Text

Proposed Tech Specs 3.0 & 4.0 & Corresponding Bases
	ML20127A361
Person / Time
Site:	Duane Arnold
Issue date:	12/31/1992
From:	; IES UTILITIES INC., (FORMERLY IOWA ELECTRIC LIGHT
To:	;
Shared Package
ML20127A339	List: ML20127A361; NG-92-5179, Request for TS Change RTS-249 to License DPR-49,adding TS 3.0 & 4.0,corresponding Bases & Modifying Existing Specs Consistent W/New Requirements Denoting Applicability of LCOs & Srs,Per STS & GLs 87-09 & 89-14;
References
	NUDOCS 9301110264
	Download: ML20127A361 (43)
	v • d • e

l DAEC-1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS PAGE NO.

1.0 Definitions 1.0-1 LIMITING SAFETY SAFETY ' LIMITS SYSTEM SETTING 1.1 Fuel Cladding Integrity 2.1 1.1 1.2- Reactor Coolant System Integrity 2.2 1.2-1 SURVEILLANCE LIMITING CONDITIONS FOR OPERATION REQUIREMENTS 3.0-1 l 3.0 Applicability 4.0 3.1 Reactor Protection System 4.1 3.1-1 3.2 Protective Instrumentation 4.2 3.2-1 A. Primary Containment Isolation Functions A 3.2-1 B. Core and Containment Cooling Systems' B 3.2-1 C. Control Rod Blvck Actuation C 3.2-2 D. Radiation Monitoring Systems D 3.2-2 E. Drywell Leak Detection E 3.2-3 F. Surveillance Information Readouts F 3.2-3 l G. Recirculation Pump Trips and Alternate

( Rod Insertion G 3.2-4 H. Accident Monitoring Instrumentation H 3.2-4 I. Explosive Gas 2:onitoring I 3.2-4A

Instrumentation 3.3 Reactivity Control 4.3 3.3-1 l A. Reactivity Limitations A 3.3 l B. Scram Discharge volume B 3.3-3 C, Reactivity Control Systema C: 3.3-4 D. Scram Insertion Times D~ 3'.3-5 E. Reactivity Anomalies' E 3.3-6 F. Recirculation Pumps F .3.3-6 l.

3.4- Standby Liquid Control System -4.4 3.4-1E A. Normal System Availability A 3.4-1 L B, Operation with Inoperable Components 3.4-2 C. Sodium Pentaborate Solution C 3.4-2

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LIMITING CONDITIONS FOR OPERATION REQUIREMENTS ' PAGE NO.

3.5 Core and Containment cooling Systems 4.5 3.5-1 A. Core Spray and LPCI Subsystems A 3.5-1 B. Containment Spray Cooling Capability B 3.5 C. Residual Heat Removal Service C 3.5 4 Water System D. HPCI-Subsystem D 3.5-6~

E. Reactor Core Isolation Cooling E 3.5-7 Subsystem-F. Automatic Depressurization System- F 3.5-9 G. Minimum Low Pressure Cooling G 3.5-10 and Diesel-Generator Availability

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H. Maintenance of Filled Discharge Pipe H 3.5-11

q. I. Engineered Safeguardo Compartments I 3.5-11 '

Cooling & Ventilation J. River Water Supply System J 3.5-12.

3.6 Priinary System Boundary 4.6 3.6-1 A. Thermal and' Pressurization A 3.6-1 Limitations B. Coolant Chemistry B 3.6-3 C. Coolant-Leakage- C 3.6-8 D. -Safety and Relief Valves D .3.6-9 E. Jet Pumps E 3.6 F. Jet Pump Flow Mismatch F .3.6-11 G. Structural Integrity G 3.6-11 H. Shock Suppressors (Snubbers) H '3.6-12 t

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DAEC-1 l

1.0 DEFINITIONS The succeeding frequently used terms are explicitly defined so that a uniform interpretation of the specifications may bn achieved.

1. SAFETY LIMIT The safety limits are limits below which the reasonable maintenance of the cladding and primary systems are assured. Exceeding such a limit requires unit shutdown and review by the Nuclear Regulatory Commission before resumption of unit operation.

Operation beyond such a limit may not in itself result in serious consequences but it indicates an operational deficiency subject to regulatory review.

2. LIMITING SAFETY SYSTEM SETTING (LSSS)

The limiting safety system settings are settings on instrumentation which initiate --

the automatic protective action at a level such that the safety limits will not be exceeded. These settings take into consideration the inntrumentation tolerances and the instruments are required to be periodically calibrated as specified in these Technical Specifications. The limiting saf ety system setting plus the tolerance of the instrument as given in the system design control document gives the limiting trip point for operation. This additional margin has been established so that with proper operation of the instrumentation the safety limits will never be exceeded. The inequality sign which may be given merely signifies the pref erred direction of operatior.al trip setting.

3. LIMITING CONDITIONS FOR OPERATION (LCO)

The limiting conditions specify the minimum acceptable levels of system performance necessary to assure safe startup and operation of the facility. When these conditions are met, the plant can be operated safely and abnormal situations can be safely controlled.

When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided: (1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system (s), subsystem (s), train (s), component (s) and devices (s) are OPERABLE, or likewise satisfy the requirements of this specification.

l 4. ACTION l ACTION shall be that part of a Specification which prescribes remedial measures l required under designated conditions.

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22. Instrumentation - Continued

h. Protective Function - A system protective action which results from the protective action of the channels monitoring a particular plant condition.

1. Simulated Automatic Actuation - Simulated automatic actuation means I applying a simulated signal to the sensor to actuate the circuit in question.

j. Logic - A logic is an arrangement of relays, contacts, and other components that produces a decision output.

3) Initiating - A logic that receives signals from channels and' produces decision outputs to the actuation logic.

2) Actuation - A logic that receives signals (either from initiating logic or channels) and produces decision outputs to accomplish a protective action.

k. Primary Source Signal - The first signal, which by plant design,' should initiate a reactor scram for the subject abnormal occurrence (see Updated-FSAR Chapters 7 and 15). l l

Source' Check - A Source check is the assessment of channel response when I 1.

the channel sensor is exposed to a source of radiation. d

.)

23. FUNCTIONAL TEST A functional test is the manual operation or initiation of a system, ,

subsystem, or component to verify that it functions within design tolerances (e.g., the manual start of a core spray pump to verify that it runs and that it pumps"the required volume of water).

24. SHUTDOWN ,

The reactor is in a shutdown condition when the reactor mode switch is in the shutdown mode position and no core alterations are being performed.

25.- ENGINEERED-SAFEGUARD An engineored safeguard is a safety system, the actions of which are '

essential to a safety action required in response toiaccidents.

l 26. DELETED ,

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27. FREQUENCY NOTATION j l The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined below. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda-shall be applicable as follows in these Technical -

Specifications.

NOTATION FREQUENCY ;

S (Shiftly) At least once per 12 houre.

D (Daily) At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

W (Weekly) At least once per 7 days.

M (Monthly) At least once per 31 days, Q (Quarterly or every 3 months) At least once per 92 dayo.

SA (Semi-annually or every 6 months) At least once per 184 days.

A (Yearly or Annually) At least once per 366 days.

R (Refuel) .

At least-once per 18 months.

S/U (Startup) Prior to each reactor-startup.

P Prior to each release.

NA Not applicable.

28. FIRE SUPPRESSION WATER SYSTEMS A fire suppression water system shall consist of a water source, pumps, and distribution piping with associated sectionalizing control or isolation valves. Such valves include yard hydrant curb valves, the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe-or deluge system riser.

29. REACTOR TRIP SYSTEM RESPONSE TIME Reactor trip system response time is the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until deenergization of the scram pilot valve solenoids. ,

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30. REPORTABLE EVENT A REPORTABLE EVENT shall be any of those conditions specified in Section '

'50.73 to 10 CFR Part 50.

31. OFFSITE DOSE ASSESSMENT MANUAL ,

The Offsite Dose As?essment Manual (ODAM) contains the methodology and. parameters to be' used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints,-and in the conduct of the Radiological Environmental Monitoring Program.

The.ODAM shall also contain (1) the Radioactive Effluent Controls and Radiological p Environmental Monitoring Program' required by Section 6.9.4 and-(2) descriptions.of-the information that should be included in the Semiannual-Radioactive Material Release Report and Annual Radiological Environmental Report required by the Technical.

Specification 6.11.1.

32. Deleted

33. PURGE - PURGING J PURGE or PURGING is the controlled process of discharging air or' gas from a confinement to maintain. temperature,. pressure, humidity, concentration or

-other operating condition, in such a manner that replacement air or gas is, required to purify the confinement.

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LIMITING CONDITIONS FOR OPERATION SURVE!LLANCE REQUIREMENTS

-i l f. 0 Applicability l 4.0 Applicability P

i A. Compliance with the LCO's A. Surveillance Requirements shall be contained in the succeeding met during the conditions ,

specified for individual LCO's ;

Specifications is required during the conditions-specified therein; unless otherwise stated.

except that upon failure to meet ;

the LCO's, the associated ACTION requirements shall be met. ,

Each Surveillance Requirement 4 D. Noncompliance with a B.

