ML20056G167

Forwards Proof & Review Advanced BWR TSs for Low Power & Shutdown.Anticipates That Formal Comments to Proof & Review Advanced BWR TS Will Be Made by 930920

ML20056G167
Person / Time
Site:	05200001
Issue date:	08/17/1993
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Marriott P GENERAL ELECTRIC CO.
References
NUDOCS 9309020176
Download: ML20056G167 (140)

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NUCLEAR REGULATORY COMMISSION

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17, 1993 Auaust Docket No.52-001 Mr. Patrick W. Marriott, Manager Licensing & Consulting Services GE Nuclear Energy 175 Curtner Avenue San Jose, California 95125

SUBJECT:

ADVANCED BOILING WATER REACTOR (ABWR) TECHNICAL SPECIFICATIONS Enclosed are the proof and review ABWR bases for the following technical specifications for low Power and Shutdown (LPS):

3.4.8 RHR - Cold Shutdown 3.5.2 ECCS - Shutdown 3.7.2 RCW/RSW and UHS - Shutdown (for P&R LC0 3.7.1.1) 3.7.3 RCW/RSW and UHS - Refueling (for P&R LC0 3.7.1.2)

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3.9.7 RHR - H gh Water Level

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3.9.8 RHR - Low Water Level 4

In addition, technical specifications and their bases for the following sections, with GE Nuclear Energy's (GE) proof and review comments incorporated, are enclosed:

2.0 Safety Limits

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3.0 Applicability

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3.1 Reactivit3 Control 3.2 Power Distribution The comments for LC0 3.1.7, " Standby Liquid Control (SLC) System," received from GE on August II, 1993, have not yet been incorporated. A revised LCO will be forwarded at a future date.

As we discussed at our management meeting on June 10, 1993, the Nuclear Regulatory Commission (NRC) staff will be providing GE, until August 31, selected sections of proof and review ABWR technical specifications. These sections are based on the NRC staff review of the GE mark-up of the BWR-6 and BWR-4 Standard Technical Specifications; the sections, as provided, are acceptable to the NRC staff. As discussed, we anticipate that GE tvill interface very closely with the staff to resolve any issues on these sections prior to August 31, 1993. Under this arrangement, we anticipate that formal coments to proof and review ABWR technical specifications made by September 20, 1993, will be few.

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9309020176 9308171

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Mr. Patrick W. Marriott August 17, 1993 The electronic text of these sections is available on the NRC Technical Specifications Branch electronic bulletin board (OTSB-BBS) in Wordperfect 5.1 format. The data telephone number for the OTSB-BBS is (301) 504-1778, and thesystem operator is Tom Dunning who is available for assistance at (301) o 504-1189. Also, in accordance with our agreements, GE will maintain these sections in Wordperfect 5 1 format and will produce subsequent issues of the ABWR technical specifications in Wordperfect 5.1 format.

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, Or$1a!SieniL4 l

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

Enclosure:

As stated I

cc w/ enclosure:

See next page r

DISTRIBUTION w/ enclosure:

Docket File CPoslusny PShea FMReinhart PDR PDST R/F DORS R/F OTSB R/F DISTRIBUTION w/o enclosure:

TEMurley/FJMiraglia WTRussell JGPartlow ACRS (11)

DMCrutchfield BMGrimes CIGrimes BABoger JWiggins ACThadani FJCongel JSWermiel RBBarrett CEMcCracken RCJones CEBerlinger AEChaffee GHMarcus WBHardin, RES LCShao, RES JA0'Brien RWBorchardt JNWilson SNinh TGody, 17G21 JEMoore, 15B18 RHLo PCHearn l

OFC:

LA:PDST:ADAR OTSB J[tSB:

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CPodus y:sg JNbson Di Crutchfield NAME:

DATE:

08/b3 08h93 08/f793 Off1CIAL RECORD COPY: DOCUMENT NAME: TSP &R7.LTR t

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s Mr. Patrick W. Harriott Docket No.52-001 General Electric Company cc: Mr. Robert Mitchell Mr. Joseph Quirk I

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 i

Regulatory Programs GE Nuclear Energy i

12300 Twinbrook Parkway Suite 315 Rockville, Maryland 20852 Director, Criteria & Standards Division Office of Radiation Programs j

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.

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1615 L Street, N.W.

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Suite 1000 Washington, D.C.

20036 i

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RHR Shutdown Cooling System-Cold Shutdoun B 3.4.8 I

B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown BASES BACKGROUND Irradiated fuel in the shutdown reactor core generates heat during the decay of fission products and increases the temperature of the reactor coolant. This decay heat must be removed to maintain the temperature of the reactor coolant i

at 5; 93' (200"F). This decay heat removal is in preparation for performing refueling or maintenance operations, or for keeping the reactor in the Cold Shutdown condition.

The three redundant, manually controlled shutdown cooling i

subsystems of the RHR System provide decay heat removal.

Each loop consists of a motor driven pump, a heat exchanger, and associated piping and valves.

Each 1 cop has its own dedicated suction from the RPV.

Each pump discharges the reactor coolant, after circulation through the respective heat exchanger, to the reactor via feedwater line "A" for RHR loop "A" and via the RHR low pressure flooder spargers for loops "B" and "C".

The RHR heat exchangers transfer heat to the Reactor Building Cooling Water System j

(LC0 3.7.2).

APPLICABLE Decay heat removal by the RHR System in the shutdown cooling j

SAFETY ANALYSES mode is not required for mitigation of any event or accident evaluated in the safety analyses. Decay heat removal is, t

however, an important safety function that must be accomplished or core damage could result. Although the RHR_

t shutdown cooling subsystem does not meet a specific criterion of the NRC Policy Statement, it was identified in the Policy Statement as a significant contributor to risk reduction. Therefore, the RHR shutdown cooling subsystem is covered by a Technical Specification.

LCO Two RHR shutdown cooling subsystems are required to be F

OPERABLE, and one shutdown cooling subsystem must be in operation. An OPERABLE RHR shutdown cooling subsystem consists of one OPERABLE RHR pump, a heat exchanger, and the l

associated piping and valves.

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(continued)

ABWR STS B 3.4-1 P&R(LPS),08/13/93 i

i RHR Shutdown Cooling System-Cold Shutdown B 3.4.8 j

BASES i

2 LCO Each shutdown cooling subsystem is considered OPERABLE if it (continued) can be manually aligned (remote or local) in the shutdown l

cooling mode for removal of decay heat.

In MODE 4, one RHR l

shutdown cooling subsystem can provide the required cooling, but two subsystems are required to be OPERABLE to provide redundancy. Operation of one subsystem can maintain and reduce the reactor coolant temperature as required.

However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required.

Note 1 permits both RHR shutdown cooling subsystems to be shut down for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, l

provided one subsystem is OPERABLE.

Note 2 allows one RHR i

shutdown cooling subsystem to be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the performance of surveillance tests provided the remaining RHR shutdown cooling subsystem is OPERABLE.

These tests may be on the affected RHR System or on some other plant system or component that necessitates placing l

i the RHR System in an inoperable status during the performance. This is permitted because the core heat l

generation can be low enough and the heatup rate slow enough i

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to allow some changes to the RHR subsystems or other j

operations requiring RHR flow interruption and loss of redundancy.

APPLICABILITY In MODES 1 and 2, and in MODE 3 with reactor steam dome i

pressure above the RHR cut in permissive pressure, this LCO j

is not applicable. Operation of the RHR System in the

)

shutdown cooling mode is not allowed above this pressure j

because the RCS pressure may exceed the design pressure of the shutdown cooling piping. Decay heat removal at reactor pressures above the RHR cut in permissive pressure is typically accomplished by condensing the steam in the main i

i condenser. Additionally, in MODE 2 below this pressure, the OPERABILITY requirements for the Emergency Core Cooling Systems (ECCS) (LC0 3.5.1, "ECCS-Operating") do not allow placing the low pressure RHR shutdown cooling subsystem into i

operation.

In MODE 4, the RHR System may be operated in the shutdown cooling mode to remove decay heat to reduce or maintain coolant temperature.

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(continued)

ABWR STS B 3.4-2 P&R(LPS), 08/13/93 j

i RHR Shutdown Cooling System-Cold Shutdown B 3.4.8 BASES i

APPLICABILITY The requirements for decay heat removal in MODE 3 below the (continued) cut in permissive pressure and in MODE 5 are discussed in LCO 3.4.7, " Residual Heat Removal (RHR) Shutdown Cooling System-Hot Shutdown"; LCO 3.9.7, " Residual Heat Removal (RHR)-High Water Level"; and LCO 3.9.8, " Residual Heat Removal (RHR)-Low Water Level."

ACTIONS A.]

With one or both required RHR shutdown cooling subsystems inoperable for decay heat removal, the inoperable subsystem (s)must be restored to OPERABLE status without delay.

In this condition, the remaining OPERABLE required subsystem can provide the necessary decay heat removal.

The overall reliability is reduced, however, because a single failure in the OPERABLE required subsystem could result in reduced RHR shutdown cooling capability.

With one of the two required RHR shutdown cooling subsystems e

inoperable, the remaining required subsystem is capable of providing the required decay heat removal. However, the overall reliability is reduced. Therefore an alternate required method of decay heat removal must be provided (such i

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 i

cooling subsystem inoperability. This re-establishes backup decay heat removal capabilities, similar to the requirements

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of the LCO. The I hour Completion Time is cased 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 provide assurance of continued heat removal capability.

The required cooling capacity of the alternate method should i

be ensured by verifying (by calculation or demonstration) its capability to maintain or reduce temperature. Decay 4

heat removal by ambient losses can be considered as contributing to the alternate method capability. Alternate methods that can be used include (but are not limited to)

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l (continued) i ABWR STS B 3.4-3 P&R(LPS),08/13/93

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l RHR Shutdoen Cooling System--Cold Shutdown l

B 3.4.8 i

BASES ACTIONS A.] (continued)

[the third RHR shutdown cooling subsystem], the Spent Fuel Pool Cooling System or the Reactor Water Cleanup System.

B.1 and B.2 With no RHR shutdown cooling subsystem in operation, except as is permitted by the LCO Note, reactor coolant circulation by the RHR shutdown cooling subsystem or [enough RIPS so that at least [4] are operating] must be restored without delay.

Until RHR operation is re-established, an alternate method of reactor coolant circulation must be placed into service.

This will provide the necessary circulation for monitoring coolant temperature and pressure.

The I hour Completion Time is based on the coolant circulation function and is modified such that the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is applicable separately for each occurrence involving a loss of coolant circulation.

Furthermore, verification of the functioning of the alternate method must be reconfirmed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter. This will provide assurance of continued temperature monitoring capability.

During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR shutdown cooling system), the reactor coolant temperature must be periodically monitored to ensure proper function of the alternate method.

The once per hour i

Completion Time is deemed appropriate.

1 SURVEILLANCE SR 3.4.8.1 REQUIREMENTS t

This Surveillance verifies that [one] RHR subsystem or at least [4] RIPS are 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.

i (continued)

ABWR STS B 3.4-4 P&R(LPS),08/13/93 1

RHR Shutdown Cooling System-Cold Shutdown B 3.4.8 BASES

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f REFERENCES None.

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ABWR STS B 3.4-S PAR (LPS), 08/13/93 i

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l ECCS--Shutdown B 3.5.2 8 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) i B 3.5.2 ECCS--Shutdown BASES BACKGROUND A description of the High Pressure Core Spray (HPCS) System, Low Pressure Core Spray (LPCS) System, and low pressure coolant injection (LPCI) mode of the Residual Heat Removal (RHR) System is provided in the Bases for LCO 3.5.1, "ECCS-Operating."

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APPLICABLE ECCS performance is evaluated for the entire spectrum of i

SAFETY ANALYSES break sizes for a postulated loss of coolant accident (LOCA). The long term cooling analysis following a design basis LOCA (Ref. 1) demonstrates that only one ECCS injection / spray subsystem is required, post LOCA, to i

maintain the peak cladding temperature below the allowable limit. To provide redundancy, a minimum of two ECCS subsystems are required to be OPERABLE in MODES 4 and 5.

l Two OPERABLE ECCS injection / spray subsystems also ensure adequate inventory makeup in the reactor pressure vessel (RPV) in the event of an inadvertent vessel draindown.

The ECCS satisfy Criterion 3 of the NRC Policy Statement.

f LCO Two ECCS injection / spray subsystems are required to be OPERABLE. The ECCS injection / spray subsystems are defined as the three LPCI subsystems, the LPCS System, and the HPCS-

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System. The LPCS System and each LPCI subsystem consist of one motor driven pump, piping, and valves to transfer water

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from the suppression pool to the RPV. The HPCS System consists of one motor driven pump, piping, and valves to transfer water from the suppression pool or condensate storage tank (CST) to the RPV.

One LPCI subsystem (A or B) may be aligned for decay heat removal in MODE 4 or 5 and considered OPERABLE for the ECCS l

function, if it can be manually realigned (remote or local) to the LPCI mode and is not otherwise inoperable. Because of low pressure and low temperature conditions in MODES 4 and 5, sufficient time will be available to manually align (continued)

ABWR STS B 3.5-1 P&R(LPS),08/13/93 i

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ECCS--Shutdown B 3.5.2 l

BASES LCO and initiate LPCI subsystem operation to provide core r

(continued) cooling prior to postulated fuel uncovery.

APPLICABILITY OPERABILITY of the ECCS injection / spray subsystems is required in MODES 4 and 5 to ensure adequate coolant inventory and sufficient heat removal capability for the irradiated fuel in the core in case of an inadvertent draindown of the vessel.

Requirements for ECCS OPERABILITY during MODES 1, 2, and 3 are discussed in the Applicability section of the Bases for LCO 3.5.1.

ECCS subsystems are not required to be OPERABLE during MODE 5 with the upper containment pool gate removed, and the water level maintained at 2 7 m (23 ft) above the RPV flange. This provides sufficient coolant inventory to allow operator action to terminate the inventory loss prior to fuel uncovery in case of an inadvertent draindown.

The Automatic Depressurization System is not required to be OPERABLE during MODES 4 and 5 because the RPV pressure is 2

< [3.5 Kg/cm g (50 psig)], and the LPFL and HPFL subsystems can provide core cooling without any depressurization of the primary system.

ACTIONS A.1 and B.1 If any one required ECCS injection subsystem is inoperable, the required inoperable ECCS injection subsystem must be

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restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

In this Condition, the remaining OPERABLE subsystem can provide sufficiert RPV flooding capability to recover from an inadvertent vessel draindown. However, overall system reliability is reduced because a single failure in the remaining OPERABLE subsystem concurrent with a vessel draindown could result in the ECCS not being able to perform its intended function. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time for restoring the required ECCS injection subsystem to OPERABLE status is based on engineering judgment that considered the availability of one subsystem and the low probability of a vessel draindown event.

With the inoperable subsystem not restored to OPERABLE status within the required Completion Time, action must be (continued)

ABWR STS B 3.5-2 P&R(LPS),08/13/93

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ECCS-Shutdown i

B 3.5.2 BASES l

l ACTIONS A.1 and B.1 (continued)

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initiated immediately to suspend operations with a potential for draining the reactor vessel (OPDRVs) to minimize the probability of a vessel draindown and the subsequent potential for fission product release.

Actions must continue until OPDRVs are suspended.

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C.I. C.2. D.1. D.2. and D.3 If both of the required ECCS injection subsystems are inoperable, all coolant inventory makeup capability may be i

unavailable.

Therefore, actions must be initiated immediately to suspend OPDRVs in order to minimize the probability of a vessel draindown and the subsequent potential for fission product release.

Actions must continue until OPDRVs are suspended. One ECCS injection subsystem must also be restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

If at least one ECCS injection subsystem is not restored to OPERABLE status within the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time, additional actions are required to minimize any potential fission product release to the environment. This includes initiating immediate action to restore the following to OPERABLE status: secondary containment, one standby gas treatment subsystem, and one isolation valve and associated instrumentation in each secondary containment penetration flow path not isolated.

This may be performed by an administrative check, by examining logs or other

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information, to determine if the components are out of service for maintenance or other reasons. Verification does not require performing the Surveillances needed to demonstrate OPERABILITY of the components.

If, however, any required component is inoperable, then it must be restored i

to OPERABLE status.

In this case, the Surveillances may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE.

The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time to restore at least one ECCS injection subsystem to OPERABLE status ensures that prompt action will be taken to provide the required cooling capacity or to initiate actions to place the plant in a I

(continued) l ABWR STS B 3.5-3 P&R(LPS),08/I3/93

ECCS--Shutdown B 3.5.2 BASES i

ACTIONS C.I. C.2. D.l. D.2. and D.3 (continued) condition that minimizes any potential fission product release to the environment.

SURVEILLANCE SR 3.5.2.1 REQUIREMENTS The minimum water level of 7 m (23 ft) required for the suppression pool is periodically verified to ensure that the suppression pool will provide adequate net positive suction head (NPSH) for the ECCS pumps, recirculation volume, and vortex prevention. With the suppression pool water level less than the required limit, all ECCS injection subsystems are inoperable.

When the suppression pool level is < 7 m (23 ft), the HPCF System is considered OPERABLE only if it can take suction from the CST and the CST water level is sufficient to provide the required NPSH for the HPCF pump. Therefore, a verification that either the suppression pool water level is 2 7 m (23 ft) or the HPCF System is aligned to take suction from the CST and the CST contains 2 [ ] liters ([ ] gallons) of water, equivalent to [ ] m ([ ] ft), ensures that the HPCF System can supply makeup water to the RPV.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency of this SR was developed considering operating experience related to suppression pool and CST water level variations and instrument drift during the applicable MODES.

Furthermore, the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is considered adequate in view of other indications in the control room, including alarms, to alert the operator to an abnormal suppression pool or CST water level condition.

SR 3.5.2.2. SR 3.5.2.4. and SR 3.5.2.5 The Bases provided for SR 3.5.1.1, SR 3.5.1.4, and SR 3.5.1.7 are applicable to SR 3.5.2.2, SR 3.5.2.4, and SR 3.5.2.5, respectively.

(Continued)

ABWR STS B 3.5-4 P&R(LPS),08/13/93

ECCS-Shutdotin B 3.5.2 BASES SURVEILLANCE SR 3.5.2.3 REQUIREMENTS (continued)

Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. The 31 day Frequency is appropriate because the valves are operated under procedural control and the probability of their being mispositioned during this time period is low.

In MODES 4 and 5, the RHR System may operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor. Therefore, RHR valves that are required for LPFL subsystem operation may be aligned for the shutdown cooling mode.

Therefore, this SR is modified by a Note that allows one LPFL subsystem of the RHR System to be considered OPERABLE for the ECCS function if all the required valves in the LPFL flow path can be manually realigned (remote or local) to allow injection into the RPV and the system is not otherwise inoperable. This will ensure adequate core

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cooling if an inadvertent vessel draindown should occur.

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REFERENCES 1.

FSAR, Section [6.3.3.4].

ABWR STS B 3.5-5 P&R(LPS),08/13/93

RCW/RSW System and UHS-Shutdown B 3.7.2 B 3.7 PLANT SYSTEMS B 3.7.2 Reactor Building Cooling Water (RCW) System, Reactor Service Water (RSW) System and Ultimate Heat Sink (UHS) - Shutdown BASES i

BACKGROUND A description of the RCW and RSW Systems and the UHS are provided in the Bases for LC0 3.7.1, " Reactor Building i

Cooling Water (RCW) System, Reactor Service Water (RSW)

System and Ultimate Heat Sink (UHS) - Operating."

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APPLICABLE The volume of water incorporated in the UHS is sized so that i

SAFETY ANALYSES sufficient water inventory is available for all RCW/RSW System post LOCA cooling requirements for a 30 day period with no additional makeup water source available (Ref.1).

The ability of the RCW/RSW System to support long term cooling of the reactor or containment is assumed in evaluations of the equipment required for safe reactor shutdown presented in the SSAR, Sections 9.2.11, 6.2.1.1.3.3.1.4, and Chapter 15, (Refs 2, 3, and 4, respectively). The long term cooling analyses following a design basis LOCA demonstrates that only one division of the RCW/RSW System is required, post LOCA, to support long term cooling of the reactor or containment.

To provide redundancy, a minimum of two RCW/RSW divisions are required i

to be OPERABLE in MODES 4 and 5 except with the reactor cavity to dryer / separator storage pool gate removed and water level 2 7.0 m (23 ft) over the top of the reactor i

pressure vessel flange.

The combined RCW/RSW System, together with the UHS, satisfy Criterion 3 of the NRC Policy Statement.

LC0 Two divisions of the RCW/RSW System and the UHS are required to be OPERABLE to ensure the effective operation of the RHR System in removing heat from the reactor, and the effective operation of other safety related equipment during a DBA or transient.

Requiring two division to be OPERABLE ensures that one' division will be available to provide adequate capability to meet cooling requirements of the equipment required for safe shutdown in the event of a single failure.

Operability of the UHS and the RCW/RSW System is defined in the Basis for LC0 3.7.1.

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(continued)

ABWR TS B 3.7-1 P&R(LPS), 08/13/93

RCW/RSB System and UHS-Shutdowa B 3.7.2 BASES APPLICABILITY In MODES 4 and 5, except with the reactor cavity to dryer / separator storage pool gate removed and water level 2 7.0 m (23 ft) over the top of the reactor pressure vessel flange, two divisions of the RCW/RSW System and the UfP" are required to be OPERABLE to support OPERABILITY o W ie equipment serviced by the RCW/RSW System and UHS, and are required to be OPERABLE in these MODES.

