ML14267A230

B 3.5.1-1, Sits, Emergency Core Cooling System Through B 3.5.5-5, Stb

ML14267A230
Person / Time
Site:	Calvert Cliffs
Issue date:	09/19/2014
From:	Calvert Cliffs, Exelon Generation Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML14267A237	List: ML14267A221 ML14267A223 ML14267A224 ML14267A225 ML14267A226 ML14267A227 ML14267A228 ML14267A229 ML14267A230 ML14267A231 ML14267A232 ML14267A233 ML14267A234
References
Download: ML14267A230 (32)

Text

SITs B 3.5.1 B 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS)

B 3.5.1 Safety Injection Tanks (SITs)

BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.1-1 Revision 2 BACKGROUND The function of the four SITs is to inject large quantities of borated water to the reactor vessel following the blowdown phase of a large break loss of coolant accident (LOCA) and to provide inventory to help accomplish the refill phase that follows thereafter.

The blowdown phase of a large break LOCA is the initial period of the transient during which the Reactor Coolant

System (RCS) departs from equilibrium conditions, and heat

from fission product decay, hot internals, and the vessel

continues to be transferred to the reactor coolant. The

blowdown phase of the transient ends when the RCS pressure

falls to a value approaching that of the containment

atmosphere.

The refill phase of a LOCA follows immediately where reactor coolant inventory has vacated the core through steam

flashing and ejection out through the break. The core is

essentially in adiabatic heatup. The rest of the SITs' inventory is then available to help fill voids in the lower

plenum and reactor vessel downcomer to establish a recovery

water level at the bottom of the core and continue reflood of the core with the addition of safety injection water.

The SITs are pressure vessels partially filled with borated water and pressurized with nitrogen gas. The SITs are

passive components, since no operator or control action is

required for them to perform their function. Internal tank

pressure is sufficient to discharge the contents to the RCS, if RCS pressure decreases below the SIT pressure.

Each SIT is piped into an RCS cold leg via the injection lines utilized by the High Pressure Safety Injection and Low

Pressure Safety Injection (HPSI and LPSI) systems. Each SIT

is isolated from the RCS by a motor

-operated isolation valve and two check valves in series. The motor

-operated isolation valves are normally open, with power removed from

the valve motor to prevent inadvertent closure prior to or

during an accident.

SITs B 3.5.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.1-2 Revision 2 Additionally, the isolation valves are interlocked with the pressurizer pressure instrumentation channels

, to ensure that the valves will automatically open as RCS pressure

increases above SIT pressure

, and to prevent inadvertent closure prior to an accident. The valves also receive a

safety injection actuation signal (SIAS) to open. These features ensure that the valves meet the requirements of Reference 1 for "operating bypasses" and that the SITs will be available for injection without reliance on operator

action.

The SIT gas and water volumes, gas pressure, and outlet pipe size are selected to allow three of the four SITs to

partially recover the core before significant clad melting

or zirconium water reaction can occur following a LOCA. The

need to ensure that three SITs are adequate for this

function is consistent with the LOCA assumption that the

entire contents of one SIT will be lost via the break during the blowdown phase of a LOCA.

APPLICABLE The large break LOCA analyses at full power (Reference 2

, SAFETY ANALYSES Section 6.3

) credits the SITs. This is the Design Basis Accident (DBA) that establishes the acceptance limits for

the SITs. Reference to the analysis for this DBA is used to

assess changes to the SITs as they relate to the acceptance

limits.

In performing the large break LOCA calculations, conservative assumptions are made concerning the

availability of safety injection flow. These assumptions include signal generation time, equipment starting times, and delivery time due to system piping. In the early stages

of a large break LOCA with a loss of offsite power, the SITs

provide the sole source of makeup water to the RCS. (The

assumption of a loss of offsite power is required by

regulations.) This is because the LPSI pumps, HPSI pumps, and charging pumps cannot deliver flow until the diesel

generators start, come to rated speed, and go through their

timed loading sequence. In cold leg breaks, the entire

contents of one SIT are assumed to be lost through the break

during the blowdown and reflood phases.

SITs B 3.5.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.1-3 Revision 43 The limiting large break LOCA is a double ended guillotine cold leg break at the discharge of the reactor coolant pump.

During this event, the SITs discharge to the RCS as soon as

RCS pressure decreases to below SIT pressure.

N o credit is taken for the high and low pressure safety injection until 30 and 45 seconds after the receipt of SIAS, respectively, assuming no offsite power is available. No operator action is assumed during the blowdown stage of a large break LOCA.

This Limiting Condition for Operation (LCO) helps to ensure that the following acceptance criteria, established by

Reference 3 for the Emergency Core Cooling System (ECCS),

will be met following a LOCA: a. Maximum fuel element cladding temperature is 2200°F; b. Maximum cladding oxidation is 0.17 times the total cladding thickness before oxidation;

c. Maximum hydrogen generation from a zirconium water reaction is 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding

cylinders surrounding the fuel, excluding the cladding

surrounding the plenum volume, were to react; and

d. The core is maintained in a coolable geometry.

Since the SITs discharge during the blowdown phase of a

LOCA, they do not contribute to the long-term cooling

requirements of 10 CFR 50.46.

Since the SITs are passive components, single active

failures are not applicable to their operation. The SIT

isolation valves, however, are not single failure proof;

therefore, whenever the valves are open, power is removed

from their operators and the switch is key locked open.

These precautions ensure that the SITs are available during

an accident (Reference 2, Section 14.17). With power supplied to the valves, a single active failure could result in a valve closure, which would render one SIT unavailable

for injection. If a second SIT is lost through the break, only two SITs would reach the core. Since the only active

failure that could affect the SITs would be the closure of a SITs B 3.5.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.1-4 Revision 43 motor-operated outlet valve, the requirement to remove power from these eliminates this failure mode.

The minimum volume requirement for the SITs ensures that three SITs can provide adequate inventory to reflood the

core and downcomer following a LOCA.

The maximum volume limit is based on maintaining an adequate gas volume to ensure proper injection and the ability of the

SITs to fully discharge, as well as limiting the maximum

amount of boron inventory in the SITs.

A minimum water level is used in the safety analysis as the volume in the SITs. To provide margin, the low level alarms

are set at 187 inches (corresponding to 1113 cubic feet) and

199 inches (corresponding to 1179 cubic feet). The analyses

are based upon the cubic feet requirements; the level (inches) figures are provided for operator use because the

level indicator provided in the Control Room is marked in

inches, not in cubic feet.