Specification shall exist when shall be performed within the the requirements of the LCO and specified time interval with a associated ACTION requirements maximum allowable extension not to ,

are not met within the specified exceed 25% of the surveillance ,

time intervals. If the Lc0 is interval.

restored prior to expiration of the specified time intervals, completion of the ACTION ~i j requirements is not required.

C. When an LCO is-not met and C. Failure to perform a Surveillance associated ACTION requirements Requirement within the allowed are not met or an associated surveillance interval, defined by-ACTION-is not provided, within Specification 4.0.B, shall'. !

one hour action shall be constitute noncompliance with the ;

initiated to place the unit in a OPERABILITY requirements for an condition in which the LCO. The time limits of the .

Specification does not' apply by ACTION requirements are applicable placing it, as applicable, ins at the time it is identified that a Surveillance Requirement has notL 2i

1. at least STARTUP within the been performed.- The ACTION next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , requirements may be delayed for-up-to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit the

2. at least HOT FHUTDOWN completion of the surveillance .

within the following 6 when-the allowable outage time '

hours, and limits.of.the. ACTION requirements are lesa than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ._

3. at.least COLD SHUTDOWN Surveillance Requirements do not-within the subsequent 24 have to be performed on inoperable hours, equipment.

Where corrective measures are

completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by Specification P 3.0.C is not' required. . _ _ .

t Exceptions-to this specification are statedLin the individual r specifications.

This specification is not applicable in COLD SHUTDOWN or ,

the REFUEL mode, 1

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DAEC-1 LIMITING CONDITIONS FOR OPERATION ,,

SURVEILLANCE REQUIREMENTS D. When an Lc0 is not met, entry D. Entry into a condition shall into a specified condition shall- not be made unless the not be made except when the Surveillance Requirement (s) associated ACTIONS to be entered associated with the LCO have permit continued operation in the been performed within the specified condition for an allowed'eurveillance unlimited period of time. This interval, defined by provision shall not prevent Specification 4.0.D, or as passage through or to conditions otherwise specified. This as required to comply with ACTION provision shall not prevent-requirements. Exceptions to this passage through or to specification are stated in the conditions as required to individual specifications, comply with ACTION requirements.

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3 3.0 and 4.0 BASES: APPLICABILITY r

specifications 3.0.A through 3.0.D establish the general requirements applicable to LCo's. These requTrements are based on the requirements for LCO's stated in the Code of Federal Regulations, 10 CFR 50.36(c)(2):

l " Limiting conditions for operation are the lowest functional capability l or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specification until the condition can be met."

l Specification 3.0.A establishes the applicability within each individual l specification as the requirement for when conf 0rmance to the LCO is required l for safe operation of the facility. The ACTION requirements establish those remedial measures that must be taken within specified time limits when the requirements of an LCO are not uet. It is not intended that the shutdoan ACTION requirements be used as an operational convenience which permits (routine) voluntary removal of a system (s) or component (s) from service in lieu of other alternatives that would not result in redundant systems or components being inoperable.

There are two basic types of ACTION requirements. The first specifies the remedial measures that permit continued operation of the facility which is not further restricted by the time limits of the ACTION requirements. In this case, conformance to the ACTION requirements provides an acceptable level of safety for unlimited continued operation as long as the ACTION requirements

, continue to be met. The second type of ACTION requirement specifies a time limit in which conformance to the conditions of the LCO must be met. This time limit is the allowable outage time to restore an inoperable system or component to OPERABLE status or for restoring parameters within specified limits. If these actions are not completed within the allowable outage time limits, a shutdown is required to place the facility in an operational condition or other specified condition in which the specification no longer applies.

l The specified time limits of the ACTION requirements are applicable from the j point in time it is identified that an LCO is not met. The time limits of the i

ACTION requirements are also applicable when a system or component is removed from service for surveillance testing or investigation of operational .

problems. Individual specifications may include a specified time limit for _

the completion of a Surveillance Requirement when equipment is removed from service. in this case, the allowable outage time limits of the ACTION requirements are applicable when this limit expires if the surveillance has not been completed. When a shutdown is required to comply with ACTION requirements, the plant may have entered an operational condition in which a new specification becomes applicable. In this case, the time limits of the ACTION requirements would apply from the point in time that the new specification becomes applicable if the requirements of the LCO are not met.

Specification 3.0.B establishes that noncompliance with a specification exists when the requirements of the LCO are not met and the associated ACTION requirements have not been implemented within the specified time interval.

The purpose of this specification is to clarify that (1) implementation of the ACTION requirements within the specified time interval constitutes compliance with a specification and (2) completion of the remedial measures of the ACTION requirements is not required when compliance with an LCO is restored within l the time interval specified in the associated ACTION requirements.

Specification 3.0.C establishes the shutdown ACTION requirements that must be implemented when an LCO is not met and the condition is not specifically addressed by the associated ACTION requiremen*s. The purpose of this specification is to delineate the time limits for placing the unit in a safe shutdown condition when plant operation cannot be maintained within the limits for safe operation defined by the LCO and its ACTION requirements. It is not l RTS-249 3.0-3 12/92

r DAEC-1 l intended to be used as an operational convenience which permits (routine) voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant systems or components being inoperable. One hour 11s allowed to prepare for an orderly shutdown ,

before' initiating a change in plant operation. This time permits the operator to coordinate the reduction in electrical generation with the load dispatcher to ensure the stability and availability of the electrical grid. The time limits specified to reach lower conditions of operation penmit the shutdown to proceed in a controlled and orderly manner that is well within the specified-maximum cooldown rate and within the cooldown capabilities of the facility assuming only the minimum required equipment is OPERABLE. This reduces thermal stresses on components of the Primary Coolant System and the potential for a plant upset that could challenge safety systems under conditions'for which this specification applies.

If remedial measures permitting limited continued operation of the facility-under the provisions of the~ ACTION requirements are completed, the shutdown may be terminated. The time limits of the ACTICN requirements are applicable from the point in time there was a failure to meet an LCO. Therefore,-the shutdown may be terminated if the ACTION requirements have been met or the time limits of the ACTION requirements have not expired, thus providing an-allowance for the completion of the required actions. .

The time limits of Specification 3.0.C allow 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months for the plant.to be in '

COLD SHUTDOWN when a chutdown is required during REACTOR POWER OPERATION. - If the plant is in a lower condition of operation when a shutdown is-required, the time limit for reaching the next lower condition of operation applies.

However, if a lower condition of operation is reached in less time than allowed, the total allowable time to reach COLD SHUTDOWN, or other. operational condition, is not reduced. For example, if STARTUP is reached in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the time allowed to reach HOT SHUTDOWN is the next 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months because the total, time-to reach HOT SHUTDOWN is not reduced from the allowable limit of 13. hours.

Therefore, if remedial measures are completed that would permit a return to REACTOR POWER OPERATION, a penalty is not incurred by having to. reach a lower condition of operation in less than the total time allowed.

The same principle applies with regard to the allowable outage time limits of the ACTION requirements, if compliance with the ACTION requirements for one specification results in entry into an operational condition or condition of operation for another specification in which the. requirements:of the LOO are not met. If the. new specification becomesl applicable in less time than specified, the difference may be added to the allowable outage time limits of the second specification. However, the allowable outage time limitsiof ACTION requirements for a higher condition of operation may not be used to extend the allowable outage-time that is applicable when an LCO is not met in a lower

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condition of operation.

The shutdown requirements of Specification 3.0.C do not apply in COLD SHUTDOWN or the REFUEL mode because'the ACTION requirements of individual specifications. define the remedial measures to.be taken.

Specification 3.0.D establishes.limitatione on a change in operational- ~

. conditions when an LCO is not met. It precludes placing the facility.in a higher condition of operation. when the requirements for an -LCO_ are not . met and ~

continued noncompliance to these conditions would' result in a shutdown;to comply with the ACTION requirements if a change in conditions were permitted.

The purpose of this specification is to ensure'that facility operation is:not

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initiated or that higher conditions of-operation are not entered when corrective action is-being taken to obtain compliance with a specification by.-

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. restoring equipment to OPERABLE status or parameters to'specified limits.

Compliance with ACTIONLrequirements that' permit' continued operation of the facility for an unlimited period of time provides an acceptable. level'of safety for continued operation without regard to the status of the. plant:

before or after a change in operational conditions. Therefore, in this case, l

entry into a condition may be made. in accordance with the provisions of the ACTION requirements. The provisions of-this specification should not, V - -

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DAEC-1 l however, be interpreted as endorsing the failure to exercise good practice in i restoring systems or components to OPERABLE status before plant startup.

i When a shutdown is required to comply with ACTION requirements, the provisions of specification 3.0.D do not apply tecause they would delay placing the facility in a lower condition of operation.

Specifications 4.0.A through 4.0.D establish the general requirements applicable to Surveillance Requirements. These requiremence are based on the Surveillance Requirements stated in the Code of Federal Regulations, 10 CFR 50.36(c)(3):

" Surveillance requirements are requirements relating to test, calibration, or inspection to ensure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions of operation will be met."

