ML20205G863

Amend 230 to License DPR-65,resolving Several Previously Identified TS Compliance Issues

ML20205G863
Person / Time
Site:	Millstone
Issue date:	03/11/1999
From:	NRC (Affiliation Not Assigned)
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ML20205G864	List: ML20205G863 ML20205G867
References
NUDOCS 9904070395
Download: ML20205G863 (18)

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• A UNITED STATES j

NUCLEAR REGULATORY COMMISSION o

WASHINGTON, D.C. 20$55-0001 -

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l NORTHEAST NUCLEAR ENERGY COMPANY l

THE CONNECTICUT LIGHT AND POWER COMPANY THE WESTERN MASSACHUSETTS ELECTRIC COMPANY DOCKET NO. 50-336 MILLSTONE NUCLEAR POWER STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.?30 License No. DPR-65

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Northeast Nuclear Energy Company, et al. (the licensee) dated September 9,1998, as supplemented February 19 and 26,1999, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; j

C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commis ion's regulations; DJ The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

4 9904070395 990311 PDR ADOCK 05000336 i

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4 2. Accordingly, the licei se is amended by changes to the Technical Specifications as I

indicated in the attacument to this license amendment, and paragraph 2.C.(2) of Facility Operating License Nc DPR-65 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 230, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effectise as of the date of issuance, to be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

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M Elinor G. Adensam, Director Project Directorate 1-2 Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

ftrch 11,1999 i

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ATTACHMENT TO LICENSE AMENDMENT NO. 230 l

FACILITY OPERATING LICENSE NO. DPR-65 DOCKET NO. 50-336 Replace the following pages of the Appendix A, Technical Specifications, with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Remove Insert 1-5 1-5 1-6 1-6 3/4 0-1 3/4 0-1 3/4 0-2 3/4 0-2 3/4 2-9 3/4 2-9 3/4325 3/4 3-25 3/4 4-1 3/4 4-1 3/4 4-23 3/4 4-23 B 3/4 0-5 8 3/4 0-5 B 3/4 0-5a B 3/4 2-1 B 3/4 21 B 3/4 2 2 8 3/4 2-2 B 3/4 3-1a B 3/4 3-1a B 3/4 4-7c B 3/4 4-7c B 3/4 4-8 8 3/4 4-8

DEFINIT' IONS

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g AXIAL SHAPE INDEX 1.23 The AXIAL SHAPE INDEX (Yr) used for normal control and indication is the power level detected by the lower excore nuclear instrument detectors (L) less the power level detected by the upper excore nuclear instrument detectors (U) divided by the sum of these power levels. Thi AXIAL SHAPE INDEX (Y )

i used for the trip and pretrip signals in the reactor protection system is the above value (Ye) modified by an appropriate multiplier (A) and a constant (B) to

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determine the true core axial power distribution for that charnel.

Y, =

Y, = A Y, + B

.{0RE OPERATING LIMITS REPORT 1.24 The CORE OPERATING LIMITS REPORT is the unit specific document that provides the core operating limits for the current operating reload cycle.

These cycle specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.8. Plant operation within these l operating limits is addressed in individual specifications.

ENCLOSURE BUILDING INTEGRITY - DELETED J

REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until electrical power is interrupted to the CEA drive mechanism.

ENGINEERED SAFETY JEATURE RESPONSE TIME l

1.27 The ENGINEERED SAFETY FEATURE RESPONSE TINE shall be that time i

interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of HILLSTONE - UNIT 2 1-5 Amendment No. #, UP, # 7, 7978 0397 J

' DEFINITIONS ENGINEERED SAFETY FEATURE RESPONSE. TIME (Continued) l performing its ' safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

PHYSICS TESTS' l.28 PHYSICS TESTS shall be those tests performed to measure the fundamental-nuclear characteristics of the reactor core and related instrumentation and

1) described in Chapter 13.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

10TALtNRODDED INTEGRATED RADIAL PEAKING FACTOR - F(

1.29 The TOTAL UNR000ED INTEGRATED RADIAL PEAKING FACTOR is the ratio of the

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peak pin power to the average pin power in an unrodded core. This value includes the effect of AZIMUTHAL POWER TILT.

