Text

Watts Bar Nuclear Plant, Unit 2, Developmental Revision B - Technical Specifications Bases B 3.4 - Reactor Coolant System
	ML100550505
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Site:	Watts Bar
Issue date:	02/02/2010
From:	; Tennessee Valley Authority
To:	; Office of Nuclear Reactor Regulation
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RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 (continued)

Watts Bar - Unit 2 B 3.4-1 (developmental)

A B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits

BASES BACKGROUND These Bases address requirements for maintaining RCS pressure, temperature, and flow rate within limits assumed in the safety analyses.

The safety analyses (Ref. 1) of normal operating conditions and

anticipated operational occurrences assume initial conditions within the

normal steady state envelope. The limits placed on RCS pressure, temperature, and flow rate ensure that the minimum departure from

nucleate boiling ratio (DNBR) will be met for each of the transients

analyzed.

The RCS pressure limit is consistent with operation within the nominal

operational envelope. Pressurizer pressure indications are averaged to

come up with a value for comparison to the limit. A lower pressure will

cause the reactor core to approach DNB limits.

The RCS coolant average temperature limit is consistent with full power

operation within the nominal operational envelope. Indications of

temperature are averaged to determine a value for comparison to the

limit. A higher average temperature will cause the core to approach DNB

limits.

The RCS flow rate normally remains constant during an operational fuel

cycle with all pumps running. The minimum RCS flow limit corresponds

to that assumed for DNB analyses. Flow rate indications are averaged to

come up with a value for comparison to the limit. A lower RCS flow will

cause the core to approach DNB limits.

Operation for significant periods of time outside these DNB limits

increases the likelihood of a fuel cladding failure in a DNB limited event.

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.4-2 (developmental)

A APPLICABLE SAFETY ANALYSES The requirements of this LCO represent the initial conditions for DNB

limited transients analyzed in the plant safety analyses (Ref. 1). The

safety analyses have shown that transients initiated from the limits of this

LCO will result in meeting the DNBR criterion. This is the acceptance

limit for the RCS DNB parameters. Changes to the unit that could impact

these parameters must be assessed for their impact on the DNBR

criteria. The transients analyzed for include loss of coolant flow events

and dropped or stuck rod events. A key assumption for the analysis of

these events is that the core power distribution is within the limits of LCO 3.1.7, "Control Bank Insertion Limits;" LCO 3.2.3, "AXIAL FLUX

DIFFERENCE (AFD);" and LCO 3.2.4, "QUADRANT POWER TILT

RATIO (QPTR)."

The pressurizer pressure limit of 2214 psig and the RCS average

temperature limit of 593.2°F correspond to analytical limits of 2185 psig

and 594.2°F used in the safety analyses, with allowance for measurement

uncertainty.

The RCS DNB parameters satisfy Criterion 2 of the NRC Policy

Statement.

LCO This LCO specifies limits on the monitored process variables - pressurizer pressure, RCS average temperature, and RCS total flow rate - to ensure

the core operates within the limits assumed in the safety analyses.

Operating within these limits will result in meeting the DNBR criterion in

the event of a DNB limited transient.

RCS total flow rate contains a measurement error of 1.6% (process

computer) or 1.8% (control board indication) based on performing a

precision heat balance and using the result to calibrate the RCS flow rate

indicators. Potential fouling of the feedwater venturi, which might not be

detected, could bias the result from the precision heat balance in a

nonconservative manner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi raises the nominal flow measurement

allowance to 1.7% (process computer) or 1.9% (control board indication).

Any fouling that might bias the flow rate measurement greater than 0.1%

can be detected by monitoring and trending various plant performance

parameters. If detected, either the effect of the fouling shall be quantified

and compensated for in the RCS flow rate measurement or the venturi

shall be cleaned to eliminate the fouling. The LCO numerical values for

pressure, temperature, and flow rate are given for the measurement

location and have been adjusted for instrument error.

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES (continued)

Watts Bar - Unit 2 B 3.4-3 (developmental)

B APPLICABILITY In MODE 1, the limits on pressurizer pressure, RCS coolant average temperature, and RCS flow rate must be maintained during steady state

operation in order to ensure DNBR criteria will be met in the event of an

unplanned loss of forced coolant flow or other DNB limited transient. In

all other MODES, the power level is low enough that DNB is not a

concern.

A Note has been added to indicate the limit on pressurizer pressure is not

applicable during short term operational transients such as a THERMAL

POWER ramp increase > 5% RTP per minute or a THERMAL POWER

step increase > 10% RTP. These conditions represent short term

perturbations where actions to control pressure variations might be

counterproductive. Also, since they represent transients initiated from

power levels < 100% RTP, an increased DNBR margin exists to offset the

temporary pressure variations.

Another set of limits on DNB related parameters is provided in SL 2.1.1, "Reactor Core SLs." Those limits are less restrictive than the limits of this

LCO, but violation of a Safety Limit (SL) merits a stricter, more severe

Required Action. Should a violation of this LCO occur, the operator must

check whether or not an SL may have been exceeded.

ACTIONS A.1

RCS pressure and RCS average temperature are controllable and

measurable parameters. With one or both of these parameters not within

LCO limits, action must be taken to restore parameter(s).

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES (continued)

Watts Bar - Unit 2 B 3.4-4 (developmental)

A ACTIONS A.1 (continued)

RCS total flow rate is not a controllable parameter and is not expected to

vary during steady state operation. If the indicated RCS total flow rate is

below the LCO limit, power must be reduced, as required by Required

Action B.1, to restore DNB margin and eliminate the potential for violation

of the accident analysis bounds.