Specification 4.0.A establishes the requirement that surveillances must be l performed during the conditions for which the requirements of the LCO apply ;

unless otherwise stated. The purpose of this specification is to ensure that surveillances are performed to verify the operational status of systems and l components and that parameters are within specified limits to ensure safe 1 operation of the facility when the plant is in a condition for which the l individual LCO is applicable. Surveillance Requirements do not have to be l performed when the facility is in a condition for which the requirements of l the associated LCO do not apply unless otherwise specified.

1 Specification 4.0.B establishes the limit for which the specified time Interval for Surveillance Requirements may be extended. It permits an l allowable extension of the normal surveillance interval to facilitate surveillance scheduling and consideration of plant operating cor.ditions that may not be suitable for conducting the surveillance; e.g., transient

, conditions or other ongoing surveillance or maintenance activities. It also l provides flexibility to accommodate the length of a fuel cycle for surveillances that are performed at each refueling outage and are specified with an 18-month surveillance interval. It is not intended that this provision be used repeatedly as a convenience to extend surveillance intervals beyond that specified for surveillances that are not performed during refueling outages. The Ibnitation of Specification 4.0.B is based on engineering judgement and the recognition that the most probable result of any particular surveillance being performed is the verification of conformance _

with the Surveillance Requirements. This prot ision is suf ficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval.

Specification 4.0.C establishes the failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by the provisions of Specification 4.0.B, as a condition that constitutes a failure to meet the OPERABILITY requirements for an LCO. Under the provisions of this specification, systems and components are assumed to be OPERABLE when Surveillance Requirements have been satisfactorily performed within the specified time interval. However, nothing in this provision is to be construed as implying that systems or components are OPERABLE when they are l found or known to be inoperable although still meeting the Sarveillance Requirements. This specification also clarifies that the ACTION requirements are applicable when Surveillance Requirements have not been completed within the allowed surveillance interval and that the time limits of the ACTION requirements apply from the point in time it is identified that a surveillance has not been performed and not at the time that the allowed surveillance interval was exceeded. Completion of the Surveillance Requirement within the allowable outage time limits of the ACTION requirements restores compliance with the requirements of Specification 4.0.C. However, this does not negate the fact that the failure to have performed the surveillance within the allowed surveillance interval, defined by the provisions of Specification l RTS-249 3.0-5 12/92

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DAEC-1 4.0.B, was a violation of the OPERABILITY requirements of an LCO that is subject to enforcement action. Further, the failure to perform a surveillance within the provisions of Specification 4.0.B constitutes a failure to meet the OPERABILITY requirements for an LCO and any reports required by 10 CFR 50.73 shall be determined based on the length of-time the surveillance interval has been exceeded, and the corresponding LCO ACTION time requirements.

l If the allowable outage time limits of the ACTION-requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or a shutdown is required to comply with ACTION requirements, e.g.,

Specifi. cation 3.0.C, a 24-hour allowance is provided to permit a delay in implementing the ACTION requirements. This provides an adequate time limit to complete Surveillance Requirements that have not been performed. - The purpose of this allowance is to permit the completion of a surveillance before a shutdown would be required to comply with ACTION-requiremente or before other remedial measures would be required that may preclude the completion of a surveillance. The basis for this allowance includes consideration for plant conditions, adequate planning, availability of personnel, the- time ' required to ,

perform the surveillance, and the safety significance of the delay in completing the required surveillance. This provision also provides a time limit for the completion of Surveillance Requirements that become applicable as a consequence of condition changes imposed by ACTION requiremento and-for completing Surveillance Requirements that are applicable when an exception to

, the requirements of Specification 4.0.D is allowed. If a surveillance is not j completed within the 24-hour allowance, the time limits of the ACTION requirements are applicable at that time. When a surveillance is performed within che 24-hour allowance and the Surveillance Requirements are not met, l the time limits of the ACTION requirements are applicable at the time that the I surveillance is terminated.

1 Surveillance Requirements do not have to be performed on inoperable equipment' because the ACTION requirements define the-remedial measures that apply..

However, the Surveillance Requirements have to be met to demonstrate that I inoperable equipment has been restored to OPERABLE statas.

Specification 4.0.D establishes the requirement that all applicable surveillances must be. met.before entry-into a specified condition. The purpose of this specification is to ensure that system and ' component OPERABILITY requirements or parameter limits are met before entry into a condition for which these systems and components ensure safe operation of the facility. This provision applies to changes-in conditions associated with plant shutdown as well as startup. i Under the provisions of this specification, the applicable Surveillance Requirements must be performed within the specified surveillance interval, defined by the provisions of Specification 4.0.B, to assume that the LCO's are met during initial plant startup.or following a plant outage.

When a shutdown is required to' comply with' ACTION requirements, the provisions of Specification 4.0.D do not apply because this would delay placing the facility in a lower condition of operation.

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I l DAEC-1 NOTES FOR TABLE 3.1-1

1. There shall be two cperable or tripped trip systems for each function.

If the minimum number of operable sensor or instrument channels for a trip system cannot be met, the affected trip system shall bc pinced in the safe (tripped) condition, or the appropriate actions listed below I shall be taken. If the effected trip system is placed in the safe l (tripped) condition, the provisions of Specification 3.0.D are not I applicable.

a. Initiate insertion of operable rods and corrplete insertion of all operable rods within four hours.

b. Reduce power level to IRM range and place mode switch in the startup position within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ,

c. Reduce turbine load and close main steam line isolation valves within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

d. Reduce power to less than 30% of rated.

2. Permissible to bypass, in refuel and shutdown positions of the reactor mode switch.

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y -TABLE 4.1-1

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'f. REACTOR PROTECTION SYSTEM (SCRAM)-INSTRUMENT FUNCTIONAL TESTS

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MINIMUM' FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTRUMENT AND CONTROL CIRCUITS ,

i Minimum Frequency (3)

! Group (2) Functional Test j Mode Switch in Shutdown' A Place Mode Switch in once/ operating cycle ,

Shutdown l i

- Manual Scram A Trip Channel and Alarm Every 3 months a

i RPS Channel Test Switch' A . Trip Channel and Alarm once/ operating cycle or after channel

, maintenance 1

IRM C Trip Channel and Alarm (4) Once'per week during refueling or startup

High Flux and before each startup unless a j satisfactory test has been accomplished j j. during the. preceding 7 days. (6) ,

I , Inoperative- C Trip Channel and Alarm (4) Once'per week during refueling or startup

?*. and before each startup unless a satisfactory test has been accomplished

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'm ] during the preceding 7 days. (6)

APRM

. High Flux in Run B Trip Output Relays (4) Once/ week (While in.Run Mode)

j. Inoperative B . Trip Output Relays (4) .Once/ Week t

Downscale.* B Trip Output'. Relays (4) Once/ month (1) ,

' Flow Bias B Trip Output Relays (4) Once/ month (1)

^High Flux in Startupf C Trip Output Relays Once per' week during refueling or startup or Refuel and before each startup'unless a .

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satisfactory. test has been accomplished l

during the preceding 7 days (6)

Hich Reactor Reactor.. .A ' Trip Channel Alarm Every 1 month-(l)

-Pressure ,

With companion IRM Hi-Hi or Inoperable.

l

'.u i 3

g. '

m l'.

I

g. .

L

DAEC-1 Al LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS D. Safety and Relief Valves D. Safety and Relief Valves

1. When in RUN, STARTUP, or HOT 1. Once per OPERATING CYCLE, at least SHUTDOWN MODE, both safety valves one safety valve and 3 relief and the safety modes of all valves shall be removed, set "

relief valves

shall be OPERABLE, pressure tested and reinstalled or except as specified in replaced with spares that have Specification 3.6.D.2. been previously set pressure tested. The safety and relief valves shall be rotated, at least once per 40 months, such that both safety and 6 relief valves are removed, set pressure tested and reinstalled or replaced with spares. Any spare that is installed must have been set ,

1 pressure tested within the ,

previous 40 months.

The setpoint of the safety valves shall be as specified in Specification 2.2.

2.a With the safety valve function of .. At least one of the relief valves one relief valve inoperable, shall be disassembled and restore the inoperable safety inspected once per OPERATING ,

valve function to OPERABLE status CYCLE.

within thirty days,

b. With the safety valve function of two relief valves inoperable, restore the inoperable safety valve function to OPERABLE status within seven days.

3. If Specification 3.6.D.1 cr l 3.a With the reactor pressure 3.6.D.2 is not met, be in at > 100 psig and turbine bypass flow [

least HOT SHUTDOWN within the to the main condenser, each relief L next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD valve shall be manually opened and SHUTDOWN within the following 24 verified open by turbine bypass hours. valve position decrease, pressure switches and thermocouple readings downstream of the relief valve to indicate steam flow from the valve once per OPERATING CYCLE. The provisions of Specification 4.0.D are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam L pressure is adequate to perform the test.

b. If OPERABILITY is not successfully demonetrated within the 12-hour

SRVs which perform an ADS function period, reduce reactor steam dome i must also satisfy the OPERABILITY pressure to less than 100 psig requirements of Specification 3.5.F, within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Core and Containment Cooling Systems.

4. The relief valve setpoints for the Low-Low Set function shall be as

(

specified in Section 2.2.1.c.

Instrumentation and system logic shall be. functionally tested, calibrated, and checked as specified in Table 4.2-B.

f RTS-249 3.6-9 12/92 s

Et .