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SOURCE CHECK 1.30 A SOURCE CHECK shall be the qualitative assessment of channel response.when the channel sensor is exposed to radiation.

RADIOLOGICAL EFFLUENT MONITORING AND OFFSITE DOSE CALCULATION MANUAL (REMODCM) 1.31 A RADIOLOGICAL EFFLUEFT MONITORING MANUAL shall be a manual containing the site and environmental sampling and analysis programs for measurements 'of radiation and radioactive materials in those exposure pethways and for those radionuclides which lead to the highest potential radiation exposures tol individuals from station operation. An 0FFSITE DOSE CALCULA110N MANUAL shall be a manual containing the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints.

Require'ments of the REM 00CM are.

provided in Specification 6.15.

l RADI0 ACTIVE WASTE TREATMENT SYSTEMS 1.33 RADI0 ACTIVE WASTE TREATMENT SYSTEMS are those liquid, gaseous and solid waste systems which are required to maintain control over radioactive material in order to meet the LCOs set forth in these specifications.

PURGE'-' PURGING 1.34 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 containment.

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MILLSTONE - UNIT 2 1-6 Amendment No. #, M. J#, N i

0397

E-c.

3/4 L'INITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIRENENTS 3/4.0 APPLICABILITY LINITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

3.0.2 Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are j

.not met within the specified time intervals, except as provided in LC0 3.0.6.

l If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

3.0.3.

When a L!miting Condition for Operation is not met, except as provided in the associated ACTION requirements, within one hour ACTION shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable, in:

1.

At least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, 2.

At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and 3.

At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Where corrective measuree are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measur i from the time it is identified that a Limi*.ing Condition for Operatio is not met. Exceptions to these requirements are stated in the individual specifications.

3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made when the conditions for the Limiting Condition for Operation are not met and the associated ACTION requires a shutdown if they are not met within a specified time interval. Entry into an OPERATIONAL MODE or specified condition may be made in accordance with ACTION requirements when conformance to them permits continued operation of the facility for an unlimited period of time. This. provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements.

3.0.5 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 (5),

subsystem (s), train (s), component (s) and device (s) are OPERABLE. :r likewise satisfy the requirements of this specification. Unless both canditions (1) and (2) are satisfied within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, ACTION shall be initiatad to place the unit in a MODE in which the appli;3ble Limiting Condition foi Operation does not apply by placing it, as applicable, in:

NILLSTONE - UNIT 2 3/4 0-1 Amendment Nos. JK, JJJ, M 0398

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PPLIC48ILITY i

LIMITING CONDITION FOR OPERATION (Continued) 1.

At least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

2.

At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and i

3.

' At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

This specification is not applicable in MODES 5 or 6.

3.0.6 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to i

perform testing required to demonstrate its OPERABILITY or the OPERABILITY of l

other equipment. This is an exception to LCO 3.0.2 for the system returned to j

service under administrative control to perform the testing required to l

demonstrate OPERABILITY.

SURVEILLANCE REQUIREMENTS-l 4.0.1 Surveillance Requirements shall be applicable during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveille.nce Requirement.

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'4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25%

of the surveillance time interval.

1 4.0.3

-Failure to perform a Surveillance Requirement within the allowed surveillance. interval, defined by ' Specification 4.0.2, shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation. The time limits of the ACTION requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed.

The ACTION reqcirements may be delayed for up to.24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to permit the completion of the surveillance when the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Surveillance Requirements do not have to be performed on inoperable equipment.

l 4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement (s) associated with the' Limiting Condition for Operation have been perforned within the stated surveillance interval or as otherwise specified. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION' requirements.

4.0.5 Surveillance Requirements for inservice inspection and testing of l'

ASME Code Class 1, 2 and 3 components shall be applicable as follows:

a.

Inservice inspection of ASME Code Class 1, 2 and 3 components and inservice testing ASME Ccde Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a.

MILLSTONE - UNIT 2 3/4 0-2 Amendment No. J7, 77, JJJ JJ7, 8 0308

a.