The 2-hour Completion Time for restoration of the parameters provides

sufficient time to adjust plant parameters, to determine the cause for the

off normal condition, and to restore the readings within limits, and is

based on plant operating experience.

B.1 If Required Action A.1 is not met 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 2

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . In MODE 2, the reduced power condition eliminates the

potential for violation of the accident analysis bounds. The Completion

Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable to reach the required plant conditions in an

orderly manner.

SURVEILLANCE

REQUIREMENTS SR 3.4.1.1

Since Required Action A.1 allows a Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to restore

parameters that are not within limits, the 12-hour Surveillance Frequency

for verifying that the pressurizer pressure is sufficient to ensure the

pressure can be restored to a normal operation, steady state condition

following load changes and other expected transient operations. The 12-

hour interval has been shown by operating practice to be sufficient to

regularly assess for potential degradation and to verify operation is within

safety analysis assumptions.

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES Watts Bar - Unit 2 B 3.4-5 (developmental)

B SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.1.2

Since Required Action A.1 allows a Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to restore

parameters that are not within limits, the 12-hour Surveillance Frequency

for verifying RCS average temperature is sufficient to ensure the

temperature can be restored to a normal operation, steady state condition

following load changes and other expected transient operations. The

12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to

regularly assess for potential degradation and to verify operation is within

safety analysis assumptions.

SR 3.4.1.3

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Surveillance Frequency to verify the RCS total flow rate is

performed using the installed flow instrumentation. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval

has been shown by operating practice to be sufficient to regularly assess

potential degradation and to verify operation within safety analysis

assumptions.

SR 3.4.1.4

Measurement of RCS total flow rate by performance of a precision

calorimetric heat balance once every 18 months allows the installed RCS flow instrumentation to be calibrated and verifies the actual RCS flow rate

is greater than or equal to the minimum required RCS flow rate.

The Frequency of 18 months reflects the importance of verifying flow after

a refueling outage when the core has been altered, which may have

caused an alteration of flow resistance.

This SR is modified by a Note that allows entry into MODE 1, without

having performed the SR, and placement of the unit in the best condition

for performing the SR. The Note states that the SR is not required to be

performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after 90% RTP. This exception is appropriate since the heat balance method requires the plant to be at a minimum of 90% RTP to obtain the stated RCS flow accuracies. The Surveillance

shall be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reaching 90% RTP.

Note: The accuracy of the instruments used for monitoring RCS pressure, temperature and flow rate is discussed in this Bases

section under LCO.

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES Watts Bar - Unit 2 B 3.4-6 (developmental)

B REFERENCES 1. Watts Bar FSAR, Section 15.0, "Accident Analysis," Section 15.2, "Condition II - Faults of Moderate Frequency," and Section 15.3.4, "Complete Loss Of Forced Reactor Coolant Flow."

RCS Minimum Temperature for Criticality B 3.4.2 (continued)

Watts Bar - Unit 2 B 3.4-7 (developmental)

A B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.2 RCS Minimum Temperature for Criticality

BASES BACKGROUND This LCO is based upon meeting several major considerations before the reactor can be made critical and while the reactor is critical.

The first consideration is moderator temperature coefficient (MTC),

LCO 3.1.4, "Moderator Temperature Coefficient (MTC)." In the transient

and accident analyses, the MTC is assumed to be in a range from slightly

positive to negative, and the operating temperature is assumed to be

within the nominal operating envelope while the reactor is critical. The

LCO on minimum temperature for criticality helps ensure the plant is

operated consistent with these assumptions.

The second consideration is the protective instrumentation. Because

certain protective instrumentation (e.g., excore neutron detectors) can be

affected by moderator temperature, a temperature value within the

nominal operating envelope is chosen to ensure proper indication and

response while the reactor is critical.

The third consideration is the pressurizer operating characteristics. The

transient and accident analyses assume that the pressurizer is within its

normal startup and operating range (i.e., saturated conditions and steam

bubble present). It is also assumed that the RCS temperature is within its

normal expected range for startup and power operation. Since the

density of the water, and hence the response of the pressurizer to

transients, depends upon the initial temperature of the moderator, a

minimum value for moderator temperature within the nominal operating

envelope is chosen.

The fourth consideration is that the reactor vessel is above its minimum

nil ductility reference temperature when the reactor is critical.

RCS Minimum Temperature for Criticality B 3.4.2 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.4-8 (developmental)

A APPLICABLE SAFETY ANALYSES Although the RCS minimum temperature for criticality is not itself an initial

condition assumed in Design Basis Accidents (DBAs), the closely aligned

temperature for hot zero power (HZP) is a process variable that is an

initial condition of DBAs, such as the rod cluster control assembly (RCCA)

withdrawal, RCCA ejection, and main steam line break accidents

performed at zero power that either assumes the failure of, or presents a

challenge to, the integrity of a fission product barrier.

All low power safety analyses assume initial RCS loop temperatures the HZP temperature of 557 F (Ref. 1). The minimum temperature for criticality limitation provides a small band, 6 F, for critical operation below HZP. This band allows critical operation below HZP during plant

startup and does not adversely affect any safety analyses since the MTC

is not significantly affected by t he small temperature difference between

HZP and the minimum temperature for criticality.

The RCS minimum temperature for criticality satisfies Criterion 2 of the

NRC Policy Statement.

LCO Compliance with the LCO ensures that the reactor will not be made or maintained critical (k eff 1.0) at a temperature less than a small band below the HZP temperature, which is assumed in the safety analysis.

Failure to meet the requirements of this LCO may produce initial

conditions inconsistent with the initial conditions assumed in the safety

analysis.

APPLICABILITY In MODE 1 and MODE 2, with k eff 1.0, LCO 3.4.2 is applicable since the reactor can only be critical (k eff 1.0) in these MODES.

