Text

Proposed Tech Specs Correcting Various TS Compliance Issues, Including Modifying Definition of Azimuthal Power Tilt, Changing Wording of SR 4.1.1.1.2 & Changing Applicability of TS 3.1.3.4
	ML20203D103
Person / Time
Site:	Millstone
Issue date:	12/08/1997
From:	; NORTHEAST NUCLEAR ENERGY CO.
To:	;
Shared Package
ML20203D094	List: B16834, Application for Amend to License DPR-65,correcting Various TS Compliance Issues,Including Modifying Definition of Azimuthal Power Tilt,Changing Wording of SR 4.1.1.1.2 & Changing Applicability of TS 3.1.3.4; ML20203D103;
References
	NUDOCS 9712160111
	Download: ML20203D103 (35)
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DEFINITION!:

A2ItATDIAL POWER TILT - T y,,,,

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1.18 AZIMUTHAL POWER TILT shall be the + - difference betecen the d.

power generated in any core quadrant (upper or lower) and the average i power 6f all quadrants in that half (rpper or lower) of the core divided by the average power of all quadrants in that half (upper or lower) of l

/ % u m p o. hu Power in any core quadrant (upper or lower)

AZ1MUTHAL POWER TILT = 9 fveragepowerofallquadrants(upperorlower)Iq

-een-DOSE EQUIVALENT,I-131 J

1.19 DOSE EQUI. JINT I-131 shall be that concentration of I-131 (micro-curie / gram) which alone would produce the same thyroid dose as the quantity cnd isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Regulatory Guide 1.109 Rev.1, " Calculation of Annual Doses to Man from Routine R.: leases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50 Appendix I."

E-AVERAGE DISINTEGRATION ENERGY 1.20 E shall be the average sum of the beta and gamma energies per dis-integration (in MEV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total noniodine cctivity in the coolant.

STAGGERED TEST BASIS 1.21 A STAGGERED TEST BASIS snall consist of:

a.

A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into a equal subinterval, and b.

The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.

TREQUENCY NOTATION 1.22 The TREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

MILLSTONE UNIT 2 1-4 AmendmentNo.//[

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SURVEILLANCE REQUIREMENTS (Continued)

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When in MODES 3 or 4, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consider-ation of tte following factors:

l 1.

Reactor coolant system boron concentration.

2.

CEA position.

3.

Reactor coolant temperature, 4.

Fuel burnup based on gross themal energy generation, 5.

Xenon concentration, and 6.

Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within + 1.0% ak/k at least once per 31 Effective Full Power Days. This comparison shall. consider at least those factors stated in Specification 4.1,1.1.1.d. above.

The predicted reactivity values 4h*H be adjusted (nomalized) to correspond-to the actual core conditionshrior to excetding Afuel burnup of 60 Effective Full Power Days afterleach refueling.

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o REACTIVITY CONTROL SYSTEMS CEA DROP TIME i

LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full length (shutdown and control) CEA drop time, from a fully withdrawn position, shall be < 2.75 seconds from I

i when electrical power is interrupted.to the CIA ifrive mechanism until the CEA reaches its 90 percent insertion position with:~

T,yg L 515'F,and a.

b.

All reactor coolant Nmps operating.

APPLICABIL11Y: "0DE '

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/HO DFS f o nJ3 ACTION:

a.

With the drop time of any full length CEA determined to exceed the above limit, restore the CEA drop time to within the above limit prior to proceecing to MODE 1 or 2.

b.

With the CEA drop times within limits but determined at less than full reactor cociant flow, operation may proceed provided THERM'L POWER is restricted to less than or equal to the maximum THEL', POWER level allowable for the reactor coolant pump comhnation operating t.t the time of CEA drop time determination, SURVEILLANCE REOUIRDiENTS r

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4.1.3.4 The CEA drop time of full length CEAs shall be demonstrated through measurement prior to reactor criticality:

l a.

For a31 CEAs following each removal of the reactor vessel

head, b.

For specifically affected individual CEAs following an" main-tenance on or modification to the CEA drive system which could affect the drop time of those specific CEAs, and C.

At least once per 18 months.

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MILLSTONE - UNIT 2 3/4 1-26 Amendment No. 38,fj?,h f

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PEACTIVITY CONTROL SYSTEMS REGULATING CEA INSERTION LIMITS (Continued)

LIMITING CONDITION FOR OPERATION c.