The minimum nitrogen cover pressure requirement ensures that the contained gas volume will generate discharge flow rates

during injection that are consistent with those assumed in

the safety analyses.

The maximum nitrogen cover pressure limit ensures that excessive amounts of gas will not be injected into the RCS

after the SITs have emptied.

A minimum pressure of 195 psia is used in the analyses. To allow for instrument accuracy, a 200 psig minimum and

250 psig maximum are specified. The maximum allowable boron

concentration of 2700 ppm is based upon boron precipitation limits in the core following a LOCA. Establishing a maximum limit for boron is necessary since the time at which boron

precipitation would occur in the core following a LOCA is a

function of break location, break size, the amount of boron

injected into the core, and the point of ECCS injection.

Post-LOCA emergency procedures directing the operator to

establish simultaneous hot and cold leg injection are based

on the worst case minimum boron precipitation time.

Maintaining the maximum SIT boron concentration within the SITs B 3.5.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.1-5 Revision 43 upper limit ensures that the SITs do not invalidate this calculation. An excessive boron concentration in any of the

borated water sources used for injection during a LOCA could

result in boron precipitation earlier than predicted.

The minimum boron requirements of 2300 ppm are based on beginning-of-life reactivity values and are selected to

ensure that the reactor will remain subcritical during the

reflood stage of a large break LOCA. During a large break

LOCA, all control element assemblies are assumed not to

insert into the core, and the initial reactor shutdown is

accomplished by void formation during blowdown. Sufficient

boron concentration must be maintained in the SITs to

prevent a return to criticality during reflood.

The SITs satisfy 10 CFR 50.36(c)(2)(ii), Criterion 3.

LCO The LCO establishes the minimum conditions required to ensure that the SITs are available to accomplish their core

cooling safety function following a LOCA. Four SITs are

required to be OPERABLE to ensure that 100% of the contents, for three of the SITs, will reach the core during a LOCA.

This is consistent with the assumption that the contents of one tank spill through the break. If the contents of fewer

than three tanks are injected during the blowdown phase of a

LOCA, the ECCS acceptance criteria of Reference 3 could be

violated.

For a SIT to be considered OPERABLE, the isolation valve must be fully open, power removed above 2000 psig, and the

limits established in the Surveillance Requirement (SR) for

contained volume, boron concentration, and nitrogen cover pressure must be met.

APPLICABILITY In MODEs 1, 2, and 3 the SIT OPERABILITY requirements are based on an assumption of full power operation. Although

cooling requirements decrease as power decreases, the SITs

are still required to provide core cooling as long as

elevated RCS pressures and temperatures exist.

In MODEs 4, 5, and 6, the SIT motor-operated isolation valves are closed to isolate the SITs from the RCS. This SITs B 3.5.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.1-6 Revision 43 allows RCS cooldown and depressurization without discharging the SITs into the RCS or requiring depressurization of the SITs. ACTIONS A.1 If the boron concentration of one SIT is not within limits

it must be returned to within the limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

In this condition, ability to maintain subcriticality or

minimum boron precipitation time may be reduced, but the

reduced concentration effects on core subcriticality during

reflood are minor. Boiling of the ECCS water in the core

during reflood concentrates the boron in the saturated

liquid that remains in the core. In addition, the volume of

the SIT is still available for injection. Since the boron

requirements are based on the average boron concentration of

the total volume of three SITs, the consequences are less

severe than they would be if an SIT were not available for

injection. Thus, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed to return the boron

concentration to within limits.

B.1 If one SIT is inoperable, for reasons other than boron

concentration, the SIT must be returned to OPERABLE status

within one hour. In this Condition, the required contents

of three SITs cannot be assumed to reach the core during a

LOCA. Due to the severity of the consequences should a LOCA

occur in these conditions, the one hour Completion Time to

open the valve, remove power from the valve, or restore

proper water volume or nitrogen cover pressure, ensures that

prompt action will be taken to return the inoperable accumulator to OPERABLE status. The Completion Time minimizes the exposure of the plant to a LOCA in these

conditions.

C.1 and C.2 If the SIT cannot be restored to OPERABLE status within the

associated Completion Time, the plant must be brought to a

MODE in which the LCO does not apply. To achieve this

status, the plant must be brought to at least MODE 3 within

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion

Times are reasonable, based on operating experience, to

reach the required plant conditions from full power SITs B 3.5.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.1-7 Revision 43 conditions in an orderly manner and without challenging plant systems.

D.1 If more than one SIT is inoperable, the unit is in a

condition outside the accident analyses. Therefore, LCO 3.0.3 must be entered immediately.

SURVEILLANCE SR 3.5.1.1 REQUIREMENTS Verification every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that each SIT isolation valve is fully open, as indicated in the Control Room, ensures that

SITs are available for injection and ensures timely

discovery if a valve should be partially closed. If an

isolation valve is not fully open, the rate of injection to

the RCS would be reduced. Although a motor-operated valve

should not change position with power removed, a closed

valve could result in not meeting accident analysis

assumptions. A 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is considered reasonable

in view of other administrative controls that ensure the

unlikelihood of a mispositioned isolation valve.

SR 3.5.1.2 and SR 3.5.1.3 Safety injection tank borated water volume and nitrogen

cover pressure should be verified to be within specified

limits every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in order to ensure adequate injection

during a LOCA. Due to the static design of the SITs, a

12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency usually allows the operator sufficient

time to identify changes before the limits are reached.

Operating experience has shown this Frequency to be appropriate for early detection and correction of off normal trends.

SR 3.5.1.4 Six months is reasonable for verification by sampling to

determine that each SIT's boron concentration is within the

required limits, because the static design of the SITs

limits the ways in which the concentration can be changed.

This Frequency is adequate to identify changes that could

occur from mechanisms, such as stratification or inleakage.

SITs B 3.5.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.1-8 Revision 43 Verification consists of monitoring inleakage or sampling.

The inleakage is monitored every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by monitoring tank

level. Sampling of each tank is done every six months. All

intentional sources of level increase are maintained

administratively to ensure SIT boron concentrations are

within technical specification limits. The boron concentration of each tank is verified prior to startup from outages. A sample of the SIT is required, to verify boron

concentration, if 10 inches or greater of inleakage has

occurred since last sampled.

