Text

Tech Spec Bases, Rev 13, for Calvert Cliffs Nuclear Power Plant
	ML023600485
Person / Time
Site:	Calvert Cliffs
Issue date:	12/19/2002
From:	Katz P; Constellation Energy Group
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML023600485 (58)
	v • d • e

Peter E. Katz Vice President Calvert Cliffs Nuclear Power Plant Constellation Generation Group, LLC 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 410 495-4455 410 495-3500 Fax 0

Constellation Energy Group December 19, 2002 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:

SUBJECT:

Document Control Desk Calvert Cliffs Nuclear Power Plant Unit Nos. 1 & 2; Docket Nos. 50-317 & 50-318 Technical Specification Bases, Revision 13 Enclosed for your use is one copy of the Calvert Cliffs Technical Specifications Bases, Revision 13. This revision was performed under the Technical Specification Bases Control Program (Technical Specification 5.5.14). This program states, "Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e)."

The List of Effective pages is included. Please replace the appropriate pages of your copies of the Technical Specification Bases with these enclosed pages.

Should you have questions regarding this matter, we will be pleased to discuss them with you.

Very tr.l yours, PEK/DLM/bjd

Enclosures:

As stated cc:

(Without Enclosures)

J. Petro, Esquire J. E. Silberg, Esquire Director, Project Directorate I-I, NRC D. M. Skay, NRC H. J. Miller, NRC Resident Inspector, NRC R. I. McLean, DNR

ý ()C) I

PAGE REPLACEMENT INSTRUCTIONS Calvert Cliffs Nuclear Power Plant Technical Specification Bases - Revision 13 Remove and Discard Insert List of Effective Pages LEP-1 through LEP-5 LEP-1 through LEP-5 List of Revisions LOR-1 LOR-1 Technical Specification Bases Pages B 3.0-5 through B 3.0-8 B 3.0-5 through B 3.0-8 B 3.0-15 through B 3.0-18 B 3.0-15 through B 3.0-19 B 3.3.1-17 and B 3.3.1-18 B 3.3.1-17 and B 3.3.1-18 B 3.3.6-3 and B 3.3.6-4 B 3.3.6-3 and B 3.3.6-4 B 3.3.10-9 and B.3.3.10-10 B 3.3.10-9 and B.3.3.10-10 SB 3.3.10-17 B 3.3.10-17 through B 3.3.10-18 B 3.4.1-1 and B 3.4.1-2 B 3.4.1-1 and B 3.4.1-2 B 3.4.4-1 and B 3.4.4-3 B 3.4.4-1 and B 3.4.4-3 B 3.7.1-3 through B 3.7.1-5 B 3.7.1-3 through B 3.7.1-5 B 3.7.3-7 and B 3.7.3-10 B 3.7.3-7 and B 3.7.3-10 B 3.7.9-1 and B 3.7.9-2 B 3.7.9-1 and B 3.7.9-2 B 3.7.15-1 and B 3.7.15-2 B 3.7.15-1 and B 3.7.15-2 B 3.8.1-27 through B 3.8.1-30 B 3.8.1-27 through B 3.8.1-30 B 3.9.1-1 and B 3.9.1-2 B 3.9.1-1 and B 3.9.1-2 B 3.9.3-1 through B 3.9.3-7 B 3.9.3-1 through B 3.9.3-7 B 3.9.4-3 and B 3.9.4-4 B 3.9.4-3 and B 3.9.4-4 B 3.9.5-3 and B 3.9.5-4 B 3.9.5-3 and B 3.9.5-4

TECHNICAL SPECIFICATION BASES LIST OF REVISIONS AND ISSUE DATES Rev.

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10 11 12 13 Date Issued August 28, 1998 August 28, 1998 October 28, 1998 March 16, 1999 October 18, 1999 April 14, 2000 May 18, 2000 June 29, 2000 October 24, 2000 February 1, 2001 March 22, 2001 November 13, 2001 September 5, 2002 Rev. 13 Date to NRC May 4, 1998 October 30, 1998 October 30, 1998 October 30, 1998 October 18, 1999 October 18, 1999 October 24, 2000 October 24, 2000 October 24, 2000 October 24, 2000 November 13, 2001 November 13, 2001 November 13, 2001 LOR-I

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September 5, 2002 TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES B 3.3.1-15 B 3.3.1-16 B 3.3.1-17 B 3.3.1-18 B 3.3.1-19 B 3.3.1-20 B 3.3.1-21 B 3.3.1-22 B 3.3.1-23 B 3.3.1-24 B 3.3.1-25 B 3.3.1-26 B 3.3.1-27 B 3.3.1-28 B 3.3.1-29 B 3.3.1-30 B 3.3.1-31 B 3.3.1-32 B 3.3.1-33 B 3.3.1-34 B 3.3.1-35 B 3.3.2-1 B 3.3.2-2 B 3.3.2-3 B 3.3.2-4 B 3.3.2-5 B 3.3.2-6 B 3.3.2-7 B 3.3.2-8 B 3.3.2-9 B 3.3.2-10 B 3.3.3-1 B 3.3.3-2 B 3.3.3-3 B 3.3.3-4 B 3.3.3-5 B 3.3.3-6 B 3.3.3-7 B 3.3.3-8 B 3.3.3-9 B 3.3.3-10 B 3.3.3-11 B 3.3.3-12 Rev.

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September 5, 2002 TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES B 3.6.2-2 Rev. 2 B 3.7.1-2 Rev. 9 B 3.7.8-7 Rev. 11 B 3.6.2-3 Rev. 2 B 3.7.1-3 Rev.

13 B 3.7.9-1 Rev. 2 B 3.6.2-4 Rev. 2 B 3.7.1-4 Rev. 13 B 3.7.9-2 Rev. 13 B 3.6.2-5 Rev. 2 B 3.7.1-5 Rev. 13 B 3.7.9-3 Rev. 11 B 3.6.2-6 Rev. 2 B 3.7.2-1 Rev. 2 B 3.7.9-4 Rev. 11 B 3.6.2-7 Rev. 2 B 3.7.2-2 Rev. 2 B 3.7.10-1 Rev. 9 B 3.6.2-8 Rev. 2 B 3.7.2-3 Rev.

2 B 3.7.10-2 Rev. 2 B 3.6.3-1 Rev. 2 B 3.7.2-4 Rev. 2 B 3.7.10-3 Rev. 2 B 3.6.3-2 Rev. 2 B 3.7.2-5 Rev. 2 B 3.7.11-1 Rev. 2 B 3.6.3-3 Rev. 2 B 3.7.3-1 Rev. 2 B 3.7.11-2 Rev. 2 B 3.6.3-4 Rev. 2 B 3.7.3-2 Rev. 2 B 3.7.11-3 Rev. 2 B 3.6.3-5 Rev. 2 B 3.7.3-3 Rev. 12 B 3.7.11-4 Rev. 2 B 3.6.3-6 Rev. 2 B 3.7.3-4 Rev. 12 B 3.7.12-1 Rev. 2 B 3.6.3-7 Rev. 2 B 3.7.3-5 Rev. 12 B 3.7.12-2 Rev. 2 B 3.6.3-8 Rev. 2 B 3.7.3-6 Rev. 12 B 3.7.12-3 Rev. 2 B 3.6.3-9 Rev. 2 B 3.7.3-7 Rev. 12 B 3.7.12-4 Rev. 2 B 3.6.3-10 Rev. 2 B 3.7.3-8 Rev. 13 B 3.7.13-1 Rev. 8 B 3.6.4-1 Rev. 2 B 3.7.3-9 Rev. 13 B 3.7.13-2 Rev. 8 B 3.6.4-2 Rev. 2 B 3.7.3-10 Rev. 12 B 3.7.13-3 Rev. 8 B 3.6.4-3 Rev. 2 B 3.7.4-1 Rev. 2 B 3.7.14-1 Rev. 2 B 3.6.5-1 Rev. 2 B 3.7.4-2 Rev. 8 B 3.7.14-2 Rev. 2 B 3.6.5-2 Rev. 2 B 3.7.4-3 Rev. 2 B 3.7.14-3 Rev. 2 B 3.6.5-3 Rev. 3 B 3.7.4-4 Rev. 2 B 3.7.15-1 Rev. 2 B 3.6.6-1 Rev. 2 B 3.7.5-1 Rev. 2 B 3.7.15-2 Rev. 13 B 3.6.6-2 Rev. 2 B 3.7.5-2 Rev. 2 B 3.7.15-3 Rev. 2 B 3.6.6-3 Rev. 2 B 3.7.5-3 Rev. 2 B 3.7.15-4 Rev. 2 B 3.6.6-4 Rev. 5 B 3.7.5-4 Rev. 2 B 3.8.1-1 Rev. 5 B 3.6.6-5 Rev. 2 B 3.7.5-5 Rev. 2 B 3.8.1-2 Rev. 12 B 3.6.6-6 Rev. 2 B 3.7.6-1 Rev. 5 B 3.8.1-3 Rev. 2 B 3.6.6-7 Rev. 2 B 3.7.6-2 Rev. 2 B 3.8.1-4 Rev. 10 B 3.6.6-8 Rev. 2 B 3.7.6-3 Rev. 5 B 3.8.1-5 Rev. 7 B 3.6.6-9 Rev. 2 B 3.7.6-4 Rev. 5 B 3.8.1-6 Rev. 7 B 3.6.7-1 Rev. 2 B 3.7.6-5 Rev. 5 B 3.8.1-7 Rev. 3 B 3.6.7-2 Rev. 2 B 3.7.7-1 Rev. 5 B 3.8.1-8 Rev. 3 B 3.6.7-3 Rev. 2 B 3.7.7-2 Rev. 12 B 3.8.1-9 Rev. 3 B 3.6.7-4 Rev. 2 B 3.7.7-3 Rev. 2 B 3.8.1-10 Rev. 3 B 3.6.7-5 Rev. 2 B 3.7.7-4 Rev. 12 B 3.8.1-11 Rev. 3 B 3.6.7-6 Rev. 2 B 3.7.8-1 Rev. 8 B 3.8.1-12 Rev. 3 B 3.6.8-1 Rev. 2 B 3.7.8-2 Rev. 11 B 3.8.1-13 Rev. 3 B 3.6.8-2 Rev. 2 B 3.7.8-3 Rev. 11 B 3.8.1-14 Rev. 3 B 3.6.8-3 Rev.

2 B 3.7.8-4 Rev. 11 B 3.8.1-15 Rev. 3 B 3.6.8-4 Rev. 2 B 3.7.8-5 Rev. 11 B 3.8.1-16 Rev. 3 B 3.7.1-1 Rev. 2 B 3.7.8-6 Rev.

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September 5, 2002 TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES B 3.8.1-18 Rev. 3 B 3.8.6-3 Rev. 2 B 3.9.4-3 Rev. 11 B 3.8.1-19 Rev. 3 B 3.8.6-4 Rev. 2 B 3.9.4-4 Rev. 13 B 3.8.1-20 Rev. 3 B 3.8.6-5 Rev. 2 B 3.9.5-1 Rev. 2 B 3.8.1-21 Rev. 3 B 3.8.6-6 Rev. 2 B 3.9.5-2 Rev. 2 B 3.8.1-22 Rev. 3 B 3.8.6-7 Rev. 2 B 3.9.5-3 Rev. 13 B 3.8.1-23 Rev. 3 B 3.8.7-1 Rev. 2 B 3.9.5-4 Rev. 2 B 3.8.1-24 Rev. 3 B 3.8.7-2 Rev. 2 B 3.9.5-5 Rev. 2 B 3.8.1-25 Rev. 3 B 3.8.7-3 Rev. 2 B 3.9.6-1 Rev. 2 B 3.8.1-26 Rev. 5 B 3.8.7-4 Rev. 2 B 3.9.6-2 Rev. 2 B 3.8.1-27 Rev. 5 B 3.8.8-1 Rev. 2 B 3.9.6-3 Rev. 2 B 3.8.1-28 Rev.