In MODES 1, 2, and 3, the OPERABILITY requirements of the RCW/RSW System and UHS are specified in LCO 3.7.1.

In MODE 5 with the reactor cavity to dryer / separator storage pool gate removed and water level 2 7.0 m (23 ft, over the top of the reactor pressure vessel flange, the OPERABILITY requirements of the RCW/RSW System and UHS are specified in LCO 3.7.3, "RCW/RSW System and UHS - Refueling."

ACTIONS A.1 If one or more required RCW/RSW division (s) or the VHS is inoperable, then immediately, those required feature (s) supported by the inoperable RCW/RSW division (s) or the UHS must be declared inoperable (i.e., Emergency Diesel Generator, RHR heat exchanger) and the applicable Conditions and Required Actions of the appropriate LCOs for the inoperable required feature (s) must be entered.

For the applicable shutdown MODES, an inoperable RCW/RSW division or UHS requires entering the Conditions of LC0 3.8.2, AC Sources-Shutdown," for a diesel generator made inoperable and either LC0 3.4.8, " Residual Heat Removal (RHR) Shutdown -

Cooling System-Cold shutdown," or LC0 3.9.8, " Residual Heat Removal (RHR) Low Water Level" fcr RHR shutdown cooling made inoperable. This is in accordance with LC0 3.0.6 and ensures the proper actions are taken for these components.

SURVEILLANCE SR 3.7.2.1 REQUIREMENTS This SR ensures adequate long term (30 days) cooling can be maintained.

With the UHS water source below the minimum level, the affected RCW/RSW division must be declared inoperable. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on operating (continued)

ABWR TS B 3.7-2 P&R(LPS), 08/13/93

RCW/RSW System and UHS-Shutdown j

B 3.7.2 l

i BASES SURVEILLANCE SR 3.7.2.2 REQUIREMENTS (continued) experience relate to trending of the parameter variations during the applicable MODES. This SR verifies the water level in each RSW pump well of the intake structure to be sufficient for the proper operation of the RSW pumps (net positive suction head and pump vortexing are considered in determining this limit). The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES.

SR 3.7.2.3 Ver',fication of the UHS temperature ensures that the heat I

removal capability of the RCW/RSW System is within the assumptions of the DBA analysis.

The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES.

SR 3.7.2.4 i

Verifying the correct alignment for each manual, power operated, and automatic valve in each RCW/RSW division flow path provides assurance that the proper flow paths will exist for RCW/RSW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing.

A valve is also allowed to be in the nonaccident position and

~~

yet considered in the correct position, provided it can be ~

automatically realigned to its accident position. This SR does not require any testing or valve manipulation; rather, i

it involves verification that those valves capable of potentially being mispositioned are in the correct position.

4 This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

This SR is modified by a Note indicating that isolation of the RCW/RSW System to components or systems may render those i

components or systems inoperable, but does not affect the i

OPERABILITY of the RCW/RSW System. As such, when all RCW/RSW pumps, valves, and piping are OPERABLE, but a branch connection off the main header is isolated, the RCW/RSW System is still OPERABLE.

i (continued) i ABWR TS B 3.7-3 P&R(LPS), 08/13/93 t

RCW/RSW System and UHS-Shutdown B 3.7.2 BASES i

SURVEILLANCE SR 3.7.2.4 (continued)

REQUIREMENTS The 31 day Frequency is based on engineering judgement, is consistent with the procedural controls governing valve operation, and ensure correct valve positions.

SR 3.7.2.5 This SR verifies that the automatic isolation valves of the RCW/RSW System will automatically switch to the safety or emergency position to provide cooling water exclusively to the safety related equipment, and limited non-safety related equipment, during an accident event.

This is demonstrated by use of an actual or simulated initiation signal.

This SR also verifies the automatic start capability of the RCW/RSW pumps that are in standby and automatic valving in each of the standby RCW/RSW heat exchangers in each division. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.5.1.4 overlaps this SR to provide complete testing of the safety function.

Operating experience has shown that these components usually pass the SR when performed on the 18 month Frequency.

i Therefore, this Frequency is concluded to be acceptable from a reliability standpoint.

REFERENCES 1.

Regulatory Guide 1.27, Revision 2, January 1976.

[

2.

ABWR SSAR, Sections 9.2.11 and 9.2.15.

3.

ABWR SSAR, Section 6.2.1.1.3.3.1.4.

i 4.

ABWR SSAR, Chapter 15.

ABWR TS B 3.7-4 P&R(LPS), 08/13/93

RCU/RSM System and UHS-Refueling B 3.7.3 B 3.7 PLANT SYSTEMS P

B 3.7.3 Reactor Building Cooling Water (RCW) System, Reactor Service Water (RSW) System and Ultimate Heat Sink (UHS) - Refueling BASES J

BACKGROUND A description of the RCW and RSW Systems and the UHS are provided in the Bases for LCO 3.7.1, " Reactor Building Cooling Water (RCW) System, Reactor Service Water (RSW)

System and Ultimate Heat Sink (URS) - Operating."

In MODE 5 with the reactor vessel water level 2 7.0 m (23 ft) over the vessel flange the unit components to which the RCW/RSW System is required to supply cooling water is greatly i

reduced from normal operation.

For example, LC0 3.8.2, "AC Sources-Shutdown" and LCO 3.9.7, "RHR-High Water Level"

~;

require one DG and one RHR subsystem to be OPERABLE, respectively, and LC0 3.5.2, "ECCS-Shutdown" does not i

require any ECCS components to be OPERABLE for this condition.

APPLICABLE The volume of water incorporated in the UHS is sized so that SAFETY ANALYSES sufficient water inventory is available for all RCW/RSW System post LOCA cooling requirements for a 30 day period l

with no additional makeup water source available (Ref.1).

The ability of the RCW/RSW System to support long term cooling of the reactor or containment is assumed in evaluations of the equipment required for safe reactor h

shutdown presented in the SSAR, Sections 9.2.11, 6.2.1.1.3.3.1.4, and Chapter 15, (Refs. 2, 3, and 4, respectively). With the unit in MODE 5 and with the reactor ~

cavity to dryer / separator storage gate removed and water level 2 7.0 m (23 ft) over the top of the reactor pressure vessel flange, the volume of water in the reactor vessel provides a heat sink for decay heat removal.

However, to provide redundancy, a minimum of one RCW/RSW division is required to be OPERABLE.

}

The combined RCW/RSW System, together with the UHS, satisfies Criterion 3 of the NRC Policy Statement.

l l

(continued)

ABWR TS B 3.7-1 P&R(LPS), 08/13/93 i

RCW/RSW System and UHS-Refueling B 3.7.3

{

BASES LCO One division of the RCW/RSW System and the UHS are required I

to be OPERABLE to ensure the effective operation of the RHR System in removing heat from the reactor.

LC0 3.9.7, "RHR-High Water Level" requires that one RHR subsystem be OPERABLE in operation 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.

Operability of the VHS and the RCW/RSW System is defined in the Basis I

for LCO 3.7.1.

APPLICABILITY In MODE 5 with the reactor cavity to dryer / separator storage pool gate removed and water level 2 7.0 m (23 ft) over the top of the reactor pressure vessel flange, one division of i

the RCW/RSW System and the UHS are required to be OPERABLE to support OPERABILITY of the equipment serviced by the PCW/RSW System and UHS, and are required to be OPERABLE in this MODE.

In MODES 1, 2, and 3, the OPERABILITY requirements of the RCW/RSW System and UHS are specified in LC0 3.7.1.

In MODES 4 and 5, except with the reactor cavity to dryer / separator storage pool gate removed and water level 2 7.0 m (23 ft) over the top of the reactor pressure vessel flange, the OPERABILITY requirements of the RCW/RSW System and UHS are specified in LC0 3.7.2, "RCW/RSW System and UHS

- Shutdown."

I ACTIONS A.1 i

If no RCW/RSW division is operable or the UHS is inoperable, i

i then, immediately, those required feature (s) supported by the inoperable required RCW/RSW division or UHS must be declared inoperable (i.e., Emergency Diesel Generator, RHR heat exchanger) and the applicable Conditions and Required i

Actions of the appropriate LCOs for the inoperable required i

feature (s) must be entered. An inoperable RCW/RSW division or UHS requires entering the Conditions of LC0 3.8.2, "AC Sources-Shutdown," for a diesel generator made inoperable and LCO 3.9.7, " Residual Heat Removal (RHR)-High Water (continued)

ABWR TS B 3.7-2 P&R(LPS), 08/13/93 j

l i

i RCW/RSW System and UHS-Refueling B 3.7.3 BASES ACTIONS L1 (continued)

Level" for RHR shutdown cooling made inoperable. This is in i

accordance with LCO 3.0.6 and ensures the proper actions are i

taken for these components.

SURVEILLANCE SR 3.7.3.1 REQUIREMENTS This SR ensures adequate long term (30 days) cooling can be maintained. With the UHS water source below the minimum level, the affected RCW/RSW division must be declared inoperable. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES.

f SR 3.7.3.2 This SR verifies the water level in each RSW pump well of the intake structure to be sufficient for the proper i

operation of the RSW pumps (net positive suction head and pump vortexing are considered in determining this limit).

The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES.

i SR 3.7.3.3 Verification of the UHS temperature ensure that the heat removal capability of the RCW/RSW System is within the assumptions of the DBA analysis. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is 4

based on operating experience related to trending of the parameter variations during the applicable MODES.

t SR 3.7.3.4 Verifying the correct alignment for each manual, power operated, and automatic valve in each RCW/RSW division flow path provides assurance that the proper flow paths will exist for RCW/RSW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the (continued)

ABWR TS B 3.7-3 P&R(LPS),08/13/93 i

RCB/RSW System and UHS-Refueling 1

B 3.7.3 BASES I

correct position prior to locking, sealing, or securing. A i

valve is also allowed to be in the nonaccident position and yet considered in the correct position, provided it can be l

automatically realigned to its accident position. This SR does not require any testing or valve manipulation; rather, j

it involves verification that those valves capable of potentially being mispositioned are in the correct position.

This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

1 This SR is modified by a Note indicating that isolation of the RCW/RSW System to components or systems may render those components or systems inoperable, but does not affect the OPERABILITY of the RCW/RSW System. As such, when all RCW/RSW pumps, valves, and piping are OPERABLE, but a branch connection off the main header is isolated, the RCW/RSW System is still OPERABLE.

The 31 day Frequency is based on engineering judgement, is consistent with the procedural controls governing valve l

operation, and ensures correct valve positions.

SR 3.7.3.5 This SR verifies that the automatic isolation valves of the RCW/RSW System will automatically switch to the safety or emergency position to provide cooling water exclusively to the safety related equipment, and limited non-safety related equipment, during an accident event.

This is demonstrated by use of an actual or simulated initiation signal.

This SR also verifies the automatic start capability of the RCW/RSW~

pumps that are in standby and automatic valving in each of the standby RCW/RSW heat exchangers in each division. The c

LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.5.1.4 overlaps this SR to provide complete testing of the safety function.

Operating experience has shown that these components usually pass the SR when performed on the 18 month Frequency.

Therefore, this Frequency is concluded to be acceptable from a reliability standpoint.

l (continued) j ABWR TS B 3.7-4 P&R(LPS), 08/13/93

i i

RCW/RSW System and UHS-Refueling B 3.7.3 BASES REFERENCES 1.

Regulatory Guide 1.27, Revision 2, January 1976.

i 2.

ABWR SSAR, Sections 9.2.11 and 9.2.15.

3.

ABWR SSAR, Section 6.2.1.1.3.3.1.4.

4.

ABWR SSAR, Chapter 15.

i h

h 6

6 6

)

a i

n b

5

?

i l

i ABWR TS B 3.7-5 P&R(LPS), 08/13/93

)

l i

RHR-High Water Level B 3.9.7 8 3.9 REFUELING OPERATIONS B 3.9.7 Residual Heat Removal (RHR)-High Water Level BASES BACKGROUND The purpose of the RHR 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 RHR 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 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 RHR inlet nozzles for subsystems B and C.

The RHR heat exchangers transfer heat to the Reactor Building Cooling System (LC0 3.7.2).

The RHR shutdown cooling mode is manually controlled.

In addition to the RHR subsystems, the volume of water above the reactor pressure vessel (RPV) flange provides a heat sink for decay heat removal.

APPLICABLE With the unit in MODE 5, the RHR System is not required to SAFETY ANALYSES mitigate any events or accidents evaluated in the safety analyses. The RHR System is required for removing decay heat to maintain the temperature of the reactor coolant.

Although the RHR System does not meet a specific criterion _

i of the NRC Policy Statement, it was identified in the NRC Policy Statement as an important contributor to risk reduction. Therefore, the RHR System is included as a Specification.

LCO Only one RHR shutdown cooling subsystem is required to be OPERABLE in MODE 5 with the water level 2 7.0 m (23 ft) i 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. RHR System requirements in MODE 5, with the water level < 7.0 m (23 ft) above the RPV i

(continued)

ABWR TS B 3.9-1 P&R(LPS), 08/13/93 l

s

i RHR-High Water Level B 3.9.7 BASES LCO flange, are given in LCO 3.9.8, " Residual Heat Removal (continued)

(RHR) Low Water Level.

j An OPERABLE RHR shutdown cooling subsystem consists of an i

RHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path.

Additionally, each RHR 2hutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or i

local) in the shutdown cooling mode for removal of decay l

heat. Operation of one subsystem can maintain and reduce the reactor coolant temperature as required. However, to ensure adequate core flow to allow for accurate average i

i reactor coolant temperature and level monitoring, continuous operation is required. A Note is provided to allow a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> exception to shut down the operating subsystem every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

l APPLICABILITY Two RHR shutdown cooling subsystems are required to be OPERABLE in MODE 5. with the water level < 7.0 m (23 ft) l above the top of the RPV flange, to provide decay heat removal.

RHR System requirements in other MODES are covered 4

by LCOs in Section 3.4, Reactor Coolant System (RCS);

i Section 3.5, Emergency Core Cooling Systems (ECCS) and Section 3.6, Containment Systems.

RHR System requirements in MODE 5, with the water level 2 7.0 m (23 ft) above the RPV flange, are given in LC0 3.9.8, " Residual Heat Removal (RHR)-High Water Level."

i 9

ACTIONS A.1 With no RHR shutdown cooling subsystem OPERABLE, an L

1 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 l

result in reduced decay heat removal capability. The I hour Ccmpletion 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 s

(continued) l ABWR TS B 3.9-2 P&R(LPS),08/13/93 i

f RHR--High Water Level B 3.9.7 BASES t

ACTIONS A.] (continued) 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.

t B.I. B.2. B.3. and B.4 If no RHR shutdown coolino subsystem is OPERABLE and an alternate method of decay heat removal is not available in accordance with Required Action A.I, 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.

i Additional actions are required to minimize any potential fission product release to the environment. This includes initiating immediate action to restore the following to OPERABLE status: secondary containment, one standby gas treatment subsystem, and one 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 component are out of service for maintenance or other reasons. It coes 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 OPERABLE.

C.1 and C.2 If no RHR Shutdown Cooling System is in operation, an i

alternate method of coolant circulation is required to be (continued)

ABWR TS B 3.9-3 P&R(LPS),08/I3/93

i RHR--High Water Level B 3.9.7 BASES

\\

ACTIONS C.1 AND C.? (continued) i 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 biing j

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.

i i

SURVEILLANCE SR 3.9.7.1 REQUIREMENTS This Surveillance demonstrates that one RHR subsystem 1s in t

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.

t i

i l

l l

1 t

l (continued)

ABWR TS B 3.9-4 P&R(LPS),08/13/93 l

t

RHR-Low Water Level B 3.9.8 83.9 REFUELING OPERATIONS B 3.9.8 Residual Heat Removal (RHR)-Low Water Level BASES i

BACKGROUND The purpose of the RHR System in MODE 5 is to remove decay heat and sensible heat from the reactor coolant, as required i

by GDC 34.

Each of the three shutdown cooling loops of the RHR 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 nozzle from the reactor vessel.

Each pump discharges the reactor coolant, after it has been j

cooled by circulation through the respective heat exchangers, to the reactor via feedwater line B for subsystem A, and via the individual RHR inlet nozzles for subsystems B and C.

The RHR heat exchangers transfer heat 1

to the Reactor Building Cooling System (LCO 3.7.2).

The RHR shutdown cooling mode is manually controlled.

APPLICABLE With the unit in MODE 5, the RHR System is not required to SAFETY ANALYSES mitigate any events or accidents evaluated in the safety analyses.

The RHR System is required for removing decay heat to maintain the temperature of the reactor coolant.

Although the RHR System does not meet a specific criterion l

of the NRC Policy Statement, it was identified in the NRC Policy Statement as an important contributor to risk reduction., Therefore, the RHR System is retained as a Specification.

LCO In MODE 5 with the water level < 7.0 m (23 ft) above the reactor pressure vessel (RPV) flange both RHR shutdown cooling subsystems must be OPERABLE.

An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path.

Additionally, each RHR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or (continued)

ABWR TS B 3.9-1 P&R(LPS), 08/13/93

RHR-Lou Water Level B 3.9.8 l

BASES LCO local) in the shutdown cooling mode for removal of decay (continued) heat. Operation of one subsystem can naintain and reduce

[

the reactor coolant temperature as required. However, to t

ensure adequate core flow to allow for accurate average reactor coolant temperature and level monitoring, continuous l

operation is required. A Note is provided to allow a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> exception to shut down the operating subsystem every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

I

~

APPLICABILITY Two RHR shutdown cooling subsystems are required to be OTERABLE in MODE 5, with the water level < 7.0 m (23 ft) above the top of the RPV flange, to provide decay heat renoval.

RHR System requirements in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS);

i Section 3.5, Emergency Core Cooling Systems (ECCS) and Section 3.6, Containment Systems. RHR System requirements i

in MODE 5, with the water level 2 7.0 m (23 ft) above the i

j RPV flange, are given in LCO 3.9.8, "Res W al Heat Removal (RHR)--High Water Level."

i ACTIONS A.1 i

With one of the two required RHR shutdown cooling subsystems inoperable, the remaining subsystem is capable of providing l

the required decay heat removal. However, the overall reliability is reduced. Therefore an alternate method of decay heat removal must be provided. With both RHR shutdown _

cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided i

3 for the initial RHR shutdown cooling subsystem i

inoperability.

This re-establishes backup decay heat

(

removal capabilities, similar to the requirements of the LCO.

The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is based on the decay heat removal function and the probability of a loss of the l

available decay heat removal capabilities.

Furthermore, i

verification of the functional availability of these i

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 /> j

thereafter. This will ensure.,ontinued heat removal capability.

l (continued)

{

ABWR TS B 3.9-2 P&R(LPS), 08/13/93 i

[

l l

RHR-Low Water Level i

B 3.9.8 l

t i

BASES ACTIONS A1 (continued)

Alternate decay heat removal methods are available to the operators for review and preplanning in the unit's Operating Procedures.

For example, this may include the use of the Reactor Water Cleanup Sye+?m, operating with the regenerative heat exchanger ;vpassed. The method used to t

remove decay heat should be the most prudent choice based on unit conditions.

i B.l. 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 l

established within I hour.

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 alternate method (other than by the required RHR Shutdown Cooling System), the reactor coolant temperature must be periodically monitored to ensure proper function of the alternate method. The once per hour Completion Time is deemed appropriate.

If at least one RHR subsystem is not restored to OPERABLE status immediately, additional actions are required to minimize any potential fission product release to the environment. This includes initiating immediate action to restore the following to OPERABLE status:

secondary containment, one standby gas treatment subsystem, and one secondary containment isolation valve and associated j

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 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.

ia this case, the surveillance may need to be performed to restore the component to OPERABLE status. Actions must continue until t

all required components are OPERABLE.

i (continued) r ABWR TS B 3.9-3 P&R(LPS),08/13/93

f-RHR--Low Water Level B 3.9.8 BASES (continued)

SURVEILLANCE SR 3.9.9.1 REQUIREMENTS 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 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.

l

\\

ABWR TS B 3.9-4 P&R(LPS),08/13/93 i

SLs 2.0 l

2.0 SAFETY LIMITS (SLs) l 2.1 SLs l

2.1.1 Reactor Core SLs i

2 2.1.1.1 With the reactor steam dome pressure < 55.2 Kg/cm g (785 psig) or core flow < 10% rated core flow:

THERMAL POWER shall be s 25% RTP.

i 2.1.1.2 With the reactor steam dome pressure 2 55.2 Kg/cm g i

(785 psig) and core flow 210% rated core flow:

j i

MCPR shall be 21.07.

t 2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be maintained s 93.1 Kg/cm g 2

(1325 psig).

2.2 SL Violations l

With any SL violation, the following actions shall be completed:

I 2.2.1 Within I hour, notify the NRC Operations Center, in accordance with 10 CFR 50.72.

)

t 2.2.2 Within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.2.1 Restore compliance with all SLs; and t

2.2.2.2 Insert all insertable control rods.

i 2.2.3 Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, notify the [ General Manager--Nuclear Plant and Vice President--Nuclear Operations) and the [offsite reviewers specified in Specification 5.5.2, "[0ffsite) Review and Audit").

i (continued) i ABWR TS 2.0-1 P&R, 8/9/93

SLs 2.0 1

2.0 SLs 5

2.2 SL Violations (continued) 2.2.4 Within 30 days, a Licensee Event Report (LER) shall be prepared pursuant to 10 CFR 50.73. The LER shall be submitted to the NRC, the [offsite reviewers specified in Specification 5.5.2], and the

[ General Manager--Nuclear Plant and Vice President--Nuclear Operations].