_ _ . . , - _ _ . . = . . . . . . . . - . _ .. . . . .. . . _ . . , _ . . . . .. >>

' ' '~ )

DAEC-1

- _ . . . ,. -1

5. The water level in the reactor _ vessel willibe perturbed and the--

corresponding level indicator changes will be monitored. -This perturbation test will be performed ~every month after completion of'the functional-test program.

- 6. During plant shutdowns the-provisions of Specification 4.0.D are not-applicable provided the surveillances are performed within-12 hours-after entering HOT STANDBY CONDITION or actions-are-being_taken to ;

proceed to HOT SHUTDOWN.

I 4

j 4

1 1

l 1

~l l

i l

.I l

l I 4

i l

l

?, .

i 4

i: . .i l

-RTS-249 3.1-11 12/92?

l , 1

l

, - . TIL .-,4 _ _ ._~ .....a .. .. .. - . . - - - . . _ _ _ _ ..r____ __z_.______

_c__.___2

DAEC-1 l

i 3.1 BASES The reactor protection system automatically initi?.tes a reactor scram tot

1. Preserve the integrity of the fuel cladding.

2. Precarve the integrity of the reactor coolant system.

3. Minimize the energy which mus' be absorbed following a loss-of-coolant _.

I accident, and prevent inadvertent criticality.

This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be oat of service because of maintenance. When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

l The exception to Specification 3.0.D denot: u in Table 3.1-1 clarifien that [

l mode changes may be made when instrument channel (s) for a trip system are in a i tripped (safe) condition.

6 The reactor protection system is of the dual channel type (Reference Subsection 7.2 of the Updated FSAR). The system is made up of two independent trip systems, each having three subchannels of tripping devices. One of the three subchannels has inputs l'

RTS-249 3.1-15 12/92

N DAEC J l-(QMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS c ')' , At.least onetpre-treatment steam air ejector offgas system radiation monitor shall be operable during' reactor power-operation.

The monitors shall be set to initiate an alarm if the <

monitor exceeds a~ trip setting equivalent to 1.0 Ci/aec.of noble gases after 30 minutes delay in the

, offgas holdup line.

o

'In the event the noble gas

flow in the air ejector offgas exceeds the equivalent of 1.0 ci/sec after 30 minutes delay in the offgas holdup line, restore the rate to less-than thic limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in-at least hot standby within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

d) In the event no pre-treatment monitor is

operable, gases.from the-steam air ejector-offgas system may be released for up to 30 days provided (1) .,

the charcoal bed of the offgas system-is not r

bypassed, (2) Grab samples i are collected and analyzed weekly, and=(3) the offgas 4 stack noble gas activity monitor is operable, or at least-1 post-treatment monitor is operable.

Otherwise, be-in at least ;

HOT' STANDBY within the

, following 24' hours.

e) The provisions of.

4 Specification 3.0.D.are not applicable.

2. -Reactor Building-Isolation and '2.

Reactor Building Isolation and Standby Gas Traatment System Standby Gas Treatment System The limiting conditions for Inctrumentation shall txa operation are given in' functionally tested, calibrated Spe'ification 3.7.B. ' and checked as indicated.in Table.

4.2-D.

System logic shall be functionally. ,

tested as: indicated in Table 4.2-D.

p IRTS-249 3.2-3 12/92 t

7 yt c--y,-5 y .,..[m J-, % , c . _ , , , -. . . - , - .- . . - -

. -. - . . - . ~ . . . . - . - - .- . . . . . . . - - - _ . _

j

.DAEC-1

-NOTES FOR TABLE 3.2-A i

1. Whenever Primary containment: integrity is required by. Subsection 3.7, there shall be two operable or tripped systems for each function.

2. If the first column cannot be met for one of the trip' systems, that trip system shall be tripped or the appropriate action listed below shall .tx3_

l l taken. If the affected trip system.is placed in the safe (tripped) J I condition, the provisions of Specification 3.0.D are not applicable.

ACTION A - Be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION B - De in at least STARTUP with the-associated isolation valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION C - Close the affected system isolation valves within one' hour and declare the affected system inoperable.

. ACTION D - De in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION E - Isolate secondary containment'and start the standby gas

~

treatment system.

3. Zero referenced to top of active fuel.*

Top of the active fuel zone is defined to be 344.5 inches above-vessel ,

zero (see Bases 3.2).

RTS-249 3.2-6 12/92.

. - . . _ _ _ _ _ _ . _ . ~ . . ~ ...c

- ~ . . - . . .

t DAEC-1 NOTES.FOR TABLE 3.2-B

1. Whanever any CSCS subsystem is required by Subsection 3.5 to be operable, there shall be two operable trip systems. If the first column cannot be met for one of-the trip systems, that trip system shall be placed in the tripped condition or the reactor shall be placed in the Cold shutdown Condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If=the affected trip system is l

l placed in the safe (tripped) condition, the provisions of Specification l 3.0.D are not applicable.

2. Close isolation valves in RCIC subsystem.

3. Close isolation valves in HPCI subsystem.

4. Zero referenced to top of active fuel.*

5. HPCI has only one trip system for these sensors.

6. There is no trip function associated with these relays. The relays provide signals to annunciators only.

7.- Four undervoltage relays.with integral' timers per 4KV bus. -The relay output contacts are connected to' form a one-out-cf-two-twice coincident logic matrix. With one 7:elay- inoperable, operation may proceed provided -

that the inoperable relay is piaced in the tripped condition within one hour.

Top of active fuel zone is defined to be 344.5 inches above vessel zero

l. (see Bases 3.2).

l-RTS-249' 3.2-15 ~ 12/92-

n. . . . . --

z.:

TABLE 3.2-D c;

. RADIATION MONITORING SYSTEMS THAT INITIATE AND/OR ISOLATE SYSTEMS i

'55 Number of valve >

os

,83 Minimum'No. Instrument Groups.

" Channels Operated' i

- .of Operable-

'# Provided by by ' Action d Instrument

'l Channels Trip Function Trip' Level Setting Design Signal. --- (1) (4)' ,

~

3 A or:B' 'I 1 Refuel Area Exhaust Monitor: Upscale, < 9 mr/hr 2 Inst. Channels 1 Reactor Building Area - Upscale, < 11 mr/hr ' 2 '. Inst. Channels' 'JL - B-Exhaust Monitors offgas Radiation Monitors Note 2 2 Inst. ' Note 2 C .

1

2. LMain' Steam Line '<3x Normal Full 4 Inst. Channels- . Note'3 Da Radiation Monitor Power Background

-i

~

NOTES'FOR TABLE-3.2-D us.

^.

'd :1. ' Action' I

w

A. Cease operation of the ,

refueling equipment ~.

B5 Isolate secondary containment' and start the standby gas' treatment system.

C. Refer to. Subsection 3.2.D.1.

. D.

Refer to Specificaticn 3.7.F. )

~2. For trip setting and valves

- isolated,.see Specification

-3.2.D.1.a.

I 3.' Trips Mechanical. Vacuum Pump which results in afsubsequent isolation of the Mechanical' Vacuum Pump suction ,

valves.

y: -

s . 4. - The provisions'of' Specification j@' - 3.0.D arefnot1 applicable.

- - P y m- g W f- Mu# y- gA$i-- .g T p .y"/tgg y-.-s*h 3- +

n--____-____-_. _ _ _ _ _ . _

j DAEC-1 NOTES FOR TABLE 3.2-F l

1. From and after the date that one of these parameters is reduced to one indication, when required, continued operation is permissible during the succeeding thirty days unless such instrumentation is sooner made l operable. The provisions of Specification 3.0.D are not applicable.

2. From and after the date that one of these parameters is not indicated in the control room, continued operation is permissible during the succeeding seven days unless such instrumentation is sooner made operable.

3. If the requirements of notes (1) and (2) cannot be met, an orderly shutdown shall be initiated and the reactor shall be in a Cold Condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4. These curveillance inctruments are considered redundant to each other.

RTS-249 3.2-22 12/92

= .

_ _ . . _ _ . - . _ _ . . _. _ - _ .__ .-_ _ _ _ . . ___m.-. ._ - . . _ . _ .

i' l DAEC-1

' timer is set to annunciate _before the values specified in Specification-3.6.c are exceeded ~ An air sampling system is also provided, as.a backup.to the

~

sump system, to detect leakage inside the primary containment.

For each parameter monitored,.as listed in Table 3.2 F, there are two (2) channels of instrumentation. By comparing readings between the two (2) channels, a near continuous surveillance of instrument performance is available. Any deviation in readings will initiate an early recalibration, thereby maintaining;the quality of the instrument 1 readings. 3

} The exception-to' Specification 3.0.D denoted in Tables 3.2-A, 3. 2-B '- 3.2-D and ,

o l 3.2-F clarifies that mode changes may be made when instrument channel (s) f or a j trip system are in a tripped (safe) condition.

on July 26, 1984 the NRC published their' final-rule on Anticipated Transients Without Scram (ATWS), (10 CFR 550.62). This rule requires all BWR's to make certain plant modifications to mitigate the consequences of the unlikely occurrence of a failure to scram during an anticipated operational transient.