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LCO Applicability "B 3.0 BASES "irradiite d~fuel,,assemblies in the-spent fuel pool."

.'Therefore, this:LCO can be applicable in any or all MODEs.

If the LCO'and,the Required Actions of LCO 3.7.13 are not met while in MODE 1, 2, or 3,; there is no safety benefit to be gained by placing the unit in a shutdown condition.

The

ý,, Required Action-ofLCO 3.7.13;of "Suspend movement of irradiated fuel assemblies in spent fuel. pool" is the

. -appropriate Required Action to complete in lieu of the actions of LCO 3.0.3.'.These exceptions are addressed in the

.individual Specificati6ns.

LCO 3.0.4 Limiting Conditionn-for Operation 3.0.4 establishes limitations on changes in MODEs (including MODEs within the

',Appl icability) 'or 'ther'specified "conditions in the

=

Applicability when an LCO-is not met.

It precludes placing

-the unit in a' MODE-or other specified condition stated in that Applicability, (e.gl.,.Ap'plicability desired to be

,,, entered) when the" following exist:

a. -

Unit conditions are-iuch that.the 'requirements of the LCO would 'not be met in -the Applicability desired to be entered; and"

b.

Continued noncompliance with the LCO requirements, if the Applicability were entered, would result in the

"unit being required-to'exit this Applicability to comply with-the Required Actions.

Compliance'with RequiredeActions that permit continued operation of the,'unit foran.unlimited period of time in a

-- IMODE"or-6ther specified condition provides-an acceptable

'--:level of safetycfor-c6ntinued~operation.

-This is without S'

regjard to the statuszof-the unit before or'after the MODE "change.,' Therefore, in such cases, entry into a MODE or

-thefspecified conditio~nsinthe-Applicability may be made

'in accordance withthe.-provisions-of the-Required Actions.

The provisions of this Specification should not be interpreted as endorsing the failure to exer-cise the good "practice of, restoringr'systems-or-components to OPERABLE status before~entering,-an:associated MODE:or other specified

, condition in the-Applicability.

oThe provisions-of LCO%3.0.4 shall not prevent changes in K>

' MODEs or-other°specified.conditions in -the-Applicability B 3.;0-5' CALVERT CLIFFS -

UNITS 1 & 2 S*Revi si on' 13

LCO Applicability B 3.0 BASES that are required to comply with ACTIONS.

In addition, the provisions of LCO 3.0.4 shall not prevent changes in MODEs or other specified conditions in the Applicability that result from any unit shutdown.

Exceptions to LCO 3.0.4 are stated in the individual Specifications.

The exceptions allow entry into MODEs or other specified conditions in the Applicability when the associated ACTIONS to be entered do not provide for continued operation for an unlimited period of time.

Exceptions may apply to all the ACTIONS or to a specific Required Action of a Specification.

Limiting Condition for Operation 3.0.4 is only applicable when entering MODE 4 from MODE 5, MODE 3 from MODE 4, MODE 2 from MODE 3, or MODE 1 from MODE 2.

Furthermore, LCO 3.0.4 is applicable when entering any other specified condition in the Applicability only while operating in MODE 1, 2, 3, or

4.

The requirements of LCO 3.0.4 do not apply in MODEs 5 and 6 or in other specified conditions of the Applicability (unless in MODE 1, 2, 3, or 4), because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken.

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

Therefore, changing MODEs or other specified conditions while in an ACTIONS Condition, in compliance with LCO 3.0.4 or where an exception to LCO 3.0.4 is stated, is not a violation of SR 3.0.1 or SR 3.0.4 for those SRs that do not have to be performed due to the associated inoperable equipment.

However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO.

LCO 3.0.5 Limiting Condition for Operation 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with ACTIONS.

The sole purpose of this Specification is to provide an exception to LCO 3.0.2 [e.g., to not comply with the CALVERT CLIFFS -

UNITS I & 2 B 3.0-6 Revision 13

LCO Applicability B 3.0 BASES applicable Required Action (s)] -to,allow~the performance of

- required testingto demonstrate:_

--'.'a.- '-The OPERABILITY of the equipment being returned to service; or-~

, b.

,The OPERABILITY of, other equipment.

,,,The administrative controls ensure the -time the equipment is

, returned to service*

(in conf i ct 'with,the -requirements of

.the N

ACTIONS), is limited to the time absolutely necessary to

-perform the required testing to demonstrate OPERABILITY.

This Specification,does not providetime-to perform any other preventive or-co-rrective maintenance.

An example of demonstrating the OPERABILITY, of the equipment

'being returned,toservice.is :reopening a containment

-isolation -valve that has been closed to-comply with Required Actions and must be reopened to performthe required testing.

, An example of demonstrating -theOPERABILITY of other Sequipmentis taking an Anpperable channel or trip system out of he",-tri pped condition to prevent the trip function from occurring during the -performance of required testing on

- another channel in the other -trip system.

A similar example of demonstrating the OPERABILITY of other-equipment is taking an inoperable channel or trip system out of the tripped condition.to permit the logic to function and indicate the appropriate response during-the performance of

.,required testing on another channel in the'same trip system.

LCO 3.0.6 Limiting Condition,for,,Operation 3.0.6 establishes an

,exception;to LCO -3.0.2 for support systems that have an LCO

.-specified in the Technical-Specifications.,

This exception is provided because LCO:3.0.2 would require the Conditions

-, *-:and Required Actions, of-,the _associated -inoperable supported system LCO to be entered solely due to the;inoperability of the support system.

This exception is justified because the

-,--actions required -to ensure the-unit is-maintained in a safe condition are specifiedJin, the -support -system LCO's Required "Actions. - These Required Actions may include entering the supported system's-Conditions-and Required Actions or may specify other Required Actions.

CAVRCIF

UNT, 2

B3

ons, Reiin1 B, 3 r 0 -;77

,Revision 13 CALVERT CLIFFS - UNITS 1 & 2

LCO Applicability B 3.0 BASES When a support system is inoperable and there is an LCO specified for it in the Technical Specification, the supported system(s) are required to be declared inoperable if they are determined to be inoperable as a result of the support system inoperability.

However, it is not necessary to enter into the supported systems' Conditions and Required Actions unless directed to do so by the support system's Required Actions.

The potential confusion and inconsistency of requirements related to the entry into multiple support and supported systems' LCOs' and Required Actions are eliminated by providing all the actions necessary to ensure the unit is maintained in a safe condition in the support system's Required Actions.

However, there are instances where a support system's Required Action may either direct a supported system to be declared inoperable or direct entry into Conditions and Required Actions for the supported system.

This may occur immediately or after some specified delay to perform some other Required Action.

Regardless of whether it is immediate or after some delay, when a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.

Specification 5.5.15, "Safety Function Determination Program (SFDP)," ensures loss of safety function is detected and appropriate actions are taken.

Upon entry into LCO 3.0.6, an evaluation shall be made to determine if a loss of safety function exists.

Additionally, other limitations, remedial actions, or compensatory actions may be identified as a result of the support system inoperability and corresponding exception to entering supported system Conditions and Required Actions.

The SFDP implements the requirements of LCO 3.0.6.

Cross-train checks to identify a loss of safety function for those support systems supporting multiple and redundant safety systems are required.

The cross-train check verifies that the supported systems of the redundant OPERABLE support system are OPERABLE, thereby ensuring safety function is CALVERT CLIFFS - UNITS 1 & 2 B 3.0-8 Revi sion 2

SR Applicability B 3.0 BASES with refueling intervals)-orperiodic Completion Time intervals beyond those specified.

SR

-- The basis for thisdelay period includes consideration of "unit conditions,,adequate planning, availability of personnel'," the-'timerequired to pery-form the surveillance "test, the'safety-significance of the delay-in completing the

__surveillance test, and.the-'recognition that-the most Sprobable" result of-any-particula'r surveillance test being performed is the verificatioh of'conformance with the requirements.

When a'Surveillahce-with:'a-Frequency basedinot on time intervals, but upon:'specified* unit-conditions, operating

-situations, or requirenients of regulationns (e.g., prior to entering MODE lafter each-fuel loading; or in accordance w-ith 10 CFR Part-50,"Appendix J, as-,modified by approved

"-exemptions,"'etc.) is discovered'to not have been performed when specified, SR'3,0.3-allows'for:the full delay period of up to the'specified Frequency to-perform the Surveillance.

Ho~iever, since there'is 'not'a time-interval specified, the CALVERT -CLIFFS -

UNITS 1 & 2 B 3.0-15 Revision 13 3.0.3-

c Surveillance.Requiiement 3.0.3 establishes-the flexibility

ý-

",-;to, deferdeclari ng affected equipment inoperable or an

.. affected variable-outside the specified limits when a surveillance test has not-been 'completed within the specified Frequency.

A delay period of up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to'the limit ofithespecified Frequency, whichever is greater;?applies from when it is discovered that the

,-,surveillance test ha-s not been performed in accordance with SR 3.0.2,' and not-at-the time that the specified Frequency was not met.

A risk-evaluation-shall be-performed for any Surveillance

-- delayed'greater,than'24,hours and the risk 'impact shall be managed-.,

This~delay periodprovides-an' adequate time to complete

-missed surveillan'ce tests; This delay period permits the

,completion of a surveillance-test before' 'complying with SR~quir&d[Actions-or -other-remedial measures that might "preclude completibn of-the surveillance test.

r

'-Revi si on. 13 CALVERT-CLIFFS - UNITS 1 & 2 B 13.0-15

SR Applicability B 3.0 BASES missed Surveillance should be performed at the first reasonable opportunity.

Surveillance Requirement 3.0.3 provides a time limit for, and allowances for the performance of, Surveillances that become applicable as a consequence of MODE changes imposed by Required Actions.

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

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

While up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the limit of the specified Frequency is provided to perform the missed Surveillance, it is expected that the missed Surveillance will be performed at the first reasonable opportunity.

The determination of the first reasonable opportunity should include consideration of the impact on plant risk (from delaying the Surveillance, as well as any plant configuration changes required or shutting the plant down to perform the Surveillance) and impact on any analysis assumptions, in addition to unit conditions, planning, availability of personnel, and the time required to perform the Surveillance.

This risk impact should be managed through the program in place to implement 10 CFR 50.65(a)(4) and its implementation guidance, NRC Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants."

This Regulatory Guide addresses consideration of temporary and aggregate risk impacts, determination of risk management action thresholds, and risk management action up to and including plant shutdown.

The missed Surveillance should be treated as an emergent condition as discussed in the Regulatory Guide.

The risk evaluation may use quantitative, qualitative, or blended methods.

The degree of depth and rigor of the evaluation should be commensurate with the importance of the component.

Missed Surveillances for important components should be analyzed quantitatively.

If the results of the risk evaluation determine the risk increase is significant, this evaluation should be used to determine the safest course of action.

All missed CALVERT CLIFFS - UNITS 1 & 2 B 3.0-16 Revision 13

"SR Applicability B 3.0 BASES Surveillances will be placed in the licensee's Corrective

- Action Program.

If a surveillance -test Ais ýnot completed'within the allowed delay period, the"equipmentis considered-inoperable or the variable is considered outside the specified limits, and the Completion Times,of the Required Actions for the applicable LCO Conditions begin immediately upon expiration of the delay period.

If a surveillance test fails within the delay period, the equipment, s Cinoperable or the variable is

,,outside the specified limits, and the 'Completion Times of the Required Actions-forthe applicable LCO Conditions begin immediately upon the-failur~eof the surveillance test.