2.2.5 Operation of the unit shall not be resumed until authorized by the NRC.

ABWR TS 2.0-2 P&R, 8/9/93 re

i Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs BASES l

BACKGROUND GDC 10 (Ref.1) requires, and SLs ensure, that specified i

acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (A00s).

The fuel cladding integrity SL is set, such that no fuel damage is calculated to occur if the limit is not violated.

Because fuel damage is not directly observable, stepback approach is used to establish an SL, such that the MCPR is i

not less than the limit specified in Specification 2.1.1.2.

MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one of the physical barriers that separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously j

measurable.

Fuel cladding perforations, however, can result from thermal stresses, which occur from reactor operation significantly above design conditions.

While fission product migration from cladding perforation is

~~

t just as measurable as that from use related cracking, the

~

thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than incremental, cladding deterioration.

Therefore, the fuel cladding SL is defined with a margin to the f

conditions that would produce onset of transition boiling (i.e., MCPR - 1.00). These conditions represent a significant departure from the condition intended by design for planned operation. The MCPR fuel cladding integrity SL ensures that during normal operation and during A00s, at least 99.9% of the fuel rods in the core do not experience transition boiling.

(continued)

ABWR TS B 2.0-1 P&R, 08/11/93

.~

Reactor Core SLs B 2.1.1 l

BASES BACKGROUND Operation above the boundary of the nucleate boiling regime (continued) could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient.

Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form.

This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

2 i

i I

i

• PPLICABLE The fuel cladding must not sustain damage as a result of I

i SAFETY ANALYSES normal operation and A00s. The reactor core SLs are established to preclude violation of the fuel design criterion that an MCPR is to be established, such that at least 99.9% of the fuel rods in the core would not be j

expected to experience the onset of transition boiling.

l 4

i The Reactor Protection System setpoints (LCO 3.3.1.1, 3

" Reactor Protection System (RPS) Instrumentation"), in combination with all the LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERMAL POWER level that would result in reaching the MCPR SL.

j 5

l 2.1.1.1 fuel Claddina Inteority (General Electric Corporation (GE) Fuel) i GE critical power correlations are applicable for all 2

critical power calculations at pressures 2 55.2 Kg/cm g (785 psig) or core flows 210% of rated flow.

For operation q

at low pressures and low flows, another basis is used, as i

follows:

i Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be >

1 2

.316 Kg/cm (4.5 psi). Analyses that witp a bundle flow of 12.7 m{Ref. 2) show

/h (28 x 10 lb/hr), bundle pressure drop is nearly (continued)

ABWR TS B 2.0-2 P&R, 08/11/93

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.1 Fuel Claddino Intearity (General Electric SAFETY ANALYSES Corocration (GE) Fuel)

(continued) independen}ofbundlepowerandhasavalueof

.246 Kg/cm Thus, the bundle flow with a.316 Kp/cm[3.5 psi).(4.5pgi)drivingheadwillbe

> 12.7 m /1 (28 x 10 lb/hr).

Full scale ATLAS test data taken at pressupes from 1 Kg/cm'a (14.7 psia) to 56.2 Kg/cm a (800 psia) indicate that the fuel assembly critical power at this ficw is approximately 3.35 Mwt. With the design peaking factors, this corresponds to a THERMAL POWER > 50% RTP. Thus, a THERMAL POWER ljmit of 25% RTP for reactor pressure < 55.2 Kg/cm g (785 psig) is conservative.

2.1.1.2 MCPR (GE Fuel)

The fuel cladding integrity SL is set, such that no significant fuel damage is calculated to occur if the limit is not violated. Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur. Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is

~

calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity SL is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties.

The MCPR SL is determined using a statistical model that combines all the uncertainties in operating (continued)

ABWR TS B 2.0-3 P&R, 08/11/93

Reactor Core SLs B 2.1.1 BASES t

APPLICABLE 2.1.1.2 MCPR (GE Fuel)

(continued)

SAFETY ANALYSES parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved General Electric Critical Power correlations.

Details of the fuel cladding integrity SL calculation are given in Ref. 2.

Ref. 2 also includes a tabulation of the uncertainties used in the determination of the MCPR SL and of the nominal values of the parameters used in the MCPR SL statistical analysis.

2.1.1.3 Reactor Vessel Water level During MODES I and 2, the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability.

With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat.

If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes < 2/3 of the core height.

The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action.

SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to the release of radioactive materials to the environs.

SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel, thus maintaining a coolable geometry.

t (continued)

ABWR TS B 2.0-4 P&R, 08/11/93 1

i Remetor Core SLs B 2.1.1 BASES APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.

However, in MODES 3, 4, and 5, with the reactor shut down, it is unlikely that fuel cladding -

integrity SLs would be violated.

SAFETY LIMIT 2.2.1 VIOLATIONS If any SL is violated, the NRC Operations Center must be notified within I hour, in accordance with 10 CFR 50.72 (Ref. 3).

2.2,2 Exceeding an SL may cause fuel damage and create a potential for radioactive releases in excess of 10 CFR 100, " Reactor Site Criteria," limits (Ref. 4).

Therefore, it is required to insert all insertable control rods and restore compliance with the SL within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.

2.2,3 If any SL is violated, the appropriate senior management of the nuclear plant and the utility shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period provides time for plant operators and staff to take the appropriate immediate action and assess the condition el the unit before reporting to the senior management.

2.2.4 If any SL is violated, a Licensee Event Report shall be prepared and submitted within 30 days to the NRC.

A copy of the report shall also be provided to the senior management of the nuclear plant, and the (continued)

ABWR TS B 2.0-5 P&R, 08/11/93

Reactor Core SLs B 2.1.1 BASES SAFETY LIMIT R.?,4 (continued))

VIOLATIONS (continued) utility Vice President--Nuclear Operations. This requirement is in accordance with 10 CFR 50.73 (Ref. 5).

2.?.5 If any SL is violated, restart of the unit shall not commence until authorized by the NRC.

This requirement ensures the NRC that all necessary reviews, analyses, and actions are completed before the unit begins its restart to normal operation.

REFERENCES 1.

10 CFR 50, Appendix A, GDC 10.

2.

NEDE-24011-P-A, (latest approved revision).

3.

10 CFR 50.72.

4.

10 CFR 100.

5.

10 CFR 50.73.

i ABWR TS B 2.0-6 P&R, 08/11/93

e RCS Pressure SL t

B 2.1.2 i

B 2.0 SAFETY LIMITS (SLs)

B 2.1.2 Reactor Coolant System (RCS) Pressure SL I

BASES l

BACKGROUND The SL on reactor steam dome pressure protects the RCS against overpressurization.

In the event of fuel cladding failure, fission products are released into the reactor coolant.

The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere.

Establishing an upper limit on reactor steam dome pressure ensures continued RCS integrity.

According to 10 CFR 50, Appendix A, GDC 14, " Reactor Coolant Pressure Boundary," and GDC 15, " Reactor Coolant System Design" (Ref.1), the reactor coolant pressure boundary (RCPB) shall be designed with sufficient margin to ensure that the design conditions are not exceeded during normal operation and anticipated operational occurrences (A00s).

During normal operation and A00s, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 2).

To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, in accordance with ASME Code requirements, prior to initial operation when there is no fuel in the core.

Any further hydrostatic testing with fuel in the core is done under LC0 3.10.1, " Inservice Leak and Hydrostatic Testing Operation." Following inception of unit operation, RCS components shall be pressure tested, in accordance with the requirements of ASME Code,Section XI (Ref. 3).

I Overpressurization of the RCS could result in a breach of the RCPB.

If this occurred in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere, raising concerns relative to limits on radioactive releases specified in 10 CFR 100, " Reactor Site Criteria" (Ref. 4).

(continued)

ABWR TS B 2.0-1 P&R, 08/11/93

RCS Pressure SL B 2.1.2 BASES (continued)

APPLICABLE The RCS safety / relief valves and the Reactor Protection SAFETY ANALYSES System Reactor Vessel Steam Dome Pressure--High Function have settings established to ensure that the RCS pressure SL will not be exceeded.

The RCS pressure SL has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered.

The reactor pressure vessel is designed to ASME, Boiler and Pressure Vessel Code,Section III, [1971 Edition] [later Edition], including Addenda through the [ winter of 1972] [later Edition]

(Ref.5),whichpermitsamaximumpressuretransientof 110%,,96.7 Kg/cm 9 (1375 psig), of design pressure 87.9 Kg/cm'g (1250 psig). The SL of 93.1 Kg/cm g (1325 psig), as measuredbythereactorpteamdomepressureindicator,is equivalent to 96.7 Kg/cm g (1375 psig) at the lowest elevation of the RCS.

The RCS is designed to ASME Code,Section III,1974 Edition [later Edition] (Ref. 6), for the reactor recirculation piping, which pen 64ts a maximum i

pressyretransientof110%ofdesignpresscresof87.9 2

Kg/cm g (1250 psig) for suction piping and 116 Kg/cm g (1650 psig) for discharge piping. The RCS pressure SL is selected to be the lowest transient overpressure allowed by the applicable codes.

SAFETY LIMITS The maximum transient pressure allowable in the RCS pressure vessel under the ASME Code,Section III, is 110% of design pressure.

The maximum transient pressure allowable in the RCSpiping,vglves,andfittingsis110%ofdesignpressures-of 87.9 Kg/cm g (1250 psig) for suction piping and 2

105.5 Kg/cm g (1500 psig) for discharge piping. The most limiting of these two allowances is the 110% of design i

pressure; therefore, the SL on maximum allowable RCS 2

pressure is established at 97.7 Kg/cm g (1375 psig).

APPLICABILITY SL 2.1.2 applies in all MODES; however, in MODE 5, because the reactor vessel head closure bolts are rot fully tightened, it is unlikely the RCS would be pressurized.

(continued) i ABWR TS B 2.0-2 P&R, 08/11/93 j

RCS Pressure SL l

B 2.1.2 I

BASES (continued)

SAFETY LIMIT 221 VIOLATIONS i

If any SL is violated, the NRC Operations Center must be notified within I hour, in accordance with 10 CFR 50.72 (Ref. 7).

l 2.2.2 Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for radioactive releases in excess of 10 CFR 100, " Reactor Site Criteria," limits (Ref. 4). Therefore, it is required to insert all insertable control rods and restore compliance with the SL within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action.

2.2.3 If any SL is violated, the appropriate senior management of the nuclear plant and the utility shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period provides time for plant operators and staff to take the appropriate immediate action and assess the condit1oe of the unit before reporting to the senior management.

2,2.4 If any SL is violated, a Licensee Event Repet shall be prepared and submitted within 30 days to the i!RC. A copy of the report shall also be provided to the senior management of the nuclear plant, and the utility Vice President-Nuclear Operations. This requirement is in accordance with 10 CFR 50.73 (Ref. 8).

2.2.5 If any SL is violated, restart of the unit shall not commence until authorized by the NRC. This requirement ensures the NRC that all necessary reviews, analyses, and actions are completed before the unit begfns its restart to normal operation.

(continued)

ABWR TS B 2.0-3 P&R, 08/11/93

RCS Pressure SL B 2.1.2 BASES REFERENCES 1.

10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28.

2.

ASME, Boiler and Pressure Vessel Code,Section III, Article NB-7000.

3.

Boiler and Presture Vessel Code,Section XI, Article IW-5000 t

4.

10 CFR 100.

5.

ASME, Boiler and Pressure Vessel Code, [1971 Edition]

[later Edition], Addenda, [ winter of 1972] [later Edition].

6.

ASME, Boiler and Pressure Vessel Code, [1974 Edition]

[later Edition].

7.

10 CFR 50.72.

8.

10 CFR 50.73.

e-M P

l

+

ABWR TS B 2.0-4 P&R, 08/11/93

LCO Applicability j

3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2 and LCO 3.0.7.

LC0 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.6.

If the LC0 is met or is no longer applicable prior to expiration of the specified Completion Time (s), completion of the Required Action (s) is not required, unless otherwise stated.

LCO 3.0.3 When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided or if directed by the associated ACTION to this LCO, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within I hour to t

place the unit, as applicable, in:

a.

MODE 2 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />; I

b.

H0DE 3 within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />; and c.

MODE 4 within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.

Exceptions to this Specification are stated in the

~

1 individual Specifications.

~

Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required.

LCO 3.0.3 is applicable in MODES 1, 2, and 3.

LC0 3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall not be made except when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the (continued)

ABWR TS 3.0-1 P&R, 8/9/93

^

LCO Applicability 3.0 3.0 LCO APPLICABILITY LCO 3.0.4 Applicability for an unlimited period of time.

This (continued)

Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS.

I Exceptions to this Specification are stated in the individual Specifications. These exceptions allow entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered allow unit operation in the MODE or other specified condition in the Applicability onlv for a limited period of 4

time.

j I

r LC0 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.

LCO 3.0.6 When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LC0 ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system.

In this event, additional evaluations and limitations may be required in

~

accordance with Specification 5.8, " Safety function Determination Program (SFDP)."

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

When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered i

in accordance with LCO 3.0.2.

l 1

(continued)

ABWR TS 3.0-2 P&R, 8/9/93 l

I

LCO Applicability 3.0 3.0 LC0 APPLICABILITY (continued)

LCO 3.0.7 Special Operations LCOs in Section 3.10 allow specified Technical Specifications (TS) requirements to be changed to permit performance of special tests and operations.

Unless otherwise specified, all other TS requirements remain unchanged.

Compliance with Special Operations LCOs is optional. When a Special Operations LCO is desired to be met but is not met, the ACTIONS of the Special Operations LCO shall be met. When a Special Operations LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with the other applicable Specifications.

W l

i i

o I

i ABWR TS 3.0-3 P&R, 8/9/93

SR Applicability l

3.0 t

3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.I SRs shall be met during the MODES or other specified

[

conditions in the Applicability for individual LCOs, unless otherwise stated in the SR.

Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO.

Failure to t

perform a Surveillance within the specified frequency shall be failure to meet the LC0 except as provided in SR 3.0.3.

Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.

j SR 3.0.2 The specified frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the frequency is met.

l For Frequencies specified as "once," the above interval extension does not apply.

If a Completion Time requires periodic performance on a i

"once per..." basis, the above Frequency extension applies to each performance after the initial performance.

i Exceptions to this Specification are stated in the individual Specifications.

i l

L SR 3.0.3 If it is discovered that a Surveillance was not performed

~

within its specified frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is less.

This delay period is permitted to allow performance of the i

Surveillance.

)

If the Surveillance is not performed within the delay l

period, the LCO must immediately be declared not met, and I

the applicable Condition (s) must be entered.

l (continued)

ABWR TS 3.0-4 P&R, 8/9/93 i

t

SR Applicability 3.0 3.0 SR APPLICABILITY i

SR 3.0.3 When the Surveillance is performed within the delay period i

(continued) and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition (s) must be i

entered.

l SR 3.0.4 Entry into a MODE or other specified condition in the Applicability of an LCO shall not be made unless the LCO's Surveillances have been met within their specified i

Frequency. This provision shall not prevent passage through or to MODES or other specified conditions in compliance with 1

Required Actions.

b I

i i

r t

v.

i l

I t

I h

ABWR TS 3.0-5 P&R, 8/9/93 i

f

j LCOs and SRs B 3.0 j

B 3.0 LIMITING CONDITION FOR OPERATION (LCOs) AND SURVEILI'4NCE REQUIREMENTS (SRs)

BASES LCOs LCO 3.0.1 through LCO 3.0.7 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated.

r LCO 3.0.1 i

LCO 3.0.1 establishes the Applicability statement within each individual Specification as the requirement for when the LCO is required to be met (i.e., when the unit is in the MODES or other specified conditions of the Applicability statement of each Specification).

i i

LCO 3.0.2 LC0 3.0.2 establishes that upon discovery of a failure to i

meet an LCO, the associated ACTIONS shall be met.

The Completion Time of each Required Action for an ACTIONS i

Condition is applicable from the point in time that an ACTIONS Condition is entered. The Required Actions establish those remedial measures that must be taken within specified Completion Times when the requirements of an LC0 are not met.

This Specification establishes that:

a.

Completion of the Required Actions within the i

specified Completion Times constitutes compliance with a Specification; and 1

b.

Completion of the Required Actions is not required when an LCO is met within the specified Completion Time, unless otherwise specified.

There are two basic types of Required Actions. The first type of Required Action specifies a time limit in which the LCO must be met. This time limit is the Completion Time to restore an inoperable system or component to OPERABLE status or to restore variables to within specified limits.

If this type of Required Action is not completed within the i

specified Completion Time, a shutdown may be required to I

(continued)

ABWR TS B 3.0-1 P8R, 08/11/93

I LCOs and SRs B 3.0 BASES L

LCOs LCO 3.0.2 (continued) place the unit in a MODE or condition in which the Specification is not applicable.

(Whether stated as a Required Action or not, correction of the entered Condition is an action that may always be considered upon entering ACTIONS.) The second type of Required Action specifies the remedial measures that permit continued operation of the unit that is not further restricted by the Completion Time.

In this case, compliance with the Required Actions provides an acceptable level of safety for continued operation.

Completing the Required Actions is not required when an LCO is met or is no loncer applicable, unless otherwise stated in the individual Specifications.

The nature of some Required Actions of some Conditions necessitates that, once the Condition is entered, the Required Actions must be completed even though the associated Conditions no longer exist. The individual LCO's ACTIONS specify the Required Actions where this is the case.

An example of this is in LCO 3.8.1, "AC Sources-0perating."

The Completion Times of the Required Actions are also applicable when a system or component is removed from service intentionally. The reasons for intentionally relying on the ACTIONS include, but are not limited to, performance of Surveillances, preventive maintenance, corrective maintenance, or investigation of operational problems.

Entering ACTIONS for these reasons must be done in a manner that does not compromise safety.

Intentional entry into ACTIONS should not be made for operational convenience. Alternatives that would not result in redundant equipment being inoperable should be used instead.

Doing so limits the time both subsystems / divisions of a safety function are inoperable and limits the time other conditions exist which result in LC0 3.0.3 being entered.

Individual Specifications may specify a time limit for performing an SR when equipment is removed from service or bypassed for testing.

In this case, the Completion Times of the Required Actions are applicable when this time limit expires, if the equipment remains removed from service or bypassed.

(continued)

ABWR TS B 3.0-2 P&R, 08/11/93

LCOs and SRs i

B 3.0 I

BASES f

LCOs LCO 3.0.2 (continued)

When a change in MODE or other specified condition is i

required to comply with Required Actions, the unit may enter a MODE or other specified condition in which another i

Specification becomes applicable.

In this case, the Completion Times of the associated Required Actions would apply from the point in time that the new Specification becomes applicable and the ACTIONS Condition (s) are entered.

a 1

LCO 3.0.3 l

LCO 3.0.3 establishes the actions that must be implemented

.)

when an LC0 is not met and-a.

An associated Required Action and Completion Time is not met and no other Condition applics; or b.

The condition of the unit is not specifically i

addressed by the associated ACTIONS. This means that no combination of Conditions stated in the ACTIONS can be made that exactly corresponds to the actual condition of the unit.

Sometimes, possible combinations of Conditions are such that entering LC0 3.0.3 is warranted; in such cases, the ACTIONS specifically state a Condition corresponding to such combinations and also that LC0 3.0.3 be entered 1

immediately.

This Specification delineates the time limits for placing the unit in a safe MODE or other specified condition when operation cannot be maintained within the limits for safe operation as defined by the LCO and its ACTIONS.

It is not-1 intended to be used as an operational convenience that permits routine voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant systems or components being i

j inoperable.

~

i t

(continued)

)

ABWR TS B 3.0-3 P&R, 08/11/93

l f

)

LCOs and SRs

.B 3.0 BASES (continued)

LCOs LCO 3.0.3 (continued)

Upon entering LCO 3.0.3, I hour is allowed to prepare for an I

orderly shutdown before initiating a change in unit operation. This includes time to permit the operator to coordinate the reduction in electrical generation with the i

load dispatcher to ensure the stability and availability of I

the electrical grid. The time limits specified to reach lower MODES of operation permit the shutdown to proceed in a controlled and orderly manner that is well within the specified maximum cooldown rate and within the capabilities of the unit, assuming that only the minimum required i

equipment is OPERABLE. This reduces thermal stresses on l

components of the Reactor Coolant System and the potential 4

I for a plant upset that could challenge safety systems under conditions to which this Specification applies. The use and interpretation of specified times to complete the actions of l

I LCO 3.0.3 are consistent with the discussion of Section 1.3, Completion Times.

A unit shutdown required in accordance with LCO 3.0.3 may be i

terminated and LC0 3.0.3 exited if any of the following occurs:

a.

The LCO is now met.

b.

A Condition exists for which the Required Actions have i

j now been performed.

c.

ACTIONS exist that do not have expired Completion i

Times. These Completion Times are applicable from the point in time that the Condition is initially entered i

and not from the time LCO 3.0.3 is exited.

i The time limits of Specification 3.0.3 allow 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> for the unit to be in MODE 4 when a shutdown is required during MODE 1 operation.