The special test exception of LCO 3.1.10, "PHYSICS TESTS Exceptions

- MODE 2," permits PHYSICS TESTS to be performed at 5% RTP with RCS loop average temperatures slightly lower than normally allowed so

that fundamental nuclear characteristics of the core can be verified. In

order for nuclear characteristics to be accurately measured, it may be

necessary to operate outside the normal restrictions of this LCO. For

example, to measure the MTC at beginning of cycle, it is necessary to

allow RCS loop average temperatures to fall below T no load , which may cause RCS loop average temperatures to fall below the temperature limit

of this LCO.

RCS Minimum Temperature for Criticality B 3.4.2 BASES (continued)

Watts Bar - Unit 2 B 3.4-9 (developmental)

B ACTIONS A.1 If the parameters that are outside the limit cannot be restored, 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

30 minutes. Rapid reactor shutdown can be readily and practically

achieved within a 30-minute period. The allowed time is reasonable, based on operating experience, to reach MODE 3 in an orderly manner

and without challenging plant systems.

SURVEILLANCE

REQUIREMENTS SR 3.4.2.1

RCS loop average temperature is required to be verified at or above

551 F (value does not account for instrument error) every 30 minutes when the Tavg - T ref deviation alarm is not reset and any RCS loop Tavg < 561 F.

The Note modifies the SR. When any RCS loop average temperature is

< 561 F and the Tavg - T ref deviation alarm is alarming, RCS loop average temperatures could fall below the LCO requirement without additional

warning. The SR to verify RCS loop average temperatures every

30 minutes is frequent enough to prevent the inadvertent violation of the

LCO.

REFERENCES 1. Watts Bar FSAR, Section 15.0, "Accident Analysis."

RCS P/T Limits B 3.4.3 (continued)

Watts Bar - Unit 2 B 3.4-10 (developmental)

B B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.3 RCS Pressure and Temperature (P/T) Limits

BASES

BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads

are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and

temperature changes during RCS heatup and cooldown, within the design

assumptions and the stress limits for cyclic operation.

The PTLR contains P/T limit curves for heatup, cooldown, inservice leak

and hydrostatic (ISLH) testing, and data for the maximum rate of change

of reactor coolant temperature (Ref. 1).

Each P/T limit curve defines an acceptable region for normal operation.

The usual use of the curves is operational guidance during heatup or

cooldown maneuvering, when pressure and temperature indications are

monitored and compared to the applicable curve to determine that

operation is within the allowable region.

The LCO establishes operating limits that provide a margin to brittle

failure of the reactor vessel and piping of the reactor coolant pressure

boundary (RCPB). The vessel is the component most subject to brittle

failure, and the LCO limits apply mainly to the vessel. The limits do not

apply to the pressurizer, which has different design characteristics and

operating functions.

10 CFR 50, Appendix G (Ref. 2), requires the establishment of P/T limits

for specific material fracture toughness requirements of the RCPB

materials. Reference 2 requires an adequate margin to brittle failure

during normal operation, anticipated operational occurrences, and system

hydrostatic tests. It mandates the use of the American Society of Mechanical Engineers (ASME) Code,Section XI, Appendix G (Ref. 3).

The neutron embrittlement effect on the material toughness is reflected by

increasing the nil ductility reference temperature (RT NDT) as exposure to neutron fluence increases.

RCS P/T Limits B 3.4.3 BASES (continued)

Watts Bar - Unit 2 B 3.4-11 (developmental)

A BACKGROUND (continued)

The actual shift in the RT NDT of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 4) and

Appendix H of 10 CFR 50 (Ref. 5). The operating P/T limit curves will be

adjusted, as necessary, based on the evaluation findings and the

recommendations of Regulatory Guide 1.99 (Ref. 6).

The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel

and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the

reactor vessel will dictate the most restrictive limit. Across the span of the

P/T limit curves, different locations are more restrictive, and, thus, the

curves are composites of the most restrictive regions.

The heatup curve represents a different set of restrictions than the

cooldown curve because the directions of the thermal gradients through

the vessel wall are reversed. The thermal gradient reversal alters the

location of the tensile stress between the outer and inner walls.

The criticality limit curve includes the Reference 2 requirement that it be 40 F above the heatup curve or the cooldown curve, and not less than the minimum permissible temperature for ISLH testing. However, the

criticality curve is not operationally limit ing; a more restrictive limit exists in LCO 3.4.2, "RCS Minimum Temperature for Criticality."

The consequence of violating the LCO limits is that the RCS has been

operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the

event these limits are exceeded, an evaluation must be performed to

determine the effect on the structural integrity of the RCPB components.

The ASME Code,Section XI, Appendix E (Ref. 7), provides a

recommended methodology for evaluating an operating event that causes

an excursion outside the limits.

RCS P/T Limits B 3.4.3 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.4-12 (developmental)

B APPLICABLE SAFETY ANALYSES The P/T limits are not derived from Design Basis Accident (DBA)

analyses. They are prescribed during normal operation to avoid

encountering pressure, temperature, and temperature rate of change

conditions that might cause undetected flaws to propagate and cause

nonductile failure of the RCPB, an unanalyzed condition. Reference 8 establishes the methodology for determining the P/T limits. Although the

P/T limits are not derived from any DBA, the P/T limits are acceptance limits since they preclude operation in an unanalyzed condition.

RCS P/T limits satisfy Criterion 2 of the NRC Policy Statement.

LCO The two elements of this LCO are:

a. The limit curves for heatup, cooldown, and ISLH testing; and b. Limits on the rate of change of temperature.

