
During these Special Operations and while in MODE 5, the one rod out interh>ck (LCO 3.9.2, " Refuel Position One-Rod-Out Interlock) 1 and scram functions (LCO 3.3.1.1, " Reactor Protection System (RPS)
I Instrumentation," and LCO 3.9.5, " Control Rod OPERABILITY-Refueling"), or the added administrative controls _
.l prescribed in the applicable Special Operations LCOs, minimize i
potential reactivity excursions.
A.1 With the requirements of the LCO not met (e.g., the control rod pattern not in compliance with the special test sequence), the testing is required to be immediately suspended. Upon suspension of the special test, the -
provisions of LCO 3.1.6 are no longer excepted, and appropriate actions
-l are to be taken either to restore the control rod sequence to the prescribed
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.i sequence of LCO 3.1.6, or to shut down the reactor, if required by LCO 3.1.6.
' SR 3.10.7.1 During performance of the special test, a second licensed operator or j
other qualified member of the technical staff is required to verify -
conformance with the approved sequence for the test. This verification must be performed during control rod movement to prevent deviations -
1 from the specified sequence. This Surveillance provides adequate i
assurance that the specified test sequence is being followed arid is also supplemented by SR 3.3.2.1.6, which requires verification of the i
bypassing of control rods in RACC and subsequent movement of these -
j control rods.
B3.10.7.1 NEDE-240ll-P-A-US, General Electric Standard Application for Reactor Fuel, Supplement for United States (as amended).
B3.10.7.2 ABWR SSAR, Section 15.4.1.
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B3.10.7.4 ???????
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i Special Operations Shutdown Afargin (SDM) Test-Refueling BASFS The purpose of this MODE 5 Special Operations LCO is to permit SDM testing to be performed for those plant configurations in which the reactor pressure vessel (RPV) head is either not in place or the head bolts are not fully tensioned.
LCO 3.1.1, "SIIUTDOWN MARGIN (SDM)," requires that adequate SDM be demonstrated following fuel movements or control rod replacement within the RPV. The demonstration must be performed prior to or within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after criticality is reached. This SDM test may be performed prior to or during the first startup following refueling. Performing the SDM test prior to startup requires the test to be performed while in MODE 5 with the vessel head bolts less than fully tensioned (and possibly with the vessel head removed). While in MODE 5, the reactor mode switch is required to be in the shutdown or refuel position, where the applicable control rod blocks ensure that the reactor will not become critical. The SDM test requires the reactor mode switch to he in the startup or hot standby position, since more than one control rod will be withdrawn for the purpose of demonstrating adequate SDM. This Special Operations LCO provides the appropriate additional controls to allow withdrawing more than one control rod from a core cell containing one or more fuel assemblies when the reactor vessel head holts are less than fully tensioned.
Prevention and mitigation of unacceptable reactivity excursions during control rod withdrawal, with the reactor mode switch in the startup or hot standby position while in MODE 5, is provided by the Startup Range Neutron Monitor (SRNM) neutron flux scram (LCO 3.3.1.1,
" Reactor Protection System (RPS) Instrumentation"), average power range monitor (APRM) neutron flux scram (LCO 3.3.1.1), and control rod block instrumentation (LCO 3.3.2.1, " Control Rod Block Instrumentation"). The limiting reactivity excursion during startup conditions while in MODE 5 is the Rod Withdrawal Error (RWE).
RWE analyses assume that the reactor operator follows prescribed withdrawal sequences. For SDM tests performed within these defined sequences, the analyses of References B3.10.8-1 and B3.10.8-2 are applicable. Ilowever, for some sequences developed for the SDM
testing, the control rod patterns assumed in the safety analyses of References B3.10.8-1 and B3.10.8-2 may not be met. Therefore, special RWE analyses, performed in accordance with an NRC approved methodology, are required to demonstrate that the SDM test sequence will not result in unacceptable consequences should a RWE occur during the testing. For the purpose of this test, protection provided by the normally required MODE S applicable LCOs, in addition to the requirements of this LCO, will maintain normal test operations as well as postulated accidents within the bounds of the appropriate safety analysts (Refs. B3.10.8-1 and B3.10.8-2). In addition to the added requirements for the RWM, SRNM, APRh!,
and control rod coupling, the notch out mode is specified for out of sequence withdrawals. Requiring the notch out mode limits withdrawal steps to a single notch, which limits inserted reactivity, and allows adequate monitoring of changes in neutron flux, which may occur during the test.