If the unit is in a lower MODE of l

operation when a shutdown is required, the time limit for reaching the next lower MODE applies.

If a lower MODE is a

reached in less time than allowed, however, the total allowable time to reach MODE 4, or other applicable MODE, is not reduced.

For example, if MODE 2 is reached in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, j

then the time allowed for reaching MODE 3 is the next 1

(continued) 1 ABWR TS B 3.0-4 P&R, 08/11/93

i LCOs and SRs B 3.0 BASES (continued)

LCOs LCO 3.0.3 (continued) 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />, because the total time for reaching MODE 3 is not reduced from the allowable limit of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />.

Therefore, if remedial measures are completed that would permit a return to MODE 1, a penalty is not incurred by having to reach a i

lower MODE of operation in less than the total time allowed.

In MODES 1, 2, and 3, LCO 3.0.3 provides actions for Conditions not covered in other Specifications.

The requirements of LCO 3.0.3 do not apply in MODES 4 and 5 because the unit is already in the most restrictive Condition required by LCO 3.0.3.

The requirements of LCO j

3.0.3 do not apply in other specified conditions of the i

Applicability (unless in MODE 1, 2, or 3) because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken.

Exceptions to LCO 3.0.3 are provided in instances where requiring a unit shutdown, in accordance with LCO 3.0.3, would not provide appropriate remedial measures for the associated condition of the unit. An example of this is in LC0 3.7.6, " Fuel Pool Water Level." LC0 3.7.6 has an Applicability of "During movement of irradiated fuel assemblies in the associated fuel storage pool." Therefore, this LCO can be applicable in any or all MODES.

If the LCO and the Required Actions of LC0 3.7.6 are not met while in i

MODE 1, 2, or 3, there is no safety benefit to be gained by placing the unit in a shutdown condition. The Required Action of LC0 3.7.6 of " Suspend movement of irradiated fuel assemblies in the associated fuel storage pool (s)" is the appropriate Required Action to complete in lieu of the actions of LC0 3.0.3.

These exceptions are addressed in the individual Specifications.

LC0 3.0.4 LCO 3.0.4 establishes limitations on changes in MODES or other specified conditions in the Applicability when an LCO is not met.

It precludes placing the unit in a different MODE or other specified condition when the following exist:

i a.

The requirements of an LCO, in the MODE or other specified condition to be entered, are not met; and (continued) f ABWR TS B 3.0-5 P&R, 08/11/93 i

t

LCOs and SRs B 3.0 BASES (continued)

LCDs LC0 3.0.4 (continued) b.

Continued noncompliance with these LCO requirements

[

would result in the unit being required to be placed in a MODE or other specified condition in which the LC0 does not apply to comply with the Required Actions.

i Compliance with Required Actions that permit continued operation of the unit for an unlimited period of time in a MODE or other specified condition provides an acceptable level of safety for continued operation. This is without regard to the status of the unit before or after the MODE change.

Therefore, in such cases, entry into a MODE or other specified condition in the Applicability may be made in accordance with the provisions of the Required Actions.

The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before unit startup.

The provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS.

In addition, the provisions of LC0 3.0.4 shall not prevent changes in MODES i

or other specified conditions in the Applicability that result from a normal shutdown.

Exceptions to LC0 3.0.4 are stated in the individual Specifications. Exceptions may apply to all the ACTIONS or i

to a specific Required Action of a Specification.

Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1.

Therefore, changing MODES or other specified conditions while in an ACTIONS Condition, either in compliance with LCO 3.0.4, or where an exception to LC0 3.0.4 is stated, is not a violation of i

SR 3.0.1 or SR 3.0.4 for those Surveillances that do not have to be performed due to the associated inoperable equipment. However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO.

j (continued)

ABWR TS B 3.0-6 P&R, 08/11/93

LCOs and SRs B 3.0 BASES (continued)

)

LCO 3.0.5 l

l LCO 3.0.5 establishes the allowance for restoring equipment t

to service under administrative controls when it has been i'

removed from service or declared inoperable to comply with ACTIONS. The sole purpose of this Specification is to l

provide an exception to LCO 3.0.2 (e.g., to not comply with i

the applicable Required Action (s)) to allow the performance i

of SRs to demonstrate:

~

a.

The OPERABILITY of the equipment being returned to service; or b.

The OPERABILITY of other equipment.

The administrative controls ensure the time the equipment is returned to service in conflict witi, the requirements of the ACTIONS is limited to the time absolutely necessary to perform the allowed SRs. This Specifi.: tion does not provide time to perform any other preveitivc or corrective

]

maintenance.

An example of demonstrating the OPERABILITY of the equipment being returned to service is reopening a containment isolatirn valve that has been closed to comply with Required l

Actions, and must be reopened to perform the SRs.

An example of demonstrating the OPERABILITY of other equipment is taking an inoperable channel or trip system out of the tripped condition to prevent the trip function from occurring during the performance of an SR on another channel _

i in the other trip system. A similar example of I

demonstrating the OPERABILITY of other equipment is taking an inoperable channel or trip system out of the tripped j

condition to permit the logic to function and indicate the appropriate response during the performance of an SR on another channel in the same trip system.

LCO 3.0.6 1

i LCO 3.0.6 establishes an exception to LC0 3.0.2 for support i

systems that have an LCO specified in the Technical Specifications (TS). This exception is provided because i

4 LCO 3.0.2 would require that the Conditions and Required e

1 j

(continued) a ABWR TS B 3.0-7 P&R, 0B/11/93 i

LCOs and SRs B 3.0 t

2 BASES (continued)

I 4

J LCOs LCO 3.0.6 (continued) i Actions of the associated inoperable supported system LCO be entered solely due to the inoperability of the support system. This exception is.iustified because the actions that are required to ensure tne plant is maintained in a L

safe condition are specified in the support system LCO's Required Actions. These Required Actions may include

(

entering the supported system's Conditions and Required Actions or may specify other Required Actions.

I When a support system is inoperable and there is an LCO specified for it in the TS, the supported system (s) are required to be declared inoperable if determined to be inoperable as a result of the support system inoperability.

However, it is not necessary to enter into the supported systems' Conditions and Required Actions unless directed to do so by the support system's Required Actions. The potential confusion and inconsistency of requirements i

related to the entry into multiple support and supported systems' LCOs' Conditions and Required Actions are 1

eliminated by providing all the actions that are necessary 4

to ensure the plant is maintained in a safe condition in the i

support system's Required Actions.

l However, there are instances where a support system's Required Action may either direct a supported system to be i

declared inoperable or direct entry into Conditions andRequired Actions for the supported system. This may i

occur immediately or after some specified delay to perform some other Required Action. Regardless of whether it is j

immediate or after some delay, when a support system's t

Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LC0 3.0.2.

Specification 5.8, " Safety Function Determination Program" (SFDP), ensures loss of safety function is detected and appropriate actions are taken. Upon failure to meet two or more LCOs concurrently, an evaluation shall be made to i

determine if loss of safety function exists. Additionally, other limitations, remedial actions, or compensatory actions may be identified as a result of the support system (continued)

ABWR TS B 3.0-8 P&R, 08/11/93 i

LCOs and SRs B 3.0 i

BASES (continued)

LCOs LCO 3.0.6 (continued) f f

inoperability and corresponding exception to entering supported system Conditions and Required Actions.

The SFDP implements the requirements of LC0 3.0.6.

Cross division checks to identify a loss of safety function for those support systems that support safety systems are required. The cross division check verifies that the supported systems of the redundant OPERABLE support system are OPERABLE, thereby ensuring safety function is retained.

If this evaluation determines that a loss of safety function exists, the appropriate Conditions and Required Actions of the LC0 in which the loss of safety function exists are required to be entered.

LCO 3.0.7 1

There are certain special tests and operations required to be performed at various times over the life of the unit.

I These special tests and operations are necessary to demonstrate select unit performance characteristics, to perform special maintenance activities, and to perform i

special evolutions. Special Operations LCOs in Section 3.10 allow specified TS requirements to be changed to permit performances of these special tests and operations, which otherwise could not be performed if required to comply with the requirements of these TS.

Unless otherwise specified, all the other TS requirements remain unchanged. This will ensure all appropriate requirements of the HDDE or other specified condition not directly associated with or required to be changed to perform the special test or operation will i

remain in effect.

The Applicability of a Special Operations LC0 represents a condition not necessarily in compliance with the normal requirements of the TS.

Compliance with Special Operations LCOs is optional. A special operation may be performed either under the provisions of the appropriate Special Operations LCO or under the other applicable TS 4

requirements.

If it is desired to perform the special j

operation under the provisions of the Special Operations LCO, the requirements of the Special Operations LC0 shall be followed. When a Special Operations LCO requires another (continued)

)

l ABWR TS B 3.0-9 P&R, 08/11/93

=_.

LCOs and SRs B 3.0 BASES (continued) l LCOs LCO 3.0.7 (continued)

LCO to be met, only the requirements of the LC0 statement are required to be met regardless of that LCO's Applicability (i.e., should the requirements of this other i

LCO not be met, the ACTIONS of the Special Operations LCO i

apply, not the ACTIONS of the other LCO).

However, there i

are instances where the Special Operations LC0 ACTIONS may i

direct the other LCOs' ACTIONS be met. The Surveillances of the other LCO are not required to be met, unless specified in the Special Operations LCO.

If conditions exist such t

that the Applicability of any other LC0 is met, all the j

other LCO's requirements (ACTIONS and SRs) are required to i

be met concurrent with the requirements of the Special Operations LCO.

j i

a SURVEILLANCE SR 3.0.1 through SR 3.0.4 establish the general requirements REQUIREMENTS applicable to all Specifications and apply at cll times, unless otherwise stated.

a e

1 SR 3.0.1 SR 3.0.1 establishes the requirement that SRs must be met l

during the MODES or other specified conditions in the Applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. This Specification is to ensure that Surveillances are performed to verify the OPERABILITY of systems and components, and that variables are within specified limits.

Failure to mee(_

a Surveillance within the specified Frequency, in accordance with SR 3.0.2, constitutes a failure to meet an LCO.

1 Systems and components are assumed to be OPERABLE when the associated SRs have been met. Nothing in this l

1 Specification, however, is to be construed as implying that l

systems or components are OPERABLE when:

j i

4 a.

The systems or components are known to be inoperable, I

although still meeting the SRs; or b.

The requirements of the Surveillance (s) are known to be not met between required Surveillance performances.

r (continued) i 2

ABWR TS B 3.0-10 P&R, 08/11/93 i

t

I J

LCOs and SRs l

B 3.0 i

i BASES 1

1 SURVEILLANCE SR 3.0.1 (continued)

REQUIREMENTS Surveillances do not have to be performed when the unit is in a MODE or other specified condition for which the a

requirements of the associated LCO are not applicable, unless otherwise specified. The SRs associated with a 1

Special Operations LC0 are only applicable when the Special Operations LC0 is used as an allowable exception to the requirements of a Specification.

Surveillances, including Surveillances invoked by Required J

Actions, do not have to be performed on inoperable equipment i

because the ACTIONS define the remedial measures that apply.

j Surveillances have to be met and performed in accordance with SR 3.0.2, prior to returning equipment to OPERABLE l

status.

I i

Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This e

includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with i

SR 3.0.2.

Post maintenance testing may not be possible in the current MODE or other specified conditions in the Applicability due to the necessary unit parameters not l

having been established.

In these situations, the equipment

]

may be considered OPERABLE provided testing has been l

satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function. This will allow operation to proceed to a MODE cr other specified condition where other necessary post maintenance tests can be completed.

Some examples of this process are:

Control rod drive maintenance during refueling that a.

requires scram testing at > 800 psi.

However, if other appropriate testing is satisfactorily completed and the scram time testing of SR 3.1.3.4 is satisfied, i

the control rod can be considered OPERABLE. This allows startup to proceed to reach 800 psi to perform other necessary testing.

l b.

High pressure core flooder (HPCF) maintenance uuring shutdown that requires system functional tests at a specified pressure.

Provided other appropriate testing is satisfactorily completed, startup can (continued)

ABWR TS B 3.0-11 P&R, 08/11/93

i LCOs and SRs

.B 3.0 BASES SURVEILLANCE SR lad (continued)

REQUIREMENTS proceed with HPCF considered OPERABLE.

This allows operation to reach the specified pressure to complete the necessary post maintenance testing.

SR 3.0.2 SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action with a Completion Time that requires the periodic performance of the Required Action on a "once per..."

interval.

SR 3.0.2 permits a 25% extension of the interval specified in the Frequency. This extension facilitates Surveillance scheduling and considers plant operating conditions that may not be suitable for conducting the Surveillance (e.g., transient conditions or other ongoing Surveillance or maintenance activities).

The 25% extension does not significantly degrade the reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs. The exceptions to SR 3.0.2 are those Surveillances for which the 25% extension of the interval specified in the Frequency does not apply. These exceptions are stated in the individual Specifications. An example of where SR 3.0.L does not apply is a Surveillance with a frequency of "in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions." The requirements of regulations take precedence over the TS. The TS cannot in and of themselves extend a test interval specified in the regulations.

Therefore, there is a Note in the Frequency stating, "SR 3.0.2 is not applicable."

As stated in SR 3.0.2, the 25% extension also does not apply to the initial portion of a periodic Completion Time that requires performance on a "once per..." basis. The 25%

extension applies to each performance after the' initial performance. The initial performance of the Required Action, whether it is a particular Surveillance or some (continued) 1 ABWR TS B 3.0-12 P&R, 08/11/93

LCOs and SRs 83.0 4

BASES SURVEILLANCE SR 3.0.2 (continued)

REQUIREMENTS other remedial action, is considered a single action with a single Completion Time. One reason for not allowing the 25%

extension to this Completion Time is that such an action i

usually verifies that no loss of function has occurred by checking the status of redundant or diverse components or accomplishes the function of the inoperable equipment in an alternative manner.

The provisions of SR 3.0.2 are not intended to be used repeatedly merely as an operational convenience to extend Surveillance intervals or periodic Completion Time intervals beyond those specified.

i SR 3.0.3 e

SR 3.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable 1

outside the specified limits when a Surveillance has not been completed within the specified frequency. A delay period of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time i

that the specified Frequency was not met. This delay period provides adequate time to complete Surveillances that have been missed. This delay period permits the completion of a Surveillance before complying with Required Actions or other remedial measures that might preclude completion of the Surveillance.

l The basis for this delay period includes consideration of i

unit cenditions, adequate planning, availability of personnel, the time required to perform the Surveillance, the safety significance of the delay in completing the required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements.

When a Surveillance with a frequency based not on time 1

intervals, but upon specified unit conditions or operational l

situations, is discovered not to have been performed when (continued) t ABWR TS B 3.0-13 P&R, 08/11/93

i LCOs and SRs B 3.0 i

BASES SURVEILLANCE SR 3.0.3 (continued)

REQUIREMENTS specified, SR 3.0.3 allows the full delay period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to perform the Surveillance.

SR 3.0.3 also provides a time a

limit for completion of Surveillances that become applicable as a consequence of MODE changes imposed by Required Actions.

Failure to comply with specified Frequencies for SRs is expected to be an infrequent occurrence.

Use of the delay period established by SR 3.0.3 is a flexibility which is not intended to be used as an operational convenience to extend Surveillance intervals.

If a Surveillance is not completed within the allowed delay i

period, then the equipment is considered inoperable or the variable then is considered outside the specified limits and the Completion Times of the Required Actions for the 2

applicable LC0 Conditions begin immediately upon expiration of the delay period.

If a Surveillance is failed within the delay period, then the equipment is inoperable, or the variable is outside the specified limits and the Completion Times of the Required Actions for the applicable Conditions begin immediately upon the failure of the Surveillance.

Completion of the Surveillance within the delay period allowed by this Specification, or within the Completion Time of the ACTIONS, restores compliance with SR 3.0.1.

SR 3.0.4 SR 3.0.4 establishes the requirement that all appliccble SRs must be met before entry into a MODE or other specified 4

condition in the Applicability.

This Specification ensures that system and component OPERABILITY requirements and variable limits are met before entry into MODES or other specified conditions in the Applicability for which these systems and components ensure safe operation of the unit. This Specification applies to changes in MODES or other specified conditions in the i

Applicability associated with unit shutdown as well as startup.

4 (continued)

)

t i

ABWR TS B 3.0-14 P&R, 08/11/93 l

LCOs and SRs B 3.0 BASES SURVEILLANCE SR 3.0.4 (continued)

{

REQUIREMENTS The provisions of SR 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability l

that are required to comply with ACTIONS.

The precise requirements for performance of SRs are specified such that exceptions to SR 3.0.4 are not necessary. The specific time frames and conditions j

necessary for meeting the SRs are specified in the frequency, in the Surveillance, or both. This allows performance of Surveillances when the prerequisite condition (s) specified in a Surveillance procedure require entry into the MODE or other specified condition in the l

Applicability of the associated LC0 prior to the performance j

or completion of a Surveillance. A Surveillance that could i

not be performed until after entering the LC0 Applicability would have its Frequency specified such that it is not "due" until the specific conditions needed are met.

Alternately, l

the Surveillance may be stated in the form of a Note as not required (to be met or performed) until a particular event, l

i condition, or time has been reached.

Further discussion of the specific formats of SRs* annotation is found in

)

Section 1.4, Frequency.

l 1

i e

E b

i 5

i 1

h i

l

}

l l

I

)

t ABWR TS B 3.0-15 P&R, 08/11/93 i

i SDM 3.1.1 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM) l LCO 3.1.1 SDM shall be:

a.

2 0.38% ak/k, with the highest worth control rod or rod pair analytically determined; or b.

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

APPLICABILITY:

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

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

limits.

l B.

Required Action and B.1 Be in MODE 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not met.

i 4

~

C.

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

insertable control

rods, e

D.

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

insertable control rods.

i AND (continued) l ABWR TS 3.1-1 P&R, 8/9/93

SDM 3.1.1 i

i ACTIONS

^

CONDITION REQUIRED ACTION COMPLETION TIME D.

(continued)

D.2 Initiate action to I hour restore secondary containment to OPERABLE status.

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

AND D.4 Initiate action to I hour restore one isolation valve and associated

^

instrumentation to OPERABLE status in each secondary containment penetration flow path not isolated.

E.

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

ALTERATIONS except for control rod insertion and fuel assembly removal.

AND (continued) l f

l 1

ABWR TS 3.1-2 P&R, 8/9/93 t

SDM 3.1.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E.

(continued)

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

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

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

AND E.5 Initiate action to I hour restore one isolation valve and associated instrumentation to OPERABLE status in each secondary

~~

containment

~

penetration flow path not isolated.

I ABWR TS 3.1-3 P&R, 8/9/93 I

l

SDM 3.1.1 SURVEILLANCE REQUIREMENTS l

SURVEILLANCE FREQUENCY t

SR 3.1.1.1 Verify SDM is:

Prior to each in vessel fuel a.

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

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

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

i i

• ~

e b

l G

E

'I O

b l

I ABWR TS 3.1-4 P&R, 8/9/93

.l

l Reactivity Anomalies 3.1.2 i

3.1 REACTIVITY CONTROL SYSTEMS 3.1.2 Reactivity Anomalies i

i LC0 3.1.2 The reactivity difference between the monitored core k,,, and the predicted core k,,, shall be within i 1% h.k/k.

l i

APPLICABILITY:

MODES I and 2.

Acil0NS CONDITION REQUIRED ACTION COMPLETION TIME A.

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

to within limit.

i B.

Required Action and B.1 Be in MODE 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> i

associated Completion Time not met.

~

l i

i ABWR TS 3.1-1 P&R, 8/9/93 f

Reactivity Anomalies 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.2.1 Verify core reactivity difference between Once within i

the monitored core k,,, and the predicted 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after core k,,, is within 1% ok/k.

reaching l

equilibrium conditions following startup after fuel movement or control rod replacement l

within the reactor pressure vessel AND i

1000 MWD /T l

thereafter 4

i 1

c d

o-i ABWR TS 3.1-2 P&R, 8/9/93

Control Rod OPERABILITY 3.1.3 l

3.1 REACTIVITY CONTROL SYSTEMS 3.I.3 Control Rod OPERABILITY LCO 3.I.3 Each control rod shall be OPERABLE.

APPLICABILITY:

MODES I and 2.

ACTIONS

---

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

Separate Condition entry is allowed for each control rod.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One withdrawn control

---

NOTE----

==

-=

rod stuck.

A stuck rod may be bypassed in the Rod Action and

. Position Information (RAPI)

System in accordance with SR 3.3.2.I.6 required to allow continued operation.

---

_ --_=------

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

(continued)

F l

l l

ABWR TS 3.1-1 P&R, 8/9/93 l

l r

Control Rod OPERABILITY 3.1.3 t

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.2

---

NOTE---------

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

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

AND A.3 Perform SR 3.1.1.1.

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

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

CRD.

6.!!Q B.2 Be in MODE 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

[

(continued) b i

ABWR TS 3.1-2 P&R, 8/9/93

Control Rod OPERABILITY 3.1.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

One or more control

---

NOTE---------

rods inoperable for 1.

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

in RAPI in accordance with SR 3.3.2.1.6, if required, to allow insertion of inoperable control rod and continued operation.

2.

Inoperable control rods with failed motor drives can only be fully inserted by individual scram.

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

D.

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

D.1 Restore compliance 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />

~

Not applicable when with GWSR.

THERMAL POWER

> 10% RTP.