Sampling the affected SIT (by taking the sample at the discharge of the operating HPSI pump) within one hour prior

to a 1% volume increase of normal tank volume, will ensure

the boron concentration of the fluid to be added to the SIT

is within the required limit prior to adding inventory to

the SIT(s).

SR 3.5.1.5 Verification every 31 days that power is removed from each

SIT isolation valve operator, by maintaining the feeder

breaker open under administrative control, when the pressurizer pressure is 2000 psig ensures that an active failure could not result in the undetected closure of an SIT motor-operated isolation valve. If this were to occur, only

two SITs would be available for injection, given a single

failure coincident with a LOCA. Since installation and

removal of power to the SIT isolation valve operators is

conducted under administrative control, the 31 day Frequency

was chosen to provide additional assurance that power is

removed.

This SR allows power to be supplied to the motor-operated isolation valves when RCS pressure is < 2000 psig, thus

allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers during unit startups or shutdowns. Even with power supplied to the valves, inadvertent closure is prevented by the RCS pressure

interlock associated with the valves. Should closure of a

valve occur in spite of the interlock, the safety injection

signal provided to the valves would open a closed valve in the event of a LOCA.

SITs B 3.5.1 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.1-9 Revision 43 REFERENCES 1. Institute of Electrical and Electronic Engineers Standard 279-1971, "IEEE Standard: Criteria for Protection Systems for Nuclear Power Generating

Stations" 2. Updated Final Safety Analysis Report (UFSAR)

3. 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plants" ECCS - Operating B 3.5.2 B 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS)

B 3.5.2 ECCS - Operating

BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.2-1 Revision 15 BACKGROUND The function of the ECCS is to provide core cooling and negative reactivity, to ensure that the reactor core is protected after any of the following accidents: a. Loss of coolant accident; b. Control element assembly ejection accident; c. Secondary event, including uncontrolled steam release or excess feedwater heat removal event; and d. Steam generator tube rupture.

The addition of negative reactivity by the ECCS during a secondary event where primary cooldown could add enough

positive reactivity to achieve criticality and return to

significant power was considered in design requirements for

the ECCS.

There are two phases of ECCS operation: injection and recirculation. In the injection phase, all injection is

initially added to the RCS via the cold legs. After the

refueling water tank (RWT) has been depleted, the ECCS

recirculation phase is entered as the ECCS suction is

automatically transferred to the containment sump.

Two redundant, 100% capacity trains are provided. In MODEs 1 and 2, and MODE 3 with pressurizer pressure 1750 psia, each train consists of HPSI and LPSI charging subsystems. In MODEs 1 and 2, and MODE 3 with pressurizer

pressure 1750 psia, both trains must be OPERABLE. This ensures that 100% of the core cooling requirements can be provided in the event of a single active failure.

A suction header supplies water from the RWT or the containment sump to the ECCS pumps. Separate piping

supplies each train. The discharge headers from each HPSI pump divide into four supply lines. Both HPSI trains feed into each of the four injection lines. The discharge header

which is fed from both LPSI pumps divides into four supply

lines, each feeding the injection line to each RCS cold leg.

ECCS - Operating B 3.5.2 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.2-2 Revision 15 For LOCAs that are too small to initially depressurize the RCS below the shutoff head of the HPSI pumps, the steam generators must provide the core cooling function.

During low temperature conditions in the RCS, limitations are placed on the maximum number of HPSI pumps that may be OPERABLE. Refer to LCO 3.4.12 Bases, for the basis of these requirements.

During a large break LOCA, RCS pressure will decrease to

< 200 psia in

< 20 seconds. The safety injection systems are actuated upon receipt of a SIAS. If offsite power is

available, the safeguard loads start immediately. If

offsite power is not available, the engineered safety

feature (ESF) buses shed normal operating loads and are

connected to the diesel generators. Safeguard loads are

then actuated in the programmed time sequence. The time

delay associated with diesel starting, sequenced loading, and pump starting determines the time required before pumped

flow is available to the core following a LOCA.

The active ECCS components, along with the passive SITs and RWT, covered in LCO 3.5.1 and LCO 3.5.4, provide the cooling

water necessary to meet Reference 1, Appendix 1C, Criterion 44.

APPLICABLE The LCO helps to ensure that the following acceptance SAFETY ANALYSES criteria, established by Reference 2 for ECCSs, will be met following a LOCA: a. Maximum fuel element cladding temperature is 2200°F; b. Maximum cladding oxidation is 0.17 times the total cladding thickness before oxidation;

c. Maximum hydrogen generation from a zirconium water reaction is 0.01 times the hypothetical amount generated if all of the metal in the cladding cylinders

surrounding the fuel, excluding the cladding

surrounding the plenum volume, were to react;

d. Core is maintained in a coolable geometry; and

e. Adequate long-term core cooling capability is maintained.

ECCS - Operating B 3.5.2 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.2-3 Revision 15 The LCO also limits the potential for a post-trip return to power, following a steam line break, and ensures that

containment temperature limits are met.

Both HPSI and LPSI subsystems are assumed to be OPERABLE in the large break LOCA analysis at full power (Reference 1, Section 14.17). This analysis establishes a minimum required runout flow for the HPSI and LPSI pumps, as well as

the maximum required response time for their actuation. The

HPSI pumps are credited in the small break LOCA analysis.

This analysis establishes the flow and discharge head

requirements at the design point for the HPSI pump. The

steam generator tube rupture and steam line break analyses

also credit the HPSI pumps, but are not limiting in their

design.

The large break LOCA event with a loss of offsite power and a single failure (disabling one ECCS train) establishes the

OPERABILITY requirements for the ECCS. During the blowdown

stage of a LOCA, the RCS depressurizes as primary coolant is

ejected through the break into Containment. The nuclear

reaction is terminated either by moderator voiding during

large breaks or control element assembly insertion during

small breaks. Following depressurization, emergency cooling

water is injected into the cold legs, flows into the

downcomer, fills the lower plenum, and refloods the core.

On smaller breaks, RCS pressure will stabilize at a value dependent upon break size, heat load, and injection flow.

The smaller the break, the higher this equilibrium pressure.

In all LOCA analyses, injection flow is not credited until

RCS pressure drops below the shutoff head of the HPSI pumps.