The bases for these modifications are described in NEDE-31096-P-A,

" Anticipated Transients Without Scram; Response to NRC ATWS Rule, 10 CFR 50.62," December,.1985. The Standby Liquid Control System (SLCS) was l-I modified for two-pump operation to provide the minimum required.flowrate and.

boron concentration required by the ATWS rule (see section 3.4 Bases). The existing ATWS Recirculation Pump Trip (RPT) was modified from a one-out-of-

~

l Ltwo-once logic to trip each recire pump to'a two-out-of-two-once logic to trip both recirc. pumps, ("Monticello" design). This logic will also~1nitiate the Alternate Rod Insertion (ARI). system, which actuates solenoid valves:that bleed the air off the scram air header, causing the control rodsito insert.

The instrument setpoints are_ chosen such that the normal reactor protection system (RPS) scram setpoints for reactor high pressure or low water ~ level will~

be exceeded before the ATWS RPT/ARI setpointo are reached. Because'ATWS is

~

(- considered a very low probability event and.is outside-the normal' design basis-for the DAEC, the surveillance frequencies and LCO requirements are less stringent than for safety-related instrumentation.

(

i L ,The-End-of-Cycle (EOC) recirculation pump trip'was added to the plant to improve the operating margin to fuel thermal limits, in'particular Minimum RTS-249 3.2-45: 12/92-l _. - . _ . _ _ .

. .- - ... -. - . -. . . ---n~ . - ,. .-

Si *L DAEC-l' Critical' Power' Ratio (MCPR). The EOC-RPT trips the recire. pumpsito i lessen the severity of the power' increases. caused by either a closure of turbine stop-valves or fast. closure of the turbine control' valves with reactor power ,

greater than 30% and a. simultaneous failure'of the. turbine bypass valves to' open. -The operating limit MCPR of section 3.12.C is calculated assuming an-operable EOC-RPT systam. If the requiremerls of Table-3.2-G are not met, then _;

the reactor power level is reduced to a level (85% of' rated) which will' ensure that the full-power MCPR limits of section 3.12-C will not be violated if such a transient were to occur.

Trip function settings'are included for Instrument AC and Uninterruptible AC and battery buses for surveillance of undervoltage relays. The._undervoltage relays are required to sense a reduction in the power source voltage so that the subject inctruments can be transferred to an alternate power source.

l Surveillance tests-other than a monthly functional check of the bus power monitors for the RHR, Core Spray, ADS, and HPCI and RCIC trip systems are not.

required since they serve as annunciators for complete' loss of powerfand do

'not monitor reduction of voltage. The subject functional check consists of opening the appropriate circuit breakers or removing the appropriate fuses.and' observing the loss of power annunciator activation.

The accident monitoring instrumentation listed in Table 3.2-H were -

specifically added to-comply with the requirements of'NUREG-0737 and Generic- -

Letto 83-36. The instrumentation listed is. designed to provide l plant status for accidents that exceed the' design basis accidents discussed in' Chapter 15 of the DAEC UFSAR.

'Tootnote 9 of Table 3.2-H deviates' f rom th'e. guidance of Generic Letter 83-36'-

as continued operation for 30 days (instead of 7 -days as recommended in :the generic letter) is allowed with one of two torus water level monitor (TWLM)-

channels inoperabla. . Continued operation 11s justifiedLby the:following considerations:-

1). Redundancy is available in that at least one channel of the-containment water level monitor (CWLM) instrumentation must be available. Since the-CWLM envelopes.the span measured.by the TWLM, the torus water. level can

-be monitored by;the CWLM system.

~

g -l RTS-24'9 3.2-45a- 12/92 1

. , ~ , - - . - - . . - - - . . . , , ,

(:

~ DAEC-1

-[ILIMITING'CONDITIONSFOROPERATION- _

-SURVEILLANCE REQUIREMENTS

f. A' control rod which is not f. Whenever the. reactor is operating-moveable with drive or scram greater than 20% powers pressure 3'etuck) shall be declared inoperable and.the-

-following actions shall:be taken.

(1) Disarm the associated control rod (i) each~ partially or-fully withdrawn-drive and operable control rod'shall be-demonstrated to be moveable by-exercising it one notch'at least once per week.

(ii)-verify compliance with (ii) if a control rod cannot be moved Specification 3.3.A.1. with drive or scram pressure, each partially or fully withdrawn OPERABLE ~ .

(iii) Whenever the reactor is less than control rodLshall be exercised one notch .

20% power, verify all inoperable at least once each 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,Lunless it-control rods not'in compliance with has been-determined that the failure is-BPWS are separated by 2 or more not a failed control rod drive mechanism-OPERABLE control rods in any. direction, collet' housing. .j including the diagonal.

(iv) within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , verify that the cause of the failure is not due to a failed control-rod drive mechanism collet housing.

(v) if.the requirements of "

Specification 3.3.A.2.f (1)-(iv)_cannot be met or more than one c;ntrol rod is stuck, be in COLD SHUTDOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

g. The provisions of Specification 3.0.D are not applicable.

3. Control Rod Drive Housing-Support 3. Control Rod Drive Housing Support The control rod drive housing :The control rod drive housing.

support. system'shall be in place- suppert system shall be inspected.

whenever the reactor. vessel is after reassembly and the results

. pressurized above atmospheric ofLthe inspection recorded.

pressure with fuel in the reactor vessel.

B. Scram Discharge Volume B. . Scram Discharge Volume-(Not Used) 1. At least once per. month, verify

~

the SDV vent and drain valves are

" open.

I 2. At least onceiper quarter verify;;

that

a. .The SDV vent and drain valves close within 30 seconds after.

receipt of L

l l, RTS-249 3.3-3 12/92 i o 4 -l' E

hS DAEC -[-LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

e. At reactor' Once/ operating ~

. pressure.of cyclo 150 +/- 10 psig demonstrate-ability to >

deliver rated-flow at.a-discharge

-pressure greater.

than or. equal to that. pressure ,

required.to accomplish vessel injection.--

2. With HPCI inoperable, provided The HPCI pump that both Core Spray subsystems, shall deliver LPCI, ADS, and RCIC are verified at:least 3000.

to be OPERABLE, restore HPCI to gpm for a OPERABLE status within 14 days, system head or be in at least HOT SHUTDOWN corresponding.

within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and' to a reactor s

reduce' reactor steam dome .

pressure of pressure to less than or equal to 1040 to 150 psig.

150 psig within-the following 24-hours. f. Verify that- Once/ operating =

the' suction for Cycle-the-HPCI system is automatically -

transferred.from the condensate ctorage tank to the suppression _ pool on a condensate storage tank water level-low

-signal-and on a suppression pool' water level-high signal.

4

-2.a .The provisions of Specification-4.0.D are not applicable provided-the surveillances are-performed

'within 12' hours after reactor-I e steam pressure.is. adequate tof s l perform these tests. >

j b. If OPERABILITY is ~not- successfukly -

demonstrated within the-12-hour.

period, reduce reactornsteam dome-

. pressure to less-than_150 psig within the following,72nhours.'

E. Reactor Core Isolation Cooling E. -Reactor Core Isolation Cooling-

-(RCIC) Subsystem- .(RCIC) Subsystem

1. The RCIC Subsystem shall be 1. RCIC Subsystem testing shall-be OPERABLE whenever'therecis' ' performed as follows::

irradiated fuel'in the reactor

, vessel, the reactor pressure is Item- Frequency:

- Jgreater than 150 psig,.and prior . . .

Annual to' reactor startup from a COLD a. -Simulated

' CONDITION.except as specified in

- Automatic 3.5.E.2 below. Actuation Test-(and restart). a i

RTS-249 3.5-7 12/92, 1

I L :O

~

f" '

DAEC SURVEILLANCE REQUIREMENTS'

- l .: LIMITING CONDITIONS' FOR' OPERATION ~

Item- = Frequency

b. l Pump. .

Once/3 month'a- t'

' Operability-

c. ~ Motor Operated Once/3' months Valve Operability

d. .At rated reactor. Once/3 months pressure-demonstrate ability to deliver rated flow at a discharge

-pressure greater.

than or equal-to that-pressure required to:

accomplish vessel injection if' vessel pressure were as'high as 1040 psig.- q

e. At reactor once/ operating; pressure.of cycle

. 150't 10 psig

-~ demonstrate;

ability to.

deliver rated flow at a discharge pressure greater-

' than-or equal-to that: pressure re-quired to accomplish

' vessel injection.

J

.The RCIC pump shall1 deliver at least--

400 gpm for.a' system; head corresponding-to 1040 to.150 psig.

' f. Verify that: the .Once/ operating suction for the-- . cycle;'

RCIC system Lis?

automatically.

-transferred from the condensate

' storage. tank to-the suppression:-

> pool on:a. condensate storage tank water-level-low signal.

The provisions ~of' Specification-

~

2.a 4.0.D are not-applicable provided-

'the surveillances are' performed ,

within~12 hours afterfreactors steam' pressure is adequate c tof .

perform these. tests. :!

! -' b . =- -If' OPERABILITY is.not-successfully-c = demonstrated within the 12-hour-period, reduce reactor steam dome.