OR

- -----__= - ____

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

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

(continued)

I ABWR TS 3.1-3 P&R, 8/9/93

Control Rod OPERABILITY 3.1.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E.

Required Action and E.1 Be in MODE 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A, C, or D not met.

OB Nine or more control rods inoperable.

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

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

SR 3.I.3.2

-NOTE-------

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

---

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

_ =.

Insert each fully withdrawn control rod 7 days two notches.

i 9

9 l

ABWR TS 3.1-4 P&R, 8/9/93 I

)

f Control Rod OPERABILITY 3.1.3 t

SURVEILLANCE REOUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.3.3

---

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

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

_______________--=- -

Insert each partially withdrawn control rod 31 days two notches.

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

SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 (continued) l SR 3.1.3.5 Verify each control rod does not go to the Once the first withdrawn overtravel position.

time the control rod is withdrawn to

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

declaring control rod OPERABLE after I

work on control rod or CRD System that could affect coupling

-1 l

1 ABWR TS 3.1-5 P&R, 8/9/93 i

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

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

in accordance with Table 3.1.4-1; and b.

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

APPLICABILITY:

MODES I and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Requirements of the A.1 Be in MODE 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> LCO not met.

1 SURVEILLANCE REQUIREMENTS

---

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

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

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

~~

SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify each control rod scram time is Prior to l

within the limits of Table 3.1.4-1 with exceeding a

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

fuel movement l

within the reactor pressure vessel eM1 (continued)

ABWR TS 3.1-1 P&R,8/9/93 i

Control Rod Scram Times t

3.1.4 SURVEILLANCE REQUIREMENTS

~

SURVEILLANCE FREQUENCY l

SR 3.1.4.1 (continued)

Prior to exceeding 40% RTP after each reactor shutdown i

2 120 days i

SR 3.1.4.2 Verify, for a representative sample, each 120 days tested control rod scram time is within the cumulative limits of Table 3.1.4-1 with reactor steam operation in dome pressure 2 66.8 Kg/cm g (950 psig).

MODE 1 i

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

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

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

reactorsteamdomepressure266.8Kg/cm}g 40% RTP after (950 psig).

work on control rod or CRD System that could affect scram time ABWR TS 3.1-2 P&R, 8/9/93

i Control Rod Scram Times 3.1.4 i

Table 3.1.4-1 Control Rod Scram Times

---

.---------------_-----------NOTES------------------

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

1.

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

2.

Control rods with scram times > [

] seconds to 60% rod insertion position are inoperable, in accordance with SR 3.3.3.4, and are not considered

" slow."

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

SCRAM TIMES (a)

(seconds)

REACTOR REACTOR REACTOR STEAM DOME STEAM DOME STEAMDOMg)

ROD POSITION PRESSURE (b)

PRESSURE (b}g PRESSURE (

2 66.8 Kg/cm 73.8 Kg/cm g PERCENT INSERTION O Kg/cm g 2

(%)

(0 psig)

(950 psig)

(1050 psig) 10 (c)

[

]

[

]

40 (c)

[

]

[

]

60

[ ]

[

]

[

]

l

?

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

(b) For immediate reactor steam dome pressures, the scram time criteria are L

determined by linear interpolation.

(c)

For reactor steam dome pressure s 66.8 Kg/cm g (950 psig), only 60% rod 2

insertion position scram time limit applies.

?

[

ABWR TS 3.1-3 P&R, 8/9/93 i

i control Rod Scram Accumulators 3.1.5 3.1 REACTIVITY CONTROL SYSTEMS 3.1.5 Control Rod Scram Accumulators LCO 3.1.5 Each control rod scram accumulator shall be OPERABLE.

APPLICABILITY:

MODES I and 2.

ACTIONS

---

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

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

CONDITION REQUIRED ACTION COMPLETION TIME A.

One control rod scram A.1 Declare the 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> accumulator associated control inoperable.

rod (s) inoperable.

(continued) l I

i i

t t

ABWR TS 3.1-1 P&R, 8/9/93 I

i

Control Rod Scram Accumulators 3.1.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.

Two or more control B.1 Declare the I hour rod scram accumulators associated control inoperable.

rod (s) inoperable.

f C.

Required Action and C.1

---

NOTE==--

=

associated Completion Not applicable if all i

Time not met.

inoperable control rod scram accumulators are associated with fully inserted control 4

rods.

---

===--

Place the reactor Immediately mode switch in the shutdown position.

b SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

~~

SR 3.1.5.1 Verify each control rod scram accumulator 7 days 2

pressure is 1130 Kg/cm g (1850 psig).

5 ABWR TS 3.1-2 P&R, 8/9/93

I Rod Pattern Control i

3.1.6 i

3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 Rod Pattern Control P

LCO 3.1.6 OPERABLE control rods shall comply with the requirements of f

the Ganged Withdrawal Sequence Restrictions (GWSR).

\\

.)

APPLICABILITY:

H0 DES I and 2 with THERMAL POWER s 10% RTP.

i

?

ACTIONS CONDITION REQUIRED ACTION CCAPLETION TIME

[

A.

One or more OPERABLE A.1

---

NOTE---------

control rods not in Affected control rods compliance with GWSR.

may be bypassed in Rod Action and Position Information (RAPI) System in accordance with SR 3.3.2.1.6.

---

=- _=__ ----

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

_0_R.

8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> l

i A.2 Declare associated

+

control rod (s) inoperable.

d j

(continued)

{

i t

i o

a I

i e

ABWR TS 3.1-1 P&R, 8/9/93 i

Rod Pottern Control 3.1.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.

Nine or more OPERABLE B.1

---

NOTE---------

control rods not in Affected control rods compliance with GWSR.

may be bypassed in RAPI in accordance with SR 3.3.2.1.6 for insertion only.

Suspend withdrawal of Immediately i

control rods.

AND B.2 Place the reactor I hour mode switch in the 4

shutdown position.

i I

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY j

i SR 3.1.6.1 Verify all OPERABLE control rods comply 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with GWSR.

i k

l i

i I

P 1

3 8

ABWR TS 3.1-2 P&R, 8/9/93 1

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

APPLICABILITY:

MODES I and 2.

ACTIONS i

CONDITION REQUIRED ACTION COMPLETION TIME l

A.

Concentration of A.1 Restore concentration 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> boron in solution of boron in solution j

not within limits to within limits.

AND

-t but > [ ).

10 days from discovery of

)

failure to meet the LCO l

B.

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

reasons other than 6ND Condition A).

10 days from discovery of t

failure to meet the LC0 C.

Two SLC subsystems C.1 Restore one SLC 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> i

4 inoperable [for subsystem to OPERABLE i

reasons other than status.

Condition A].

i D.

Required Action and D.1 Be in MODE 3.

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

ABWR TS 3.1-1 P&R, 8/9/93-

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify available volume of sodium 24 nours 2

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

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

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

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

AND Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after water or boron is added to solution AED Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after solution temperature is restored within the limits of Figure 3.1.7-1 (continued) i ABWR TS 3.1-2 P&R, 8/9/93

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)

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

SR 3.1.7.6 Verify each pump develops a flow rate In accordance 2 [41.2] gpm at a discharge pressure with the 2 [1300] psig.

Inservice Testing Program or 92 days SR 3.1.7.7 Verify flow through one SLC subsystem from

[18] months on pump into reactor pressure vessel, a STAGGERED TEST BASIS SR 3.1.7.8 Verify that simultaneous operation of both

[18] months pumps develop a flow rate 6.301/s (100 gpm) at a pressure of SI units (1223 psig).

SR 3.1.7.9 Verify all heat traced piping between

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

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

[ Figure 3.1.7-1]

ABWR TS 3.1-3 P&R, 8/9/93

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY t

(continued)

SR 3.1.7.10 Verify sodium pentaborate enrichment is Prior to i

2 [60.0) atom percent B-10.

addition to SLC tank i

'1 i

i i

M*

d b

a t

I 1

ABh'R TS 3.1 4 P&R, 8/9/93 1

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i SDM B 3.1.1 i

B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM)

BASES BACKGROUND SDM requirements are specified to ensure:

The reactor can be made subcritical from all operating a.

conditions and transients and Design Basis Events; b.

The reactivity transients associated with postulated accident conditions are controllable within acceptable limits; and c.

The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

These requirements are satisfied by the control rods, as described in GDC 26 (Ref. 1), which can compensate for the reactivity effects of the fuel and water temperature changes e7perienced during all operating conditions.

APPLICABLE The control rod removal error during refueling accident SAFETY ANALYSES analysis (Ref. 2) assumes the core is subtritical with the highest worth control rod withdrawn. The analysis of this reactivity insertion event assumes the refueling interlocks are OPERABLE when the reactor is in the refueling mode of operation.

These interlocks prevent the withdrawal of more than one control rod, or control rod pair, from the core l

~

during refueling.

(Special consideration and requirements for multiple control rod withdrawal during refueling are covered in Special Operations LCO 3.10.6, " Multiple Control Rod Withdrawal-Refueling.") The analysis assumes this condition is acceptable since the core will be shut down with the highest worth control rod or rod pair withdrawn, if adequate SDM has been demonstrated.

Prevention or mitigation of reactivity insertion events is necessary to limit energy deposition in the fuel to prevent significant fuel damage, which could result in undue release of radioactivity (see Bases for LCO 3.1.7, " Standby Liquid (continued)

P I

ABWR TS B 3.1-1 P&P, 08/11/93

SDM 4

B 3.1.1 BASES APPLICABLE Control (SLC) System"). Adequate SDM ensures inadvertent SAFETY ANALYSES criticalities will not cause significant fuel damage.

(continued)

SDM satisfies Criterion 2 of the NRC Policy Statement.

LC0 The specified SDM limit accounts for the uncertainty in the demonstration of SDM by testing. Separate SDM limits are provided for testing where the highest worth control rod or rod pair is determined analytically or by measurement. This is due to the reduced uncertainty in the SDM test when the highest worth control rod or rod pair is determined by measurement. When SDM is demonstrated by calculations not associated with a test, additional margin must be added to l

the specified SDM limit to account for uncertainties in the l

calculation.

To ensure adequate SDM during the design process, a design margin is included to account for uncertainties in the design calculations (Ref. 3).

4 APPLICABILITY In MODES 1 and 2, SDM must be provided because subtriticality with the highest worth control rod or rod pair withdrawn is assumed in the analysis (Ref. 4).

In MODES 3 and 4, SDM is required to ensure the reactor will be held sub:ritical with margin for a single withdrawn control rod or rod pair. SDM is required in MODE S to prevent an inadvertent criticality during the withdrawal of a single control rod from a core cell containing one or more fuel assemblies or of a control rod pair from loaded core cells during scram time testing.

~

4 i

ACTIONS Ad With SDM not within the limits of the LCO in MODE 1 or 2, SDM must be restored within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Failure to meet the specified SDM may be caused by a control rod that cannot be inserted. The 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Completion time is acceptable, considering that the reactor can still be shut down, assuming no additional failures of control rods to insert, 1

and the low probability of an event occurring during this interval.

(continued)

ABWR TS B 3.1-2 P&R, 08/11/93

SDM l

B 3.1.1 BASES ACTIONS B.1 (continued)

If the SDM cannot be restored, the plant must be brought to MODE 3 within 12 hoars, to prevent the potential for further reductions in available SDM (e.g., additional stuck control rods).

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

El With SDM not within limits in MODE 3, the operator must fully insert all insertable control rods within I hour.

This action results in the least reactive condition for the core.

The allowed Completion Time of I hour is acceptable, considering the reactor can still be shut down, assuming no failures of additional control rods to insert.

D.I. D.2. D.3. and D.4 With SDM not within limits in MODE 4, the operator must insert all insertable control rods in I hour. This action results in the least reactive condition for the core.

The I hour Completion Time provides sufficient time to take corrective action and is acceptable, considering the reactor can still be shut down assuming no failures of additional 43 control rods to insert.

Actions must also be initiated

._ jy within I hour to orovide means for control of potential

<adioactive releases. This includes ensuring secondary containment (LCO 3.6.4.1, " Secondary Containment") is OPERABLE; at least one Standby Gas Treatment (SGT)

(LCO 3.6.4.3, " Standby Gas Treatment (SGT) System")

subsystem is OPERABLE; cnd at least one secondary containment isolation valve (LC0 3.6.4.2, " Secondary Containment Isolation Valves (SCIVs)") and associated instrumentation (LCO 3.3.6.1, " isolation Instrumentation")

are OPERABLE in each associated penetration flow path not isolated.

This may be performed as an administrative check, by examining logs or other information, to determine if the components are out of service for maintenance or other reasons.

It is not necessary to perform the SRs needed to demonstrate the OPERABILITY of the components.

If, however, any required component is inoperable, then it must be (continued)

ABWR TS B 3.1-3 P&R, 08/11/93

SDM B 3.1.1 BASES i

t F

ACTIONS D.1. D.2. D.3. and D.4 (continued) restored to OPERABLE status.

In this case, SRs may need to be performed to restore the component to OPERABLE status.

l Actions must continue until all required components are OPERABLE.

E.1. E.2. E.3. E.4. and E.5 With SDM not within limits in MODE 5, the operator must immediately suspend CORE ALTERATIONS that could reduce SDM.

The suspensions are on insertion of fuel in the core or the withdrawal of control rods.

Suspension of these activities shall not preclude completion of movement of a component to J

t a safe condition.

Inserting control rods or removing fuel from the core will reduce the total reactivity and are therefore excluded from the suspended actions.

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

Action must also be initiated within I hour to provide means for control of potential radioactive releases.

This includes ensuring secondary containment (LC0 3.6.4.1) is OPERABLE; at least one SGT subsystem (LC0 3.6.4.3) is

~

I OPERABLE; and at least one secondary containment isolation valve (LCO 3.6.4.2) and associated instrumentation i

(LCO 3.3.6.1) are OPERABLE in each associated penetration flow path not isolated. This may be performed as an administrative check, by examining logs or other information, to determine if the components are out of service for maintenance or other reasons.

It is not necessary to perform the SRs needed to demonstrate the t

OPERABILITY of the components.

If, however, any required component is inoperable, then it must be restored to OPERABLE status.

In this case, SRs may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are-OPERABLE.

(continued)

~

ABWR TS B 3.1-4 P&R, 08/11/93

i i

SDM B 3.1.1 BASES i

l SURVEILLANCE SR 3.1.1.1 REQUIREMENTS 4

Adequate SDM must be demonstrated to ensure the reactor can f

be made subtritical from any initial operating condition.

l Adequate SDM is demonstrated by testing before or during the i

first startup after fuel movement, control rod replacement, l

or shuffling within the reactor pressure vessel.

Control l

rod replacement refers to the decoupling and removal of a l

control rod from a core location, and subsequent replacement i

with a new control rod or a control rod from another core location. Since core reactivity will vary during the cycle as a function of fuel depletion and poison burnup, the beginning of cycle (B0C) test must also account for changes

)

in core reactivity during the cycle.

Therefore, to obtain i

the SDM, the initial measured value must be increased by an adder. "R", which is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated BOC core reactivity.

If the value of R t

is negative (i.e., BOC is the most reactive point in the cycle), no correction to the B0C measured value is required 3

(Ref. 4).

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

testing. Local critical tests require the withdrawal of out of sequence control rods. This testing would therefore require bypassing of the Rod Worth Minimizer to allow the out of sequence withdrawal, and therefore additional requirements must be met (see LC0 3.10.7, " Control Rod Testing--Operating").

~

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

During MODE 5, adequate SDM is also required to ensure the reactor does not reach criticality during control rod withdrawals. An evaluation of each in vessel fuel movement during fuel loading (including shuffling fuel within the core) shall be performed to ensure adequate SDM is maintained during refueling. This evaluation ensures the intermediate loading patterns are bounded by the safety analyses for the final core loading pattern.

For example, bounding analyses that demonstrate adequate SDM for the most (continued)

ASWR TS B 3.1-5 PAR, 08/11/93

SDM B 3.1.1 BASES i

i SURVEILLANCE SR 3.1.1.1 (continued)

REQUIREMENTS t

i reactive configurations during the refueling may be performed to demonstrate acceptability of the entire fuel movement sequence.

For the SDM demonstrations that rely solely on calculation, additional margin (0.10% ok/k) must be added to the SDM limit of 0.28% ok/k to account for uncertainties in the calculation. Spiral offload or reload sequences inherently satisfy the SR, provided the fuel assemblies are reloaded in the same configuration analyzed for the new cycle.

Removing fuel from the core will always result in an increase in SDM.

l 4

r i

f REFERENCES 1.

10 CFR 50, Appendix A, GDC 26.

I 1

2.

ABWR SSAR, Section 15.4.1.

4 i

3.

ABWR SSAR, Section 4.3.2.

I 4.

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

.j Fuel," Section 3.2.4.1, Sept. 1988.

l 4

j t

j i

l l

4 5

i

^

r i

1 ABWR TS B 3.1-6 P&R, 08/11/93

Reactivity Anomalies B 3.1.2 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.2 Reactivity Anomalies BASES BACKGROUND In accordance with GDC 26, GDC 28, and GDC 29 (Ref. 1),

reactivity shall be controllable such that subcriticality is maintained under cold conditions and acceptable fuel design limits are not exceeded during normal operation and anticipated operational occurrences.

Reactivity anomaly is used as a measure of the predicted versus measured core reactivity during power operation. The continual confirmation of core reactivity is necessary to ensure that the Design Basis Accident (DBA) and transient safety analyses remain valid. A large reactivity anomaly could be the result of unanticipated changes in fuel reactivity, control rod worth, or operation at conditions not consistent with those assumed in the predictions of core reactivity, and could potentially result in a loss of SDM or violation of acceptable fuel design limits. Comparing predicted

-versus measured core reactivity validates the nuclear methods used in the safety analysis and supports the SDM demonstrations (LCO 3.1.1, " SHUTDOWN MARGIN (SDM)") in ensuring the reactor can be brought safely to cold, subcritical conditions.

When the reactor core is critical or in normal power operation, a reactivity balance exists and the net reactivity is zero. A comparison of predicted and measured reactivity is convenient under such a balance, since parameters are being maintained relatively stable under steady state power conditions. The positive reactivity inherent in the core design is balanced by the negative reactivity of the control components, thermal feedback, neutron leakage, and materials in the core that absorb neutrons, such as burnable absorbers, producing zero net reactivity.

In order to achieve the required fuel cycle energy output, the uranium enrichment in the new fuel loading and the fuel loaded in the previous cycles provide excess positive reactivity beyond that required to sustain steady state operation at the beginning of cycle (BOC).

When the reactor is critical at RTP and operating moderator temperature, the excess positive reactivity is compensated by burnable (continued)

ABWR TS B 3.1-1 P&R, 08/11/93

Reactivity Anomalies B 3.1.2 BASES i

BACKGROUND absorbers (if any), control rods, and whatever neutron l

4 (continued) poisons (mainly xenon and samarium) are present in the fuel.

The predicted core reactivity, as represented by k effective (k,), is calculated by a 3D core simulator code as a fu,n,ction of cycle exposure. This calculation is performed r

for projected operating states and conditions throughout the

{

cycle. The monitored k,, is calculated by the core monitoring system for a,ctual plant conditions and is then 3

compared to the predicted value for the cycle exposure.

APPLICABLE Accurate prediction of core reactivity is either an explicit SAFETY ANALYSES or implicit assumption in the accident analysis evaluations.

(

3 Every accident evaluation (Ref. 2) is, therefore, dependent upon accurate evaluation of core reactivity.

In particular, SDM and reactivity transients, such as control rod withdrawal accidents or rod drop accidents, are very sensitive to accurate prediction of core reactivity. These accident analysis evaluations rely on computer codes that have been qualified against available test data, operating plant data, and analytical benchmarks. Monitoring i

i reactivity anomaly provides additional assurance that the nuclear methods provide an accurate representation of the core reactivity.

j The comparison between measured and predicted initial core reactivity provides a normalization for the calculational models used to predict core reactivity.

If the measured and predicted k,,, for identical core conditions at BOC do not

~

J reasonably agree, then the assumptions used in the reload j

cycle design analysis or the calculation models used to predict k,,, may not be accurate.

If reasonable agreement i

between measured and predicted core reactivity exists at BOC, then the prediction may be normalized to the measured i

value. Thereafter, any significant deviations in the measured k,, from the predicted k,,, that develop during l

fuel deplef.lon may be an indication that the assumptions of

]

the DBA and transient analyses are no longer valid, or that an unexpected change in core conditions has occurred.

J t

Reactivity anomalies satisfy Criterion 2 of the NRC Policy l

Statement.

)

(continued)

ABWR TS B 3.1-2 P&R, 08/11/93

Reactivity Anomalies B 3.1.2 BASES (continued)

LC0 The reactivity anomaly limit is established to ensure plant operation is maintained within the assumptions of the safety analyses.

Large differences betwe6n monitored and predicted core reactivity may indicate that the assumptions of the DBA and transient analyses are no longer valid, or that the uncertainties in the Nuclear Design Methodology are larger than expected. A limit on the difference between the monitored core k and the predicted core k of 1% ok/k has been establisNed based on engineering ju,Ngment. A > 1%

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

APPLICABILITY In MODE 1, most of the control rods are withdrawn and steady state operation is typically achieved.

Under these conditions, the comparison between predicted and monitored core reactivity provides an effective measure of the reactivity anomaly.

In MODE 2, control rods are typically being withdrawn during a startup.