The LCO ensures that an ECCS train will deliver sufficient water to match decay heat boiloff rates soon enough to

minimize core uncovery for a large LOCA. It also ensures

that the accident assumptions are met for the small break

LOCA and steam line break. For smaller LOCAs the steam generators serve as the heat sink to provide core cooling.

Emergency Core Cooling System - Operating satisfies 10 CFR 50.36(c)(2)(ii), Criterion 3.

ECCS - Operating B 3.5.2 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.2-4 Revision 15 LCO In MODEs 1, 2, and 3, with pressurizer pressure 1750 psia, two independent (and redundant) ECCS trains are required to ensure that sufficient ECCS flow is available, assuming there is a single failure affecting

either train. Additionally, individual components within

the ECCS trains may be called upon to mitigate the consequences of other transients and accidents.

In MODEs 1 and 2, and in MODE 3 with pressurizer pressure 1750 psia, an ECCS train consists of a HPSI subsystem, and a LPSI subsystem.

Each HPSI and LPSI train includes the piping, instruments, and controls to ensure the availability of an OPERABLE flow

path capable of taking suction from the RWT on a SIAS and

automatically transferring suction to the containment sump

upon a recirculation actuation signal.

During an event requiring ECCS actuation, a flow path is provided to ensure an abundant supply of water from the RWT

to the RCS, via the HPSI and LPSI pumps and their respective

supply headers, to each of the four cold leg injection

nozzles. In the long-term, this flow path may be switched

to take its supply from the containment sump and to supply

part of its flow to the RCS hot legs via the pressurizer or

the shutdown cooling (SDC) suction nozzles.

The flow path for each train must maintain its designed

independence to ensure that no single failure can disable

both ECCS trains.

In addition for the HPSI pump system to be considered OPERABLE, each HPSI pump system (consisting of a HPSI pump

and one of two safety injection headers) must have balanced

flows, such that the sum of the flow rates of the three lowest flow legs is

> 470 gpm.

APPLICABILITY In MODEs 1 and 2, and in MODE 3 with RCS pressure 1750 psia, the ECCS OPERABILITY requirements for the limiting DBA large break LOCA are based on full power

operation. Although reduced power would not require the

same level of performance, the accident analysis does not

provide for reduced cooling requirements in the lower MODEs.

ECCS - Operating B 3.5.2 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.2-5 Revision 15 The HPSI pump performance is based on the small break LOCA, which establishes the pump performance curve and has less

dependence on power. The requirements of MODE 2, and MODE 3 with RCS pressure 1750 psia, are bounded by the MODE 1 analysis.

The ECCS functional requirements of MODE 3, with RCS pressure < 1750 psia, and MODE 4 are described in LCO 3.5.3.

In MODEs 5 and 6, unit conditions are such that the probability of an event requiring ECCS injection is

extremely low. Core cooling requirements in MODE 5 are

addressed by LCO 3.4.7 and LCO 3.4.8. MODE 6 core cooling requirements are addressed by LCO 3.9.4 and LCO 3.9.5.

ACTIONS A.1 If one or more trains are inoperable and at least 100% of

the ECCS flow equivalent to a single OPERABLE ECCS train is

available, the inoperable components must be returned to

OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion

Time is based on an Nuclear Regulatory Commission study (Reference 3) using a reliability evaluation and is a

reasonable amount of time to effect many repairs.

An ECCS train is inoperable if it is not capable of delivering the design flow to the RCS. The individual

components are inoperable if they are not capable of

performing their design function, or if supporting systems

are not available.

The LCO requires the OPERABILITY of a number of independent subsystems. Due to the redundancy of trains and the

diversity of subsystems, the inoperability of one component

in a train does not render the ECCS incapable of performing

its function. Neither does the inoperability of two

different components, each in a different train, necessarily

result in a loss of function for the ECCS. The intent of

this Condition is to maintain a combination of OPERABLE

equipment such that 100% of the ECCS flow equivalent to 100%

of a single OPERABLE train remains available. This allows

increased flexibility in plant operations when components in

opposite trains are inoperable.

ECCS - Operating B 3.5.2 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.2-6 Revision 46 An event accompanied by a loss of offsite power and the failure of an emergency diesel generator can disable one

ECCS train until power is restored. A reliability analysis (Reference 3) has shown that the impact with one full ECCS

train inoperable is sufficiently small to justify continued operation for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Reference 4 describes situations in which one component, such as a SDC total flow control valve, can disable both

ECCS trains. With one or more components inoperable, such

that 100% of the equivalent flow to a single OPERABLE ECCS

train is not available, the facility is in a condition

outside the accident analyses. Therefore, LCO 3.0.3 must be

immediately entered.

B.1 and B.2 If the inoperable train cannot be restored to OPERABLE

status within the associated Completion Time, the plant must

be brought to a MODE in which the LCO does not apply. To

achieve this status, the plant must be brought to at least

MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and pressurizer pressure reduced to

< 1750 psia within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.5.2.1 REQUIREMENTS Verification of proper valve position ensures that the flow path from the ECCS pumps to the RCS is maintained.

Misalignment of these valves could render both ECCS trains

inoperable. MOV-659 and MOV-660 are secured in position by interrupting the control signal to the valve operator via a key switch in the Control Room. Power is removed from the valve operator for CV-306 by isolating the air supply to the valve positioner. These actions ensure that the valves cannot be inadvertently misaligned. These valves are of the type described in Reference 4, which can disable the

function of both ECCS trains and invalidate the accident

analysis. A 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is considered reasonable in

view of other administrative controls ensuring that a

mispositioned valve is an unlikely possibility.

ECCS - Operating B 3.5.2 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.2-7 Revision 38 SR 3.5.2.2 Verifying the correct alignment for manual, power-operated, and automatic valves in the ECCS flow paths provides

assurance that the proper flow paths will exist for ECCS

operation. This SR does not apply to valves that are

locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that

receives an actuation signal is allowed to be in a

non-accident position provided the valve automatically

repositions within the proper stroke time. This SR does not

require any testing or valve manipulation. Rather, it

involves verification that those valves capable of being

mispositioned are in the correct position.

The 31 day Frequency is appropriate because the valves are operated under procedural control and an improper valve

position would only affect a single train. This Frequency

has been shown to be acceptable through operating

experience.