Completion of the surveillance test within the delay period allowed by this Specification, or within the Completion Time

'of the 'ACTIONS,' restores compliance with SR '3.0.1.

SR 3.0.4 Surveillance Requirement :3.0.4'establishes'the requirement that,all applicable'-SRs must -be met before entry into a MODE K>

-or other specified Condition in the Applicability.

This'Specification-ensures system and component OPERABILITY

".requirem~ents and variable'-limits aie met before entry into MODEs,-or'other:specified'conditions in the'Applicability, for which these sjs~iems'and components ensure safe operation

-of'the unit.

-The provisions of,-this Specification should'not be interpreted as endorsing the failure to exercise the good practice of restoring'systems or componentsito OPERABLE status before entering an associated MODE'or other specified condition"in therApý1licabiiity.

',IL'*,

However, in certain circumstances, failing to meet an SR "will not result in-SR'3.0.4:restricting a'MODE change or

-other-specified conditionhchange.

When a system, subsystem, division, component, device,' or variable-is inoperable or

-:outsideitsspecified limits the associated SR(s) are not required to be'performed"'per',SR 3.0.1', which states S,

' '-surveillance testsIdonot'have-to'be performed on inoperable equipment.'

Whien-equipment-is inoperable, SR 3.0.4 does not K

apply to the associated&SR(s) since the requirement for the CALVERT CLIFFS - UNITS I & 2 B 3.0217 r Re'Vision 13'

SR Applicability B 3.0 BASES SR(s) to be performed is removed.

Therefore, failing to perform the surveillance test(s) within the specified Frequency does not result in an SR 3.0.4 restriction to changing MODEs or other specified conditions of the Applicability.

However, since the LCO is not met in this instance, LCO 3.0.4 will govern any restrictions that may (or may not) apply to MODE or other specified condition changes.

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

In addition, the provisions of SR 3.0.4 shall not prevent changes in MODEs or other specified conditions in, the Applicability that result from any unit shutdown.

The precise requirements for performance of SRs are specified such that exceptions to SR 3.0.4 are not necessary.

The specific time frames and conditions necessary for meeting the SRs are specified in the Frequency, in the SR, or both.

This allows performance of surveillance tests when the prerequisite condition(s) specified in a surveillance test procedure requires entry into the MODE or other specified condition in the Applicability of the associated LCO prior to the performance or completion of a surveillance test.

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

Alternately, the surveillance test may be stated in the form of a Note as not required (to be met or performed) until a particular event, condition, or time has been reached.

Further discussion of the specific formats of SRs' annotation is found in Section 1.4, Frequency.

Surveillance Requirement 3.0.4 is only applicable when entering MODE 4 from MODE 5, MODE 3 from MODE 4, or MODE 2 from MODE 3, or MODE 1 from MODE 2.

Furthermore, SR 3.0.4 is applicable when entering any other specified condition in the Applicability only while operating in MODEs 1, 2, 3, or

4.

The requirements of SR 3.0.4 do not apply in MODEs 5 and 6, or in other specified conditions of the Applicability (unless in MODEs 1, 2, 3, or 4), because the ACTIONS of CALVERT CLIFFS - UNITS 1 & 2 B 3.0-18 Revision 13

SR Applicability B 3.0 BASES individual Specifications sufficiently define the remedial measures to be taken.

CALVERT CLIFFS UNITS 1 & 2 B 3.0-19 Revision 13 CALVERT CLIFFS - UNITS I & 2 B 3.0-19 Revision 13

RPS Instrumentation-Operating B 3.3.1 BASES POWER is increased. -The trip setpoint is automatically decreased as THERMAL POWER decreases.

The trip setpoint has a maximum and a minimum

_ setpoint.:o Adding to this maximum value the possible variation in trip setpoint due to calibration and instrument errors, the maximum actual steady state THERMAL POWER

-level at which a-trip would be actuated is 109% RTP, whichis the-value used inf the safety analyses.

To account for these errors, the safety analysis minimum value is,40% RTP.. The 10% step increase in trip setpoint-is-a maximum value assumed in the safety analysis.

Therýe.is no' uncertainty applied to the step in the safety analys~es._

2.

Rate of Chanqe of Power-High Trip

-This LCO requires.fourinstrument channels of Rate of Change of Power-High trip to be OPERABLE in MODEs 1 and 2.

The high power rateof change trip serves as a backup

-to the administratively-enforced startup rate limit.

The Function is not'credited in the accident analyses; therefore, the Allowable Value for the trip is not derived from analytical limits.

3.-

Reactor Coolant' Flow-Low Trip This LCO requires four instrument channels of Reactor Coolant Flow-Low trip to be OPERABLE in MODEs 1 and 2.

The trip~may bemanually bypassed when NUCLEAR "INSTRUMENT.POWER falls below 1E-4% RTP.

This

°operating bypass-isý part of the ZPMB circuitry, which also bypasses theTM/LP trip and provides a AT power

-block signalýto the Q 'power select logic.

The ZPMB allows low power physics-testing at reduced RCS temperatures and pressures.

It also allows heatup and cooldown with shutdown CEAs withdrawn.

CALVERT CLIFFS UNITS 1 & 2 B 3.-3.1-17 Revision 1.1 CALVERT CLIFFS -

UNITS I & 2 SRevision "Ii B 3'3.-3.1-17

RPS Instrumentation-Operating B 3.3.1 BASES This trip is set high enough to maintain fuel integrity during a loss of flow condition.

The setting is low enough to allow for normal operating fluctuations from offsite power.

Reactor Coolant System flow is maintained above design flow by LCO 3.4.1.

4.

Pressurizer Pressure-High Trip This LCO requires four instrument channels of Pressurizer Pressure-High trip to be OPERABLE in MODEs 1 and 2.

The Allowable Value is set high enough to allow for pressure increases in the RCS during normal operation (i.e., plant transients) not indicative of an abnormal condition.

The setting is below the lift setpoint of the pressurizer safety valves and low enough to initiate a reactor trip when an abnormal condition is indicated.

The analysis setpoint includes allowance for harsh environment, where appropriate.

The Pressurizer Pressure-High trip concurrent with power-operated relief valve operation avoids unnecessary operation of the pressurizer safety valves (Reference 5).

5.

Containment Pressure-High Trip This LCO requires four instrument channels of Containment Pressure-High trip to be OPERABLE in MODEs 1 and 2.

The Allowable Value is high enough to allow for small pressure increases in Containment, expected during normal operation (i.e., plant heatup) that are not indicative of an abnormal condition.

The setting is low enough to initiate a reactor trip to prevent containment pressure from exceeding design pressure following a DBA.

CALVERT CLIFFS -

UNITS 1 & 2 B 3.3.1-18 Revision 13 CALVERT CLIFFS -

UNITS 1 & 2 B 3.3.1-18 Revision 13

DG-LOVS B 3.3.6 BASES explicitly account for each-individual component of the loss of power detection and subsequent actions.

This delay time includes contributions from the DG start, DG loading, and Safety InjectionUSystemIcoIiponent actuati6n.I The response of the DG'to a loss of power'must be demonstrated to fall within this analysis response time when 'including the contributions of 'al -portions of the delay.

The required channels 6f'LOVS. in conjunction with the ESF "systems powered 'from the DGs,' provide plantprotection in

- th6 event of any of. the -analyzed accidents discussed in 7 Reference 1; Chapter'18:in wh'ich a loss'of offsite power is ass'umed.

Loss of voltfage-'start 'channels' are required to

'meet the redundancy and "testability requirements of Reference 1, Appendix IC.

--,-The delay times ass'unied in 'the safety an'alysis for the ESF

."equipment includethb 1O-second DG start 'delay and the S-appropriate sequen'cing delay,'.if applicable.

The response times for ESFAS-actuated 'ecuipment include 'the appropriate "DG loading and sequent'ing'delay.

The DG-LOVS channels satisfy 1 CFR 50.36(c)(2)(ii),

Criterion 3.

LC()

.The LCO for the LOVS reqiires'that four-channels per bus of each LOVS instrumentation Function be OPERABLE in MODEs 1, 2,-3, and 4.

The LOVS supports safety systems associated, with the ESFAS.

.. Actions allow maintenance 'bypass of individual sensor

"-channels.

The plant isirstricted to '48, hours in a

'- *maintenance bypass-'condition-'before either. restoring the S

,Function to four channel operation (two-out-of-four logic)

, or placing! the channel -in tri ' (one-out-6f-three logic).

"Loss of LOVS Function could 'result in the 'delay of safety system actuation when required.- This could.:lead to

", unacceptable consequences.during accidents.

During the loss

, of'offsite power,,whichista 'AOO, the DG-powers the motor driven AFW pump.

Failure of thistpump to start would leave two turbine-driven pumps as well as an increased potential CALVERT CLIFFS -

UNITS 1 & 2

'-' B 3.3.6-3 Revision 2 I I I

CALVERT CLIFFS - UNITS 1 & 2

""1

--- B 3.3.6-3 Revision-2

DG-LOVS B 3.3.6 BASES for a loss of decay heat removal through the secondary system.

Only Allowable Values are specified for each Function in the LCO.

Nominal trip settings are specified in the plant specific procedures.

The nominal settings are selected to ensure that the setting measured by CHANNEL FUNCTIONAL TESTS does not exceed the Allowable Value if the bistable is performing as required.

Operation with a trip setting less conservative than the nominal trip setting, but within the Allowable Value, is acceptable, provided that operation and testing are consistent with the assumptions of the plant specific setting calculation.

A channel is inoperable if its actual trip setting is not within its required Allowable Value.

The Allowable Values and trip settings are established in order to start the DGs at the appropriate time, in response to plant conditions, in order to provide emergency power to start and supply the essential electrical loads necessary to safely shut down the plant and maintain it in a safe shutdown condition.

APPLICABILITY The DG-LOVS actuation Function is required in MODEs 1, 2, 3, and 4 because ESF Functions are designed to provide protection in these MODEs.

ACTIONS A LOVS sensor channel is inoperable when it does not satisfy the OPERABILITY criteria for the channel's Function.

The most common cause of sensor channel inoperability is outright failure of the bistable (sensor module) or outright failure or drift of the measurement channel sufficient to exceed the tolerance allowed by the plant-specific setting analysis.

Determination of setting drift is generally made during the performance of a CHANNEL CALIBRATION when the process instrument is set up for adjustment to bring it to within specification.

CHANNEL FUNCTIONAL TESTS check that the sensor modules are functioning properly.

If the actual trip setting is not within the Allowable Value or not functioning, the channel is inoperable and the appropriate Conditions must be entered.

CALVERT CLIFFS - UNITS 1 & 2 B 3.3.6-4 Revision 13

PAM Instrumentation B 3.3.10 BASES

_-14.

Steam-Generator Water Level Transmitters

Steam Generator Water Level transmitters are provided o

to monitoroperati6n"ofdecay heat removal via the "steam generatois-The-Category'I"indication of steam

'generator levelis'the-e-tended startup range level instrumentation.

The extended startup range level covers a span of -40 inches to -63 inches (relative to normal operating 1level)'- above the lower tubesheet.

The measured differential pressure is displayed in S-inches.of water atprocessconditions of the fluid.

Redundant monitoring capability is provided by four transmitters.

The uncompensated level signal is input to the plant~computerrand a control room indicator.

-- Steam generator, water level instrumentation consists of two level transmitters.

Operator actionris-based on the control room indication of steam generator water level.

The RCS response during'a design-basissmall break LOCA is dependent on

-the break size.--.For. a certain range of break sizes, the boiler,condenser-mode of heat-transfer is necessary to remove decay-heat. _Extended startup range level is a Type A variable,because the operator must manually

-raise and control the -steam generator level to establish boiler condenser heat transfer.

Feedwater

flow is increased until indication is in range.

15.