In MODES 3 and 4, all control rods are fully inserted, and, therefore, the reactor 1

is in the least reactive state, where monitoring core reactivity is not necessary.

In MODE 5, fuel loading results in a continually changing core reactivity.

SDM requirements (LCO 3.1.1) ensure that fuel movements are performed within the bounds of the safety analysis, and an l

SDM demonstration is required during the first startup following operations that could have altered core reactivity (e.g., fuel movement, control rod replacement, control rod shuffling). The SDM test, required by LC0 3.1.1, provides a

~

direct comparison of the predicted and monitored core

~

reactivity at cold conditions; therefore, reactivity anomaly is not required during these conditions.

ACTIONS A1 Should an anomaly develop between measured and predicted core reactivity, the core reactivity difference must be restored to within the limit to ensure continued operation is within the core design assumptions.

Restoration to within the limit could be performed by an evaluation of the core design and safety analysis to determine the reason for the anomaly. This evaluation normally reviews the core (continued) i ABWR TS B 3.1-3 P&R, 08/11/93

Reactivity Anomalies 4

B 3.1.2 i

BASES 1

ACTIONS A.1 (continued) conditions to determine their consistency with input to design calculations. Measured core and process parameters are also normally evaluated to determine that they are within the bounds of the safety analysis, and safety analysis calculational models may be reviewed to verify that they are adequate for representation of the core conditions.

The required Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is based on the low probability of a DBA during this period, and allows sufficient time to assess the physical condition of the reactor and complete the evaluation of the core design and safety analysis.

U i

If the core reactivity cannot be restored to within the 1% Ak/k limit, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant t

must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The l

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

i r

SURVEILLANCE SR 3.1.2.1 i

REQUIREMENTS Verifying the reactivity difference between the monitored

~

and predicted core k is within the limits of the LCO

~

provides further ass,,r,a:,ce that plant operation is u

maintained within the assumptions of the DBA and transient i

analyses. The Core Monitoring System calculates the core k,,, for the reactor conditions obtained from plant j

instrumentation.

A comparison of the monitored core k,,,dto the predicted core k,,,ivity difference.

at the same cycle exposure is use to calculate the react The comparison is required when the core reactivity has potentially changed by a significant amount. This may occur following a refueling in which new fuel assemblies are loaded, fuel assemblies are shuffled within the core, or control rods are replaced or shuffled. Control rod replacement refers to the decoupling and removal of a control rod from a core location, and subsequent replacement with a new control rod or a control (continued)

ABWR TS B 3.1-4 P&R, 08/11/93

)

.-.___. -.~

Reactivity Anomalies B 3.1.2 1

l BASES SURVEILLANCE SR 3.1.2.1 (continued)

REQUIREMENTS rod from another core location. Also, core reactivity changes during the cycle. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval after reaching equilibrium conditions following a startup is based on the need for equilibrium xenon concentrations in the core,-such that an accurate comparison between the monitored and predicted core k values can be made.

For the i

purposes of this SR,,'the reactor is assumed to be at equilibrium conditions when steady state operations (no control rod movement or core flow changes) at 2 75% RTP have been obtained.

The 1000 MWD /T Frequency was developed, considering the relatively slow change in core reactivity with exposure and operating experience related to variations in core reactivity.

l REFERENCES 1.

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

l 2.

ABWR SSAR, Chapter 15.

a I

J 7

ABWR TS B 3.1-5 P&R, 08/11/93 i

h

Control Rod OPERABILITY B 3.1.3 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.3 Control Rod CPERABILITY BASES BACKGROUND Control rods are components of the Control Rod Drive (CRD)

System, which is the primary reactivity control system for the reactor.

In conjunction with the Reactor Protection System, the CRD System provides the means for the reliable control of reactivity changes to ensure that under conditions of normal operation, including anticipated operational occurrences, specified acceptable fuel design limits are not exceeded.

In addition, the control rods provide the capability to hold the reactor core subtritical 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 is designed to satisfy the requirements of GDC 26, GDC 27, GDC 28, and GDC 29, (Ref. 1).

The CRD System consists of 205 fine motion control rod drive (FMCRD) mechanisms and a 103 hydraulic control unit (HCU) assemblies. The FMRCD is an electro-hydraulic actuated mechanism that provides normal positioning of the control rods using an electric motor, and scram insertion of the control rods using hydraulic power. The hydraulic power for scram is provided by high pressure water stored in the individual HCD accumulators, each of which supplies sufficient volume to scram two FMRCSs.

Normal control rod positioning is performed using a ball-nut and rotating ball-screw arrangement driven by an electric stepping motor. A

~

hollow piston, which is coupled at the upper end to the control rod, rests on the ball-nut. The ball-nut inserts the hollow piston and connected control rod into the core or withdraws.them depending on the direction of rotation of the stepping motor. An electromechanical brake mechanism engages the motor drive shaft when the motor is deenergized to prevent inadvertent withdrawal of the control rod, but does not restrict scram insertion.

This Specification, along with LC0 3.1.4, " Control Rod Scram Times," and LC0 3.1.5, " Control Rod Scram Accumulators,"

ensure that the performance of the control rods in the event of a Design Basis Accident (DBA) or transient meets the assumptions used in the safety analyses of References 2, 3, 4, and 5.

(continued)

ABWR TS B 3.1-1 P&R, 08/10/93 e

Control Rod OPERABILITY

{

B 3.I.3 BASES f

APPLICABLE The analytical methods and assumptions used in the SAFETY ANALYSES evaluations involving control rods are presented in References 2, 3, 4, and 5.

The control rods provide the primary means for rapid reactivity control (reactor scram),

for maintaining the reactor subtritical, and for limiting potential effects of reactivity insertion events caused by malfunctions in the CRD System.

1 The capability of inserting the control rods ensures that L

the assumptions for scram reactivity in the DBA and transient analyses are not violated.

Since the SDM ensures the reactor will be subcritical with the highest worth control rod pair withdrawn (assumed single failure) of an HCU, the additional failure of a second control rod to insert could invalidate the demonstrated SDM and potentially limit the ability of the CRD System to hold the reactor l

subcritical. Therefore, the requirement that all control rods be OPERABLE ensures the CRD System can perform its intended function.

The control rods also protect the fuel from damage that

{

could result in release of radioactivity. The limits protected are the MCPR Safety Limit (SL) (see Bases for LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)"), the 1%

cladding plastic strain fuel design limit (see Bases for LCO 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLGHR)," and LCO 3.2.3, " LINEAR HEAT GENERATION RATE (LHGR)"), and the fuel damage limit (see Bases for LC0 3.1.6, " Rod Pattern Control") during reactivity l

insertion events.

~

The negative reactivity insertion (scrrm) provided by the CRD System provides the analytical basis for determination i

of plant thermal limits and provides protection against fuel damage limits during a Rod Withdrawal Error (RWE) event.

Bases for LCO 3.1.4, LC0 3.1.5, and LCO 3.1.6 discuss in more detail how the SLs are protected by the CRD System.

Control rod OPERABILITY satisfies Criterion 3 of the NRC l

Policy Statement.

l LCO OPERABILITY of an individual control rod is based on a combination of factors, primarily the scram insertion times, the control rod coupling integrity, and the ability to 4

(continued)

ABWR TS B 3.1-2 P&R, 08/10/93 i

2

Control Rod OPERABILITV B 3.1.3 l

r P

BASES LCO determine the control rod position. Accumulator OPERABILITY i

(continued) is addressed by LCO 3.1.5.

Although not all control rods are required to be OPERABLE to satisfy the intended reactivity control requirements, strict control over the number and distribution of inoperable control rods is required to satisfy the assumptions of the DBA and transient i

analyses.

APPLICABILITY In MODES 1 and 2, the control rods are assumed to function during a DBA or transient and are therefore required to be OPERABLE in these MODES.

In MODES 3 and 4, control rods are only allowed to be withdrawn under Special Operations LC0 3.10.3, " Control Rod Withdrawal--Hot Shutdown," and LC0 3.10.4, " Control Rod Withdrawal--Cold Shutdown," which provide adequate requirements for control rod OPERABILITY during these conditions. Control rod requirements in MODE 5 are located in LCO 3.9.5, " Rod OPERABILITY--Refueling."

e ACTIONS The ACTIONS table is modified by a Note indicating that a separate Condition entry is allowed for each control rod.

This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable control rod.

Complying with the Required Actions may allow for continued operation, and subsequent inoperable control rods are governed by subsequent Condition entry and application of associated Required Actions.

A.I. A.2. and A.3 A control rod is considered stuck if it will not insert by either FMCRD drive motor torque or scram pressure. The failure of a control rod to insert during SR 3.1.3.2 or SR 3.1.3.3 alone, however, does not necessarily mean that the control rod is stuck, since failure of the motor drive would also result in a failure of these tests.

Verification of a stuck rod can be made by attempting to withdraw the rod.

If

+

the motor is working and the rod is actually stuck, the traveling nut will back down from the bottom of the drive and a rod separation alarm and rod block will result (see LCO 3.3.1).

Conversely, if the motor drive is known to be failed, the rod is not necessarily inoperable since it is (continued)

ABWR TS B 3.1-3 P&R, 08/10/93

L A

Control Rod OPERABILITV B 3.1.3 i

BASES ACTIONS A.1. A.2. and A.3 (continued)

[

probably still capable of scram. However, at the next required performance of SR 3.1.3.2 or 3.1.3.3, there would be no way of verifying insertability, except by scram.

In this case, an individual scram should be attempted.

If the i

rod scrams, the rod is not stuck but should be considered inoperable and bypassed in RC&IS since it cannot be withdrawn and a separation situation will exist until the i

motor is repaired and the traveling nut is run-in to the full in position.

If the rod fails to insert by individual scram, it should be considered stuck and the appropriate ACTIONS taken. The failure of a centrol rod pair to insert is assumed in the design basis transient and accident analyses and therefore, with one withirawn control rod stuck, some time is allowed to make t!e control rod insertable.

With a fully inserted control rod stuck, no actions are required as long as the control rod remains fully inserted.

The Required Actions are modified by a Note that allows a stuck control rod to be bypassed in the Rod Action and i

Position Information System (RAPI) to allow continued operation.

SR 3.3.2.1.6 provides additional requirements when control rods are bypassed in RAPI to ensure compliance with the RWE analysis. With one withdrawn control rod stuck, the control rod must be disarmed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The motor drive may be disarmed by placing the rod in RAPI bypass or by manually disconnecting its power supply. The l

allowed Completion Time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is acceptable, j

i considering the reactor can still be shut down, assuming no 2

~~

additional control rods fail to insert, and provides a I

reasonable amount of time to perform the Required Action in an orderly manner.

Isolating the control rod from scram i

prevents damage to the CRD and surrounding fuel assemblies should a scram occur. The control rod can be isolated from scram by isolating its associate hydraulic control unit.

Two CRDs sharing an HCU can be individually isolated from i

scram.

Monitoring of the insertion capability of withdrawn control rods must also be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SR 3.1.3.2 and 4

SR 3.1.3.3 perform periodic tests of the control rod l

insertion capability of withdrawn control rods. Testing withdrawn control rods ensures that a generic problem does not exist. The allowed Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provides (continued) l ABWR TS B 3.1-4 P&R, 08/10/93 4

l

i Control Rod OPERABILITV i

B 3.1.3 1

BASES ACTIONS A.1. A.2. and A.3 (continued) a reasonable time to test the control rods, considering the potential for a need to reduce power to perform the tests.

Required Action A.2 is modified by a Note that states the requirement is not applicable when below the actual low power setpoint (LPSP) of the RC&IS, since the notch insertions may not be compatible with the requirements of i

rod pattern control (LC0 3.1.6) and the RC&IS (LC0 3.3.1).

To allow continued operation with a withdrawn control rod stuck, an evaluation of adequate SDM is also required within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Should a DBA or transient require a shutdown, to preserve the single failure criterion an additional control rod would have to be assumed to have failed to insert when required. Therefore, the original SDM demonstration may not be valid. The SDM must therefore be evaluated (by measurement or analysis) with the stuck control rod at its stuck position and the highest worth OPERABLE control rod l

assumed to be fully withdrawn.

With a single control rod stuck in a withdrawn position, the remaining OPERABLE control rods are capable of providing the required scram and shutdown reactivity.

Failure to reach MODE 4 is only likely if an additional control rod adjacent to the stuck control rod also fails to insert during a required scram.

Even with the postulated additional single failure of an adjacent control rod to insert, sufficient reactivity control remains to reach and maintain MODE 3 conditions (Ref. 6).

Required action A.2 performs a step test on each remaining withdrawn control rod to ensure that no additional control rods are stuck. Therefore, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ~

1 is allowed to perform the analysis to test in Required Action A.3.

B.1 and B.2 With two or more withdrawn control rods stuck, the stuck control rods should be isolated from scram pressure within i

1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and the plant brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The occurrence of more than one control rod stuck at a withdrawn position increases the probability that the reactor cannot be shut down if required.

Insertion of all insertable control rods eliminates the possibility of an additional failure of a control rod to insert. The allowed Completion (continued) l ABWR TS B 3.1-5 P&R, 08/10/93

Control Rod OPERABIL1TY B 3.1.3 BASES ACTIONS B.1 and B.2 (continued)

Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an I

orderly manner and without challenging plant systems.

C.1 and C.2 With one or more control rods inoperable for reasons other than being stuck in the withdrawn position, operation may continue, provided the control rods are fully inserted within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and disarmed (however, they do not need to be isolated from scram) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected. The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations. The control rods can be disarmed by disconnecting power to the

[

motor drive or by placing the rod in RC&IS INOP Bypass.

Required Action C.1 is modified by a Note that allows control rods to be bypassed in the RC&IS if required to allow insertion of the inoperable control rods and continued operation. Also, as noted, control rods declared inoperable with a failed motor drive can only be inserted by scram.

Control rods with failed motor drives are not inoperable for j

this reason alone, but must be considered so upon failure of SR 3.1.3.2 or SR 3.1.3.3, or when not in compliance with GWSR (see LC0 3.1.6).

This does not conflict with SR 3.0.1 since the ability to move the control rod via the FMCRD, as discussed in the bases for SR 3.1.3.2 and SR 3.1.3.3, is

~

required to prove that the rod is not stuck.

Likewise, losr of position indication, assuming no rod movement, would not result in control rod (s) inoperability until failure of SR 3.1.3.1.

SR 3.3.1.6 provides additional requirements when the control rods are bypassed to ensure compliance with the RWE analysis.

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

t (continued) 4 ABWR TS B 3.1-6 P&R, 08/10/93

1 Control Rod OPERABILITY B 3.1.3 i

BASES ACTIONS D.] and 0.2 (continued)

Out of sequence control rods may increase the potential l

reactivity worth of a control rod, or gang of control rods, during a RWE and therefore, the distribution of inoperable control rods must be controlled. Below 10% RTP, the generic ganged withdrawal sequence restrictions (GWSR) (which is equivalent to previous banked position withdrawal sequence (BPWS) analysis (Ref. 6) requires inserted control rods not in compliance with GWSR to be separated by at least two I

OPERABLE control rods in all directions, including the diagonal.

Therefore, if one or more inoperable control rods t

are not in compliance with GWSR and not separated by at i

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

restore compliance with GWSR or restore the control rods to OPERABLE status. A Note has been added to the Condition to i

clarify that the Condition is not applicable when > 10% RTP since the GWSR is not required to be followed under these l

conditions, as described in the Bases for LC0 3.1.6.

L 121 i

i If any Required Action and associated Completion Time of Condition A, C, D, or E are not met or nine or more inoperable control rods exist, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

i This ensures all insertable control rods are inserted and places the reactor in a condition that does not require the active function (i.e., scrim) of the control rods. The t

~

number of control rods permitted to be inoperable when

~

operating above 10% RTP (i.e., no CRDA considerations) could i

be more than the value specified, but the occurrence of a large number of ineperable control rods could be indicative of a generic problem, and investigation and resolution of the potential problem should be undertaken. The allowed Completion. Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

I i

i i

(continued)

ABWR TS B 3.1-7 P&R, 08/10/93

~L

~- - - - -

Control Rod OPERABILITV B 3.1.3 BASES SURVEILLANCE SR 3.1.3.1

[

REQUIREMENTS The position of each control rod must be determined, to l

ensure adequate information on control rod position i; i

available to the operator for determining CRD OPERABILITY and controlling rod patterns.

Control rod position may be determined by the use of OPERABLE position indicators, by moving control rods to a position with an OPERABLE indicator, or by the use of other appropriate methods. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency of this SR is based on operating experience related to expected changes in control rod position and the availability of control rod position indications in the contrcl room.

SR 3.1.3.2 and SR 3.1.3.3 Control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at l

two notches and observing that the control rod moves.

The control rod may then be returned to its original position.

This ensures the control rod is not stuck and is free to insert on a scram signal.

These Surveillances are not required when below the actual LPSP of the RC&IS since the notch insertions may not be compatible with the requirements of rod pattern control (LCO 3.1.6) and the RC&IS (LCO 3.3.2.1).

The 7 day Frequency of SR 3.1.3.2 is based on operating experience related to the changes in CRD performance and the ease of performing notch testing for fully withdrawn control rods.

Partially withdrawn control rods are tested at a 31 day Frequency, based on the

~~

potential power reduction required to allow the control rod ~

movement, and considering the large testing sample of SR 3.1.3.2.

Furthermore, the 31 day Frequency takes into account operating experience related to changes in CRD performance. At any time, if a control rod is immovable, a determination of that control rod's trippability (OPERABILITY) must be made and appropriate action must be taken.

i i

SR 3.3.3.4 Verifying the scram time for each control rod to 60% rod insertion position is $; [ ] seconds provides reasonable l

assurance that the control rod will insert when required (continued) l ASWR TS B 3.1-8 P&R, 08/10/93

\\

Control Rod OPERABILITY B 3.1.3 L

BASES SURVEILLANCE SR 3.1.3.5 (continued)

REQUIREMENTS r

during a DBA or transient, thereby completing its shutdown 4

function. This SR is performed in conjunction with the control rod scram time testing of SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4.

The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1 and LCO 3.3.1.2 overlap this Surveillance to provide complete testing of the assumed safety function. The associated Frequencies are acceptable, i

considering the more frequent testing performed to i

demonstrate other aspects of control rod OPERABILITY and j

operating experience, which shows scram times do not

]

significantly change over an operating cycle.

[

L SR 3.1.3.5 l

Coupling verification is performed to ensure the control rod l

is connected to the CRDM and will perform its intended function when necessary.

The Surveillance requires verifying that a control rod does not go to the withdrawn overtravel position when it is fully withdrawn. The overtravel position feature provides a positive check on the coupling integrity, since only an uncoupled hollow piston can reach the overtravel position. The verification is required to be performed once the first time a control rod is withdrawn to the " full out" position after fuel movement r

3 within the RPV or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that l

could affect coupling. This includes control rods inserted two notches and then returned to the " full out" position during the performance of SR 3.1.3.2.

This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved and l

operating experience related to uncoupling events.

n 4

i REFERENCES 1.

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

2.

ABWR SSAR, Section 4.6.2.

(continued) i ABWR TS B 3.1-9 P&R, 08/10/93 e

1 Control Rod OPERABfLITY B 3.1.3 l

l BASES REFERENCES 3.

ABWR SSAR, Section 5.2.2.

(continued) 4.

ABWR SSAR, Section 15.4.1.

r 5.

ABWR SSAR, Section 15.4.9.

t 6.

NEDD-21231, " Banked Position Withdrawal Sequence,"

Section 7.2, January 1977.

F l

I i

?

i r

i i

[

f F

4 I

[

[

9 l

ABWR TS B 3.1-10 P&R, 08/10/93 i

Control Rod Scram Times B 3.1.4 i

B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.4 Control Rod Scram Times i

BASES i

The scram function of the Control Rod Drive (CRD) System I

BACKGROUND controls reactivity ch:nges during abnormal operational transients to ensure chat specified acceptable fuel design limits are not excerded (Ref. 1). The control rods are e

scrammed by positite means, using hydraulic pressure exerted on the CRD piston.

A single hydraulic control unit (HCU) powers the scram action of two fine motion control rod drives (FMCRDs). When 3

a scram signal is initiated, control air is vented from the scram valve in each HCU, allowing it to open by spring action. High pressure nitrogen then raises the piston within the HCU accumulator and forces the displaced wter through the scram piping to the connected FMCRDs.

Inside each FMCRD, the high pressure water lifts the hollow piston 1

off the ball-nut and drives the control rod into the core.

A buffer assembly stops the hollow piston at the end of its stroke. Departure from the ball-nut releases spring-loaded latches in the hollow piston that engage slots in the guide tube.

These latches support the control rod in the inserted position. The control rod cannot be withdrawn until the ball-nut is driven up and engaged with the hollow piston.

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

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the control rod scram function are presented in References 2, 3, and 4.

The Design Basis Accident (DBA) and transient analyses assume that all of the control rods scram at a specified insertion rate. The resulting negative scram i

reactivity forms the basis for the determination of plant thermal limits (e.g., the MCPR). Other distributions of scram times (e.g., several control rods scraming slower than the average time, with several control rods scramning faster than the average time) can also provide sufficient scram reactivity.