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POWER ~ DISTRIBUTION LIMITS TOTAL UNR000ED INTEGRATED RADIAL PEAKING FACTOR - F',

LINITING CONDITION FOR OPERATION 3.2.3 The calculated value of F', shall be within the 100% power limit specified in the CORE OPERATING LIMITS REPORT. The F, value shall include the T

effect of AZIMUTHAL POWER TILT.

APPLICABILITY: MODE 1 with THERMAL POWER >20% RTP*.

l ACTION:

With F', exceeding the 100% power limit within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:

a.

Reduce THERMAL POWER to bring the combination of THERMAL POWER and F', to within the power dependent limit specified in the CORE OPERATING LIMITS REPORT and withdraw the full length CEAs to or beyond the Long Term Steady State Insertion Limits of Specificatioi) 3.1.3.6; or b.

Be in at least H0T STANDBY.

SURVEILLANCE REQUIRENENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

T 4.2.3.2 F, shall be determined to be within the 100% power limit at the l

following intervals:

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a. Prior to operation above 70 percent of RATED THERMAL POWER after each fuel loading,

b. At least once per 31 days of accumulated operation in Mode 1, and

c. Within four hours if the AZIMUTHAL POWER TILT (T ) is > 0.020.

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4.2.3.3 F, shall be determined by using the incore detectors to obtain a power distribution map with all full length CEAs at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump Combination.

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• See Special Test Exception 3.10.2

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MILLSTONE - UNIT 2 3/4 2-9 Amendment No. 75, pg, 77, SP, 77,

((/, //f, [ff, [ff, [ff, M 0399 i

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9 TABLE 4.3-2 (Continuedl TABLE NOTATION (1) The coincident logic circuits shall ne tested automatically or manually at least once per 31 days.

The automatic test feature shall be verified OPERABLE at least once per 31 days.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or other specified conditions for surveillance testing of the j

following:

i a.

Pressurizer Pressure Safety Injection Automatic Actuation Logic; and b.

Pressurizer Pressure Containment Isolation Automatic Actuation Logic; and c.

Steam Generator Pressure Main Steam Line Isolation Automatic Actuation Logic; and d.

Pressurizer Pressure Enclosure Building Filtration Automatic Actuation Logic.

Testing of the automatic actuation logic for Pressurizer Pressure Safety _ Injection, Pressurizer Pressure Containment Isolation, and Pressurizer Pressure Enclosure Building Filtration shall be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after exceeding a pressurizer pressure of 1850 psia in MODE 3.

Testing of the automatic actuation logic for Steam Generator Pressure Main Steam Line Isolation shall be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after exceeding a steam generator pressure of 700 psia in MODE 3.

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MILLSTONE - UNIT 2 3/4 3-25 Amendment.No. [7,230 0400

a.

REACT 0'R f.30LANT SYSTEN COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION a

LINITING CONDITION FOR OPERATION 1

3.4.1.1 Both reactor coolant loops and both reactor coolant pumps in each loop shall be in operation.

APPLICABILITY: MODES 1 and 2*.

ACTION:

With less than the above required reactor coolant pumps in operation, be in at least HOT STAND 8Y wtthin I hour.

SURVEILLANCE REQUIRENENTS

'4.4.1.1 The above required reactor coolant loops shall be verifiea to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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• See Special Test Exception 3.10.4.

NILLSTONE - UNIT 2 3/4 4-1 Amendment No. pp, J7,231 0401

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a.

REACTOR C09kANT SYSTEN REACTOR COOLANT SYSTEN VENTS 4

LINITING CONDITION FOR OPERATION 3'.4.11 At least 'one reactor coolant system vent path consisting of at least two valves in series capable of being powered from emergency buses shall be OPERABLE and closed at each of the following locations:

a.

Reactor Vessel head b.

Pressurizer steam space APPLICABillTY: MODES 1, 2, 3, and 4.

ACTION:

a.

With the Pressurizer vent path inoperable, STARTUP and/or POWER OPERATION may continue provided that i) the inoperable vent path is maintained closed with power removed from the valve actuator of all the valves in the inoperable vent path and ii) one power operated relief valve (PORV) and its associated block valve is OPERABLE; otherwise, restore either the inoperable vent path or one PORV and its associated block valve to OPERABLE status within 30 days, or submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the path to OPERABLE status, b.