As described in LCO 3.0.7, compliance with Special Operations LCOs is optional, and therefore, no criteria of the NRC Policy Statement apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.
As described in LCO 3.0.7, compliance with this Special Operations LCO is optional. SDM tests may be performed while in MODE 2, in accordance with Table 1.1-1, without meeting this Special Operations LCO or its ACTIONS. For SDM tests performed while in MODE 5, additional requirements must be met to ensure that adequate protection against potential reactivity excursions is available. Because multiple control rods will be withdrawn and the reactor will potentially become critical, the approved control rod withdrawal sequence must be enforced by the RWM (LCO 3.3.2.1, Function Ib, MODE 2), or must be verified by a second licensed operator or other qualified member of the technical staff. To provide additional protection against an inadvertent criticality, control rod withdrawals that do not conform to the ganged withdrawal sequence restrictions specified in LCO 3.1.6, " Rod Pattern Control" (i.e., out of sequence control rod withdrawals) must he made in the notched withdrawal mode to minimize the potential reactivity insertion associated with
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cach movement. Coupling integrity of withdrawn control rods is required to minimize the probability of a RWE and ensure proper functioning of the withdrawn control rods, if they are required to scram. Because the reactor vessel head may be removed during these tests, no other CORE ALTERATIONS may be in progress. This l
Special Operations LCO then allows changing the Table 1.1-1 reactor mode switch position requirements to include the startup/ hot standby position, such that the SDh! tests may be performed while in h10DE 5.
These SDM test Special Operations requirements are only applicable if the SDM tests are to be performed while in MODE 5 with the reactor vessel head removed or the head bolts not fully tensioned. Additional requirements during these tests to enforce control rod withdrawal sequences and restrict other CORE ALTERATIONS provide protection against potential reactivity excursions. Operations in all other MODES are unaffected by this LCO.
A.1 With one or more of the requirements of this LCO not met, the testing should be immediately stopped by placing the reactor mode switch in the shutdown or refuel position. This results in a condition that is consistent with the requirements for MODE 5 where the provisions of this Special Operations LCO are no longer required.
SR 3.10.8.1 and SR 3.10.8.2 The control rod withdrawal sequences during the SDM tests may be enforced by the RWM (LCO 3.3.2.1, Function Ib, MODE 2 requirements) or by a second licensed operator or other qualified member of the technical staff. As noted, either the applicable SRs for the RWM (LCO 3.3.2.1) must be satisfied according to the applicable Frequencies (SR 3.10.8.1 and SR 3.10.8.2), or the proper movement of control rods must be verified. This latter verification (i.e., SR 3.10.8.2) must be performed during control rod movement to prevent deviations from the specified sequence. These surveillances provide adequate assurance that the specified test sequence is being followed.
SR 3.10.8.3 Periodic verification of the administrative controls established by this LCO will ensure that the reactor is operated within the bounds of the safety analysis. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is intended to provide appropriate assurance that each operating shift is aware of and verifies compliance with these Special Operations LCO requirements.
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SR 3.10.8.4 Coupling verification is performed to ensure the control rod is connected I
to the control rod drive mechanism and will perform its intended function when necessary. The verification is required to be performed any time a control rod is withdrawn to the " full out" notch position or prior to declaring the control rod OPER ABLE after work on the control rod or 4
CRD System that could affect coupling. This Frequency is acceptable, considering the low probability that a control rod will become uncoupled l
when it is not being moved as well as operating experience related to uncoupling events.
B3.10.8.1 NEDE-24011-P-A-US, General Electric Standard Application for Reactor Fuel, Supplement For United States (as amended).
B3.10.8.2 AEWR SSAR, Section 15.4.1.