SR 3.5.2.3 Periodic surveillance testing of the HPSI and LPSI pumps to

detect gross degradation caused by impeller structural

damage or other hydraulic component problems is required by

the American Society of Mechanical Engineers Code. This type of testing may be accomplished by measuring the pump

developed head at only one point of the pump characteristic

curve. This verifies both that the measured performance is

within an acceptable tolerance of the original pump baseline

performance and that the performance at the test flow is

greater than or equal to the performance assumed in the unit

safety analysis. Surveillance Requirements are specified in

the Inservice Testing Program, which encompasses American Society of Mechanical Engineers Code. American Society of Mechanical Engineers Code provides the activities and Frequencies necessary to satisfy the requirements.

SR 3.5.2.4 The Surveillance Requirement was deleted in Amendment

Nos. 260/237.

ECCS - Operating B 3.5.2 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.2-8 Revision 38 SR 3.5.2.5, SR 3.5.2.6, and SR 3.5.2.7 These SRs demonstrate that each automatic ECCS valve actuates to the required position on an actual, or simulated

SIAS, and on a recirculation actuation signal; that each

ECCS pump starts on receipt of an actual or simulated SIAS;

and that the LPSI pumps stop on receipt of an actual or simulated recirculation actuation signal. This Surveillance is not required for valves that are locked, sealed, or

otherwise secured in the required position under

administrative controls. In order to assure the results of

the low temperature overpressure protection analysis remain

bounding, whenever flow testing into the RCS is required at RCS temperatures 365°F (Unit 1), 301°F (Unit 2), the HPSI pump shall recirculate RCS water (suction from the RWT isolated) or the requirements of LCO 3.4.12, shall be

satisfied. The 24 month Frequency is based on the need to

perform these surveillance tests under the conditions that

apply during a plant outage and the potential for unplanned

transients if the surveillance tests were performed with the

reactor at power. The 24 month Frequency is also acceptable

based on consideration of the design reliability (and

confirming operating experience) of the equipment. The

actuation logic is tested as part of the Engineered Safety

Feature Actuation System testing, and equipment performance

is monitored as part of the Inservice Testing Program.

SR 3.5.2.8 Periodic inspection of the containment sump ensures that it

is unrestricted and stays in proper operating condition.

The 24 month Frequency is based on the need to perform this

surveillance test under the conditions that apply during an

outage, on the need to have access to the location, and on

the potential for unplanned transients if the surveillance

test were performed with the reactor at power. This

Frequency is sufficient to detect abnormal degradation and is confirmed by operating experience.

SR 3.5.2.9 Verifying that the SDC System open-permissive interlock is

OPERABLE ensures that the SDC suction isolation valves are

prevented from being remotely opened when RCS pressure, is

at or above, the SDC System design suction pressure of ECCS - Operating B 3.5.2 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.2-9 Revision 38 350 psia. The suction piping of the LPSI pumps, is the SDC component with the limiting design pressure rating. The

interlock provides assurance that double isolation of the

SDC System from the RCS is preserved whenever RCS pressure, is at or above, the design pressure. The 309 psia value

specified in the Surveillance is the actual pressurizer pressure at the instrument tap elevation for PT-103 and PT-103-1 when the SDC System suction pressure is 350 psia.

The procedure for this surveillance test contains the

required compensation to be applied to this value to account

for instrument uncertainties. This surveillance test is

normally performed using a simulated RCS pressure input

signal. The 24 month Frequency is based on the need to

perform this surveillance test under conditions that apply

during an outage. The 24 month Frequency is also acceptable

based on consideration of the design reliability (and confirming operating experience) for the equipment.

REFERENCES 1. UFSAR

2. 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plants" 3. Nuclear Regulatory Commission Memorandum to V. Stello, Jr., from R. L. Baer, "Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975 4. Inspection and Enforcement Information Notice No. 87-01, "RHR Valve Misalignment Causes Degradation of ECCS in PWRs," January 6, 1987

ECCS - Shutdown B 3.5.3 B 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS)

B 3.5.3 ECCS - Shutdown

BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.3-1 Revision 2 BACKGROUND The Background S ection for B 3.5.2, is applicable to these Bases, with the following modification.

In MODE 3 with pressurizer pressure 1750 psia and in MODE 4, an ECCS train is defined as one HPSI subsystem. The HPSI flow path consists of piping, valves, and pumps that enable water from the RWT to be injected into the RCS following the accidents described in B 3.5.2.

APPLICABLE The Applicable Safety Analyses section of B 3.5.2 is SAFETY ANALYSES applicable to these Bases.

Due to the stable conditions associated with operation in MODE 3 with RCS pressure 1750 psia and MODE 4, and the reduced probability of a DBA , the ECCS operational requirements are reduced. Included in these reductions is that certain automatic SIASs are not available. In this MODE, sufficient time exists for manual actuation of the

required ECCS to mitigate the consequences of a DBA.

Only one train of ECCS is required for MODE 3 with RCS pressure 1750 psia and MODE 4. Protection against single failures is not relied on for this MODE of operation.

Emergency Core Cooling System - Shutdown satisfies 10 CFR 50.36(c)(2)(ii), Criterion 3.

LCO In MODE 3 with pressurizer pressure 1750 psia and MODE 4, an ECCS subsystem is composed of a single HPSI subsystem.

Each HPSI subsystem includes the piping, instruments, and

controls to ensure an OPERABLE flow path capable of taking

suction from the RWT and transferring suction to the

containment sump.

During an event requiring ECCS actuation, a flow path is required to supply water from the RWT to the RCS via the

HPSI pumps and their respective supply headers to each of

the four cold leg injection nozzles. In the long

-term, this flow path may be switched to take its supply from the ECCS - Shutdown B 3.5.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.3-2 Revision 2 containment sump and to deliver its flow to the RCS hot and cold legs.

With RCS pressure

< 1750 psia, one HPSI pump is acceptable without single failure consideration, based on the stable reactivity condition of the reactor

, and the limited core cooling requirements. The LPSI pumps may therefore be released from the ECCS train for use in SDC. In MODE 3 with RCS cold leg temperature 365°F (Unit 1), 301°F (Unit 2), a maximum of one HPSI pump is allowed to be OPERABLE in

accordance with LCO 3.4.12.