The LCO limits apply to all components of the RCS, except the

pressurizer. These limits define allowable operating regions and permit a

large number of operating cycles while providing a wide margin to

nonductile failure.

The limits for the rate of change of temperature control and the thermal

gradient through the vessel wall are used as inputs for calculating the

heatup, cooldown, and ISLH testing P/T limit curves. Thus, the LCO for

the rate of change of temperature restricts stresses caused by thermal

gradients and also ensures the validity of the P/T limit curves.

Violating the LCO limits places the reactor vessel outside of the bounds of

the stress analyses and can increase stresses in other RCPB

components. The consequences depend on several factors, as follow:

a. The severity of the departure from the allowable operating P/T regime or the severity of the rate of change of temperature; b. The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more

pronounced); and c. The existences, sizes, and orientations of flaws in the vessel material.

RCS P/T Limits B 3.4.3 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.4-13 (developmental)

A APPLICABILITY The RCS P/T limits LCO provides a definition of acceptable operation for prevention of nonductile failure in accordance with 10 CFR 50, Appendix G (Ref. 2). Although the P/T limits were developed to provide

guidance for operation during heatup or cooldown (MODES 3, 4, and 5)

or ISLH testing, their Applicability is at all times in keeping with the

concern for nonductile failure. The limits do not apply to the pressurizer.

During MODES 1 and 2, other Technical Specifications provide limits for

operation that can be more restrictive than or can supplement these P/T

limits. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure

from Nucleate Boiling (DNB) Limits"; LCO 3.4.2, "RCS Minimum

Temperature for Criticality"; and Safety Limit 2.1, "Safety Limits," also

provide operational restrictions for pressure and temperature and

maximum pressure. Furthermore, MODES 1 and 2 are above the

temperature range of concern for nonductile failure, and stress analyses

have been performed for normal maneuvering profiles, such as power

ascension or descent.

ACTIONS A.1 and A.2

Operation outside the P/T limits during MODE 1, 2, 3, or 4 must be

corrected so that the RCPB is returned to a condition that has been

verified by stress analyses.

The 30 minute Completion Time reflects the urgency of restoring the

parameters to within the analyzed range. Most violations will not be

severe, and the activity can be accomplished in this time in a controlled

manner.

Besides restoring operation within limits, an evaluation is required to

determine if RCS operation can continue. The evaluation must verify the

RCPB integrity remains acceptable and must be completed before

continuing operation. Several methods may be used, including

comparison with pre-analyzed transients in the stress analyses, new

analyses, or inspection of the components.

ASME Code,Section XI, Appendix E (Ref. 7), may be used to support the

evaluation. However, its use is restricted to evaluation of the vessel

beltline.

RCS P/T Limits B 3.4.3 BASES (continued)

Watts Bar - Unit 2 B 3.4-14 (developmental)

A ACTIONS A.1 and A.2 (continued)

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is reasonable to accomplish the evaluation.

The evaluation for a mild violation is possible within this time, but more

severe violations may require special, event specific stress analyses or

inspections. A favorable evaluation must be completed before continuing

to operate.

Condition A is modified by a Note requiring Required Action A.2 to be

completed whenever the Condition is entered. The Note emphasizes the

need to perform the evaluation of the effects of the excursion outside the

allowable limits. Restoration alone per Required Action A.1 is insufficient

because higher than analyzed stresses may have occurred and may have

affected the RCPB integrity.

B.1 and B.2

If a Required Action and associated Completion Time of Condition A are

not met, the plant must be placed in a lower MODE because either the

RCS remained in an unacceptable P/T region for an extended period of

increased stress or a sufficiently severe event caused entry into an

unacceptable region. Either possibility indicates a need for more careful

examination of the event, best accomplished with the RCS at reduced

pressure and temperature. In reduced pressure and temperature

conditions, the possibility of propagation with undetected flaws is

decreased.

If the required restoration activity cannot be accomplished within

30 minutes, Required Action B.1 and Required Action B.2 must be

implemented to reduce pressure and temperature.

If the required evaluation for continued operation cannot be accomplished

within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the results are indeterminate or unfavorable, action

must proceed to reduce pressure and temperature as specified in

Required Action B.1 and Required Action B.2. A favorable evaluation

must be completed and documented before returning to operating

pressure and temperature conditions.

Pressure and temperature are reduced by bringing the plant to MODE 3

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 with RCS pressure < 500 psig within

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

RCS P/T Limits B 3.4.3 BASES (continued)

Watts Bar - Unit 2 B 3.4-15 (developmental)

A ACTIONS B.1 and B.2 (continued)

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.

C.1 and C.2

Actions must be initiated immediately to correct operation outside of the

P/T limits at times other than when in MODE 1, 2, 3, or 4, so that the

RCPB is returned to a condition that has been verified by stress analysis.

The immediate Completion Time reflects the urgency of initiating action to

restore the parameters to within the analyzed range. Most violations will

not be severe, and the activity can be accomplished in this time in a

controlled manner.

Besides restoring operation within limits, an evaluation is required to

determine if RCS operation can continue. The evaluation must verify that

the RCPB integrity remains acceptable and must be completed prior to

entry into MODE 4. Several methods may be used, including comparison

with pre-analyzed transients in the stress analyses, or inspection of the

components.

ASME Code,Section XI, Appendix E (Ref. 7), may be used to support the

evaluation. However, its use is restricted to evaluation of the vessel

beltline.

Condition C is modified by a Note requiring Required Action C.2 to be

completed whenever the Condition is entered. The Note emphasizes the

need to perform the evaluation of the effects of the excursion outside the

allowable limits. Restoration alone per Required Action C.1 is insufficient

because higher than analyzed stresses may have occurred and may have

affected the RCPB integrity.