Surveillance of each individual control r

(continued)

ABWR TS B 3.1-1 P&R, 08/10/93 j

1 Control Rod Scram Times B 3.1.4 i

BASES APPLICABLE rod's scram time ensures the scram reactivity assumed in the SAFETY ANALYSES DBA and transient analyses can be met.

t (continued)

The scram function of the CRD System protects the MCPR i

Safety Limit (SL) (see Bases for LCO 3.2.2, " MINIMUM i

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

PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," and LC0 3.2.3,

" LINEAR HEAT GENERATION RATE (LHGR)"), which ensure that no fuel damage will occur if these limits are not exceeded.

2 Above 66.8 Kg/cm g (950 psig), the scram function is designed to insert negative reactivity at a rate fast enough to prevent the actual MCPR from becoming less than the MCPR SL during the analyzed limiting power transient.

Below 66.8 a

Kg/cm g (950 psig), the scram function is assumed to perform during the Rod Withdrawal Error (RWE) event (Ref. 4) and, therefore, also provides protection against violating fuel damage limits during reactivity insertion accidents (see l

Bases for LCO 3.1.6, " Rod Pattern Control").

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

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

LCO The scram times specified in Table 3.1.4-1 (in the

~

accompanying LCO) are required to ensure that the scram

~

reactivity assumed in the DBA and transient analysis is met.

To account for single failure and " slow" scramming control rods, the scram times specified in Table 3.1.4-1 are faster i

than those assumed in the design basis analysis. The scram times have a margin to allow up to [8.0]% of the control rods to have scram times that exceed the specified limits (i.e., " slow" control rods) assuming a single stuck control rod (as allowed by LCO 3.1.3, " Control Rod OPERABILITY") and an additional control rod failing to scram per the single failure criterion.

The scram times are specified as a function of reactor steam dome pressure to account for the pressure dependence of the scram times. The scram times are specified relative to measurements based on reed switch positions, which provide the control rod position (continued)

ABWR TS B 3.1-2 P&R, 08/10/93

Control Rod Scram Times B 3.1.4 BASES i

LC0 indication. The reed switch closes (" pickup") when the (continued) hollow piston passes a specific location and then opens

(" dropout") as the index tube travels upward.

Verification of the specified scram times in Table 3.1.4-1 is accomplished through measurement of the " dropout" times.

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

" slow" control rods may occupy adjacent locations.

Table 3.1.4-1 is modified by two Notes, which state control rods with scram times not within the limits of the Table are considered " slow" and that control rods with scram times

>[

] seconds to 60% rod insertion position are considered inoperable as required by SR 3.1.3.4.

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

In MODES 3 and 4, the control rods are only allowed to be withdrawn under Special Operations LCO 3.10.3, " Single Control Rod Withdrawal--Hot Shutdown," and LC0 3.10.4,

" Single Control Rod Withdrawal--Cold Shutdown," which provide adequate requirements for control rod scram capability during these conditions.

Scram requirements in MODE 5 are contained in LC0 3.9.5, " Control Rod OPERABILITY--Refueling. "

ACTIONS A.1 When the requirements of this LC0 are not met, the plant must be brought to a MODE in which the LCO does not apply.

I To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 i

from full power conditions in an orderly manner and without challenging plant systems.

i (continued) l d

ABWR TS B 3.1-3 P&R, 08/10/93 l

i

1 Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE The four SRs of this LC0 are modified by a Note stating that REQUIREMENTS during a single or pair control rod scram time surveillance, the CRD pumps shall be isolated from the associated scram

~

accumulator. With the CRD pump isolated (i.e., charging valve closed), the influence of the CRD pump head does not affect the single or pair control rod scram times. During a full core scram, the CRD pump head would be seen by all control rods and would have a negligible effect on the scram insertion times.

SR 3.1.4.1 The scram reactivity used in DBA and transient analyses is based on assumed control rod scram time.

Measurement of he scram times with reactor steam dome pressure 2 66.8 Kg/cm}g s

t (950 psig) demonstrates acceptable scram times for the transients analyzed in References 2 and 3.

Scram insertion times increase with increasing reactor J

pressure because of the competing effects of reactor steam dome pressure and stored accumulator energy. Therefore, demonstrationofadequatescrampimesatreactorsteamdome pressure greater than 66.8 Kg/cm g (950 psig) ensures that the scram times will be within the specified limits at higher pressures. Limits are specified as a function of reactor pressure to account for the sensitivity of the scram insertion times with pressure and to allow a range of pressures over which scram time testing can be performed.

To ensure scram time testing is performed within a reasonable time following a refueling or after a shutdown 2120 days, all control rods are required to be tested before exceeding 40% RTP following a shutdown. This Frequency is acceptable, considering the additional i

surveillances performed for control rod OPERABILITY, the frequent verification of adequate accumulator pressure, and i

the required testing of control rods affected by work on control rods or the CRD System.

SR 3.1.4.2 Additional testing of a sample of control rods is required to verify the continued performance of the scram function during the cycle. A representative sample contains at least 10% of the control rods, with no more than 20% of the i

(continued) i ABWR TS B 3.1-4 P&R, 08/10/93

Control Rod Scram Times B 3.1.4 i

BASES SURVEILLANCE SR 3.1.4.2 (continued)

REQUIREMENTS control rods in the sample " slow."

If more than 20% of the sample is declared to be " slow" per the criteria in Table 3.1.4-1, additional control rods are tested until this 20% criterion is satisfied, or Required Action A.1 must be taken.

For planned testing, the control rods selected for the sample should be different for each test.

Data from inadvertent scrams should be used whenever possible to avoid unnecessary testing at power, even if the control rods with data were previously tested in a sample.

The 120 day Frequency is based on operating experience that has shown control rod scram times do not significantly change over an operating cycle.

This Frequency is also reasonable, based on the additional Surveillances done on the CRDs at more frequent intervals in accordance with LCO 3.1.3 and LC0 3.1.5, " Control Rod Scram Accumulators."

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

permissible pressure. The scram testing must be performed once before declaring the control rod OPERABLE. The required scram time testing must demonstrate that the affected control rod is still within the limits of Table 3.1.4-1, for startup conditions.

Specific examples of work that could affect the scram times include (but are not limited to) the following: removal of any CRD for maintenance or modification; replacement of a control rod; and maintenance or modification of a scram j

solenoid pilot valve, scram valve, accumulator isolation valve, or check valves in the piping required for scram.

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

i (continued) 1 i

ABWR TS B 3.1-5 P&R, 08/10/93 j

1 i

i Control Rod Scram Times j

B 3.1.4 BASES i

SURVEILLANCE SR 3.1.4.4 l

l REQUIREMENTS (continued)

When work that could affect the scram insertion time is performed on a control rod or CRD System, testing must be l

done to demonstrate each affected control rod is still within the limits of Table 3.1.4-1 with the reactor steam l

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

Where work has I

been performed at high reactor pressure, the requirements of SR 3.1.4.3 and SR 3.1.4.4 will be satisfied with one test.

For a control rod affected by work performed while shut i

down, however, a zero pressure and a high pressure test may 4

be required. This testing ensures that the control rod scram performance is acceptable for operating reactor pressure conditions prior to withdrawing the control rod for continued operation. Alternatively, a test during hydrostatic pressure testing could also satisfy both criteria.

The Frequency of once prior to exceeding 40% RTP is acceptable because of the capability of testing the control i

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

i 4

l i

i REFERENCES 1.

10 CFR 50, Appendix A, GDC 10.

2.

ABWR SSAR, Section 4.6.2.

3.

ABWR SSAR, Section 5.2.2.

I 4.

ABWAR, Section 15.4.1.

~

4 i

l.

\\

I i

l ABWR TS B 3.1-6 P&R, 08/10/93

Control Rod Scram Accumulators B 3.1.5 8 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.5 Control Rod Scram Accumulators BASES BACKGROUND The control rod scram accumulators are part of the Control Rod Drive (CRD) System and are provided to ensure that the control rods scram under varying reactor conditions. The control rod scram accumulators store sufficient energy to fully insert a pair of control rods associated with a specific hydraulic control unit (HCU) at any reactor vessel pressure. The accumulator is a hydraulic cylinder with a free floating piston. The piston separates the water used to scram the control rods from the nitrogen, which provides the required energy. The scram accumulators are necessary to scram the control rods within the required insertion times of LC0 3.1.4, " Control Rod Scram Times."

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the control rod scram function are presented in References 1, 2, 3, and 4.

The Design Basis Accident (DBA) and transient analyses assume that all of the control rods scram at a specified insertion rate. OPERABILITY of each individual control rod scram accumulator, along with LCO 3.1.3, " Control Rod OPERABILITY," and LCO 3.1.4, ensures that the scram reactivity assumed in the DBA and transient analyses can be met. The existence of an inoperable accumulator may invalidate prior scram time measurements for the associated control rod.

The scram function of the CRD System, and, therefore, the OPERABILITY of the accumulators, protects the MCPR Safety Limit (see Bases for LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)") and the 1% cladding plastic strain fuel design limit (see Bases for LCO 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," and LC0 3.2.3,

" LINEAR HEAT GENERATION RATE (LHGR)"), which ensure that no fuel damage will occur if these limits are not exceeded (see Bases for LCO 3.1.4).

Also, the scram function at low reactor vessel pressure (i.e., startup conditions) provides protection against violating fuel design limits during reactivity insertion accidents (see Bases for LCO 3.1.6,

" Rod Pattern Control").

(continued)

ABWR TS B 3.1-1 P&R, 08/11/93

Control Rod Scram Accumulators B 3.1.5 BASES APPLICABLE Control rod scram accumulators satisfy Criterion 3 of the SAFETY ANALYSES NRC Policy Statement.

(continued)

LC0 The OPERABILITY of the control rod scram accumulators is required to ensure that adequate scram insertion capability exists when needed over the entire range of reactor pressures. The OPERABILITY of the scram accumulators is based on maintaining adequate accumulator pressure.

APPLICABILITY In MODES I and 2, the scram function is required for mitigation of DBAs and transients and, therefore, the scram accumulators must be OPERABLE to support the scram function, t

In MODES 3 and 4, control rods are only allowed to be i

withdrawn under Special Operations LC0 3.10.3, " Control Rod W1'hdrawal--Hot Shutdown," and LC0 3.10.4, " Control Rod Witndrawal--Cold Shutdown," which provide adequate requirements for control rod scram accumulator OPERABILITY under these conditions. Requirements for scram accumulators in MODE 5 are contained in LC0 3.9.5, " Control Rod OPERABILITY--Refueling. "

f 5

ACTIONS The ACTIONS table is modified by a Note indicating that a separate Condition entry is allowed for each control rod.

This is acceptable since the Required Actions for each

~~

Condition provide appropriate compensatory action for each

~

inoperable control rod. Complying with the Required Actions may allow for continued operation and subsequent inoperable control rods governed by subsequent Condition entry and application of associated Required Actions.

ail With one control rod scram accumulator inoperable, the scram function could become severely degraded because the accumulator is the primary source of scram force for the associated control rod or rod pair at all reactor pressures.

In this event, the associated control rod or rod pair is declared inoperable and LCO 3.I.3 entered. This would (continued) i ABWR TS B 3.1-2 P&R, 08/11/93 i

)

k

f i

l Control Rod Scram Accumulators i

B 3.1.5 l

BASES r

ACTIONS L1 (continued) result in requiring the affected control rod or rod pair to l

be fully inserted and disarmed, thereby satisfying its intended function in accordance with ACTIONS of LC0 3.1.3.

The allowed Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is considered reasonable, based on the large number of control rods available to provide the scram function. Additionally, an automatic reactor scram function is provided on sensed low pressure in the CRD charging water header (see LC0 3.3.1.1, "RPS Instrumentation").

This anticipatory reactor trip a

protects against the possibility of significant pressure i

degradation (and thus reduced scram force) concurrently in multiple control rod scram accumulators due to a transient in the CRD hydraulic system.

IL1 With two or more control rod scram accumulators inoperable, the scram function could become severely degraded because the accumulators are the primary source of scram force for the control rods at all reactor pressures.

In this event, the associated control rods are declared inoperable and LC0 3.1.3 entered. This would result in requiring the affected control rods to be fully inserted and disarmed, thereby satisfying its intended function in accordance with ACTIONS

)

of LC0 3.1.3.

1 The allowed Completion Time of I hours is considered reasonable, based on the capability to drive in the control _

j rods by the FMCRD motors and the low probability of a DBA or transient occurring while the affected accumulators are i

inoperable.

L.1 The reactor mode switch must be immediately placed in the shutdown position if any Required Action and associated i

4 Completion Time cannot be met. This ensures that all insertable control rods are inserted and that the reactor is in a condition that does not require the active function (i.e., scram) of the control rods. This Required Action is modified by a Note stating that the Required Action is not 1

applicable if all control rods associated with the J

l (continued) i ABWR TS B 3.1-3 P&R, 08/11/93

Control Rod Scram Accumulators i

8 3.1.5 BASES ACTIONS C.]

(continued) inoperable scram accumulators are fully inserted, since the function of the control rods has been performed.

SURVEILLANCE SR 3.1.5.1 REQUIREMENTS l

SR 3.1.5.1 requires that the accumulator pressure be checked every 7 days to ensure adequate accumulator pressure exists to provide sufficient scram force. The primary indicator of accumulator OPERABILITY is the accumulator pressure. A minimum accumulator pressure is specified, below which the capability of the accumulator to perform its intended function becomes degraded and the accumulator is considered inoperable. The minimum accumulator pressure of 2

130.1 Kg/cm g (1850 psig) is well below the expected pressure of 151.2 Kg/cm g (2150 psig) (Ref. 2).

Declaring the accumulator inoperable when the minimum pressure is not maintained ensures that significant degradation in scram t

times does not occur. The 7 day Frequency has been shown to be acceptable through operating experience and takes into account other indications available in the control room.

REFERENCES 1.

NEDE-24011-P-A, " General Electric Standard Application Fuel," September 1988.

2.

ABWR SSAR, Section 4.6.1.

3.

ABWR SSAR, Section 5.2.2.

4.

ABWR SSAR, Section 15.4.1 I

i 4

ABWR TS B 3.1-4 P&R,08/11/93

Rod Pattern Control B 3.1.6 l

B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.6 Rod Pattern Control i

BASES BACKGROUND Control rod patterns during startup conditions are controlled by the operator and the rod worth minimizer (RWM)

(LCO 3.3.2.1, " Control Rod Block Instrumentation"), so that

+

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

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the RWE are summarized in References I and 2.

RWE analyses assume that the reactor operator follows prescribed withdrawal sequences.

These sequences define the potential initial conditions for the RWE analysis. The RWM (LCO 3.3.2.1) provides backup to operator control of the withdrawal sequences to ensure that the initial conditions of the RWE analysis are not violated.

Prevention or mitigation of positive reactivity insertion events is necessary to limit the energy deposition in the fuel, thereby preventing significant fuel damage, which could result in undue release of radioactivity (Reference 4).

Since the failure consequences for U0 have been shown g

~

to be insignificant below fuel energy depositions of t

300 cal /gm, the fuel damage limit of 280 cal /gm provides a margin of safety from significant core damage, which would result in release of radioactivity (Reference 3).

Generic l

analysis of the GWSR (equivalent to the BPWS, see Reference

5) has demonstrated that the 380 cal /gm fuel damage limit will not be violated during a postulated reactivity transient while following the GWSR mode of operation.

Control rod patterns analyzed in References 1 and 2, follow the GWSR which is the same as the banked position withdrawal sequence (BPWS) described in Reference 5.

The GWSR is applicable from the condition of all control rods fully inserted to 10% RTP.

For the GWSR, the control rods are required to be moved in groups, with all control rods (continued)

ABWR TS B 3.1-1 P&R, 08/11/93

i Rod Pattern Control B 3.1.6 i

i BASES i

l APPLICABLE assigned to a specific group required to be within specified l

SAFETY ANALYSES banked positions. The banked positions are defined to (continued) minimize the maximum incremental control rod worths without being overly restrictive during normal plant operation.

The generic BPWS analysis (Reference 5) also evaluated the effect of fully inserted, inoperable control rods not in compliance with the sequence, to allow a limited number (i.e., eight) and distribution of fully inserted, inoperable control rods.

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

LCO Compliance with the prescribed control rod sequences minimizes the potential consequences of a RWE by limiting the initial conditions to those consistent with the GWSR.

This LCO only applies to OPERABLE control rods.

For inoperable control rods required to be inserted, separate requirements are specified in LC0 3.1.3, "Centrol Rod OPERABILITY," consistent with the allowances for inoperable control rods in the GWSR.

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

When THERMAL POWER is

> 10% RTP, there is no credible control rod configuration that results in a control rod worth that could exceed the i

280 cal /gm fuel damage limit during a RWE.

In MODES 3, 4, i

and 5, since only a total of one control rod or control rod _

pair can be withdrawn from core cells containing fuel l

1 assemblies, adequate SDM ensures that the reactor will remain subcritical.

i i

i ACTIONS A.] and A.2 l

With one or more OPERABLE control rods not in compliance with the prescribed control rod sequence, action may be taken to either correct the control rod pattern or declare the associated control rods inoperable within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Noncompliance with the prescribed sequence may be the result l

(continued)

)

ABWR TS B 3.1-2 P&R, 08/11/93 i

l l

i Rod Pattern Control B 3.1.6 i

BASES (continued)

ACTIONS A.] aad A.2 (continued) of failed synchros, drifting from a control rod drive purge water transient, leaking scram valves, or a power reduction to s 10% RTP before establishing the correct control rod pattern. The number of OPERABLE control rods not in compliance with the prescribed sequence is limited to eight

~

to prevent the operator from attempting to correct a control e

rod pattern that significantly deviates from the prescribed sequence. When the control rod pattern is not in compliance with the prescribed sequence, all control rod movement should be stopped except for moves needed to correct the control rod pattern, or scram if warranted.

Required Action A.1 is modified by a Note, which allows control rods to be bypassed in RAPI to allow the affected control rods to be returned to their correct position.

This ensures that the control rods will be moved to the correct position. A control rod not in compliance with the prescribed sequence is not considered inoperable except as required by Required Action A.2.

OPERABILITY of control rods is determined by compliance with LC0 3.1.3; LCO 3.1.4,

" Control Rod Scram Times"; and LCO 3.1.5, " Control Rod Scram Accumulators." The allowed Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is reasonable, considering the restrictions on the number of allowed out of sequence control rods and the low probability of a RWE occurring during the time the control rods are out of sequence.

B.1 and B.2 If nine or more OPERABLE control rods are out of sequence, the control rod pattern significantly deviates from the prescribed sequence. Control rod withdrawal should be suspended immediately to prevent the potential for further deviation from the prescribed sequence. Control rod insertion to correct control rods withdrawn beyond their allowed position is allowed since, in general, insertion of control rods has less impact on control rod worth than withdrawals have. Required Action 8.1 is modified by a Note that allows the affected control rods to be bypassed in RAPI in accordance with SR 3.3.2.1.6 to allow insertion only.

I With nine or more OPERABLE control rods not in compliance with GWSR, the reactor mode switch must be placed in the l

(continued) j 2

ABWR TS B 3.1-3 P&R, 08/11/93 i

4 Rod Pattern Control B 3.1.6 BASES ACTIONS B.1 and B.2 (continued) shutdown position within I hour. With the reactor modeswitch in shutdown, the reactor is shut down, and therefore does not meet the applicability requirements of

~

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

l SURVEILLANCE SR 3.1.6.1 REQUIREMENTS The control rod pattern is verified to be in compliance with the GWSR at a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency, ensuring the assumptions of the RWE analyses are met. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency of this Surveillance was developed considering that the primary check of the control rod pattern compliance with the GWSR is performed by the RWM (LC0 3.3.2.1).

The RWM provides control rod blocks to enforce the required control rod sequence and is required to be OPERABLE when operating at s 10% RTP.

J t

REFERENCES 1.

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

~~

2.

ABWR SSAR, Section 15.4.1.

l

~

3.

NUREG-0800, " Standard Review Plan," Section 15.4.1,

" Uncontrolled Control Rod Assembly Witndrawal from a Subtritical or Low Rod Power Startup Condition,"

l Revision 2, July 1981.

)

4.

10 CFR 100.11, " Determination of Exclusion Area low Population Zone and Population Center Distance."

5.

NED0-21231, " Banked Position Withdrawal Sequence,"

January 1977.

I f

ABWR TS B 3.1-4 P&R, 08/11/93 l

r SLC System B 3.1.7 8 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.7 Standby Liquid Control (SLC) System BASES i

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

The SLC System satisfies the requirements of 10 CFR 50.62 (Ref.1) on anticipated transient without scram (ATWS).

The SLC System consists of a boron solution storage tank, two pos!tive displacement pumps, two motor operated injection valves, which are provided in parallel for redundancy, and associated piping and valves used to transfer borated water from the storage tank to the reactor pressure vessel (RPV). The borated solution is discharged through the high pressure core flooder (HPCF) subsystem.

a i

APPLICABLE The SLC System is automatically initiated. The SCL System SAFETY ANALYSES is used in the event that not enough control rods can be inserted to accomplish shutdown and cooldown in the normal manner. The SLC System injects borated water into the reactor core to compensate for all of the various reactivity effects that could occur during plant operation. To meet this objective, it is necessary to inject a quantity of boron that produces a concentration of 850 ppm of natural boron in the reactor core at 21*C (70*F). To allow for potential leakage and imperfect mixing in the reactor system, an additional amount of boron equal to 25% of the amount cited above is added (Ref. 2). The temperature versus concentration limits in Figure 3.1.7-1 (in the accompanying LCO) are calculated such that the required concentration is achieved accounting for dilution in the RPV with normal water level and including the water volume in the residual heat removal shutdown cooling piping. This quantity of borated solution is the amount that is above the pump suction shutoff level in the boron solution storage tank. No credit is taken for the portion of the tank volume that cannot be injected.