The LCO is modified by a Note which allows the HPSI train to

not be capable of automatically starting on an actuation signal when RCS cold leg temperature is

< 385°F (Unit 1), < 325°F (Unit 2), during heatup and cooldown and when

< 365°F (Unit 1), < 301°F (Unit 2), during other conditions.

This allowance is necessary to ensure low temperature overpressure protection analysis assumptions are maintained.

The LCO Note provides a transition period

[between 385

°F and 365°F (Unit 1), between 325

°F and 301°F (Unit 2)

] where the OPERABLE HPSI pump will be placed in pull-to-lock on a cooldown and restored to automatic status on heatup (see LCO 3.4.12

). At 365°F and less (Unit 1), 301

°F and less (Unit 2), the required HPSI pump shall be placed in pull-to-lock and will not start automatically. The HPSI pumps and

HPSI header isolation valves are required to be out of

automatic when o perating within the MODE s of Applicability for the Low Temperature Overpressure Protection System (LCO 3.4.12).

APPLICABILITY In MODE s 1, 2, and 3 with RCS pressure 1750 psia, the OPERABILITY requirements for ECCS are covered by LCO 3.5.2.

In MODE 3 with RCS pressure

< 1750 psia and in MODE 4, one OPERABLE ECCS train is acceptable without single failure consideration, based on the stable reactivity condition of

the reactor

, and the limited core cooling requirements.

In MODE s 5 and 6, unit conditions are such that the probability of an event requiring ECCS injection is

extremely low. Core cooling requirements in MODE 5 are ECCS - Shutdown B 3.5.3 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.3-3 Revision 26 addressed by LCO 3.4.7 and LCO 3.4.8. MODE 6 core cooling requirements are addressed by LCO 3.9.4 and LCO 3.9.5.

ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable ECCS HPSI subsystem. There is an increased risk associated with entering MODE 4 from MODE 5 with an inoperable ECCS HPSI subsystem and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.

A.1 With no HPSI pump OPERABLE, the unit is not prepared to

respond to a LOCA. The one hour Completion Time to restore

at least one HPSI train to OPERABLE status, ensures that

prompt action is taken to restore the required cooling

capacity or to initiate actions to place the unit in MODE 5, where an ECCS train is not required.

B.1 When the Required Action cannot be completed within the

required Completion Time, a controlled shutdown should be

initiated. Twenty-four hours is reasonable, based on

operating experience, to reach MODE 5 in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.5.3.1 REQUIREMENTS The applicable SR descriptions from B 3.5.2 apply.

REFERENCES The applicable references from B 3.5.2 apply.

RWT B 3.5.4 B 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS)

B 3.5.4 Refueling Water Tank (RWT)

BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.4-1 Revision 2 BACKGROUND The RWT supports the ECCS and the Containment Spray System by providing a source of borated water for ESF pump operation.

The RWT supplies two ECCS trains by separate, redundant supply headers. Each header also supplies one train of the Containment Spray System. A motor

-operated isolation valve is provided in each header

, to allow the operator to isolate the usable volume of the RWT from the ECCS after the ESF

pump suction has been transferred to the containment sump

, following depletion of the RWT during a LOCA. A separate header is used to supply the Chemical and Volume Control

System from the RWT. Use of a single RWT to supply both

trains of the ECCS is acceptable

, since the RWT is a passive component, and passive failures are not assumed to occur

coincidentally with the Design Basis Event during the

injection phase of an accident. Not all the water stored in

the RWT is available for injection following a LOCA; the

location of the ECCS suction piping in the RWT will result

in some portion of the stored volume being unavailable.

The HPSI , LPSI , and containment spray pumps are provided with recirculation lines that ensure each pump can maintain

minimum flow requirements when operating at shutoff head

conditions. These lines discharge back to the RWT, which

vents to the atmosphere. When the suction for the HPSI and

containment spray pumps is transferred to the containment

sump, this flow path must be isolated to prevent a release

of the containment sump contents to the RWT. If not

isolated, this flow path could result in a release of

contaminants to the atmosphere and the eventual loss of

suction head for the ESF pumps.

This LCO ensures that: a. The RWT contains sufficient borated water to support the ECCS during the injection phase; b. Sufficient water volume exists in the containment sump to support continued operation of the ESF pumps at the RWT B 3.5.4 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.4-2 Revision 41 time of transfer to the recirculation mode of cooling; and c. The reactor remains subcritical following a LOCA.

Insufficient water inventory in the RWT could result in

insufficient cooling capacity of the ECCS when the transfer to the recirculation mode occurs. Improper boron concentrations could result in a reduction of SDM or

excessive boric acid precipitation in the core following a

LOCA, as well as excessive caustic stress corrosion of mechanical components and systems inside Containment.

APPLICABLE During accident conditions, the RWT provides a source of SAFETY ANALYSES borated water to the HPSI, LPSI, containment spray, and charging pumps when level is low in the boric acid tanks.

As such, it provides containment cooling and

depressurization, core cooling, and replacement inventory, and is a source of negative reactivity for reactor shutdown (Reference 1). The design basis transients and applicable

safety analyses concerning each of these systems are

discussed in the Applicable Safety Analyses Section of

B 3.5.2 and B 3.6.6. These analyses are used to assess

changes to the RWT in order to evaluate their effects in

relation to the acceptance limits.

The volume limit of 400,000 gallons is based on two factors: a. Sufficient deliverable volume must be available to provide at least 30 minutes of full flow from all ESF pumps prior to reaching a low level switchover to the containment sump for recirculation; and b. The containment sump water volume must be sufficient to support continued ESF pump operation after the

switchover to recirculation occurs. This sump volume

water inventory is supplied by the RWT borated water

inventory.

When ESF pump suction is transferred to the sump, there must be sufficient water in the sump to ensure adequate net

positive suction head for the HPSI and containment spray

pumps. The RWT capacity must be sufficient to supply this RWT B 3.5.4 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.4-3 Revision 2 amount of water without considering the inventory added from the SITs or RCS , but accounting for loss of inventory to containment subcompartments and reservoirs due to

containment spray operation and to areas outside containment

due to leakage from ECCS injection and recirculation

equipment.

The 2300 ppm limit for minimum boron concentration was

established to ensure that, following a LOCA with a minimum

level in the RWT, the reactor will remain subcritical in the

cold condition following mixing of the RWT and RCS water

volumes with all control rods inserted, except for the

control element assembly of highest worth, which is withdrawn from the core. The most limiting case occurs at

beginning of core life.