RCS P/T Limits B 3.4.3 BASES (continued)

Watts Bar - Unit 2 B 3.4-16 (developmental)

B SURVEILLANCE REQUIREMENTS SR 3.4.3.1

Verification that operation is within the PTLR limits is required every

30 minutes when RCS pressure and temperature conditions are

undergoing planned changes. This Frequency is considered reasonable

in view of the control room indication available to monitor RCS status.

Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permit assessment and correction for minor

deviations within a reasonable time.

Surveillance for heatup, cooldown, or ISLH testing may be discontinued

when the definition given in the relevant plant procedure for ending the

activity is satisfied.

This SR is modified by a Note that only requires this SR to be performed

during system heatup, cooldown, and ISLH testing. No SR is given for

criticality operations because LCO 3.4.2 contains a more restrictive

requirement.

REFERENCES 1. Appendix "B" to RCS System Description N3-68-4001, "Watts Bar Unit 2 RCS Pressure and Temperature Limits Report." 2. Title 10, Code of Federal Regulations, Part 50, Appendix G, "Fracture Toughness Requirements." 3. ASME Boiler and Pressure Vessel Code,Section XI, Appendix G, "Fracture Toughness Criteria for Protection Against Failure." 4. ASTM E 185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels,"

July 1982.

5. Title 10, Code of Federal Regulations, Part 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements." 6. Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.

RCS P/T Limits B 3.4.3 BASES (continued)

Watts Bar - Unit 2 B 3.4-17 (developmental)

B REFERENCES (continued) 7. ASME Boiler and Pressure Vessel Code,Section XI, Appendix E, "Evaluation of Unanticipated Operating Events." 8. WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004.

RCS Loops - MODES 1 and 2 B 3.4.4 (continued)

Watts Bar - Unit 2 B 3.4-18 (developmental)

A B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.4 RCS Loops - MODES 1 and 2

BASES BACKGROUND The primary function of the RCS is removal of the heat generated in the fuel due to the fission process, and transfer of this heat, via the steam

generators (SGs), to the secondary plant.

The secondary functions of the RCS include:

a. Moderating the neutron energy level to the thermal state, to increase the probability of fission;

b. Improving the neutron economy by acting as a reflector;

c. Carrying the soluble neutron poison, boric acid;

d. Providing a second barrier against fission product release to the environment; and

e. Removing the heat generated in the fuel due to fission product decay following a unit shutdown.

The reactor coolant is circulated through four loops connected in parallel

to the reactor vessel, each containing an SG, a reactor coolant pump (RCP), and appropriate flow and temperature instrumentation for both

control and protection. The reactor vessel contains the clad fuel. The

SGs provide the heat sink to the isolated secondary coolant. The RCPs

circulate the coolant through the reactor vessel and SGs at a sufficient

rate to ensure proper heat transfer and prevent fuel damage. This forced

circulation of the reactor coolant ensures mixing of the coolant for proper

boration and chemistry control.

APPLICABLE

SAFETY ANALYSES Safety analyses contain various assumptions for the design bases

accident initial conditions including RCS pressure, RCS temperature, reactor power level, core parameters, and safety system setpoints. The

important aspect for this LCO is the reactor coolant forced flow rate, which is represented by the number of RCS loops in service.

RCS Loops - MODES 1 and 2 B 3.4.4 BASES (continued)

Watts Bar - Unit 2 B 3.4-19 (developmental)

A APPLICABLE SAFETY ANALYSES (continued)

Both transient and steady state analyses have been performed to

establish the effect of flow on the departure from nucleate boiling (DNB).

The transient and accident analyses for the plant have been performed

assuming four RCS loops are in operation. The majority of the plant

safety analyses are based on initial conditions at high core power or zero

power. The accident analyses that are most important to RCP operation

are the four pump coastdown, single pump locked rotor, single pump (broken shaft or coastdown), and rod withdrawal events (Ref. 1).

Steady state DNB analysis has been performed for the four RCS loop

operation. For four RCS loop operation, the steady state DNB analysis, which generates the pressure and temperature Safety Limit (SL) (i.e., the

departure from nucleate boiling ratio (DNBR) limit) assumes a maximum

power level of 118% RTP. This is the design overpower condition for four

RCS loop operation. The value for the accident analysis setpoint of the nuclear overpower (high flux) trip is 118% and is based on an analysis assumption that bounds possible instrumentation errors. The DNBR limit

defines a locus of pressure and temperature points that result in a

minimum DNBR greater than or equal to the critical heat flux correlation

limit.

The plant is designed to operate with all RCS loops in operation to

maintain DNBR above the SL, during all normal operations and

anticipated transients. By ensuring heat transfer in the nucleate boiling

region, adequate heat transfer is provided between the fuel cladding and

the reactor coolant.

RCS Loops - MODES 1 and 2 satisfy Criterion 2 of the NRC Policy

Statement.

LCO The purpose of this LCO is to require an adequate forced flow rate for core heat removal. Flow is represented by the number of RCPs in

operation for removal of heat by the SGs. To meet safety analysis

acceptance criteria for DNB, four pumps are required at rated power.

An OPERABLE RCS loop consists of an OPERABLE RCP in operation

providing forced flow for heat transport and an OPERABLE SG.

RCS Loops - MODES 1 and 2 B 3.4.4 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.4-20 (developmental)

A APPLICABILITY In MODES 1 and 2, the reactor is critical and thus has the potential to produce maximum THERMAL POWER. Thus, to ensure that the assumptions of the accident analyses remain valid, all RCS loops are

required to be OPERABLE and in operation in these MODES to prevent

DNB and core damage.