, pressure to less than-150_psig within-the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

i

. l E RTS-249 3.5-8 12/92 w - 'e -

'^

ev,, ,,.m,N,n,N , , ,

A, ,--ym y - s s e>wns- 4 p- ~ g ,

n. . -. , . .- . ,, . -

n: . .

DAEC-1:

operability:of the redundant-and: diversified low pressure core cooling systems

~

?and the RCIC system.

.The-HPCI and RCIC as well as all.other-Core Standby Cooling Systems must be operable when-starting-up from a. Cold Condition. It is realized that the HPCI ~

is-not_ designed to operate until reactor pressure exceeds 150'psig and is automatically' isolated before the reactor pressure decreases belowl100'psig.

It is the intent of this specification to assure that when the reactor is-

'being started up from a Cold Condition, the HPCI is not known to be inoperable.

] A~ time period of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is-given after reaching 150 psig vessel pressure to g _ { demonstrate that HPCI and RCIC are OPERABLE. If OPERABILITY is not'

l. euccessfully demonatrated within the; 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months period, a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months period is given l to allow any remedial measures to be taken and time for operators to safely l reduce reactor pressure below 150 psig. .

4

)

E. RCIC System i

The RCIC is designed to provide makaup to the nuclear system as part of the planned operation.for periods when the main condenser is unavailable. .RCIC-

_ also_ serves for decay heat removal when feedwater is lost. 'In1all other-

[ postulated accidents and transients, the ADS provides redundancy for the.HPCI.

- Based on this, an' allowable. repair time.of 1 month is-justified, however, a.

s maximem allowable repair time of 14 days is_ selected for conservatism.

4 F. ' Automatic Depressurization System (ADS)

~

The operability of the ADS'under all conditions'of depressurization of the

, xnuclear. system automatically or manually,-insuresfan; essential response to station abnormalities.

The nuclear system pressure relief system provides automatic nuclear system depressurization'for small breaksLin the nuclear system so that the low

~

. pressure coolant ~ injection (LPCI) and the_ core-spray = subsystems _canLoperate-to protect.the, fuel barrier.

Because the Automatic Depressurization System'does not1 provide makeup to the .

reactor primary vessel,=no credit is,taken for theLeteam cooling of the core- i

. caused by the system actuation to provice further conservatism to the CSCS.

Performance _ analysis _of thefAutomatic Depressurization System _is_ considered

- only with respect-to its depressurizing effect in conjunction with LPCI and-Core Spray and:is based on.3 valves. There are-four valves in the ADS andn each has a capacity of.approximately 810,000 lb/hr at a set pressure of 1125

.psig.

~

.RTS-249 3.5-18 '12/92;

---..:- 2. z ;; , -_ a . . a _ .. .- . . - . a. _.

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.14.B Liquid Holdup Tank 4.14.B Liquid Holdup Tank Instrumentation Instrumentation 3.14.B.1 A minimum of one LLRPSF 4.14.B.1 Each liquid holdup tank level Sample Tank level indicating instrument shall be demonstrated channel and one LLRPSP Surge Tank OPERABLE by level indicating channel shall be OPERABLE.

Applicability: At all times. a. Daily channel check during liquid additions to the tank (s).

Action: b. A channel calibration once per 18 months.

a. With no channel operable, liquid c. A quarterly channel functional additions to the tank may test. --

continue for up to 30 days provided that the tank level is estimated during all liquid additions to the tank.

b. If the minimum required instrumentation is not returned to OPERABLE status within 30 days, prepare and submit to the commission within 30 days, pursuant to Specification 6.11.3, a Special Report, in lieu of any other report, why the instrument was not made OPERABLE in a timely manner.

I c. The provisions of Specification 3.0.c are not applicable. {

l W

RTS-249 3.14-2 12/92

, , . .- - . -.. - . ~ . , ~ - - . . . ..

DAEC-1

LIMITING CONDITIONS FOR-OPERATION SURVEILLANCE REQUIREMENTS 13.14 RADIOACTIVE EFFLUENTS 4.14 RADIOACTIVE EFFLUENTS j

'I 3.14.A Liquid Holdup Tanks

4.14.A. Liquid Holdup.Tanksi 3.14.A.1 The quantity of radioactive 4 .14 . A .' 1 The quantity-of radioactive material contained in the material contained-'in the tanks; unprotected outdoor tanke shall shall be determined to be within f be limited to loss than or' equal _ the_50 curie' limit by. analyzing a.

~

to 50 curies, excluding tritium .ropresentative sample of the

- e.nd dissolved or entrained noble tanks' contenta at_leaut once.per gases. (The' liquid radwasto 7 days when radioactive mater'.ala- o R

storage tanks in the Low-Level are being added to a tank.

Radwaste Processing and Storage Facility are considered. ~

unprotected outdoor tanks.) i

)

Applicabilit_y At all times. _

Action: -!

a. With the quantity of radioactive i material in the tanks exceeding the above limit, immediately suepond all-additions of radioactive material to the tankn, within 40 houra reduco she tank contents to within the limit, and describe-the events leading to thin condition in the next Semiannual Radioactive Effluent Release beport.

j b. The provisions of Specification L 3.0.C are not applicable.

i--

l-1 i

l

-Tanko included-in. thin specification are those outdoor. tanks that are not surrounded by liners, dikes, or walla capable of-holding the tanks' contents and that do not have tank overflows and currounding area draine' connected to the liquid radwante treatment-system.

RTSA 249 ,

1 3.14-1 - 12/92 e n ne - d w. w w.a.- ---*-rrw ,,-e4, , gm.-, +g.. .,,,,,w@ ppg.p,,_, _ g 9.,,,. -

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l.- LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4'

27 .If at any-time during REACTOR POWER OPERATION at 225% : RATED

-POWER,it-is determined byinormal-

-surveillance that the limiting-value for LHGH is being exceeded, action shall-then be initiated within'15 minutes to restore.

operation'to within the prescribed limits. If the-LHGR-

~

is not returned:to within the prescribed limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , reduce reactor power to 525% of RATED POWER, or to such a. power -

level that the' limits'are again being met, within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Surveillance and corresponding action shall continue until'.the prescribed

~

limits are again being met.

~

C. ' Minimum Critical Power Ratio C. -Minimum Critical Power Ratio:

(MCPR) (MCPR)

~

'1. MCPR shall be greater than-or l '. Verify MCPR is greater than or equal to'the MCPR limit specified equal to-the' required ilmit.

in the CORE OPERATING LIMITS REPORT a. At'least once per day during-REACTOR POWER OPERATION AT 225%?

-L RATED POWER and

b. Following'any.significant change L in power.--levol or_ distribution.

2. .The. provisions of Specification

'4.0.~D are not. applicable.

f i'

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l-f, g JRTS-249- - 3.12-3 12/92

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l DAEC-1 l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

2. If at any time during REACTOR POWER OPERATION (one or two loop) at c25% RATED POWER, it is determined by normal surveillance that the limiting value for MAPLHGR (LAPLHGR) is being exceeded, action shall then be initiated within 15 minutes to restore operation to within the prescribed limits. If the MAPLHGR (LAPLHGR) is not returned to within the prescribed limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , reduce reactor power to $25% of RATED POWER, or to such a power level that the limits are again being met, within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . _

3. If the reactor is being operated in SLO and cannot be returned to within prescribed limitr. within this 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period, thr reactor shall be brought to tre COLD SHUTDOWN condition wichin 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

4. For either the one or two loop operating condition surveillance and corresponding action shall continue until the prescribed action is met.

B. Linear Heat Generation Rate D. Linear Heat Generation Rate (LHGR)

(LHGR)

1. All LHGRa shall be less than or At least once per day during equal to the limits specified in reactor power operation at 225%

the CORE OPERATING LIMITS REPORT. rated power, verify all LHGRs are less than or equal to the required -

limits. _

The provisions of Specifice 3n 4.0.D are not applicable.

RTS-249 3.12--2 12/92

~

DAEC-1.

l LIMITING _ CONDITIONS FOR OPERATION -SURVEILLANCE REQUIREMENTS 3.12 CORE THERMAL LIMITS' 4.12 CORE-THERMAL LIMITS Applicability. Applicability TheLLimiting conditions for The Surveillance-Requirements operation associated with the _ apply to the-parameters which fuel' rods apply-to those monitor the fuel rod operating _

parameters which monitor the fuel conditions, rod operating conditions.

Objective Objective The Objective of the Limiting The Objective' of the' Surveillance '

Conditions for Operation is to Requirements is to specify'the assure.the performance of the- type and frequency of surveillance-fuel rods.' to be applied to the fuel _ rods.

Specification Specifications, A. Maximum Average Planar Linear A. Maximum Average-Planar' Linear Heat Heat Generation Rate (MAPLHGR1 Generation Rate (MAPLHGR)- >

1. All MAPLHGRs shall be less than At least once per_dayfduring or equal to the limits specified reactor power operation at 225% of in the CORE OPERATING-LIMITS. rated power,-verify _-all.MAPLHGRs REPORT. z are less than or equal _to required'

' limits.

The provisions of specificationi

- 4.0.D-are not applicable.

J 0l 4

'i 4

RTS-249 -3.12-1 12/92:-

?