Condensate StoraqeTankrLevel Monitor Condensate storageitankV(CST) level monitoring is provided to ensure water supply for AFW.

Condensate Storage Taik 12Irovide's rhi6e` ensured safety grade water supply forlthe;AFW*S~siem.-'

Inventory in CST 12 is J'

monitored byIle-vel; injdication covering the full range 7

'of rdquiredlusalle water level.

Condensate storage tanklevel'is displayed on control room indicators and

" the plantcdmiiputer' "In'addition, a control room annunciator alarms on low'level.'-I Co_ e' C

nsate-sirag'iý ankale~el'is considered a Type A

....variable-becauseth' control room meter and annunciator are considered'the primary indication used by the Operator.

The DBAs that require AFW are the steam line B-3.3.,10-9 IRevision-8 ;

CALVERT-CLIFFS - UNITS 1 & 2

PAM Instrumentation B 3.3.10 BASES break and loss of main feedwater.

Condensate Storage Tank 12 is the initial source of water for the AFW System.

However, as the CST is depleted, manual operator action is necessary to replenish the CST or align suction to the AFW pumps from an alternate source.

16, 17, 18,

19.

Core Exit Temperature Monitor Core Exit Temperature monitors are provided for verification and long-term surveillance of core cooling.

An evaluation was made of the minimum number of valid core exit thermocouples necessary for inadequate core cooling detection.

The evaluation determined the reduced complement of core exit thermocouples necessary to detect initial core uncovery and trend the ensuing core heatup.

The evaluations account for core nonuniformities, including incore effects of the radial decay power distribution and excore effects of condensate runback in the hot legs and nonuniform inlet temperatures.

Based on these evaluations, adequate or inadequate core cooling detection is ensured with two valid core exit thermocouples per quadrant.

The design of the Incore Instrumentation System includes a Type K (chromel alumel) thermocouple within each of the 45 (35 in Unit 2) incore instrument detector assemblies.

The junction of each thermocouple is located more than a foot above the fuel assembly, inside a structure that supports and shields the incore instrument detector assembly string from flow forces in the outlet plenum region.

These core exit thermocouples monitor the temperature of the reactor coolant as it exits the fuel assemblies.

The core exit thermocouples have a usable temperature range from 40°F to 2300 0F, although accuracy is reduced at temperatures above 18007F.

ULVLRI LL1Ii-

UNITS 1 & 2 B 3.3.10-10 Revision 13 CALVERT CLIFFS -

UNITS 1 & 2 B 3.3.10-10

PAM Instrumentation B 3.3.10 BASES frequent, checks of channel during normal operational use of the displays associated with this LCO's required channels.

SR 3.3.10.2 A CHANNEL CALIBRATION is performed every 46 days on a staggered test basis for the Containment Hydrogen Analyzers.

The CHANNEL CALIBRATION is performed using sample gases in accordance with manufacturer's recommendations.

SR 3.3.10.3 A CHANNEL CALIBRATION is performed every 24 months or approximately every refueling.

CHANNEL CALIBRATION is a check of the indication channel including the sensor.

The SR verifies the channel responds to the measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATION of the CIV position indication channels will consist of verification that the position indication changes from not-closed to closed when the valve is exercised to the isolation position as required by Technical Specification 5.5.8, Inservice Testing Program.

The position switch is the sensor for the CIV position indication channels.

A Note allows exclusion of neutron detectors, Core Exit Thermocouples, and Reactor Vessel Level Monitor System from the CHANNEL CALIBRATION.

The Frequency is based upon operating experience and consistency with the typical industry refueling cycle and is justified by an 24 month calibration interval for the determination of the magnitude of equipment drift.

REFERENCES

1. Letter from Mr. R. E. Denton (BGE) to NRC Document Control Desk, dated June 6, 1995, "License Amendment Request; Extension of Instrument Surveillance Intervals"

2.

Letter from Mr. J. A. Tiernan (BGE) to NRC Document Control Desk, dated August 9, 1988, "Regulatory Guide 1.97 Review Update"

3.

Regulatory Guide 1.97, "Instrumentation for Light Water-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and Following an Accident (Errata Published July 1981),

December 1975 B 3.31.10-17 Revision 13 CALVERT CLIFFS - UNITS I & 2

PAM Instrumentation B 3.3.10 BASES

4.

NUREG-0737, Supplement 1, Requirements for Emergency Response Capabilities (Generic Letter 82-33),

December 17, 1982

5.

UFSAR, Chapter 7, "Instrumentation and Control" CALVERT CLIFFS - UNITS 1 & 2 B 3.3. 10-18 Revision 13

RCS-Pressure, Temperature, and Flow DNB Limits B 3.4.1 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-(Reference-i)

-..of normal operating conditions and anticipated operational occurrences, assume initial conditions-within the normal steady-state envelope.

The limits placed on departure from

. icleate boiling(DNB).-related parameters ensure that these

.parameters will not be less conservative-than were assumed in the analyses, and thereby provide assurance that the minimum departure from nucleate boiling ratio (DNBR) will Smeet the required criteria for each of the transients analyzed.

'4 The Limiting Condition for Operation (LCO) limit for minimum RCS pressure -as-measured -at the pressurizer is consistent with operation within the nominal operating envelope and is bounded by' the initi alpres~sure in the analyses.

The LCO limit for maximumRCS coldleg temperature is dnsi stent with o~eration atthe indicated power level and is bounded by the initial :temperature in'the analyses.

The LCO limit for minimum RCS flow rate-i* bounded by initial flow rate in the analyses.

TheRCS flow rate expected&to vary during plantoperation with all pumps

-running.

the i s not APPLICABLE

' The requirementsfLCO 3A4.1 represent the initial SAFETY ANALYSES conditions for DNB limited tr anslents analyzed in the safety analyses (Reference 1),.

.The'safety analyses have shown that triasihI en itiated' fromthe-limits of this LCO will meet the'.DNBR-critrj6n.- Changes to the'facility that could impact these parametersmu'st be assessedfor their impact on the DNBR criteri-n.-Th transients analyzed include loss of coolant flow events and dropped or stuck control element assembly events.

A key assumption for the analysis-of these events is that the core power distribution is-within the limits of LCO 3.1.6, LCO 3.2.4, and LCO-3.2.5.

The safety analyses are performed over the following range of initial Revision 2, CALVERT:CLIFFS - UNITS 1 & 2 I

I I I

B _3A.*.

1-4

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES values:

RCS pressure 2154-2300 psia, core inlet temperature

5481F, and reactor vessel inlet coolant flow rate

370,000 gpm**.

The RCS DNB limits satisfy 10 CFR 50.36(c)(2)(ii),

Criterion 2.

LCO This LCO specifies limits on the monitored process variables

- RCS pressurizer pressure, RCS cold leg temperature, and RCS total flow rate - to ensure that the core operates within the limits assumed for the plant safety analyses.

Operating within these limits will result in meeting the DNBR criterion in the event of a DNB limited transient.

The LCO numerical values for pressure and temperature (P/T) are given for the measurement location and have been adjusted for instrument error.

Reactor Coolant System flow rate is given as an analytical value.

APPLICABILITY In MODE 1, the limits on RCS pressurizer pressure, RCS cold leg temperature, and RCS flow rate must be maintained during steady-state operation in order to ensure that 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 so that DNBR is not a concern.

A Note has been added to indicate the limit on pressurizer pressure may be exceeded during short-term operational transients such as a THERMAL POWER ramp increase of

> 5% RATED THERMAL POWER (RTP) per minute or a THERMAL POWER step increase of > 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.

The Reactor Coolant System Total Flow Rate limit shall be Ž 340,000 gpm through Unit 2, Cycle 14.

CALVERT CLIFFS -

UNITS I & 2 B 3.4.1-2 Revision 13

RCS Loops - MODEs 1 and 2 B 3.4.4 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'i'-ia the SGs, to the secondary plant.

The secondary functions of the RCS include:

a.. Moderating the-neutron6 energy level to the thermal

.state, to increase rthe-probability of fission;

" zb.

Improving theneutron-economy by acting as a reflector;

c.

Carrying theý'solubleneuetron poison,-boric acid;

d.

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

e.

'Removing the,eatn6erated'in the fuel due to fission

-product decay following a unit shutdown.

-The RCS-configuration for heat'transport uses two RCS loops.

-Each RCS loop contaihs-aaSG and two-reactor coolant pumps

" (RCPs);

-An RCPis'lbated in-each of the'two SG cold legs.

-*Thepump flow rate'has-been sized to provide core heat

'removal with'appropriate margin to.DNB'during power operation and for anticipated transients originating from power operation.

This Specification requires two RCS loops

, with both RCPs~in operation in each loop. 'The intent of the Specification is to require coreheat removal with forced

"-flow-durihg power 'operation.,-Specifying two RCS loops

-provides the minimumnecessary paths (two SGs) for heat removal.

APPLICABLE SAFETY ANALYSES K-Safety analyseshcontain various-'assumptions for the DBA initial conditiohs'ihcluding RCS pressure, RCS

-temperature,

'reactor" power -level, core parameters, and safety system setpoints.

The important aspect for this LCO is, the reactor"boblabht forced flow rate,-Awhich is represented by _thei-unjber of" RC-S loops -in service.

Both transient and steady-state analyses have been performed to establish the effect of flow on DNB. -The-transient-or-

'accident analysis'for the-plant has been performed assuming four RCPs are in operation.

The majority ofthe'plant I

I I

CALVERT CLIFFS - UNITS I & 2 B *3.4.4-1 "

Revision-2

RCS Loops - MODEs 1 and 2 B 3.4.4 BASES safety analyses are based on initial conditions at high core power or zero power.

The accident analyses that are of most importance to RCP operation are loss of coolant flow and seized rotor (Reference 1).

RCS Loops -

MODEs 1 and 2 satisfy 10 CFR 50.36(c)(2)(ii),

Criteria 2 and 3.

LCO The purpose of this LCO is to require adequate forced flow for core heat removal.

Flow is represented by having both RCS loops with both RCPs in each loop in operation for removal of heat by the two SGs.

To meet safety analysis acceptance criteria for DNB, four pumps are required at rated power.

Each OPERABLE loop consists of two RCPs providing forced flow for heat transport to an SG that is OPERABLE in accordance with the Steam Generator Tube Surveillance Program.

Steam generator, and hence RCS loop, OPERABILITY with regard to SG water level is ensured by the RPS in MODEs 1 and 2.

A reactor trip places the plant in MODE 3 if any SG level is Ž 50 inches below normal water level* as sensed by the RPS.

The minimum water level to declare the SG OPERABLE is < 50 inches below normal water level*.

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, 5, and 6.

Operation in other MODEs is covered by:

LCO 3.4.5, LCO 3.4.6, LCO 3.4.7, LCO 3.4.8, LCO 3.9.4, and LCO 3.9.5.

For Unit 2, the value shall remain 10 inches below the top of the feed ring through Cycle 14.

CALVERT CLIFFS - UNITS I & 2 B 3.4.4-2 Revision 13

RCS Loops - MODEs 1 and 2 B 3.4.4 BASES 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.

It should be noted that the reactor will trip and place the plant in MODE 3 as soon as the RPS senses less than 370,000 gpm** RCS flow.

The Completion Time of six hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging safety systems.

SURVEILLANCE REQUIREMENTS REFERENCES 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 of the required number of loops in operation.

Verification includes flow rate, temperature, or pump status monitoring, which help to 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 has been shown by operating practice to be sufficient to regularly assess degradation and verify operation within safety analyses assumptions.

In addition, Control Room indication and alarms will normally indicate loop status.

1. UFSAR, Chapter 14, "Safety Analysis" System Flow Rate limit shall be - 340,000 gpm through CALVERT CLIFFS UNITS 1 & 2 B 3.4.4-3 Revision 13 ACTIONS The Reactor Coolant Unit 2, Cycle 14.