The decay heat production rate is much lower than the full power heat

rate. As such, the forced circulation flow and heat sink requirements are

reduced for lower, noncritical MODES as indicated by the LCOs for

MODES 3, 4, and 5.

Operation in other MODES is covered by:

LCO 3.4.5, "RCS Loops - MODE 3";

LCO 3.4.6, "RCS Loops - MODE 4";

LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled";

LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled";

LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation - High Water Level" (MODE 6); and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation -

Low Water Level" (MODE 6).

ACTIONS A.1 If the requirements of the LCO are not met, the Required Action is to

reduce power and bring the plant to MODE 3. This lowers power level

and thus reduces the core heat removal needs and minimizes the

possibility of violating DNB limits.

The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating

experience, to reach MODE 3 from full power conditions in an orderly

manner and without challenging safety systems.

RCS Loops - MODES 1 and 2 B 3.4.4 BASES (continued)

Watts Bar - Unit 2 B 3.4-21 (developmental)

A SURVEILLANCE REQUIREMENTS SR 3.4.4.1

This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that each RCS loop is in

operation. Verification includes flow rate, temperature, or pump status

monitoring, which help ensure that forced flow is providing heat removal

while maintaining the margin to DNB. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is

sufficient considering other indications and alarms available to the

operator in the control room to monitor RCS loop performance.

REFERENCES

1. Watts Bar FSAR, Section 15.0, "Accident Analysis."

RCS Loops - MODE 3 B 3.4.5 (continued)

Watts Bar - Unit 2 B 3.4-22 (developmental)

A B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.5 RCS Loops - MODE 3

BASES BACKGROUND In MODE 3, the primary function of the reactor coolant is removal of decay heat and transfer of this heat, via the steam generators (SGs), to

the secondary plant fluid. The secondary function of the reactor coolant

is to act as a carrier for soluble neutron poison, boric acid.

The reactor coolant is circulated through four RCS loops, connected in

parallel to the reactor vessel, each containing an SG, a reactor coolant

pump (RCP), and appropriate flow, pressure, level, and temperature

instrumentation for control, protection, and indication. The reactor vessel

contains the clad fuel. The SGs provide the heat sink. The RCPs

circulate the water through the reactor vessel and SGs at a sufficient rate

to ensure proper heat transfer and prevent fuel damage.

In MODE 3, RCPs are used to provide forced circulation for heat removal

during heatup and cooldown. The MODE 3 decay heat removal

requirements are low enough that a single RCS loop with one RCP

running is sufficient to remove core decay heat. However, two RCS loops

are required to be OPERABLE to ensure redundant capability for decay

heat removal.

APPLICABLE

SAFETY ANALYSES Whenever the reactor trip breakers (RTBs) are in the closed position and the control rod drive mechanisms (CRDMs) are energized, an inadvertent

rod withdrawal from subcritical, resulting in a power excursion, is

possible. Such a transient could be caused by a malfunction of the rod

control system. In addition, the possibility of a power excursion due to the

ejection of an inserted control rod is possible with the breakers closed or

open. Such a transient could be caused by the mechanical failure of a

CRDM.

RCS Loops - MODE 3 B 3.4.5 BASES (continued)

Watts Bar - Unit 2 B 3.4-23 (developmental)

A APPLICABLE SAFETY ANALYSES (continued)

Therefore, in MODE 3 with RTBs in the closed position and Rod Control

System capable of rod withdrawal, accidental control rod withdrawal from

subcritical is postulated and requires at least two RCS loops to be

OPERABLE and in operation to ensure that the accident analyses limits

are met. For those conditions when the Rod Control System is not

capable of rod withdrawal, two RCS loops are required to be OPERABLE, but only one RCS loop is required to be in operation to be consistent with

MODE 3 accident analyses.

Failure to provide decay heat removal may result in challenges to a

fission product barrier. The RCS loops are part of the primary success

path that functions or actuates to prevent or mitigate a Design Basis

Accident or transient that either assumes the failure of, or presents a

challenge to, the integrity of a fission product barrier.

RCS Loops - MODE 3 satisfy Criterion 3 of the NRC Policy Statement.

LCO The purpose of this LCO is to require that at least two RCS loops be OPERABLE. In MODE 3 with the RTBs in the closed position and Rod

Control System capable of rod withdrawal, two RCS loops must be in

operation. Two RCS loops are required to be in operation in MODE 3

with RTBs closed and Rod Control System capable of rod withdrawal due

to the postulation of a power excursion because of an inadvertent control

rod withdrawal. The required number of RCS loops in operation ensures

that the Safety Limit criteria will be met for all of the postulated accidents.

With the RTBs in the open position, or the CRDMs de-energized, the Rod

Control System is not capable of rod withdrawal; therefore, only one RCS

loop in operation is necessary to ensure removal of decay heat from the

core and homogenous boron concentration throughout the RCS. An

additional RCS loop is required to be OPERABLE to ensure adequate

decay heat removal capability.

The Note permits all RCPs to be de-energized for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. The purpose of the Note is to perform tests that are designed to

validate various accident analyses values. One of these tests is

validation of the pump coastdown curve used as input to a number of

accident analyses including a loss of flow accident. This test is generally

performed in MODE 3 during the initial startup testing program, and as

such should only be performed once.

RCS Loops - MODE 3 B 3.4.5 BASES (continued)

Watts Bar - Unit 2 B 3.4-24 (developmental)

A LCO (continued)

If, however, changes are made to the RCS that would cause a change to the flow characteristics of the RCS, the input values of the coastdown

curve must be revalidated by conducting the test again. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months time

period specified is adequate to perform the desired tests, and operating

experience has shown that boron stratification is not a problem during this

short period with no forced flow.