The maximum boron limit of 2700 ppm in the RWT is based on

boron precipitation in the core following a LOCA. With the

reactor vessel at saturated conditions, the core dissipates

heat by pool nucleate boiling. Because of this boiling

phenomenon in the core, the boric acid concentration will

increase in this region. If allowed to proceed in this

manner, a point will be reached where boron precipitation

will occur in the core. Post

-LOCA emergency procedures direct the operator to establish simultaneous hot and cold

leg injection to prevent this condition by establishing a

forced flow path through the core regardless of break

location. These procedures are based on the minimum time in

which precipitation could occur, assuming that maximum boron

concentrations exist in the borated water sources used for

injection following a LOCA. Boron concentrations in the RWT

in excess of the limit could result in precipitation earlier

than assumed in the analysis.

The upper limit of 100

°F (only required for MODE 1 operation) and the lower limit of 40

°F (RWT temperature

), are the limits assumed in the accident analysis.

The RWT satisfies 10 CFR 50.36(c)(2)(ii), Criterion 3.

LCO The RWT ensures that an adequate supply of borated water is

available to

cool and depressurize the C ontainment in the event of a DBA , to cool and cover the core in the event of a RWT B 3.5.4 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.4-4 Revision 2 LOCA, ensure that the reactor remains subcritical following a DBA, and ensure that an adequate level exists in the containment sump to support ESF pump operation in the recirculation mode.

To be considered OPERABLE, the RWT must meet the limits established in the SRs for water volume, boron

concentration, and temperature.

APPLICABILITY In MODE s 1, 2, 3, and 4, the RWT OPERABILITY requirements are dictated by the ECCS and Containment Spray System

OPERABILITY requirements. Since both the ECCS and the

Containment Spray System must be OPERABLE in MODE s 1, 2, 3, and 4, the RWT must be OPERABLE to support their operation.

Core cooling requirements in MODE 5 are addressed by

LCO 3.4.7 and LCO 3.4.8

. MODE 6 core cooling requirements are addressed by LCO 3.9.4 and LCO 3.9.5

.

ACTIONS A.1 With RWT boron concentration or borated water temperature

not within limits, it must be returned to within limits

within eight hours. In this condition neither the ECCS nor the Containment Spray System can perform their design

functions; therefore, prompt action must be taken to restore the tank to OPERABLE condition. The allowed Completion Time of eight hours to restore the RWT to within limits was developed considering the time required to change boron

concentration or temperature

, and that the contents of the tank are still available for injection.

Required Action A.1 only applies to the maximum borated

water temperature in MODE 1.

B.1 With RWT borated water volume not within limits, it must be

returned to within limits within one hour. In this condition, neither the ECCS nor Containment Spray System can

perform their design functions; therefore, prompt action

must be taken to restore the tank to OPERABLE status or to

place the unit in a MODE in which these systems are not

required. The allowed Completion Time of one hour to RWT B 3.5.4 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.4-5 Revision 2 restore the RWT to OPERABLE status is based on this condition simultaneously affecting multiple redundant

trains.

C.1 and C.2 If the RWT cannot be restored to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this

status, the plant must be brought to at least MODE 3 within

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed

Completion Times are reasonable, based on operating

experience, to reach the required plant conditions from full

power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.5.4.1 and SR 3.5.4.2 REQUIREMENTS

Refueling water tank borated water temperature shall be verified every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to be within the limits assumed in

the accident analysis. This Frequency has been shown to be

sufficient to identify temperature changes that approach

either acceptable limit.

The SRs are modified by a Note that eliminates the

requirement to perform this s urveillance test when ambient air temperatures are within the operating temperature limits

of the RWT. With ambient temperatures within this range, the RWT temperature should not exceed the limits.

Surveillance Requirement 3.5.4.2 is modified by an additional N ote which requires the S R to be met in MODE 1 only. A S R is "met" only when the acceptance criteria are satisfied. Known failure of the requirements of a S R , even without a s urveillance test specifically being "performed," constitutes a S R not "met." This reflects the maximum coolant temperature assumptions in the LOCA analysis.

SR 3.5.4.3 Above minimum RWT water volume level shall be verified every

seven days. This Frequency ensures that a sufficient initial water supply is available for injection and to

support continued ESF pump operation on recirculation.

RWT B 3.5.4 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.4-6 Revision 2 Since the RWT volume is normally stable and is provided with a Low Level Alarm, a seven day Frequency is appropriate and has been shown to be acceptable through operating

experience.

SR 3.5.4.4 Boron concentration of the RWT shall be verified every seven days to be within the required range. This Frequency ensures that the reactor will remain subcritical following a

LOCA. Further, it ensures that the resulting sump pH will

be maintained in an acceptable range such that boron

precipitation in the core will not occur earlier than

predicted , and the effect of chloride and caustic stress corrosion on mechanical systems and components will be

minimized. Since the RWT volume is normally stable, a

seven day sampling Frequency is appropriate and has been shown through operating experience to be acceptable.

REFERENCES 1.

UFSAR , Chapter s 6 , "Engineered Safety Features," and 14 , "Safety Analysis" STB B 3.5.5 B 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS)

B 3.5.5 Sodium Tetraborate (STB)

BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.5-1 Revision 37 BACKGROUND Sodium tetraborate decahydrate is placed in baskets on the floor of the Containment Building to ensure that iodine, which may be dissolved in the recirculated reactor cooling

water following a LOCA, remains in solution.

Sodium tetraborate also helps inhibit stress corrosion cracking (SCC) of austenitic stainless steel components in

Containment during the recirculation phase following an

accident.

Fuel that is damaged during a LOCA will release iodine in

several chemical forms to the reactor coolant and to the

containment atmosphere. A portion of the iodine in the

containment atmosphere is washed to the sump by containment

sprays. The emergency core cooling water is borated for

reactivity control. This borated water causes the sump

solution to be acidic. In a low pH (acidic) solution, dissolved iodine will be converted to a volatile form. The

volatile iodine will evolve out of solution into the

containment atmosphere, significantly increasing the levels

of airborne iodine. The increased levels of airborne iodine

in Containment contribute to the radiological releases and

increase the consequences from the accident due to

containment atmosphere leakage.