I Revision 13 CALVERT CLIFFS - UNITS 1 & 2 B 3.4.4-3

MSSVs B 3.7.1 BASES

,-This LCO provides assurance that the MSSVs will perform their designed safety function to mitigate the consequences

, of-accidents that couldresult in a challenge to the reactor

- coolant pressure~boundary.,:-

2 APPLICABILITY In MODEs 1, 2, and 3, a minimum of five MSSVs per steam generator are required to be OPERABLE, according to S.-Table 3.7.1-1 in the accompanying LCO, which is limiting and

-,bounds all lower MODEs.

in MODEs 4 and 5, there are no credible transients requiring the MSSVs.

.,-The steam generators are not normally used for heat removal in MODEs 5 and 6, and thus cannot be overpressurized; there is no requirement for 1the.MSSVs to be OPERABLE in these MODEs.

ACTIONS The ACTIONS table is modified by a Note -indicating that separate Condition entry is allowed for each MSSV.

A.1 and A.2

'An'alternative to restoring'the inoperable MSSV(s) to OPERABLE'status isto reduce power so that'the available MSSV relieving capacity' meets Code requirements for the power level, The:number of.inoperable MSSVs will determine the necessary level*V6f reduction in secondary system steam flow and THERMAL POWER-required by the reduced reactor trip settings of the ppwei'level-high channels." The setpoints in Table 3.7.1-1 haVe~bien' verified by transient analyses.

The operator should limit thermaximum steady state power level to some value slightly below this setpoint to avoid an inadvertent overpower trip.

The four-hour Completion-Time for Required Action A.1 is a reasonable time period-torieduce power level and is based on the low prdbability of an event occurring'during this period that would require activation of the MSSVs.

An additional

'eiht-h 1ous is'ailowedi;

' Required Action A.2 to reduce the "setpoints..

The;Completio6jTime of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for Required Actioh A.2 is based *on -a reasonable time to correct the MSSV CALVERT-CLIFFS

- UNITS 1 & 2 B 3.7.1-3' iRevision 13,

MSSVs B 3.7.1 BASES inoperability, the time required to perform the power reduction, operating experience in resetting all channels of a protective function, and on the low probability of the occurrence of a transient that could result in steam generator overpressure during this period.

B.1 and B.2 If the MSSVs cannot be restored to OPERABLE status in the associated Completion Time, or if one or more steam generators have less than five MSSVs OPERABLE, the unit must be placed in a MODE in which the LCO does not apply.

To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.1.1 REQUIREMENTS This Surveillance Requirement (SR) verifies the OPERABILITY of the MSSVs by the verification of each MSSV lift setpoints in accordance with the Inservice Testing Program.

Reference 2,Section XI, Article IWV-3500, requires that safety and relief valve tests be performed in accordance with Reference 3.

According to Reference 3, the following tests are required for MSSVs:

a.

Visual examination;

b.

Seat tightness determination;

c.

Setpoint pressure determination (lift setting);

d.

Compliance with owner's seat tightness criteria; and

e.

Verification of the balancing device integrity on balanced valves.

The ANSI/American Society of Mechanical Engineers (ASME)

Standard requires that all valves be tested every five years, and a minimum of 20% of the valves be tested every 24 months.

The ASME Code specifies the activities, as found lift acceptance range, and frequencies necessary to satisfy the requirements.

Table 3.7.1-2 defines the lift CALVERT CLIFFS - UNITS I & 2 B 3.7.1-4 Revi sion 13

MSSVs B 3.7.1 BASES setting range for each MSSV for OPERABILITY; however, the valves are reset to + 15 during the surveillance test to allow for drift.

This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR.

This is to allow testing of the MSSVs at hot conditions.

The MSSVs may be either bench tested or tested in situ at hot conditions, using an assist device to simulate lift pressure.

If the MSSVs are not tested at hot conditions, the lift setting pressure shall be corrected to ambient conditions of the valve at operating temperature and pressure.

REFERENCES

1.

Updated Final Safety Analysis Report (UFSAR)

2.

ASME, Boiler and Pressure Vessel Code

3.

ANSI/ASME OM-1-1987, Code for the Operation and Maintenance of Nuclear Power Plants, 1987 CALVERT CLIFFS -

UNITS 1 & 2 B 3.7.1-5 Revision 13 Revision 13 CALVERT CLIFFS -

UNITS I & 2 B 3.7.1-5

AFW System B 3.7.3 BAE BASES achieve this status,: the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without chal l engi ng unit systems.

F. 1 SRequired-Action'F.-is modified by a Note'indicating that all required MODE-changes or power reductions are suspended until one AFW train-is restoredto OPERABLE status.

With two AFW trains inoperable-in MODEs 1,-.2, and 3, the unit may be in a~seriouslyidegraded condition with only non safety-related means for conducting~a cooldown.

In such a condition, the unit should not be perturbed by any action, including a power change, that might result' in a trip.

However, a powerchange.is.not precluded if it is determined to be the most pruddent action.

The seriousness of this "condition requires that, action be'started immediately to restore one AFW train to OPERABlE status.

While other plant conditions may requirenehtry'into LCO 3.0,3, the ACTIONS requ~ired by LCO 3.0.3 do not have'to be completed because

-they could force'the unit -into a less safe condition.

SURVEILLANCE REQUIREMENTS SR 3.7.3.1 Verifying 'the cofre~t alignment for manual, power-operated,

'and automatic valves'-in the-AFW water and steam supply flow paths, provides-assurance-that'the properflow paths exist for AFW operation6i.* Thiý SR-does'not apply to valves that are locked, sealed, or-6therwise secured in position, since these valves are verified to be in the correct position prior to, locking, sealing, or securing.

This SR also does not apply to valves that. 'ain6t be inadvertently misaligned, "such-as check' alves." This SR-does not require any testing

.or valve manipiultionsi.

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

CALVERT CLIFFS -

UNITS 1 & 2 B 3.7.3-7 Revision 12 CALVERT CLIFFS - UNITS 1 & 2 Revision 12 -

B 3-.7.3-7

AFW System B 3.7.3 BASES The 31 day Frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions.

SR 3.7.3.2 Cycling each testable, remote-operated valve that is not in its operating position, provides assurance that the valves will perform as required.

Operating position is the position that the valve is in during normal plant operation.

This is accomplished by cycling each valve at least one cycle.

This SR ensures that valves required to function during certain scenarios, will be capable of being properly positioned.

The Frequency is based on engineering judgment that when cycled in accordance with the Inservice Testing Program, these valves can be placed in the desired position when required.

SR 3.7.3.3 Verifying that each AFW pump's developed head at the flow test point is greater than or equal to the required developed head (__ 2800 ft for the steam-driven pump and

>__ 3100 ft for the motor-driven pump),

ensures that AFW pump performance has not degraded during the cycle.

Flow and differential head are normal tests of pump performance required by Reference 2.

Because it is undesirable to introduce cold AFW into the steam generators while they are operating, this testing is performed on recirculation flow.

This test confirms one point on the pump design curve and is indicative of overall performance.

Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance.

Performance of inservice testing, discussed in Reference 2, at three month intervals satisfies this requirement.

This SR is modified by a Note indicating that the SR should be deferred up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months until suitable test conditions are established.

This deferral is required because there is an insufficient steam pressure to perform the test.

CALVERT CLIFFS - UNITS 1 & 2 B 3.7.3-8 Revision 13

AFW System B 3.7.3 BASES SR 3.7.3.4 This SR ensures that AFW can be delivered',to the appropriate steam generator, in the event of any accident or transient that generates an AFAS signal, by demonstrating that each automatic valve in the flow path actuates to its correct position on an actual or-simulated actuati6n signal (verification of flow-modulating characteristics is not required).

This SR is'not required for valves that are locked, sealed, or otherwise secured in the'required position under administrative-controls.'

The 24 month Frequency is based on the heed to perform this surveillance test under the'conditions that apply during a unit outage

-'-"and the potential for*an-unplanned transient if the surveillance test were performed with the reactor at power.

The 24 month Frequehcy is acceptable,;based on the design reliability and.operating'experience of the equipment.

This SR is modifiedlbyý,a Note indicating that the SR should be deferred up to24 hours until suitable test conditions have been established.

SR 3.7.3.5 This SR ensures-that-,the-AFW:pumps will start in the event of any accident-or transient that generates an AFAS signal by demonstrating that each AFW-pump starts automatically on an actual or simulated actuation signal.

The 24 month Frequency is acceptable, based on the design reliability and operating experience of the equipment.

This SR is modified by a Note.

The Note indicates that the SR should be deferred up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months until suitable test conditions are established.

SR 3.7.3.6 This SR ensures that the AFW system is capable of providing a minimum nominal flow to each flow leg.

This ensures that the minimum required flow is capable of feeding each flow leg.

The test may be performed on one flow leg at a time.

The SR is modified by a Note which states, the SR is not required to be performed for the AFW train with the turbine driven AFW pump until up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reaching 800 psig in the steam generators.

The Note ensures that proper test CALVERT{LIFFS UNITS 1 & 2 B 3.7.3-9 Revisibn 13 Revision 13-,

CALVERT;CLIFFS - UNITS I & 2 B 3.7.3-9

AFW System B 3.7.3 BASES conditions exist prior to performing the test using the turbine-driven AFW pumps.

The 24 month Frequency coincides with performing the test during refueling outages.

SR 3.7.3.7 This SR ensures that the AFW System is properly aligned by verifying the flow path to each steam generator prior to entering MODE 2 operation, after 30 days in MODEs 5 or 6.

OPERABILITY of AFW flow paths must be verified before sufficient core heat is generated that would require the operation of the AFW System during a subsequent shutdown.

The Frequency is reasonable, based on engineering judgment, and other administrative controls to ensure that flow paths remain OPERABLE.

To further ensure AFW System alignment, the OPERABILITY of the flow paths is verified following extended outages to determine that no misalignment of valves has occurred.

This SR ensures that the flow path from the CST to the steam generators is properly aligned.

Minimum nominal flow to each flow leg is ensured by performance of SR 3.7.3.6.

REFERENCES

1.

UFSAR, Section 10.3

2.

ASME, Boiler and Pressure Vessel Code,Section XI, Inservice Inspection, Article IWV-3400 CALVERT CLIFFS -

UNITS 1 & 2 B 3.7.3-10 Revision 12 B 3.7.3-10 CALVERT CLIFFS -

UNITS I & 2 Revision 12

CRETS B 3.7.9 B 3.7 PLANT SYSTEMS B 3.7.9 BASES Control Room Emergency Temperature System (CRETS)

BACKGROUND.

.,The CRETS provides temperature control for the Control Room following isolation-of the Control Room.--The CRETS is a shared system which is-supported by the CREVS, since the CREVS must be operating in the emergency recirculation mode for CRETS to perform its safety function.

The CRETS consists of two independent,-redundant trains that provide cooling of recirculated Control Room air.

Each train consists of cooling coils, instrumentation, and controls to provide for Control Room temperature control.

-The CRETS is a subsystem providing air temperature control for the Control Room.

r The CRETS-is an emergency s'ystem, parts of which may also S:-operate during normal unit operations in the standby mode of operation.

A single train will provide therequired temperature control to maintain the Control-Room below

.104 0F.- The CRETS operation to maintain the Control Room temperature is discussedin Reference 1.

APPLICABLE--

SAFETY ANALYSES The design basis-of the CRETS is to maintain temperature

-of the-Control. Room-environment throughout,30 days of continuous occupancy..

A The CRETS components are arranged in redundant safety related trains.

During emergency operation, the CRETS

-.- maintains the temperature-below, 1lO40F.