With the Reactor Vessel Head vent path inoperable, STARTUP and/or POWER OPERATION may continue provided that the inoperable vent path is maintained closed with power removed from the, valve actuator of all the valves in the inoperable vent-path; restore the Reactor Vessel Head vent path to OPERABLE status within 30 days or. submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the path -

to OPERABLE status.

SURVEILLANCE REQUIRENENTS 4.4.11 Each reactor coolant system vent path shall be demonstrated OPERABLE at least once per 18 months by:

1.

Verifying all manual isolation valves in each vent path are locked in the open position.

2.

Cycling each valve in the vent path through at least once complete cycle of full travel from the control room during COLD

-SHUTDOWN or REFUELING.

3.

Verifying flow through the reactor coolant vent system vent paths during COLD SHUTDOWN or REFUELING.

g NILLSTONE - UNIT 2 3/4 4-23 AmendmentNo.79,77,Ig, om 151,Ille

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BASES (Con't)

I be consistent with the ACTION statement for the inoperable normal power sources. instead, provided the other specified conditions are satisfied.

In case, this would mean that for one division the emergency power source must be

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'0PERABLE (as must be the components supplied by the emergency power source) and all redundant systems, subsystems, trains, components and devices in the other divisions must be OPERABLE, or likewise satisfy Specification 3.0.5 (i.e., be capable of performing their design functions and have an emergency power source OPERABLE).

In'other words, both emergency power sources must be OPERABLE and all' redundant systems, subsystems, trains, components and devices in both divisions must also be OPERABLE. If these conditions are not satis-fied, action is required in accordance with this specification.

In MODES 5 and 6 Specification 3.0.5 is not applicable, and thus the

-individual ACTION statements for each applicable Limiting Condition for Operation in these MODES must be adherad to.

Specification 3.0.6 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 LC0 3.0.2 (e.g., to not comply with the applicable Required Action (s)) to allow the performance of surveillance requirements to demonstrate:

a.

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

The OPERABILITY of other equipment.

The administrative controls ensure the time the equipment is returned to service in conflict with the requirements of the ACTIONS is limited to the time absolutely necessary to perform the allowed surveillance requirements.

The Specification does not provide time to perform any other preventive or corrective maintenance.

-An example of demonstrating the OPERABILITY of equipment being returned to service is reopening a containment isolation valve that has been closed to comply with the Required Actions and must be reopened to perform the surveillance requirements.

An example of demonstrating the OPERABILITY of other equipment is taking an inoperable channel or trip system out of the tripped condition to prevent the trip function from occurring during the performance of a surveillance requirement on another channel in the other trip system. A similar example of demonstrating the OPERABILITY of other equinment 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 a

. surveillance requirement on another. channel in the same trip system.

MILLSTONE - UNIT 2 B 3/4 0-5 Amendment No. JJJ, Jpg 23n 0403

• BASES (Con't)

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Soecification 4.0.1 throuch 4.0.5 establish the general requirements applic-able to Surveillance Requirements. These requirements are based on the Surveillance Requirements stated in the Code of Federal Regulations, 10CFR50.36(c)(3):~

" Surveillance requirements are requirements relating to test, cali-bration, 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."

Soecification 4.0.1 establishes the requirement that'surveillances must.be performed during the OPERATIONAL MODES or other conditions for which the 4

requirements of the Limiting Conditions for Operation apply unless otherwise I

stated in an individual Surveillance Requirements. The purpose of this specification is to ensure that surveillances are performed to verify the operational status of systems and components and that parameters are within

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specified limits to ensure safe operation of the facility when the plant is in j

a MODE or other specified condition for which the associated Limiting Conditions for Operation are applicable. Surveillance Requirements do not have to be performed when the facility is in an OPERATIONAL MODE for which the requirements of the associated Limiting Condition for Operation do not apply unless otherwise specified. The Surveillance Requirements associated with a Special Test Exception are only applicable when the Special Test Exception is used as an allowable exception.to the requirements of a specification.