Utilization of the Note is permitted provided the following conditions are

met, along with any other conditions imposed by initial startup test

procedures:

a. No operations are permitted that would dilute the RCS boron concentration, thereby maintaining the margin to criticality. Boron

reduction is prohibited because a uniform concentration distribution

throughout the RCS cannot be ensured when in natural circulation;

and

b. Core outlet temperature is maintained at least 10 F below saturation temperature, so that no vapor bubble may form and possibly cause a

natural circulation flow obstruction.

An OPERABLE RCS loop consists of one OPERABLE RCP and one

OPERABLE SG, which has the minimum water level specified in

SR 3.4.5.2. An RCP is OPERABLE if it is capable of being powered and

is able to provide forced flow if required.

RCS Loops - MODE 3 B 3.4.5 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.4-25 (developmental)

A APPLICABILITY In MODE 3, this LCO ensures forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing.

The most stringent condition of the LCO, that is, two RCS loops

OPERABLE and two RCS loops in operation, applies to MODE 3 with

RTBs in the closed position. The least stringent condition, that is, two

RCS loops OPERABLE and one RCS loop in operation, applies to

MODE 3 with the RTBs open.

Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops - MODES 1 and 2";

LCO 3.4.6, "RCS Loops - MODE 4";

LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled";

LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled";

LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation - High Water Level" (MODE 6); and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation -

Low Water Level" (MODE 6).

ACTIONS A.1

If one required RCS loop is inoperable, redundancy for heat removal is

lost. The Required Action is restoration of the required RCS loop to

OPERABLE status within the Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . This time

allowance is a justified period to be without the redundant, non-operating

loop because a single loop in operation has a heat transfer capability

greater than that needed to remove the decay heat produced in the

reactor core and because of the low probability of a failure in the

remaining loop occurring during this period.

B.1 If restoration is not possible within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the unit must be brought to

MODE 4. In MODE 4, the unit may be placed on the Residual Heat

Removal System. The additional Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is

compatible with required operations to achieve cooldown and

depressurization from the existing plant conditions in an orderly manner

and without challenging plant systems.

RCS Loops - MODE 3 B 3.4.5 BASES (continued)

Watts Bar - Unit 2 B 3.4-26 (developmental)

A ACTIONS (continued)

C.1 and C.2 If the required RCS loop is not in operation, and the RTBs are closed and

Rod Control System capable of rod withdrawal, the Required Action is

either to restore the required RCS loop to operation or to de-energize all

CRDMs by opening the RTBs or de-energizing the motor generator (MG)

sets. When the RTBs are in the closed position and Rod Control System

capable of rod withdrawal, it is postulated that a power excursion could

occur in the event of an inadvertent control rod withdrawal. This

mandates having the heat transfer capacity of two RCS loops in

operation. If only one loop is in operation, the RTBs must be opened.

The Completion Times of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to restore the required RCS loop to

operation or de-energize all CRDMs is adequate to perform these

operations in an orderly manner without exposing the unit to risk for an

undue time period.

D.1, D.2, and D.3

If all RCS loops are inoperable or no RCS loop is in operation, except as

during conditions permitted by the Note in the LCO section, all CRDMs

must be de-energized by opening the RTBs or de-energizing the MG

sets. All operations involving a reduction of RCS boron concentration

must be suspended, and action to restore one of the RCS loops to

OPERABLE status and operation must be initiated. Boron dilution

requires forced circulation for proper mixing, and opening the RTBs or

de-energizing the MG sets removes the possibility of an inadvertent rod

withdrawal. The immediate Completion Time reflects the importance of

maintaining operation for heat removal. The action to restore must be

continued until one loop is restored to OPERABLE status and operation.

SURVEILLANCE

REQUIREMENTS SR 3.4.5.1

This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the required loops are in

operation. Verification includes flow rate, temperature, and pump status

monitoring, which help ensure that forced flow is providing heat removal.

The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and

alarms available to the operator in the control room to monitor RCS loop

performance.

RCS Loops - MODE 3 B 3.4.5 BASES Watts Bar - Unit 2 B 3.4-27 (developmental)

B SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.5.2

SR 3.4.5.2 requires verification of SG OPERABILITY. SG OPERABILITY

is verified by ensuring that the secondary side narrow range water level is 6 % (value does not account for instrument error) for required RCS loops. If the SG secondary side narrow range water level is less than

6 %, the tubes may become uncovered and the associated loop may not

be capable of providing the heat sink for removal of the decay heat. The

12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications

available in the control room to alert the operator to a loss of SG level.

SR 3.4.5.3

Verification that the required RCPs are OPERABLE ensures that safety

analyses limits are met. The requirement also ensures that an additional

RCP can be placed in operation, if needed, to maintain decay heat

removal and reactor coolant circulation. Verification is performed by

verifying proper breaker alignment and power availability to the required

RCPs.

REFERENCES None RCS Loops - MODE 4 B 3.4.6 (continued)

Watts Bar - Unit 2 B 3.4-28 (developmental)

A B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.6 RCS Loops - MODE 4

BASES BACKGROUND In MODE 4, the primary function of the reactor coolant is the removal of decay heat and the transfer of this heat to either the steam generator (SG) secondary side coolant or the component cooling water via the

residual heat removal (RHR) heat exchangers. The secondary function of

the reactor coolant is to act as a carrier for soluble neutron poison, boric

acid.

The reactor coolant is circulated through four RCS loops connected in

parallel to the reactor vessel, each loop containing an SG, a reactor

coolant pump (RCP), and appropriate flow, pressure, level, and

temperature instrumentation for control, protection, and indication. The

RCPs circulate the coolant through the reactor vessel and SGs at a

sufficient rate to ensure proper heat transfer and to prevent boric acid

stratification.