A single active 1failure of a component of the CRETS, assuming a loss of 0offsite power, -,does -not impaoir the ability,of the system to perform its design.function.,. Redundant detectors and

- controls are provided-for Control Room temperature control.

A The CRETS is designed -in accordance with,Seismic Category I requirements.

TheCRETS is capable of removing sensible and latent heat loads from the Control Room, considering Lo.,equipment heat,loads and personnel occupancy requirements, to *ensure equipment OPERABILITY.

The CRETSsatisfiesiO CFR 50.36(c)(2) (ii), Criterion 3.

CALVERT CLIFFS -

UNITS 1 & 2 B3.7.9-1

- Revision 2 I I I

CALVERT CLIFFS - UNITS I & 2 Revision 2 B -3.-,7,.-9 -,I ý.

CRETS B 3.7.9 BASES LCO Two independent and redundant trains of the CRETS are required to be OPERABLE to ensure that at least one is available, assuming a single failure disables the other train following isolation of the Control Room.

Total system failure could result in the equipment operating temperature exceeding limits in the event of an accident requiring isolation of the Control Room.

The CRETS is considered OPERABLE when the individual components that are necessary to maintain the Control Room temperature are OPERABLE.

The required components include the cooling coils and associated temperature control instrumentation.

In addition, the CRETS must be OPERABLE to the extent that air circulation can be maintained.

For MODEs 1, 2, 3, and 4, redundancy is required and both trains must be OPERABLE.

The LCO is modified by a Note which indicates that only one CRETS train is required to be OPERABLE during movement of irradiated fuel assemblies.

Therefore, redundancy is not required for movement of irradiated fuel assemblies and only one CRETS train is required to be OPERABLE.

APPLICABILITY In MODEs 1, 2, 3, and 4, and during movement of irradiated fuel assemblies, the CRETS must be OPERABLE to ensure that the Control Room temperature will not exceed equipment OPERABILITY requirements following isolation of the Control Room.

The additional Applicability for the movement of irradiated fuel assemblies was added because the CRETS is credited during a fuel handling accident.

For clarity, an LCO Note was added to require only one train of CRETS to be OPERABLE when movement of irradiated fuel is in progress.

The fuel handling accident does not assume a single failure to occur.

The LCO Note does not identify this requirement on a CRETS train basis so that it is also applicable to redundant components within a train, such as the outside air intake isolation valves.

In conjunction with this Note, Action C was added.

CALVERT CLIFFS - UNITS 1 & 2 Revision 13 B 3.7.9-2

MFIVs B 3.7.15 B 3.7 PLANT SYSTEMS B 3.7.15 Main Feedwater Isolation Valves (MFIVs)

BASES BACKGROUND The MFIVs isolate MFW flow to the secondary side of the

-steam-generators following a HELB.

The consequences of

, HELBs occurring in the main steam lines or.-in the MFW lines

- i -, downstream of,the MFIVs will be mitigated by their closure.

Closure of'the MFIVs effectively terminates the addition of feedwater to an affected steam generator,,limiting the mass and energy release for SLBs /or feedwater line breaks (FWLBs) inside theContainment Structure upstream of the

.reverse flow check valve.-and reducing the cooldown effects

-for SLBs.

The MFIVs isolate-the non-safety-related portions from the 2,safety-related portion of-the system.

Inthe event of a secondary side pipe rupture.inside,the Containment Structure upstream of the reverse flow check valve,.the valves limit the quantity of, high energy -fluid that enters the Containment Structure through the break.

SOne MFIV is located on,-each MFW line, outside, but close to, the Containment Structure., The MFIVs are located so that AFW may be supplied toa steam generator following MFIV closure.

The piping.yolume from the valve to the steam generator must be accounted for in calculating mass and energy releases.-------------

The MFIVs close on.receipt of a-steam generator isolation signal generated by low steam generator pressure.

The steam c_ generator isolation signal -also actuates, the MSIVs to close.

,TheeMFIVs may al'sobe actuated manually:.,In addition, the MFIVs reverse flowcheck'valve inside the Containment Structure is available to isolate-the feedwater line penetrating the Containment Structure, annd-to ensure that the consequences of events do not exceed the capacity of the Containment Cooling-System. --

A description of the MFIVs operation on receipt of an steam generator isolation signal is found in Reference 1.

1 Rev1s1on

CALVERT. CLIFFS -

UNITS 1 & 2 B 3.715-1 CALVERT-CLIFFS - UNITS 1 & 2 Kevision~z,-

B 3.7'.15-1,

MFIVs B 3.7.15 BASES APPLICABLE SAFETY ANALYSES The design basis of the MFIVs is established by the analysis for the large SLB.

It is also influenced by the accident analysis for the large FWLB.

Failure of an MFIV to close following an SLB or FWLB can result in additional mass and energy to the steam generator's contributing to cooldown.

This failure also results in additional mass and energy releases following an SLB or FWLB event.

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

LCO This LCO ensures that the MFIVs will isolate MFW flow to the steam generators.

Following an FWLB or SLB, these valves will also isolate the non-safety-related portions from the safety-related portions of the system.

This LCO requires that one MFIV in each feedwater line be OPERABLE.

The MFIVs are considered OPERABLE when the isolation times are within limits, and are closed on an isolation actuation signal.

Failure to meet the LCO requirements can result in additional mass and energy being released to the Containment Structure following an SLB or FWLB inside the Containment Structure.

Failure to meet the LCO can also add additional mass and energy to the steam generators contributing to cooldown.

APPLICABILITY The MFIVs must be OPERABLE whenever there is significant mass and energy in the RCS and steam generators.

In MODEs 1, 2, and 3, the MFIVs are required to be OPERABLE in order to limit the amount of available fluid that could be added to the Containment Structure in the case of a secondary system pipe break inside the Containment Structure.

In MODEs 4, 5, and 6, steam generator energy is low.

CALVERT CLIFFS -

UNITS 1 & 2 B 3.7.15-2 Revision 13 CALVERT CLIFFS -

UNITS I & 2 B 3.7.15-2 Revision 13

AC Sources-Operating B 3.8.1 BASES perform the surveillance-test, and is intended to be

... consistent-with expected fuelcycle lengths.

Operating experience has shown that-these-components usually pass the SR when performed at the 24 month Frequency.

Therefore, the Frequency was concluded-to be acceptable from a reliability standpoint.

This Frequency-is consistent-with Reference 2, Chapter 8.

SR 3.8.1.14 This-SR ensures that the manual synchronization and load transfer from the DG-to.the-offsite source Ian be made and that the DG can be-returned to-ready-to-load status when power is restored. -The DG is considered to be in ready-to-load status when the DG is at rated speed and voltage, the output breaker islopen and can receive an auto

-close signal on bus undervoltage, and the load sequence timers are.reset.

The Frequency of 24 months takes into 6onsideration unit

. conditions requiredstoperform the surveillance test.

,).

SR 3.8.1.15 In the event of a DBA coincident with a-loss of offsite power, the DGs are required to supply the necessary power to

"-ESF, systems so that the-fuel 'RCS, and-containment design

'limits-are not exceeded.-,<

-'This SR demonstratesthe!DG"operation during a loss of

-- offsite power'actuationitest signal in-conjunction with an ESF-(i.e.,-safety injection)actuation signal.

In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the DG system to perform these functions-is acceptable.

This testing may include any. series of sequential, overlapping, or-total steps so that the entire connection and loading sequence is verified.

e It is not necessary-to energize loads which'fare dependent on

-terperature,to load (i.e., heat tracing, switchgear HVAC compressor, compute&rroom'HVAC compressorY,.

Also, it is acceptable to transfer'-the -instrument AC bus to the non K

tested train to maintai ii'safe operation of the plant during CALVERT CLIFFS -

UNITS 1 & 2 B -3N;8.1-27 1 Revi sion AC Sources-Operating B 3.8.1 BASES testing.

Loads (both permanent and auto connect) < 15 kW do not require loading onto the diesel since these are insignificant loads for the DG.

Permanently-and auto-connected loads to the emergency diesel generators are defined as follows:

Permanently-Connected Load - Equipment that is not shed by an undervoltage or safety injection actuation signal and is normally operating, i.e., loads that are manually started, selected, or process signal controlled are not considered permanently-connected loads.

Auto-Connected Loads - Emergency equipment required for mitigating the events described in UFSAR Chapter 14 that are energized by loss-of-coolant incident sequencer actions after step zero and within the first minute of emergency diesel generator operation after the initiation of an undervoltage signal.

The Frequency of 24 months takes into consideration unit conditions required to perform the surveillance test and is intended to be consistent with an expected fuel cycle length of 24 months.

This SR is modified by a Note.

The reason for the Note is to minimize mechanical wear and stress on the DGs during testing.

For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations for DGs.

SR 3.8.1.16 This SR lists the SRs that are applicable to the LCO 3.8.1.c (SRs 3.8.1.1, 3.8.1.2, 3.8.1.3, 3.8.1.5, 3.8.1.6, and 3.8.1.7).

Performance of any SR for the LCO 3.8.1.c will satisfy both Unit 1 and Unit 2 requirements for those SRs.

Surveillance Requirements 3.8.1.4, 3.8.1.8, 3.8.1.9, 3.8.1.10, 3.8.1.11, 3.8.1.12, 3.8.1.13, 3.8.1.14, and 3.8.1.15, are not required to be performed for the LCO 3.8.1.c.

Surveillance Requirement 3.8.1.10 is not CALVERT CLIFFS -

UNITS 1 & 2 B 3.8. 1-28 Revision 13

AC Sources-Operating B 3.8.1 BASES

-required because this SR verifies manual transfer of AC

-*-power sources from the~norl -offsite circutoth alternate offsite circuit, but only one qualified offsite circuit is necessary for the LCO 3.8.1.c.

Surveillance Requirements 3.8.1.4,.3.8.1.11, and 3.1.8.12 are not required because they'are tests that deal with loads.

Surveillance Requirement 3.8.1.8 verifies the interval between sequenced loads.

Surveillance Requirement 3.8.1.14 verifies the proper sequencing with offsite power.

Surveillance Requirement 3.8.1.9 verifies that the DG starts within 10 seconds.

These SRs are not required because they do not support the function of the LCO 3.8.1.c to provide power to the CREVS,

CRETS, and H. Analyzer.

Surveillance Requirements 3.8.1.13 and 3.8.1.15 are not required to be performed because these SRs verify the emergency loads are actuated on an ESFAS signal for the Unit in which the test is being performed.

The LCO 3.8.1.c DG will not start on an ESFAS signal for this Unit.

REFERENCES

1.

10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants"

2.

UFSAR

3.

Regulatory Guide 1.9, Revision 3, "Selection, Design, Qualification, and Testing of Emergency Diesel Generator Units Used as Class 1E Onsite Electric Power Systems at Nuclear Power Plants," July 1993

4.

Safety Guide 9, Revision 0, March 1971

5.

NRC Safety Evaluation for Amendment Nos. 19 and 5 for Calvert Cliffs Nuclear Power Plant Unit Nos.

1 and 2, dated January 14, 1977

6.

Regulatory Guide 1.93, Revision 0, "Availability of Electric Power Sources," December 1974

ý

7.

Generic Letter 84-15, Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability, July 2, 1984

8.

Regulatory Guide 1.137, Revision 1, "Fuel-Oil Systems for Standby Diesel Generators," October 1979 CALVERT CLIFFS -

UNITS 1 & 2 B :3.8.1-29 2 Revision 13 B -3.8.1-29 **!

Revision 13, CALVERT CLIFFS - UNITS 1 & 2

AC Sources-Operating B 3.8.1 BASES

9.

Letter from Mr. D. G. McDonald, Jr. (NRC) to Mr. C. H. Cruse (BGE),

dated April 2, 1996, Issuance of Amendments for Calvert Cliffs Nuclear Power Plant, Unit 1 (TAC No.