(continued)

ABWR TS B 3.1-1 PAR, 08/11/93 i

SLC System B 3.1.7 BASES APPLICABLE The SLC System satisfies the requirements of the NRC Policy SAFETY ANALYSES Statement because operating xperience and probabilistic (continued) riskassessmenthavegeneralIyshown'.ttobeimportantto public health and safety.

s LC0 The OPERABILITY of the SLC System provides backup capability I

for reactivity control, independent of normal reactivity control provisions provided by the control rods. The OPERABILITY of the SLC System is based on the conditions of the borated solution in the storage tank and the availability of a flow path to the RPV, including the OPERABILITY of the pumps and valves.

Because the minimum required boron solution concentration is the same for both ATVS mitigation and cold shutdown (unlike some previous reactor designs) then if the boron solution concentration is less than the required limit, both SLC subsystems shall be declared inoperable.

Two SLC subsystems are required to be OPERABLE, each containing an OPERABLE pump, a motor operated i

injection valve, and associated piping, valves, and instruments and controls to ensure an OPERABLE flow path.

i i

APPLICABILITY In MODES 1 and 2, shutdown capability is required.

In I

MODES 3 and 4, control rods are only allowed to be withdrawn under Special Operatione. LCO 3.10.3, " Control Rod Withdrawal-Hot Shutdow 1," and LCO 3.10.4, " Control Rod i

Withdrawal-Cold Shutdown," which provide adequate controls to ensure the reactor remains subcritical.

In MODE 5, only i

a single control rod or control rod pair can be withdrawn from a core cell containing fuel assemblies. Demonstration of adequate SDM (LCO 3.1.1, " SHUTDOWN MARGIN (SDM)") ensures l

that the reactor will not become critical. Therefore, the SLC System is not required to be OPERABLE during these conditions, when only a single control rod or control rod pair can be withdrawn.

i i

ACTIONS A.1 If one SLC System subsystem is inoperable, the inoperable.

subsystem must be restored to OPERABLE status within 7 days.

In this condition, the remaining OPERABLE subsystem is i

(continued) l ABWR TS B 3.1-2 PAR, 08/11/93

SLC System B 3.1.7 BASES ACTIONS A.1 (continued) adequate to perform the shutdown function. However, the overall reliability is reduced because a single failure in the remaining OPERABLE subsystem could result in reduced SLC System shutdown capability. The 7 day Completion Time is based on the availability of an OPERABLE subsystem capable of performing the intended SLC System function and the low probability of a Design Basis Accident (DBA) or severe transient occurring concurrent with the failure of the Control Rod Drive System to shut down the plant. The maximum Completion Time of 10 days is allowed for this LCO in the event of multiple Condition entry.

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

C.)

If any Required Action and associated Completion Time is not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The allowed Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating

~

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

SURVEILLANCE SR 3.1.7.1. SR 3.1.7.2. and SR 3.1.7.3 REQUIREMENTS SR 3.1.7.1 through SR 3.1.7.3 are 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Surveillances, verifying certain characteristics of the SLC System (e.g.,

the volume and temperature of the borated solution in the storage tank), thereby ensuring the SLC System OPERABILITY L

without disturbing normal plant operation.

These i

Surveillances ensure the proper borated solution and temperature, including the temperature of the pump suction (continued) i ABWR TS B 3.1-3 P&R, 08/11/93 l

i

SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.1. SR 3.1.7.2. and SR 3.1.7.3 (continued)

REQUIREMENTS piping, are maintained.

Maintaining a minimum specified borated solution temperature is important in ensuring that the boron remains in solution and does not precipitate out in the storage tank or in the pump suction piping.

The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency of these SRs is based on operating experience that has shown there are relatively slow variations in the measured parameters of volume and temperature.

SR 3.1.7.4 This Surveillance requires an examination of the sodium pentaborate solution by using chemical analysis to ensure the proper concentration of boron exists in the storage I

tank. SR 3.1.7.5 must be performed anytime baron or water is added to the storage tank solution to establish that the boron solution concentration is within the specified limits.

This Surveillance must be performed anytime the temperature is restored to within the limits of Figure 3.1.7-1, to ensure no significant boron precipitation occurred. The 31 day Frequency of this Surveillance is appropriate because i

of the relatively slow variation of baron concentration l

between surveillances.

SR 3.1.7.5 l

l SR 3.1.7.5 verifies each valve in the system is in its l

correct position. Verifying the correct alignment for i

manual, power operated, and automatic valves in the SLC System flow path ensures that the proper flow paths will exist for system operation. This Surveillance does not apply to valves that are locked, sealed, or otherwise secured in position, since they were verified to be in the correct position prior to locking, sealing, or securing.

This verification of valve alignment does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation; rather, it involves verification that those l

valves capable of being mispositioned are in the correct i

positions. The 31 day Frequency is based on engineering t

(continued)

ABWR TS B 3.1-4 P&R, 08/11/93 l

1 L.

SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.5 (continued)

REQUIREMENTS judgment and is consistent with the procedural controls governing valve operation that ensure correct valve positions.

SR 3.1.7.6 Demonstrating each SLC System pump develops a flow rate 3

a 211.4 m /h (50 gpm) at a discharge pressure 2 86.0 Kg/cm g (1223 psig) ensures that pump performance has not degraded during the fuel cycle. This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay.

This test confirms one point on the pump design curve, and is indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this Surveillance is [in accordance with the Inservice Testing Program or 92 days].

SR 3.1.7.7 and SR 3.1.7.8 These Surveillances ensure that there is a functioning flow path from the boron solution storage tank to the RPV.

The pump and injection valve tested should be alternated such

~

that both complete flow paths are tested every 36 months, at alternating 18 month intervals. The Surveillance may be performed in separate steps to prevent injecting boron into the RPV. An acceptable method for verifying flow from the pump to the RPV is to pump demineralized water from a test tank through one SLC subsystem and into the RPV. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and thm potential for an unplanned transient if the Surveillance were performed with the reactor at power.

i Operating experience has shown these components usually pass the Surveillance test when performed at the 18 month Frequency; therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

(continued)

ABWR TS B 3.1-5 P&R, 08/11/93 2

SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.7 and SR 3.1.7.8 (continued)

REQUIREMENTS Demonstrating that all heat traced piping between the boron solution storage tank and the suction inlet to the injection pumps is unblocked ensures that there is a functioning flow path for injecting the sodium pentaborate solution. An acceptable method for verifying that the suction piping is unblocked is to pump from the storage tank to the test tank.

The 18 month Frequency is acceptable since there is a low probability that the subject piping will be blocked due to precipitation of the boron from solution in the heat traced piping. This is especially true in light of the daily temperature verification of this piping required by SR 3.1.7.3.

However, if, in performing SR 3.1.7.8, it is determined that the temperature of this piping has fallen below the specified minimum, this Surveillance must be performed once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the piping temperature is restored within the limits of Figure 3.1.7-1.

I i

REFERENCES 1.

10 CFR 50.62.

t 2.

ABWR SSAR, Section 9.3.5.3.

l 1

I.

i l

ABWR TS B 3.1-6 P&R, 08/11/93

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

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

APPLICABILITY:

THERMAL POWER 2 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

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

within limits.

B.

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

Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

~

SR 3.2.1.1 Verify all APLHGRs are less than or equal Once within to the limits specified in the COLR.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after 1 25% RTP AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter i

i ABWR TS 3.2-1 P&R, 8/9/93

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

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

l APPLICABILITY:

THERMAL POWER 2 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

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

within limits.

B.

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

Time not met.

3 l

SURVEILLANCE REQUIREMENTS

~

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

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

thereafter 1

ABWR TS 3.2-1 P&R, 8/9/93 i

LHGR (Non-GE Fuel) 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) (Non-GE Fuel)

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

APPLICABILITY:

THERMAL POWER 2 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

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

within limits.

B.

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

Time not met.

SURVEILLANCE REQUIREMENTS

~~

SURVEILLANCE FREQUENCY

~

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

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after 1 25% RTP AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter ABWR TS 3.2-1 P&R, 8/9/93

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

BASES i

BACKGROUND The APLHGR is a measure of the average LHGR of all the fuel rods in a fuel assembly at any axial location.

Limits on the APLHGR are specified to ensure that the fuel design limits identified in Reference 1 are not exceeded during i

anticipated operational occurrences (A00s) and that the peak cladding temperature (PCT) during the postulated design basis loss of coolant accident (LOCA) does not exceed the limits specified in 10 CFR 50.46.

i APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the fuel design limits are presented in the SSAR, Chapter 4, and in Reference 1.

The analytical methods and assumptions used in evaluating Design Basis Accidents (DBAs),

anticipated operational transients, and normal operations that determine APLHGR limits are presented in SSAR, i

Chapters 4, 6, and 15, and in Reference 1.

Fuel design evaluations are performed to demonstrate that the 1% limit on the fuel cladding plastic strain and other fuel design limits described in Reference 1 are not exceeded during A00s for operation with LHGR up to the operating i

limit LHGR. APLHGR limits are equivalent to the LHGR limit for each fuel rod divided by the local peaking factor of the i

fuel assembly. APLHGR limits are developed as a function of i

exposure and the various operating core flow and power states to ensure adherence to fuel design limits during the limiting A00s.

Flow dependent APLHGR limits are determined using the three dimensional BWR simulator code (Ref. 2) to i

analyze slow flow runout transients. The flow dependent multiplier, MAPFACv, is dependent on the maximum core flow runout capability. The maximum runout flow is dependent on t

the existing setting of the core flow limiter in the Recirculation Flow Control System.

Based on analyses of limiting plant transients (other than core flow increases) over a range of power and flow I

conditions, power dependent multipliers, MAPFACp, are also generated. Due to the sensitivity of the transient response to initial core flow levels at power levels below those at (continued)

\\

i ABWR TS B 3.2-1 P&R, 08/11/93 I

I

APLHGR B 3.2.1 l

BASES APPLICABLE which turbine stop valve closure and turbine control valve SAFETY ANALYSES fast closure scram signals are bypassed, both high and low (continued) core flow MAPFACp limits are provided for operation at power levels between 25% RTP and the previously mentioned bypass 7

power level. The exposure dependent APLHGR limits are reduced by MAPFACp and MAPFAC, at various operating conditions to ensure that all fuel design criteria are met for normal operation and A00s. A complete discussion of the analysis code is provided in Reference 3.

LOCA analyses are then performed to ensure that the above determined APLHGR limits are adequate to meet the PCT and maximum oxidation limits of 10 CFR 50.46. The analysis is performed using calculational models that are consistent with the requirements of 10 CFR 50, Appendix K.

A complete discussion of the analysis code is provided in Reference 1.

The PCT following a postulated LOCA is a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is not strongly influenced by the rod to rod power distribution within an assembly.

The APLHGR limits specified are equivalent to the i

LHGR of the highest powered fuel rod assumed in the LOCA analysis divided by its local peaking factor. A conservative multiplier is applied to the LHGR assumed in-j the LOCA analysis to account for the uncertainty associated with the measurement of the APLHGR.

The APLHGR satisfies Criterion 2 of the NRC Policy l

4 Statement.

J q

LCO The APLHGR limits specified in the COLR are the result of j

fuel design, DBA, and transient analyses. The limit is determined by multiplying the smaller of the MAPFAC, and l

MAPFAC factors times the exposure dependent APLHGR limits.

p APPLICABILITY The APLHGR limits are primarily derived from fuel design i

evaluations and LOCA and transient analyses that are assumed to occur at high power levels. Design calculations (Ref. 3) and operating experience have shown that as power is reduced, the margin to the required APLHGR limits increases.

This trend continues down to the power range of 5% to 15% RTP when entry into MODE 2 occurs.

When in MODE 2, the (continued)

ABWR TS B 3.2-2 P&R, 08/11/93

APLHGR B 3.2.1 i

BASES l

APPLICABILITY Startup Range Neutron Monitor (SRNM) scram function provides (continued) prompt scram initiation during any significant transient, thereby effectively removing any APLHGR limit compliance concern in MODE 2.

Therefore, at THERMAL POWER levels

$; 25% RTP, the reactor operates with substantial margin to l

the APLHGR limits; thus, this LCO is not required, ACTIONS

.A_d t

If any APLHGR exceeds the required limits, an assumption f

regarding an initial condition of the DBA and transient analyses may not be met.

Therefore, prompt action is taken to restore the APLHGR(s) to within the required limits such that the plant will be operating within analyzed conditions and within the design limits of the fuel rods. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is sufficient to restore the APLHGR(s) to within its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the APLHGR out of specification.

161 4

l If the APLHGR cannot be restored to within its required f

limits within the associated Completion Time, the plant must l

be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 25% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The allowed Completion Time is reasonable, based on operating

~

experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.1.1 REQU?REMENTS 4

APLHGRs are required to be initially calculated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER is 2 25% RTP and then every i

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter. They are compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis.

The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution under normal conditions.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance after (continued) i ABWR TS B 3.2-3 P&R, 08/11/93

~

APLHGR B 3.2.1 BASES SURVEILLANCE SR 3.2.1.1 (continued)

REQUIREMENTS THERMAL POWER 2 25% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels.

REFERENCES 1.

NED0-240ll-P-A, " General Electric Str..,dard A,pplication for Reactor Fuel," September 1988, 2.

NED0-301300-A, " Steady State Nuclear Methods," May 1985.

3.

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

l t

1 h

h i

i ABWR TS B 3.2-4 P&R, 08/11/93 L

i

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

BASES BACKGROUND The MCPR is the ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The MCPR Safety Limit (SL) is set such that 99.9% of the fuel rods avoid boiling transition if the limit is not violated (refer to the Bases for SL 2.1.2).

The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (A00s). Although fuel damage does not necessarily occur if a fuel rod actually experiences boiling transition (Ref.1),

the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion.

The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle i

designs.

Based on these experimental data, correlations have been developed to predict critical bundle power (i.e.,

the bundle power level at the onset of transition boiling) i for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling). Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.

APPLICABLE The analytical methods and assumptions used in evaluating 4

SAFETY ANALYSES the A00s to establish the operating limit MCPR are presented in the SSAR, Chapters 4, 6, and 15, and Reference 2.

To ensure that the MCPR SL is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR). The types of j

transients evaluated are loss of feedwater flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest change in CPR (E PR). When the largest E PR is added to the MCPR SL, the required operating limit MCPR is obtained.

b (continued) t ABWR TS B 3.2-1 P&R, 08/11/93

MCPR B 3.2.2 1,

BASES APPLICABLE The MCPR operating limits derived from the transient 4

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

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

Power dependent MCPR limits (MCPR ) are determined by the p

one dimensional transient code (Reference 4) for anticipated transients that are sigr.ificantly affected by power.

Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scram trips are bypassed, high and low flow MCPRp operating limits are provided for operating between 25% RTP and the i

previously mentioned bypass power level.

The MCPR satisfies Criterion 2 of the NRC Policy Statement.

LC0 The MCPR operating limits specified in the COLR are the i

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

]

i APPLICABILITY The MCPP, operating limits are primarily derived from i

transient analyses that are assumed to occur at high power levels. Below 25% RTP, the reactor is operating at a minimum reactor internal pump speed and the moderator void ratio is small.

Surveillance of thermal limits below 25% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs.

Statistical analyses documented in Reference 5 indicate that the nominal value of the initial MCPR expected at 25% RTP is

> 3.5.

Studies of the variation of limiting transient behavior have been performed over the range of power and (continued) i ABWR TS B 3.2-2 P&R, 08/11/93

MCPR B 3.2.2 i

l BASES 1

APPLICABILITY flow conditions. These studies encompass. the range of key

]

(continued) actual plant parameter values important to typically limiting transients. The results of these studies l

demonstrate that a margin is expected between performance and the MCPR requirements, and that margins increase as 4

power is reduced to 25% RTP. This trend is expected to continue to the 5% to 15% power range when entry into MODE 2 j

occurs. When in MODE 2, the Startup Range Neutron Monitor 3

(SRNM) provides rapid scram initiation for any significant j

power increase transient, which effectively eliminates any j

MCPR compliance concern. Therefore, at THERMAL POWER levels

< 25% RTP, the reactor is operating with substantial margin l

j to the MCPR limits and this LCO is not required.

]

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

taken to restore the MCPR(s) to within the required limits such that the plant remains operating within analyzed a

conditions. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is normally i

sufficient to restore the MCPR(s) to within its limits and j

is acceptable based on the low probability of a transient or DBA occurring simultaneously with the MCPR out of i

4 specification.

l 4

h Em1 If the MCPR cannot be restored to within the required limits within the associated Completion Time, the plant must be i

brought to a MODE or other specified condition in which the LC0 does not apply.

To achieve this status, THERMAL POWER must be reduced to < 25% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

The allowed Completion Time is reasonable, based on operating l

experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems.

s s

6 i

(continued)

ABWR TS B 3.2-3 P&R, 08/11/93 J

MCPR B 3.2.2 BASES (continued) l SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The MCPR is required to be initially calculated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter THERMAL POWER is 2 25% RTP and then every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.

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

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance after i

THERMAL POWER reaches 2 25% RTP is acceptable given the

~

large inherent margin to operating limits at low power levels.

i 1

REFERENCES 1.

NUREG-0562, June 1979.

1 2.

NED0-24011-P-A, " General Electric Standard Application I

for Reactor fuel," September 1988.

l 3.

NE00-30131-A, " Steady State Nuclear Methods," May 1985.

i

)

4.

NED0-24154, "Qualificatirsn of the One-Dimensional Core Transient Model for Boiling Water Reactors," October a

1978.

5.

"BWR/6 Generic Rod Withdrawa! Error Analysis,"

i Appendix 158, General Electric Standard Safety Analysis Report, GESSAR.

i l

i i

l l

I i

i i

ABWR TS B 3.2-4 P&R, 08/11/93 l

i LHGR (Non-GE Fuel)

B 3.2.3 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) (Non-GE Fuel)

BASES BACKGROUND The LHGR is a measure of the heat generation rate of a fuel rod in a fuel assembly at any axial location.

Limits on the LHGR are specified to ensure that fuel design limits are not exceeded anywhere in the core during normal operation, including anticipated operational occurrences (A00s).

4 Exceeding the LHGR limit could potentially result in fuel damage and subsequent release of radioactive materials.

fuel design limits are specified to ensure that fuel system a

damage, fuel rod failure or inability to cool the fuel does l

not occur during the anticipated operating conditions identified in Reference 1.

i APPLICABLE The analytical methods cnd assumptions used in evaluating SAFETY ANALYSES the fuel system design are presented in References 1 and 2.

The fuel assembly is designed to ensure (in conjunction with the core nuclear and thermal hydraulic design, plant equipment, instrumentation, and protection system) that fuel damage will not result in the release of radioactive materials in excess of the guidelines of 10 CFR, Parts 20, t

50, and 100. The mechanisms that could cause fuel damage during operational transients and that are considered in i

fuel evaluations are:

Rupture of the fuel rod cladding caused by strain from a.

the relative expansion of the U0, pellet; and b.

Severe overheating of the fuel rod cladding caused by inadequate cooling.

A value of 1% plastic strain of the Zircaloy cladding has been defined as the limit below which fuel damage caused by i

overstraining of the fuel cladding is not expected to occur (Reference 3).

The MCPR Safety Limit ensures that fuel damage caused by severe overheating of the fuel rod claddicy is avoided.

Fuel desigr evaluations have been performed and demonstrate that the 1% fuel cladding plastic strain design limit is not exceeded during continuous operation with LHGRs up to the i

4 (continued)

ABWR TS B 3.2-1 P&R, 08/11/93 C

?

LHGR (Non-GE Fuel)

B 3.2.3 BASES APPLICABLE operating limit specified in the COLR.

The analysis also SAFETY ANALYSES includes allowances for short term transient operation above (continued) the operating limit to account for A00s, plus an allowance for densification power spiking.

The LHGR satisfies Criterion 2 of the NRC Policy Statement.

LCO The LHGR is a basic assumption in the fuel design analysis.

The fuel has been designed to operate at rated core power with suffici nt design margin to the LHGR calculated to cause a 1% lut

- ' adding plastic strain. The operating limit to accomp.

this objective is specified in the COLR.

APPLICABILITY The LHGR limits are derived from fuel design analysis thet is limiting at high power level conditions. At core thermal power levels < 25% RTP, the reactor is operating with a substantial nargin to the LHGR limits and, therefore, the Specification is only required when the reactor is operating at 2 25% RTP.

ACTIONS A.I If any LHGR exceeds its required limit, an assumption regarding an initial condition of the fuel design analysis is not met. Therefore, prompt action should be taken to a

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

Etl If the LHGR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 25% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The allowed (continued)

ABWR TS B 3.2-2 P&R, 08/11/93

LHGR (Non-GE Fuel)

B 3.2.3 f

BASES ACTIONS H21 (continued)

Completion Time is reasonable, based on operating 3

i experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.3.1 REQUIREMENTS The LMGR is required to be initially calculated within

[

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER is 2 25% RTP and then every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.

It is compared with the specified j

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

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

i REFERENCES 1.

[Non GE Fuel Analysis].

2.

ABWR SSAR, Chapter 4.

l 3.

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

l a

i i

[

I i

i ABWR TS B 3.2-3 PAR, 08/11/93