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l Special Operations
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Recirculation Imops-Testing BASES The purpose of this Special Operations LCO in MODES 1 and 2 is to allow either the PIIYSICS TESTS or the Startup Test Program to be performed with less than nine reactor Internal pumps in operation.
Teting periormed as part of the Startup Test Program (Ref. B3.10.9-1), or PIIYSICS TESTS authorized under the provisions of 10 CFR 50.59 (Ref. B3.10.9-2) or otherwise approved by the NRC, may be required to be performed under natural circulation conditions with the reactor critical. LCO 3.4.1, " Reactor Internal Pumps (RIP) Operating," requires that [ ten] or nine reactor internal pumps be in open. ion during MODES I and 2. This Special Operations LCO provides the appropriate additional restrictions to f
allow testing at nature drculation conditions or with less than nine reactor internal pumps in operation with the reactor critical.
The operation of the Reactor Coolant Recirculation System is an I
initial condition assumed in the design basis loss of coolant accident (Ref. B3.10.9-3). During a LOCA the operating RIPS are all assumed l
to trip at time zero due to a coincident loss of offsite power. The subsequent mean core flow coastdown will be immediate and rapid
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because of the relatively low inertin of the pumps. During PIIYSICS TESTS l
s 57, RTP, or limited testing during the Startup Tc;t Program for the initial I
cycle, the decay heat ir '.a reactor coolant is sufficiently low, such that the consequences of an acFent are reduced and the coastdown charactesistics of the RIPS are not important. En addition, the probability of a Design Basis Accident (DBA) or other accidents occurring during the limited time allowed at natural circulation or with less than nine RIPS in operation is low.
As described in LCO 3.0.7, compliance with Special Operations LCOs f
is optional, and tl erefore, no criteria of the NRC Policy Statement apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in l
their respective Bases.
l As described in LCO 3.0.7, compliance with this Special Operations LCO is optional. Ilowever, to perform testing at natural circulation c<mditions or with less than nine RIPS operating, operations must be limited to those tests defined in the Startup Test Program or approved PIIYSICS TESTS performed at s 57c RTP. To minimize the i
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probability of an accident, while operating at natural circulation cond*tions or with less than nine operating RIPS, the duration of these tests is limited to :s; 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This Special OperaGons LCO then allows suspension of the requirements of LCO ?,.4.1 during such testing. In addition to the requirements of thb LCO, the normally required MODE 1 or MODE 2 applicable I COs must be met.
This Special Operations LCO may only be used _while performing testing at natural circulation conditions or while operating with less than nine RIPS, as may be required as part of the Startup Test Program or during low power PIIYSICS TESTS. Additional requirements during these tests to limit the operating time at natural circulation conditions reduce the probability that a DBA may occur with both recirculation loops not in operation. Operations in all other MODES are unaffected by this LCO.
With the testing performed at natural circulation conditions or with less than nine RIPS operating, and the duration of the test exceeding the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time limit, actions should be taken to promptly shut down.
Inserting all insertable control rods will result in a condition that does not require all nine RIPS to be in operation. The allowed Completion Time of I hour provides sufficient time to insert the withdrawn control rods.
B.1 Wih the requirements of this LCO not met for reasons other than those specified in Condition A (e.g., low power PHYSICS TESTS exceeding l
5% RTP, or unapproved testing at natural circulation), the reactor mode switch should immediately be placed in the shutdown position. This results -
in a condition that does not require all nine RIPS ta be in operation. The action to immediately place the reactor mode switch in the shutdown position prevents unacceptable consequences from an accident initiated from outside the analysis bounds. Also, operation beyond authorized SR 3.10.9.1 and SR 3.10.9.2 Periodic verification of the administrative controls established by this LCO will ensure that the reactor is operated within the bounds of this LCO.
l Because the I hour Frequency provides frequent checks of the LCO requirements during the allowed 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> testing interval, the probability of operation outside the limits concurrent with a postulated accident is reduced even further.
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1 B3.10.9.1 ABWR SSAR, Chapter 14.