M94030) and Unit 2 (TAC No.

M94031)

10. IEEE Standard 308-1991, "IEEE Standard Criteria for Class 1E Power Systems for Nuclear Power Generating Stations" CALVERT CLIFFS -

UNITS 1 & 2 B 3.8.1-30 Revision 13 CALVERT CLIFFS - UNITS I & 2 B 3.8. 1-30 Revision 13

Boron Concentration B 3.9.1 B 3.9 REFUELING OPERATIONS KB 3.9.1 Boron Concentration BASES BACKGROUND

.;K' The limiton the-boron concentrations of the Reactor Coolant System,(RCS) and refueling pool during refueling, ensures

-.. that-the-reactdr~remains subcritical during MODE 6.

, 'Refueling boroniconcentration is the soluble boron concentration in.the-coolant in each of these volumes that have direct access-to the.'reactor core during refueling.

The solubleboron concentration offsets the core reactivity

--- and is-measured-by chemical-analysis of a representative sample of-the coolantin-each of the volumes.

The refueling

- boron concentration limitris specified in the Core Operating

.LimitsReport (COLR)-Unit pr6cedures ensure the specified Sboron-concentrationjin-order to maintain an~overall core reactivity'of keff -0.95 during fuel handling, with.control

-element assemblies and fuel assemblies assumed to be in the most adverse configuration (least negative reactivity) allowed by unit procedures.

The negative worth of the CEAs may be credited whenrdetermining the refueling boron concentrations.'

Unit-procedures maintain the number and position of credited CEAs-during fuel handling operations.

-Reference 1, Appendix 1C, Criterion 27, requires that two

-. =--independent reactivity control systems-of different design principles be-provided..--One of-these systems must be capable of holding the reactor core subcritical under cold

, -conditions.

The Chemical and Volume Control System is the

- system capable of maintaining the reactor subcritical in

- cold conditions bymaintaining the boron concentration.

The reactor is brought to shutdown conditions before beginning operations to open the reactor vessel for

ý_,refueling.

After, the RCS is cooled and depressurized and the vessel head is-unbolted,,the head is slowly removed to form the refueling pool." The refueling -pool is then flooded with borated water from the refueling water tank into the r

open, reactorr vessel by,grayity feeding or by the use of the Shutdown Cooling (SDC)

System pumps.

CALVERT CLIFFS - UNITS 1 & 2 B 3.9.111-1 f

Revision 11

Boron Concentration B 3.9.1 BASES The pumping action of the SDC System in the RCS and the natural circulation due to thermal driving heads in the reactor vessel and the refueling pool mix the added concentrated boric acid with the water in the RCS.

The SDC System is in operation during refueling [see Limiting Condition of Operation (LCO) 3.9.4 and LCO 3.9.5], to provide forced circulation in the RCS and assist in maintaining the boron concentrations in the RCS and the refueling pool above the COLR limit.

The COLR includes a MODE 6 temperature limitation of

< 140 0F.

This restriction ensures assumptions made for calculating boron concentration and assumptions made in the boron dilution analysis for MODE 6 are preserved.

APPLICABLE During refueling operations, the reactivity condition of the SAFETY ANALYSES core is consistent with the initial conditions assumed for the boron dilution accident in the accident analysis and is conservative for MODE 6.

The required boron concentration and the unit refueling procedures that demonstrate the correct fuel loading plan (including full core mapping) ensure the keff of the core will remain

0.95 during the refueling operation.

Hence, at least a 5% Ak/k margin of safety is established during refueling.

The boron concentration limit specified in the COLR includes an uncertainty allowance.

During refueling, the water volume in the spent fuel pool, the transfer tube, the refueling pool, and the reactor vessel form a single mass.

As a result, the soluble boron concentration is relatively the same in each of these volumes.

The limiting boron dilution accident analyzed occurs in MODE 5 (Reference 1, Chapter 14).

A detailed discussion of this event is provided in B 3.1.1.

The RCS boron concentration satisfies 10 CFR 50.36(c)(2)(ii),

Criterion 2.

CALVERT CLIFFS UNITS 1 & 2 B 3.9.1-2 Revision 13 Revision 13 CALVERT CLIFFS - UNITS 1 & 2 B 3.9.1-2

Containment Penetrations B 3.9.3 B 3.9 REFUELING OPERATIONS B 3.9.3 Containment Penetrations BASES BACKGROUND During-CORE ALTERATIONS or movement of irradiated fuel assemblies within the Containment Structure, a release of fisslion product radioactivity within the Containment

-Structure will be restricted from escaping to the environment when the LCD-requirements aremet.

In MODEs 1,

- 2, 13, and 4, this is accomplished by maintaining Containment OPERABLE, as described in LCO-3.6.1.

In MODE 6, the

.potential for containment pressurization as a result of an accident is notlikely; -therefore, requirements to isolate the containment atmosphere-from the outside atmosphere can be less stringent., The LCO-requirements are referred to as "containment closure",rather than "containment OPERABILITY."

Containment closure~means that~all potential filtered or unfiltered escape paths.are closed or capable of being

.closed.

Since.there-is no-design basis accident potential

-for containment-pressurization,ithe Appendix J leakage 7

criteria and tests-are-not required.

K-The Containment Structure serves to contain fission product radioactivity that-may be released from the reactor core following-an accident-such that offsite radiation exposures are maintained.well-within the requirement'sof 10 CFR Part 100.

Additionally, the-,Containment Structure provides radiation shielding fromIthe fission products that may be present in the containment atmosphere following accident conditions.

The containment equipment-hatch opening provides a means f4 movinglarge equipmrent;and components into -and out of the.

Containment Structure..During CORE ALTERATIONS or movemen

-- of-irradiated fuel assemblies within Containment, the equipment hatch must, be held-,in place by at least four bol or the containment outage door must be capable of being closed.

Good engineering practice dictates that the bolts required by this LCObe approximately equally spaced.

or t

ts The:containment airlocks,.which-are part of the containment pressure boundary,' providie meaans -for personnel access during MODEs 1, 2, 3, and 4 operation in accordance with LCO 3.6.2.

Each air lock has a door at both ends.

The CALVERT-CLIFFS - UNITS 1 & 2

"_Revi sion13, B--3;g*3 Containment Penetrations B 3.9.3 BASES doors are normally interlocked to prevent simultaneous opening when Containment OPERABILITY is required.

In other situations, the potential for containment pressurization as a result of an accident is not present, therefore, less stringent requirements are needed to isolate the containment atmosphere from the outside atmosphere.

Both containment personnel air lock doors and the containment outage door may be open during the movement of irradiated fuel assemblies in containment and during CORE ALTERATIONS; provided one air lock door and the containment outage door are OPERABLE, the plant is in MODE 6 with at least 23 ft of water above the fuel, and a designated individual is continuously available to close each door.

The designated individuals must be stationed at the Auxiliary Building side of the outer air lock door and at the outside of the containment equipment hatch.

OPERABILITY of a containment personnel air lock door requires that the door is capable of being closed, that the door is unblocked, and no cables or hoses are run through the doorway.

OPERABILITY of the containment outage door requires that the door is capable of being closed, that the door is unblocked (i.e., capable of being closed within 30 minutes), and no cables and hoses are run through the doorway.

Containment outage door grating or truck ramps may be installed if the grating or truck ramps can be removed with the use of a forklift and the door closed within 30 minutes.

During CORE ALTERATIONS or movement of irradiated fuel assemblies in containment, the requirement for at least 23 ft of water above the fuel, ensures that there is sufficient time to close the personnel air lock and the containment outage door following a loss of SDC before boiling occurs and minimizes activity release after a fuel handling accident.

The personnel air lock door and the containment outage door may be operated independently of each other (i.e., they do not have to be open or shut at the same time).

The requirements on containment penetration closure, ensure that a release of fission product radioactivity within the Containment Structure will be restricted to within regulatory limits.

CALVERT CLIFFS UNITS 1 & 2 B 3.9.3-2 Revision 13 CALVERT CLIFFS - UNITS 1 & 2 B 3.9.3-2 Revision 13

Containment Penetrations B 3.9.3 BASES

°*The Containment -Purge -Valve 'Isolation System, for the

-purposes of compliance with LCO 3.9.3, item d.2, includes a 48 inchipurge penetration-anda 48 inch exhaust penetration.

For the purposes of'compliance with LCO 3.9.3, the containment vent isolation valves are not considered part of

-,theContainment Purge=Valve IsolationSystem since they may

notbe capable'of-being closed automatically.

The

containment ventiincludes a fourinch purge penetration and a four inch exhaust penetration.

DuringMODEs 1, 2, 3, and 4; the normal purge.and exhaust-penetrations are isolated

-- (via'a-blind flange,*if installed or by the purge valves).

The containment vent~valves can be opened, intermittently, but-are closed'automatically-by the Engineered Safety Features Actuation System.

Neither of-the subsystems is subject to a Specification in MODE 5.

In MODE 6, large air exchanges are desiredito conduct

--refueling operatibni-9 T-e'rniormal 48 inch purge system-is used,,for this purpose and allvalves are closed by the

-Engineered Safety Features-Actuation System in accordance with LCO,3.3.7.

The containment~ventjisolation valves arerequired to be closed during CORE-ALTERATIONS or.movement-of irradiated fuel within Containment.-- These valves are connected to the penetration room Technical Specification emergency air cleanup systems,'which exhaust to the outside atmosphere

'through high efficiencyparticulate air andcharcoal

-J-ilters.-

.The other containment-penetrations that provide direct access from containment atmosphere to outside atmosphere through a'filtered or unfiltered pathway must be isolated on atleast one side.,Isolation may. be achieved by an OPERABLE automatic isolation,yalve,.or by a manual isolation valve,

- blind flange, or equivalent. -Equivalent isolation methods must be approved in accordance with appropriate American Society of Mechanical Engineers / American National Standards InstituterCodes, and may include-use of a material that'can provide a~-temporary-ventilation~barrier for the

,-,other'containment-penetrations during-fuel -movements.

CALVFRT CLIFFS -

UNITS 1 & 2 Br,3;9;3 Revision 13 Vl W.................

Containment Penetrations B 3.9.3 BASES APPLICABLE During CORE ALTERATIONS or movement of irradiated fuel SAFETY ANALYSES assemblies within the Containment Structure, the most severe radiological consequences result from a fuel handling accident.

The fuel handling accident is a postulated event that involves damage to irradiated fuel (Reference 1).

The fuel handling accident, described in Reference 1, includes dropping a single irradiated fuel assembly which would then rotate to a horizontal position, strike a protruding structure, and rupture the fuel pins.

The requirements of LCO 3.9.6, and the minimum decay time of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to CORE ALTERATIONS, ensure that the release of fission product radioactivity, subsequent to a fuel handling accident, results in doses that are within the acceptance limits given in Reference 1.

Containment penetrations satisfy 10 CFR 50.36(c) (2) (ii),

Criterion 3.

LCO This LCO limits the consequences of a fuel handling accident in Containment Structure, by limiting the potential escape paths for fission product radioactivity released within Containment.

The LCO requires any penetration providing direct access from the Containment Structure atmosphere to the outside atmosphere through a filtered or unfiltered pathway (including the containment vent isolation valves) to be closed, except for the OPERABLE containment purge and exhaust penetrations, the containment personnel air locks and the containment outage door.

For the OPERABLE containment purge and exhaust penetrations, this LCO ensures that these penetrations are isolable by the Containment Purge Valve Isolation System.

The OPERABILITY requirements for this LCO ensure that the automatic purge and exhaust valve closure times specified in the UFSAR can be achieved, and therefore meet the assumptions used in the safety analysis to ensure releases through the valves are terminated, such that the radiological doses are within the acceptance limit.