DAEC-1

^ l1 LIMITING CONDITIONS FOR OPERATION - SURVEILLANCE REQUIREMENTS

-b. -In the COLD SHUTDOWN or REFUELING mode, with one main' control' room __

ventilation standby' filter unit-filtration subsystem inoperable, testore the inoperable subsystemL to OPERABLE status within 7 days or initiate and maintain operation of the OPERABLE subsystem in the isolation mode of operation or suspend CORE ALTERATIONS, handling of irradiated:1uel in the becondary containment and operations with a potential for draining the reactor vessel.

.c . In the COLD SHUTDOWN or REFUELING t mode,'with both main control room ventilation standby filter unit subsystems inoperable, IMMEDIATELY suspend CORE ALTERATIONS, handling of irradiated; fuel in the secondary containment and operations with a potential for draining the reactor vessel. 4 B. REMOTE SHUTDOWN PANELS B. REMOTE' SHUTDOWN PANELS

1. At all times when not in use or 1. The Remote Shutdown Panels:(Bay 4' being. maintained the Remote "A" Door) and local. control panels' Shutdown Panels-(Bay--"A" Door) shall be visuallycchecked once per-

<and local control panels shall be week-to verify they are' locked.

locked.

i

!' 2. -The provisions.of Specification 2. Operability of the switches:on_the-l- 3.0.C 'are not applicable. Remote Shutdown Panels . shall: be =

l- functionally tested once.per-operating cycle.

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s RTS-249- 3.lO-2a 12/92

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-DREC-1 l.-- LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS-

'2. With one or more of the primary

containment isolation valves: .

inoperable, maintain at least one isolation valve OPERABLE

or.

ISOLATED ** and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

a. Restore the inoperable valve (s) to OPERABLE status, or ,

b. Isolate each affected. penetration by use-of at least one-automatic valve locked or electrically deactivated in the isolated position,**'or 5

c. Isolate each affected penetration by use of at least one. manual ,

valve locked in the isolated position or blind flange.**

The provisions of Specification 3.0.D are not applicable provided that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the affected penetration is isolated-in accordance with Specification 3.7 D.2.b or 3.7.D 2.c, and provided thatcthe associated. r system, if applicable, is declared inoperable and the . i.

appropriate ACTION statements for that system are performed.

3. If Specification 3.7.D.1, and f 3;7.D.2 cannot be met, an orderly 3r' shutdown shall be initiated and the reactor shall be in the Cold Shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . t 4

.G

.c

This valve may be locked or T*

electrically deactivated as noted in.

Subsection 3.7.D'2.b.'

- Isolation valves closed.to satisfy

these requirements may be reopened on -

an intermittent basis under ,

administrative control. ,

RTS-249 3.7-19 12/92-

-c - - *,

?- L' DAEC-1 l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

3) Type C Tests Type C tests shall b performed during each reactor shutdown:for major refueling or other convenient interval but in no case at intervals greater.than two years. -The provisions.of :

Specification 4.0.B are-not applicable. !

4) Additional-Periodic Tests-Additional-purge system isolation 1 valve leakage integrity testing-shall be performed at_least once !

every three_ months in' order to' .i' detect' excessive leakage of the purge isolation valve resilienti. ,

seats. '_The purge system isolation 'l 1

valves will be. tested in three groups, by: penetration: . drywell purge exhaust group (CV-4302 and j' CV-4303), torus purge exhaust '

group (CV-4300 and..CV-4301), and.

drywell/ torus = purge ' supply group (CV-4307, CV-4308 and CV-4306). .

1

e. Seal Replacement & Mechanical. j Limiter- q The T-ring inflatable seals for: ;

purgeLisolation valves CV-4300, 1 CV-4301, CV-4302, CV-4303,JcV-4306, CV-4307.and CV-4308 shall be replaced at intervals notitoi l exceed four years. .The provisionsi

) of Specification'4.0.B are.not ,

. applicable.

l During Typ'e C testing, it chall be. j verified-that'the mechanical ^ j modification which' limits'the 1 maximum opening _. angle for-purge .

isolation valves CV-4300,-CV-4301,.

CV-4302, CV-4303',.CV-4306,'CV-4307 .

and CV-4308'is intact.

Ths b'aseline'for this_ requirement H shall be: established during.the cycle _6/7.refuelEcutage.

f . Containment Modification 1 l

'E f

.Any major modification,' _

-i replacement of a component which.

is part'of the primary reactor- !

containment boundary, or resealing; a seal-welded door, performed s after-the preoperational leakage -!

rate test shall be followed by ~

either a Type:A, Type B, or Type C:

test, as applicable,_for.the~ area i

RTS-249 3.7-7c 12/92 a

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't 14 L 'i ..

DAEC-1 l LIMITING CONDITIONS FOR OPERATION *?!RVEILLANCE REQUIREMENTS

d. Periodic Fetest - Schedth e-

1) Type A Test u After-the preoperational leakage rate tests,Ea set'of three Type A; tests.shall be performed, at approximately equal intervals.

.during each 30-year service?

period. (These intervals may be extended up to_eight monthecif necessary to' coincide with refueling. outages.)--The third test of each set shall be conducted when the plant-is shut

' down for the'10-year plant in-

-service inspections. The provisions of Specification 4.0.B are not applicable.

'The performance of; Type A tests shall be limited-to periods when>

the plant facility is nonoperational and secured in'the shutdown condition under-administrative' control and in accordance with the plant safety' procedures.

2) f"ype B Tests a). Penetrations and. seals of this" type;(except' air locks) shall--be leak tested-at greater _than or equal _to 43 paig (P,) during each reactor shutdown for major. fueling or other convenient-interval 1but int no case _at intervals-greater-.,.

than two years. 'The provisions of Specification 4.0.5 are not L applicabic, b) Thelpersonnel airlock;shall be-i pressurized-to greater;than or equal to '43 psig . (P,)1and leak

. tested at least once every six-(6),

months. This test interval may be' extended to~the next: refueling outage (up'to a maximumtinterval; between P, . tests of 24 : montho) -

providedLthere have been no airlock openings since.the last successful test at P . The: I provisions of Specif1 cation 54.0'B .

are not applicable to the'24-month-surveil lance ' interval . '

l 1

RTS-249-. 3.7-6 12/92 I

, . m. _ __ . . . _ _ _ _.-.

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DAEC-1 The first-10-yearJinterval for inservice testing of pumps-and valves-in accordance with the'ASME Code,'Section XI_ commenced on February 1,.1975 and The second 10 year inservice testing intervali ended-on January 31,.1985.

commenced on February 1,-1985 and is scheduled'to end on January 131,_1995.

The second-10 year testing program. addresses the requirements of the ASME

- Ecode,Section XI,11980 Edition with Addenda through Winter:1981, subject to-

- the limitations and modifications of 10 CFR 50.55a.. Section 3.9.6 of the Updated FSAR describes the inservice testir.g program.

-This specification includes a clarification of the. frequencies for performing- :;

the inservice inspection and-testing activities-required'by.Section XI of_the ASME Boiler and Pressure Vessel Code and' applicable Addenda. This clarification-is provided to ensure consistency in surveillance intervals .;

throughout the Technical Specifications and to remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing activities.

Under the terms of this specification, the more restrictive requirements of s the Technical Specifications take precedence over.the ASME Boiler and Pressure !

l Vessel Code and applicable Addenda.' The requirements of Specification _4.0.D ,

to' perform surveillance activities before entry'into'a CONDITION takes .l

' precedence over the ASME Boiler and Pressure Vessel Code provision'that allows I t- pumps'and valves to'be tested up to one week after return to normal operation. l The Technical Specification definition of OPERABLE does.not allow a grace.

period before a component, which is not' capable _of performing its specified i

-function, is declared' inoperable and takes precedence over the ASME-Boiler and l- Pressure Vessel Code provision that allows a valve to be incapable of.

-l performing its specified function for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months before being declared -

l inoperable.

i ,

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1

l 1.

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L RTS-249 3.6-29 s -12/92-a- ,

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DAEC-1 3.6.D & 4.6.D BASES:

l Safety and Relief Valves The pressure relief system has been sized to meet two design bases. the First, the total safety / relief valve capacity has been established to meet overpressure protection criteria of the ASME Code. Second, the distribution of this required capacity between safety valves and relief valves has been set to meet power generation design basis #1 of Section 5.4.13.1 of the Updated FSAR, which states that the nuclear system relief valves shall prevent opening of the safety valves during normal plant isolations and load rejections.

The details of the analysis which shows compliance with the ASME Code requirements is presented in Subsection 5.4.13 of the Updated FSAR and is reverified in individual reload analyses.

Six relief valves and two safety valves are installed. The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a peak vessel pressure less than the Code allowable overpressure limit of 1375 psig if a flux rcram is assumed.

The relief valve setpoints given in Section 2.2.1.B have been optimized to maximtze the simmer margin, i.e., the difference between the normal operating pressure and the lowest relief valve setpoint. The Reference 2 analysis shows that the six relief valves assure margin below the setting of the safety valves such that the safety valves would not be expected to open during any normal operating transient.* This analysis verifies that the peak system pressure during such an event is limited to greater than the 60_ psi design margin to the lowest spring safety valve setpoint.