1 B3.10.9.2 -10 CFR 50.59.
A B3.10.9.3 ABWR SSAR, Section 6.3.3.
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.Special Operations Training Startups BASES The purpose of this Special Operations LCO is to permit training startups to be performed while in MODE 2 to provide plant startup experience for reactor operators. This training involves withdrawal of control rods to achieve criticality and then further withdrawal of control rods, as would be experienced during an actual plant startup.
During these training startups, if the reactor coolant is allowed to heat up, maintenance of a constant reactor vessel water level requires the passage of reactor coolant through the Reactor Water Cleanup System, as the reactor coolant specific volume increases. Since this j
results in reactor water discharge to the radioactive waste disposal system, the amount of this discharge should be minimized. This Special Operations LCO provides the appropriate additional controls 1
to allow one residual heat removal (RIIR) subsystem to be aligned in the shutdown cooling mode, so that the reactor coolant temperature can be controlled during the training startups, thereby minimizing the discharge of reactor water to the radioactive waste disposal system.
The Emergency Core Cooling System (ECCS) is designed to provide core cooling following a loss of coolant accident (LOCA). The low pressure core flooder (LPFL) mode of the RIIR System is one of the
.j ECCS subsystems assumed to function during a LOCA. With reactor power :s; 1% RTP and average reactor coolant temperature < 93*C
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(200*I9, the stored energy in the reactor core and coolant system is i
1 very low, and a reduced complement of ECCS can provide the required core cooling, thereby a!!owing operation with one RIIR
)
subsystem !n the shutdown cooling mode (Ref. B3.10.10-1).
As described in LCO 3.0.7, compliance with Special Operations LCOs is optional, and therefore, no criteria of the NRC Policy Statement apply. Special Operations LCOs provide flexibility to perform certain
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operations by appropriately modifying requirements of other LCOs. A j
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discussion of the criteria satisfied for the other LCOs is provided in l
I their respective Bases.
As described in LCO 3.0.7, compliance with this Special Operations LCO is optional. Training startups may be performed while in j
MODE 2 with no RIIR subsystems aligned in the shutdown cooling j
mode and, therefore, without meeting this Special Operations LCO or j
its ACTIONS. Ilowever, to minimize the discharge of reactor coolant
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e to the radioactive waste disposal system, performance of the training startups may be performed with one RIIR subsystem aligned.tn the shutdown cooling mode to maintain reactor coolant temperature.
< 93*C (200*F). Under thase conditions, the TIIERMAL POWER must be maintained s 1% RTP and the reactor coolant temperature must be s 93*C (200*F). This Special Operations LCO then allows changing the LPFL OPERABILITY requirements. In addition to the requirements of this LCO, the normally required MODE 2 applicable LCOs must also be met.
Training startups while in MODE 2 may be performed with one RIIR subsystem aligned in the shutdown cooling mode to control the reactor coolant temperature. Additional requirements during these tests to restrict the reactor power and reactor coolant temperature provide protection against potential conditions that could require operation of both RIIR subsystems in the LPFL mode of operation. Operations in all other MODES are unaffected by this LCO.
A.1 With one or more of the requirements of this LCO not met, (i.e., reactor power s 1% RTP, or average reactor coolant temperature h 93*C (200*F) the reactor may be in a condition that requires the full complement of ECCS subsystems, and the reactor mode switch must be immediately placed in the shutdown position. This results in a condition that does not require all RHR subsystems to be OPERABLE in the LPFL mode of operation. This action may restore compliance with the requirements of this Special Operatiors LCO or may result in placing the plant in either MODE 3 or MODE 4.
SR 3.10.10.1 and SR 3.10.10.2 Periodic verification that the THERMAL POWER and reactor coolant temperature limits of this Special Operations LCO are satisfied will ensee that the stored energy in the reactor core and reactor coolant are sufficiently low to preclude the need for all RHR subsystems to be aligned in the LPFL mode of operation. The I hour Frequency provides frequent checks of these LCO requirements during the training startup.
B3.10.10.1 ABWR SSAR, Section 6.3.3.
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