Both containment personnel air lock doors and the containment outage door may be open under administrative controls during movement of irradiated fuel in Containment and during CORE ALTERATIONS provided that one OPERABLE personnel air lock door and an OPERABLE containment outage CALVERT CLIFFS - UNITS 1 & 2 B 3.9.3-4 Revision 13

Containment Penetrations B 3.9.3 BASES door are capable of being closed in the event of a fuel handling accident.

The administrative controls consist of designated individuals available immediately outside the-,

I":--personnel -air lock and the containment outage door to close theOPERABLE doors., -Should a fuel,handling accident occur inside the.ContainmentStructure, one personnel air lock

,door-and-the containment outage door will be closed

- following an evacuation.of the ContainmentfStructure.

S,:*

TheLCO is modified by.a Note which allows the emergency air Jlock temporary closure device to replace an emergency air lock door.

The temporary closure device provides an adequate barrier to shield the environment from the containment-,atmosphere in-case of a design basis event that

_ does not-create a-pressure increase inside Containment.

APPLICABILITY The containment-penetration requirements are applicable

duririg.CORE ALTERATIONS or movement of irradiated fuel

- :'assemblies withinthe.Containment Structure because this is when there-is-a-potential for a fuel handling accident.

In

"'MODEs'1, 2, 3,land-4, containment penetration requirements

--are addressed by'LCO'3"6.1.

In-MODEs 5 and 6, when CORE

-ALTERATIONS or movement of irradiated fuel assemblies within the ContainmentiStrbcture are not being conducted, the "potential for a fuel. handling accident does not exist.

Therefore, under these-conditions no requirements are placed on containment penetration status.

ACTIONS A.1 and A.2, With the containment equipment hatch,, air locks, or any S containnient penetrations that provides.direct access from the S6-ntainment atmospIhere'to the outside atmosphere through a

- filtered '6r~unfiitered-pathway not in the'required status,

-(includihig the*,Co"tainm'ent Purge'and Exhaust Isolation

".,System not capable of automaticractuation.when the purge and xe x -hadstvalves-arer open) the unit must be-placed in a condition in which-the isolationý-function is not needed.

'7"-This.is accomplisf-shed byimmediately susI-ending CORE ALTERATIONS ard movdmenibof-irradiated fuel assemblies withiin-."the ContainnmentStructure.', Performance of these "a-ations shall notlpreclude-completionof'mbvement of a component to a safefposition.

CALVERT.CLIFFS

- UNITS 1 & 2 B,,3.9.3=5$

Revisi6on 1.3

Containment Penetrations B 3.9.3 BASES SURVEILLANCE SR 3.9.3.1 REQUIREMENTS This SR demonstrates that each of the containment penetrations required to be in its closed position, is in that position.

The surveillance test on the open purge and exhaust valves will demonstrate that the valves are not blocked from closing.

Also, the surveillance test will demonstrate that each purge and exhaust valve operator has motive power, which will ensure each valve is capable of being closed by an OPERABLE automatic Containment Purge Valve Isolation System.

The surveillance test is performed every seven days during CORE ALTERATIONS or movement of irradiated fuel assemblies within the Containment Structure.

The surveillance test interval is selected to be commensurate with the normal duration of time to complete fuel handling operations.

A surveillance test before the start of refueling operations will provide two or three verifications during the applicable period for this LCO.

As such, this SR ensures that a postulated fuel handling accident, that releases fission product radioactivity within the Containment Structure, will not result in a release of fission product radioactivity to the environment in excess of those described in Reference 1.

SR 3.9.3.2 This SR demonstrates that each containment purge and exhaust valve actuates to its isolation position on manual initiation or on an actual or simulated high radiation signal.

The once each refueling outage Frequency, maintains consistency with other similar Engineered Safety Features Actuation System instrumentation and valve testing requirements.

However, in order to ensure the SR Frequency is satisfied, this surveillance test is typically performed once per refueling outage prior to the start of CORE ALTERATIONS or movement of irradiated fuel assemblies within Containment.

In LCO 3.3.7, the Containment Radiation Signal System requires a CHANNEL CHECK every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and a CHANNEL FUNCTIONAL TEST every 92 days to ensure the channel OPERABILITY during refueling operations.

Every 24 months a CALVERT CLIFFS -

UNITS 1 & 2 B 3.9.3-6 Revision 13

Containment Penetrations B 3.9.3 BASES CHANNEL CALIBRATION is performed.

The system actuation response time is demonstrated every 24 months during refueling on a STAGGERED TEST BASIS.

Surveillance Requirement 3.6.3.4 demonstrates that the isolation time of each valve is in accordance with the Inservice Testing Program requirements.

These surveillance tests performed during MODE 6 will ensure that the valves are capable of closing after a postulated fuel handling accident to limit a release of fission product radioactivity from the Containment Structure.

REFERENCES

1. UFSAR, Section 14.18, "Fuel Handling Incident" 1

Kevs1on

+/-3 CALVERT CLIFFS -

UNITS 1 & 2 B 3.9.3-7 KeVlSlOn 13 B 3.9.3-7 CALVERT CLIFFS - UNITS 1 & 2

SDC ard C6olant Circulation-High Water Level B 3.9.4 BASES.

concentration, CORE ALTERATIONS are suspended and all

-containment penetrations must be in the-,status described in "LCO.3.9.3. This -allowance is necessary to perform required maintenance and,testing.

APPLICABILITY

-One-SDC1oop must be in operation.in-MODE 6, with the water level 23 ft above the'top'of the irradiated fuel

-.,assemblies seatedin~the reactor vessel, to-provide decay

---heat removal.

-The-23-ft.level-was selected because it

-,corresponds to the:23.ft requirement established for fuel

,1 1',movement in LCO 3;9;6.

Requirements for the SDC System in other MODEs are covered

-by LCOs in Section,3.4_and Section 3.5.

Shutdown cooling

-loop' -requirements ini MODE-6, with'the water level < 23 ft above'the'toppofthe'_ir-rdiated fuel assemblies seated in the reactor vessel, areýlocated in LCO 3.9.5.

ACTIONS Shutdown cooling loop requirements are met by having one SDC loop OPERABLE and in operation, except as permitted in the Note to the LCO...

If one required SDC loop is inoperable or not in operation, action shall be immediately-initiated and,continued until the SDC loop is restored to OPERABLE stat-us and to.

operation.

An immediate Completion Time is necessary for an operator to initiate corrective actions.

A.2-.

i I-If SDC'loop req uiremnents are"_n6t met' there will be no forced circul ati-on t&provi de mixiing tioestabl i sh uni form

. boron'concentrations-Red6"ced boron concce6trations can

,occur-through the addition 'of water-withailower boron concentration1tha6tlat contained "in the-RCS.

Therefore, a-cti n that *reduce-boron-concentrationshal 1 be -suspended immediately.

In addi-tion, to ensure compliance with the

-action is maintiained,-the-chargingpumps shall be de-I energized and charging flow paths closed as part of Required Action A.2.

CALVERT CLIFFS - UNITS 1 & 2 B 3.9.4-3

ReVision-11,

SDC and Coolant Circulation-High Water Level B 3.9.4 BASES A.3 If SDC loop requirements are not met, actions shall be taken immediately to suspend loading irradiated fuel assemblies in the core.

With no forced circulation cooling, decay heat removal from the core occurs by natural convection to the heat sink provided by the water above the core.

A minimum refueling water level of 23 ft above the irradiated fuel assemblies seated in the reactor vessel provides an adequate available heat sink.

Suspending any operation that would increase the decay heat load, such as loading a fuel assembly, is a prudent action under this condition.

A.4 If SDC loop requirements are not met, all containment penetrations providing direct access from containment atmosphere to the outside atmosphere through a filtered or unfiltered pathway must be closed to prevent fission products, if released by a loss of decay heat event, from escaping the Containment Structure.

The four hour Completion Time allows fixing most SDC problems without incurring the additional action of violating the containment atmosphere.

The emergency air lock temporary closure device cannot be credited for containment closure for a loss of shutdown cooling event.

At least one door in the emergency air lock must be closed to satisfy this action statement.

SURVEILLANCE SR 3.9.4.1 REQUIREMENTS This SR demonstrates that the SDC loop is in operation and circulating reactor coolant.

The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability, and to prevent thermal and boron stratification in the core.

The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the flow, temperature, pump control, and alarm indications available to the operator in the Control Room for monitoring the SDC System.

REFERENCES

1. UFSAR, Section 9.2, "Shutdown Cooling System" CALVERT CLIFFS - UNITS 1 & 2 B 3.9.4-4 Revision 13

-,SDC and Coolant Circulation-Low Water Level B 3.9.5 BASES seated in the reactor vessel, the Applicability will change to that of LCO 3.9.4, and only one SDC loop is required to" T

.*-be OPERABLE and in operation.

An immediate.Completion Time

-- is.necessary for.an operator-to initiate corrective actions.

If hoSDC loop is-in'operation or no SDC loops are OPERABLE, there will be n6-f6r'cd circulation to provide mixing to establish uniform boron concentrations.

Reduced boron concentrations can-occur by-the addition of water with lower bo*on concentration than-that containedin the RCS.

Therefore, ictions~that-reduce boron concentration shall be

-suspended immediately.

In addition, to ensure compliance

'-with the action is maintained, the charging pumps shall be de-energized and charging flow paths closed as part of Required Action B. 1'.,

-B.__2

.If no SDC loop is in' operation or'no SDC loops are OPERABLE, action shall be ififiated'immediately and -continued without i nterruption to restor& one SDC loop to 'OPERABLE status and

. operation.

Since6'the-unit is in Conditions A and B concurrently, the restoration of two OPERABLE SDC loops and one operating SDC loop should be accomplished expeditiously.

B.3 If no SDC loop, is'inoperation, all containment penetrations

- providing direct access -from the containment atmosphere to "the outside' atm6spherethrough a filtered 'or unfiltered

.,,pathway.must be closed within four'hours.

'With the SDC loop requirements not met,the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere.

Closingcontainmentpenetrations that are open to the outside' ati6sph'ere through'a fi'ltered or unfiltered pathway ensures that dose limits are not exceeded.

The emergency air lock temporary closure device cannot be credited for ontaiainment-closure-fora*, loss of shutdown cooling pvent-, IAtileastLobne'door in'the-emergency air lock umst"ý.closed-t6°*atisfy-this action statement.

The-Completion Time of fobr'hours is reasonable, based on the low probability-of the coolant boiling in that time.

CALVERTCLIFFS - UNITS 1 & 2

. -3.9-5_

Revision 13'

SDC and Coolant Circulation-Low Water Level B 3.9.5 BASES SURVEILLANCE SR 3.9.5.1 REQUIREMENTS This SR demonstrates that one SDC loop is operating and circulating reactor coolant.

The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core.

This SR also demonstrates that the other SDC loop is OPERABLE.

In addition, during operation of the SOC loop with the water level in the vicinity of the reactor vessel nozzles, the SDC loop flow rate determination must also consider the SDC pump suction requirements.

The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the flow, temperature, pump control, and alarm indications available to the operator to monitor the SDC System in the Control Room.

Verification that the required loops are OPERABLE and in operation ensures that loops can be placed in operation as needed, to maintain decay heat and retain forced circulation.

The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is considered reasonable, since other administrative controls are available and have proven to be acceptable by operating experience.

SR 3.9.5.2 This SR demonstrates that the SDC loop is in operation and circulating reactor coolant.

The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core.

The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the flow, temperature, pump control, and alarm indications available for the operator in the Control Room for monitoring the SDC System.

SR 3.9.5.3 Verification that the required pump and valves are OPERABLE ensures that an additional SDC loop 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 available to the required pump and valves.

The Frequency of seven days CALVERT CLIFFS -

UNITS 1 & 2 B 3.9.5-4 Revision 2

ML023600485

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