Specification 4.0.2 This specification establishes the limit for which the specified time interval for Surveillance Requirements may be extended.

It permits an allowable extension of the normal surveillance interval to facilitate surveillance scheduling and consideration of plant operating conditions that may not be suitable for conducting the surveillance; e.g.,

transient conditions or other ongoing surveillance or maintenance activities.

It also 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 t.".at are not performed during refueling outages. The limitation of Specification 4.0.2 is based on engineering judgment and the recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Requirements. This provision is sufficient to ensure that the reliability ensured through MILLSTONE. UNIT 2 B 3/4 0-5a Amendment No. 2m

.0403

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3/4:2 POWER DISTRIBUTION LIMITS BASES 3/4.2.1 LINEAR HEAT RATE The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200*F.

'Either of the two core power distribution monitoring systems, the Excore Detector Monitoring. System and the Incore Detector Monitoring System, provide adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its limits.

The Excore Detector Monitoring System performs this function by continuously monitoring the AXIAL SHAPE INDEX with two OPERABLE excore neutron flux detectors and verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits specified in the Core Operating Limits Report using the Power Ratio Recorder.

The power dependent limits of the Power Ratio Recorder are less than or equal to the limits specified in the Core Operating Limits Report.

In conjunction with the use of the excore monitoring system and in establishing the AXIAL SHAPE INDEX limits, the following assumptions are made:

1) the CEA insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are satisfied, 2) the AZIMUTHAL POWER TILT restrictions of Specification 3.2.4 are satisfied, and 3) the TOTAL UNRODDED INTEGRATED RADIAL PEAKING FACTOR does not exceed the limits of Specification 3.2.3.

i The Incore Detector Monitoring System continuously provides a direct measure of the peaking factors and the alarms which have been established for the individual incore detector segments ensure that the peak linear heat rt.tes will be maintained within the allowable limits specified in the core Operating Limits Report.

The setpoints for these alarms include allowances, set in the conservative directions, for 1) a flux peaking augmentation factor,

2) a measurement-calculational uncertainty factor, 3) an engineering uncertainty factor, 4) an allowance for axial fuel densification and thermal expansion, and

5) a THERMAL POWER measurement uncertainty factor specified in the Core Operating Limits Report.

Note the Items (1) and (4) above are only applicable to fuel batches "A" through "L".

The Incore Detector Monitoring System is not used to monitor linear heat rate below 20% of. RATED THERMAL POWER. The accuracy of the neutron flux information from the incore detectors is not reliable at THERMAL POWER < 20% RATED THERMAL POWER.

and 3/4.2.4 TOTAL UNRODDED INTEGRUf] RADIAL PEAKING FACTORS F, M 3/4.2.3 T

AZIMUTHAL POWER TILT - Tq T

The limitations on F, and T are provided to 1) ensure that the assump-q tions used in the analysis for establishing the Linear Heat Rate and Local power Density - High LCOs and LSSS setpoints remain valid during operation at the various allowable CEA group insertion limits, and, 2) ensure that the assumptions used in the analysis establishing the DNB Margin LCO, and Thermal Margin / Low Pressure LSSS setpoints remain valid during operation at the various allowable CEA group insertion limits.

If F',

or T exceed their basic o

limitations, operation may continue under the additional restrictions imposed i

HILLSTONE-- UNIT 2 B 3/4 2-1 Amendment No. 77, 97, Jgg, Mo*

119, 199. 199 199, 230 j

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POWER DISTRIBUTION LIMITS BASES by the ACTION statements since these additional restrictions provide adequate provisions to assure that.the assumptions used in establishing the Linear Heat Rate, Thermal Margin / Low Pressure and Local Power Density - High LCOs and LSSS setpoints remain valid.

An AZIMUTHAL POWER TILT > 0.10 is not expected and if it should occur, subsequent operation would be restricted to only those operations required to identify the cause of this unexpected tilt.

Data from the incore detectors are used for determining the measured :adial peaking factors.

Technical Specification 3.2.3 is not applicable below 20% of RATED THERMAL POWER because the accuracy of the neutron flux information from the incore detectors is not reliable at THERMAL POWER < 20% RATED THERMAL POWER.