Experience iu relief and safety valve operation shows that a testing of 50 percent of the valves per OPERATING CYCLE is adequate to detect failures or deteriorations. The relief and safety valves are bench tested every second OPERATING CYCLE to ensure that their setpoints are within the 1 percent tolerance. Additionally, once per OPERATING CYCLE, each relief valve is tested manually with reactor pressure above 100 peig and with turbine bypass flow to the main condenser to demonstrate its ability to pass steam. By observation of the change in position of the tur bine bypass valve, the relief valve operation is verified. A time period of : 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is given to complete this surveillance. If it is not successfully cumpleted within the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months period, a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months period is given to allow any recedial measures to be taken j and time for operators to saf ely reduce reactor pressure below 100 psig.

The requirements established above apply when the nuclear system can be pressurized above ambient conditions. These requirements are applicable at nuclear system pressures below normal operating pressures be :ause abnormal operational transients could possibly start at these conditions such that aventual overpressure relief would be needed. However, these transients are much less severe, in terms of pressure, than those starting at rated conditions. The valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurized.

The surveillance requires that at least once per OPERATING CYCLE at least one safety valve and 3 relief valves shall be removed, set pressure tested and.

reinstalled or replaced with spares that have been previously set pressure tested. For the most part, these valves will be set pressure tested and stored in accordance with the manufacturer's recommendations. There may be conditions where DAEC may not be notified by the manufacturer of new storage requirements or DAEC may tako exception with the requirements. In these isolated cases, DAEC and the manufacturer will come to resolution on an acceptable position.

A normal operating transient is defined as an event whose probability of-occurrence is greater than once per 40 years, e.g., Turbine Trip with Bypass, MSIV closure with direct scran.

RTS-249 3.6~24 12/92

-v 18*J = - TABLE 4.6.H '

SNUBBER VISUAL INSPECTION INTERVAL

. NUMBER OF UNACCEPTABLE SNUBBERS-population Column- A column B Column C

or category _ Extend-Interval Repeat Interval Reduce Interval l (Notes 1 and 2) -(Notes :3- and 6) (Notes 4 and 6)- (Notes 5 and 6) 1 0 0 1-80 0 0 2 100 0 1- 4 150 0 3 -8 200' 2 5. 13

'-- 300 5 12 25 400 8 18 36 500 12 24- 48 750 20 40 ~ 78 1000 or greater. 29 56 109 Note 1: The next visual-inspection interval for a snubber population or

- category size shall be determined based upon the previous

' inspection interval and the number of unacceptable snubbers-found during that interval. Snubbers may be categorized, based upon their accessibility during power operation, as accessible-or inaccessible. . These categories may be examined-separotely or jointly. However, the licensee must make and document that decision before any inspection and shall use that decision as the basis upon which to determine the next inspection interval for that category.

Note 2: Interpolation between. population or category. sizes and the number of unacceptable snubbers is permissible. .Use'next-lower _ integer-for the value of the limit for Columns A, B, or CLif that integer includes a fractional.value of unacceptable snubbers no determined by interpolation.

' Note 3: If the number of unacceptable snubbers is equal to[or less-than the number in' Column A', the next inspection interval may be twice

. the previous interval but.not greater than.48 months. -

f.

-Note 4: If'theEnumber or unacceptable. snubbers'is equal-to or less than. .1 the number in Column B but greater tban the number in Column A,_ ^!

the.next inspection' interval shallibe the;oame as'the previous 7 l intervali _.

w

'. Note 5 :- Ifthenumberofunacceptableonubbersis. equal'toorgreater$than

the number in Column C, the'next~ inspection. interval shall be two- -

c thirda of the previous interval. 'However, if the' number of.

unacceptable snubbers l1s less than the number in Column C but

. - greater than the number in' Column;B, the next interval shall be:

reduced proportionally' by interpolation, thatils, the' previous interval.shall'be,. reduced'byza factor that'isione-third of-the

- ratio of the difference between the.numberfof. unacceptable.

J anubbers found during the1 previous interval'and the. number-in;

. column B to the differonce in the numbers in. Columns B and_C..

, -Note 6: The provisions of SpecificationL4;0.B are applicable to.all

= inspection intervals up to'and including _48_ months, i

i t

P RTS-249 3.6-13 12/92 4

4 y , - , m ,...--._m. . . ,%.,,,-m'1r .,+.i._,, 7 ,. u.,. .rv,. , y .r- 4 y -, , ,. e- 9 ,

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

4. In RUN, STARTUP, or HOT SHUTDOWN 3. Performance of the above inservice MODE with Specif icat ion 3.6.G.2, ,

inspection and testing activities or 3.6.G.3 not met: l shall be in addition to other j specified Surveillance

a. perform an engineering l Requirements.

evaluation to determine the effectn of the component (s) l 4. Nothing in the ASME Boiler and conditjon for continued 1 Pressure Vessel Code shall be operation; and l conctrued to supersede the I requirements of any Technical

b. determLne that the l Specification.

component (s) remain acceptable for continued 1 5. The augmented inspection program operation, for piping identified in NRC Generic Letter 88-01 shall be If the above requirements cannot performed in accordance with the -

be met, icolate the affected stuff poaitions on schedule, ~

component (s) and follow the methods, personnel, and cample applicable syctem LCO. expansion included in this Generic Letter.

H. Shock Suppresaors (Snubberal H. Shock Suppresacra (Snubbers)

1. During RUN, STARTUP, and HOT Each safety-related snubber shall SHUTDOWN MODES all safety-related be demonstrated OPERABLE by snubbora shall be OPERABLE. In performance of the following COLD SHU^iDOWN and REFUELING MODES augmented inspection program and safety-related snubbers, located the Surveillance Requirements of on those systems required to be 4.6.H.5 and 4.6.H.6.

OPERABLE, must be OPERAELE.

2. WLth one or more snubbers 1. Visual Inspections inoperable, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace or restore the incperable Snubbers are categorized as e anubber(s) to OPERABLE status and inaccessible or accessible during perform an engineering evaluation reactor operation. Each of these per Surveillance Requirement categories (inaccessible and 4.6.H.4 on the supported accessible) may be inspected com ponent oc declare the independently according to the supported system inoperable and schedule determineu by -

follow the appropriate LCO for Table 4.6.H-1. The visual that ayatem. inspection interval for each type of snubber shall be determined based upon the criteria provided in Table 4.6.H-1 and the first inspection interval determined using this criteria shall be based upon the previous inspection interval as established by the requirements in effect before amendment No. (NRC to assign no.).

lRTS-249 3.6-12 12/92

?o**.=' DAEC-1 LIMITING CONDITIONS-POR OPERATION SURVEILLANCE REQUIREMENTS F. Jet Pump Flow Mismatch F. Jet Pump Flow Mismatch

1. With core power greater than or 1. Recirculation pump speed mismatch equal to 80% RATED POWER with shall be verified at least once both recirculation pumps at per day, steady state operation, the speed of the faster pump may not exceed 122% of the speed of the slower pump.

2. With core power.less than 80%

RATED POWER with both recirculation pumps at steady state operation, the-speed of_the faster pump may not exceed 135%

of the speed of the slower pump.

3. With the recirculation pump speeds different by more than the specified limits:

a. restore the recirculation pump speeds to within the specified limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ,' or

b. one recirculation pump 2. See Surveillance Requirement shall he tripped. See 4.3.F.4 tor SLO _ requirements.

Specification 3.3.F.4 for SLO requirements.

G. Structural Integrity G. Structural Integrity

1. At all times, the structural 1. Inservice inspection-of ASME.

integrity of the ASME Section XI- Section XI-Code Class 1, Class 2, j Code Class 1, 2, and 3 components and' Class 3 components and-l- shall be maintained-in accordance inservice testing of_ASME Section

! with surveillance Requirement XI code Class 1, Class 2, and 4.6.G.I. Class'3 pumps-and-valves shall be' performed in accordance with:

2. With the structural integrity of Section XI.of-the ASME--Boiler and I any ASME.Section XI' Code Class 1- -Pressure' Vessel Code and or class 2 component (s) not ' applicable Addenda as required by conforming to the'above 10CFR50, Section 50.55a(g), except

!- requirements, rostore the where specLfic written relief.has1 structural integrity of the been granted by the NRC pursuant affected component (s).to within to 10CFR50, Section its limit or Leolate.the affected 50.55a(g)(6)(1).

component (s) prior;to increasing the Reactor Coolant System 2. ' Surveillance frequencies specified temperature above 212*F. in Section XI of the ASME P,oller.

and Pressure Vessel Code and

3. With the structural integrity cf applicable Addenda-for the any ASME Section XI' Code Class 3 inservice inspection and testing Lcomponent(s) not conforming to ' activities are defined in the above requirements, restore .

Specification 1.0'(FREQUENCY-the structural integrity of the J NOTAT. ION) . The provisions of' l

l affected component (s) to within Specification'4.0.B are' applicable l its llmit or isolate the affected to these defined frequencies for-component (s) from service. performing inservice inspection and' testing activities.

RTS-249 3.6-11 12/92 g - w y-6m- p- - - - - - - -ynyq-#~r^ 1-

ML20127A361

Text

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