After a LOCA, the components of the core cooling and containment spray systems will be exposed to high

temperature borated water. Prolonged exposure to the core

cooling water combined with stresses imposed on the

components can cause SCC. The SCC is a function of stress, oxygen and chloride concentrations, pH, temperature, and

alloy composition of the components. High temperatures and

low pH, which would be present after a LOCA, tend to promote SCC. This can lead to the failure of necessary safety systems or components.

Adjusting the pH of the recirculation solution to levels 7.0 prevents a significant fraction of the dissolved iodine from converting to a volatile form. The higher pH

thus decreases the level of airborne iodine in Containment

and reduces the radiological consequences from containment STB B 3.5.5 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.5-2 Revision 37 atmosphere leakage following a LOCA. Maintaining the solution pH above 7.0 also reduces the occurrence of SCC of

austenitic stainless steel components in Containment.

Reducing SCC reduces the probability of failure of

components.

Granular STB decahydrate is employed as a passive form of pH control for post-LOCA containment spray and core cooling

water. Baskets of STB are placed on the floor in the Containment Building to dissolve from released reactor

coolant water and containment sprays after a LOCA.

Recirculation of the water for core cooling and containment sprays then provides mixing to achieve a uniform solution

pH. The decahydrate form of STB is used because of the high humidity in the Containment Building during normal

operation. Since the STB is hydrated, it is not likely to absorb large amounts of water from the humid atmosphere and

will undergo less physical and chemical change than the anhydrous form of STB. APPLICABLE The LOCA radiological consequences analysis takes credit for SAFETY ANALYSES iodine retention in the sump solution based on the recirculation water pH being 7.0. The radionuclide releases from the containment atmosphere and the

consequences of a LOCA would be increased if the pH of the

recirculation water were not adjusted to 7.0 or above.

Sodium tetraborate satisfies 10 CFR 50.36(c)(2)(ii), Criterion 3.

LCO The STB is required to adjust the pH of the recirculation water to 7.0 after a LOCA. A pH 7.0 is necessary to prevent significant amounts of iodine released from fuel failures and dissolved in the recirculation water from

converting to a volatile form and evolving into the

containment atmosphere. Higher levels of airborne iodine in

Containment may increase the release of radionuclides and the consequences of the accident. A pH

> 7.0 is also necessary to prevent SCC of austenitic stainless steel components in Containment. Stress corrosion cracking

increases the probability of failure of components.

STB B 3.5.5 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.5-3 Revision 37 The required amount of STB is based upon the extreme cases of water volume and pH possible in the containment sump after a large break LOCA. The minimum required mass is the mass of STB that will achieve a sump solution pH of 7.0 when taking into consideration the maximum possible sump

water volume, and the minimum possible pH. The amount of

STB needed in the Containment Building is based on the equivalent weight of STB required to achieve the desired pH.

The equivalent weight of STB is obtained using a measured volume of STB in Containment and the manufacturer's specified density.

Since STB can have a tendency to agglomerate from high humidity in the Containment Building, the density may increase and the volume decrease during

normal plant operation. Due to possible agglomeration and

increase in density, estimating the minimum equivalent weight of STB in Containment is conservative with respect to achieving a minimum required pH.

APPLICABILITY In MODEs 1, 2, 3, and 4, the RCS is at elevated temperature and pressure, providing an energy potential for a LOCA. The

potential for a LOCA results in a need for the ability to

control the pH of the recirculated coolant.

In MODEs 5 and 6, the potential for a LOCA is reduced or

non-existent due to the reduced pressure and temperature limitations of these MODEs, and STB is not required.

ACTIONS A.1 If it is discovered that the STB in the Containment Building sump is not within limits, action must be taken to restore the STB to within limits. During plant operation the containment sump is not accessible and corrections may not be possible.

The Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed for restoring the

STB within limits, where possible, because 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is the same time allowed for restoration of other ECCS components.

B.1 and B.2 If the STB cannot be restored within limits within the Completion Time of Required Action A.1, the plant must be STB B 3.5.5 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.5-4 Revision 37 brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The

allowed Completion Times are reasonable, based on operating

experience, to reach the required plant conditions from full

power in an orderly manner and without challenging plant

systems. SURVEILLANCE SR 3.5.5.1 REQUIREMENTS

Periodic determination of the equivalent weight of STB in Containment must be performed due to the possibility of

leaking valves and components in the Containment Building

that could cause dissolution of the STB during normal operation. A Frequency of 24 months is required to

determine visually that a minimum of 13,750 lbm is contained in the STB baskets. This requirement ensures that there is an adequate mass of STB to adjust the pH of the post-LOCA sump solution to a value 7.0.

The periodic verification is required every 24 months, since

access to the STB baskets is only feasible during outages, and normal fuel cycles are scheduled for 24 months.

Operating experience has shown this SR Frequency acceptable, due to the margin in the weight of STB placed in the Containment Building.

SR 3.5.5.2 Testing must be performed to ensure the solubility and

buffering ability of the STB after exposure to the containment environment. A representative sample of 2.74 +/- 0.05 grams of STB , from one of the baskets in Containment is submerged in 1.0

+/- 0.01 liters of water at a boron concentration of 3074 +/- 50 ppm, and at the standard temperature of 120

+/- 5°F. Within four hours without agitation, the solution is decanted and mixed, the

temperature adjusted to 77

+/- 2°F, and the pH measured. The solution pH should be 7.0. The representative sample weight is based on the minimum required STB equivalent weight of 13,750 lbm, and maximum possible post-LOCA sump water mass of 4,608,356 lbm, normalized to buffer a 1.0 +/- 0.01 liter sample. The boron concentration of the STB B 3.5.5 BASES CALVERT CLIFFS - UNITS 1 & 2 B 3.5.5-5 Revision 37 test water is representative of the maximum possible boron concentration corresponding to the maximum possible post-

LOCA sump volume. Agitation of the test solution is

prohibited, since an adequate standard for the agitation

intensity cannot be specified. A test time of four hours

would allow time for the dissolved STB to naturally diffuse through the sample solution. A test time of less than four

hours is more conservative than a test time of longer than

four hours because the longer time could permit additional

STB to dissolve, if excess STB was available. In the post-LOCA containment sump, rapid mixing would occur, significantly decreasing the actual amount of time before the required pH is achieved. This would ensure compliance with the Standard Review Plan requirement of a pH 7.0 by the onset of recirculation after a LOCA.

REFERENCES None