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The surveillance requirements for verifying that F',

and T are within

]

their limits provide assurance that the actual values of F, and T do not i

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exceed the assumed values.

Verifying f', after each fuel loading prior to, I

exceeding 70% of RATED THERMAL POWER provides additional assurance that the l core was properly loaded.

3/4.2.6 DNB MARGIN The limitations provided in this specification en'sure that the assumed margins to DNB are maintained.

The limiting values of the parameters in this specification are those assumed as the initial conditions in the accident and transient analyses; therefore, operation must be maintained within the specified limits for the accident and transient analyses to remain valid.

i MILLSTONE - UNIT 2 B 3/4 2-2 Amendment Fo. 77, 52, Jgg,&

MM J)?,155, 1

, 3/4.3 INSTRUNENTATION BASES 3/4.3.1 AND 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF) INSTRUMENTATION (continued) ladders, testing one ladder matrix at a time will not remove an RPS channel from the overall logic matrix.

Therefore, matrix testing will not remove an RPS channel from service or make the RPS channel inoperable.

It is not necessary to enter an action statement while performing matrix testing.

This also applies when testing the reactor trip circuit breakers since this test will not remove an RPS channel from service or make the RPS channel inoperable.

The provisions of Specification 4.0.4 are not applicable for the CHANNEL FUNCTIONAL TEST of the Engineered Safety Feature Actuation System automatic actuation logic associated with Pressurizer Pressure Safety Injection, Pressurizer Pressure Containment Isolation, Steam Generator Pressure Main Steam Line Isolation, and Pressurizer Pressure Enclosure Building Filtration for entry into MODE 3 or other specified conditions.

After entering MODE 3, pressurizer pressure and steam generator pressure will be increased and the blocks of the ESF actuations on low pressurizer pressure and low steam generator pressure will be automatically removed. After the blocks have been removed, the CHANNEL FUNCTIONAL TEST of the ESF automatic actuation logic can be performed. The CHANNEL FUNCTIONAL TEST of the ESF automatic actuation logic must be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishing the appropriate plant conditions,-and prior to entry into MODE 2.

The measurement of response time at the specified frequencies provides assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the accident analyses.

No credit was taken in the analyse:; for those channels with response times indicated as not applicable.

The Reactor Protective and Engineered Safety Feature response times are contained in the Millstone Unit No. 2 Technical Requirements Manual.

Changes to the Technical Requirements Manual require a 10CFR50.59 review as well as a review by the Plant Operations Review Committee.

The containment airborne radioactivity monitors (gaseous and particulate) are provided to initiate closure of the containment purge valves upon detection of high radioactivity levels in the containment.

Closure of these valves prevents excessive amounts of radioactivity from being released to the environs in the event of an accident.

The actuation logic for this function is 1 out of 4.

Action Stater.sent 3 of Table 3.3-3 addresses inoperable containment purge channels.

. MILLSTONE - UNIT 2 8 3/4 3-la Amendment No. 77),230 0440

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REACTbRCOOLANTSYS'.EM BASES An exception to Technical Specification 3.0.4 is specified for Technical Specification 3.4.9.3 to allow a plant cooldown to MODE 5 if one or both PORVs are inoperable. MODE 5 conditions may be necessary to repair the PORV(s).

3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that 'the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant.

These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required.by 10 CFR Part 50.55a.

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MILLSTONE - UNIT 2 8 3/4 4-7c Amendment No. Up 230 0406

1 BASES

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3/4.4.11 Reactor Coolant System Vents Reactor Coolant System Vents are provided to exhaust i

noncondensible gases and/or steam from the primary system that could inhibit natural circulation core cooling. The OPERABILITY of at least one reactor coolant system vent path from the reactor vessel head and the pressurizer steam space ensures the capability exists to perform this function.

The valve redundancy of the reactor coolant system vent paths serves

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to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply or control system does not prevent isolation of the vent path.

The flow test verifies that each flowpath through the two solenoid valves is OPERABLE.

This verification can be performed by using a series of overlapping tests to ensure flow is verified through all parts of the system.

MILLSTONE - UNIT 2 B 3/4 4-8 Amendment No. J/p,2T 0406