In MODE 4, with the reactor trip breakers open and the rods not capable

of withdrawal, either RCPs or RHR loops can be used to provide forced

circulation. The intent in this case is to provide forced flow from at least

one RCP or one RHR loop for decay heat removal and transport. The

flow provided by one RCP loop or RHR loop is adequate for decay heat

removal. The other intent is to require that two paths be available to

provide redundancy for decay heat removal.

In MODE 4, with the reactor trip breakers closed and the rods capable of

withdrawal, two RCPs must be OPERABLE and in operation to provide

forced circulation.

APPLICABLE

SAFETY ANALYSES In MODE 4, with the reactor trip breakers open and the rods not capable

of withdrawal, RCS circulation is considered in determination of the time

available for mitigation of the accidental boron dilution event. The RCS

and RHR loops provide this circulation.

RCS Loops - MODE 4 B 3.4.6 BASES (continued)

Watts Bar - Unit 2 B 3.4-29 (developmental)

A APPLICABLE SAFETY ANALYSES (continued)

Whenever the reactor trip breakers (RTBs) are in the closed position and

the control rod drive mechanisms (CRDMs) are energized, an inadvertent

rod withdrawal from subcritical, resulting in a power excursion, is

possible. Such a transient could be caused by a malfunction of the rod

control system. In addition, the possibility of a power excursion due to the

ejection of an inserted control rod is possible with the breakers closed or

open. Such a transient could be caused by the mechanical failure of a

CRDM.

Therefore, in MODE 4 with RTBs in the closed position and Rod Control

System capable of rod withdrawal, accidental control rod withdrawal from

subcritical is postulated and requires at least two RCS loops to be

OPERABLE and in operation to ensure that the accident analyses limits

are met. For those conditions when the Rod Control System is not

capable of rod withdrawal, any combination of two RCS or RHR loops are

required to be OPERABLE, but only one loop is required to be in

operation to meet decay heat removal requirements.

RCS Loops - MODE 4 have been identified in the NRC Policy Statement

as important contributors to risk reduction.

LCO The purpose of this LCO is to require that at least two loops be OPERABLE. In MODE 4 with the RTBs in the closed position and Rod

Control System capable of rod withdrawal, two RCS loops must be

OPERABLE and in operation. Two RCS loops are required to be in

operation in MODE 4 with RTBs closed and Rod Control System capable

of rod withdrawal due to the postulation of a power excursion because of

an inadvertent control rod withdrawal. The required number of RCS loops

in operation ensures that the Safety Limit criteria will be met for all of the

postulated accidents.

With the RTBs in the open position, or the CRDMs de-energized, the Rod

Control System is not capable of rod withdrawal; therefore, only one loop

in operation is necessary to ensure removal of decay heat from the core

and homogenous boron concentration throughout the RCS. In this case, the LCO allows the two loops that are required to be OPERABLE to

consist of any combination of RCS loops and RHR loops. An additional

loop is required to be OPERABLE to provide redundancy for heat

removal.

RCS Loops - MODE 4 B 3.4.6 BASES (continued)

Watts Bar - Unit 2 B 3.4-30 (developmental)

B LCO (continued)

The Note requires that the secondary side water temperature of each SG be 50 F above each of the RCS cold leg temperatures before the start of an RCP with any RCS cold leg temperature the COMS arming temperature as specified in the PTLR. This restraint is to prevent a low temperature overpressure event due to a thermal transient when an RCP

is started.

An OPERABLE RCS loop comprises an OPERABLE RCP and an

OPERABLE SG, which has the minimum water level specified in

SR 3.4.6.3.

Similarly for the RHR System, an OPERABLE RHR loop comprises an

OPERABLE RHR pump capable of providing forced flow to an

OPERABLE RHR heat exchanger. RCPs and RHR pumps are

OPERABLE if they are capable of being powered and are able to provide

forced flow if required.

APPLICABILITY In MODE 4, this LCO ensures forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing.

One loop of either RCS or RHR provides sufficient circulation for these

purposes. However, two loops consisting of any combination of RCS and

RHR loops are required to be OPERABLE to meet single failure

considerations.

Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops - MODES 1 and 2";

LCO 3.4.5, "RCS Loops - MODE 3";

LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled";

LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled";

LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation - High Water Level" (MODE 6); and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation -

Low Water Level" (MODE 6).

RCS Loops - MODE 4 B 3.4.6 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.4-31 (developmental)

A ACTIONS A.1 If only one RCS loop is OPERABLE and both RHR loops are inoperable, redundancy for heat removal is lost. Action must be initiated to restore a

second RCS or RHR loop to OPERABLE status. The immediate

Completion Time reflects the importance of maintaining the availability of

two paths for heat removal.

B.1 If one required RHR loop is OPERABLE and in operation and there are

no RCS loops OPERABLE, an inoperable RCS or RHR loop must be

restored to OPERABLE status to provide a redundant means for decay

heat removal.

If the parameters that are outside the limits cannot be restored, the plant

must be brought to MODE 5 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Bringing the plant to

MODE 5 is a conservative action with regard to decay heat removal. With

only one RHR loop OPERABLE, redundancy for decay heat removal is

lost and, in the event of a loss of the remaining RHR loop, it would be

safer to initiate that loss from MODE 5 ( 200 F) rather than MODE 4 (200 to 350 F). The Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a reasonable time, based on operating experience, to reach MODE 5 from MODE 4 in an

orderly manner and without challenging plant systems.

C.1 and C.2

If one required RCS loop is not in operation, and the RTBs are closed and