With the regulating CEA groups inserted between the Long Tenn Steady State Insertion Limits and the Transient Insertion Limits specified in the CORE OPERATING LlHITS REPORT for irtervals > 5 EFPD per 30 EFPD interval or > 14 EFPD per calendar year, except during operations pursuant to the provisions of ACTION items c. and d. of Specification 3.1.3.1, either 1.

Restore the regulating groups to within the Long Term Steady State Insertion Limits provided in the CORE OPERATING LlHITS REPORT within two hours, or 2.

Be in HOT STANDBY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SVRVElllANCE REOUIREMENTO 4.1.3.6 The position of each regulating CEA group shall be detennined to be within the Transient Insertion Limits provided in the CORE OPERATING LIMITS REPORT at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the PDIL i*

-Attetleneer-Alarm-Cirettit inoperable, then verify the individual CEA positions at least once pe 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The accumu'.sted times during which the regulatory CEA groups are nserted between the Long Term Steady State Insertion Limits and the ransient insertion Limits specified in the CORE OPERATING LIMITS REPORT shall be determined at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months specified.

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-Febwery 10,1005-EMERGENCY CORE COOLING SYSTEMS ECCS SUB5YSTLMS - T _ < 300'I LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a.

One ' OPERABLE high-pressure safety injection pump, and b.

An OPERABLE flow path capable of taking suction from the refuel-ing water storage tank on a safety injection actuation signal and at:tomatically transferring suction to the containment sump on a xomp recirculation actuation signal.

APPLICA%1LIII: MODES 3* and 4.

g.g j ACTION:

a.

With no ECCS subsystem OPERABLE, restore at least one ECCS subsystem to OPERABLE status within one hour or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

With two or more high pressure safety injection pumps OPEFiABLE c.

and the temperature of one or more of the RCS cold legs s 275'F take innediate action to have a sayimum af one high pressure safety injection pump OPERABLE.

SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the appi.icable Surveillance Requirements of 4.5.2.

With pressurizer pressure < 1750 psia, f

A maximum of one high-pressure safety injection pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is 1275'F.

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l lasert A - Page 3/4 5-7 In MODE 4, the requirement for OPERABLE safety injection and sump recirculation actuation signals is satisfied by use of the safety injection and sump recirculation trip pushbuttons.

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s CONTAINMENT $YSTEMS CONTAINMENT VENTit.ATION SYSTEM LIMITING CONDITION FOR OPERATION closed :r.d :10:t '..ge supply and exhaust isolation valves shall b The containment pu 3.6.3.2

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rnle4 APPLICABILITY:

MODES 1. 2. 3. and 4.

ACTION:

With one containment purge supply and/or one exhaust isolation valve open

nd/:r :le;ti;6.117 ectiv.WC, close the open valve (s)-e-d e!::te4teHy

-detet4 tete within one hour or be in at least NOT STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

i SURVE1LLANCE REOU1REMENTS

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m'. '.1. 7 The containment purge supply and exhaust isolation valves shall be i

determined 10:ted losed - nd :10:t '::11y d::cti.eted pic,r te sach-ceeetow start;;.

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ELECTP!CALPOWERSYSTF1 3/448.2 OWSIT1 P % R DISTRIBUTION $YSTEMS i

i A.C. O!$7RIBUT!0N - OPERATING 4

LIMITING CONDIT!0N FOR OPERATION 4

3.8.2.1 The following A.C. electrical busses shall be OPERAeLE and energized from sources of power other than the diesel generators with tie breakers open between redundar.t busses:

4160 volt Emergency tus i 24 C 4160 volt Emdateg Ms # 24 0 480 volt Emergency iad Center f 22 E 480 volt Emergency' Lead Center # 22 F i

120 volt A.C. Vital Bus # TIAC-1 120 volt A.C. Vital Bus # "1AC-2 i

120 volt A.C. Vital Bus # 11AC, 9,A -30) 1 l

120 volt A.C. Vital Bus # TIAC-4 APPLICABILITY: MODES 1, 2. -3 and 4.

ACTION:

With less than the above complement of A.C. busses OPERABLE restore the inoperable bus and/or associated load center to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

SURVE!LLANCE RE0VIREMENTS 4.8.2 1 The specified A.C. busses shall be determined OPERABLE and energized from normal A.C. sources with tie breakers open between redundant busses at least once per 7 days by verifying correct breaker alignment and indicated power availability.

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ELECTRICAL POWft SYSTEMS 2/4.8.2 ON11TE POWER DISTRIBt1 TION SYSTEMS A.C. DISTR 18UTION - OPERATIN8 LINITINGCONDITIONFOROPERATION(Continued) 3.8.2.lA Inverters 5 and 6 shall be OPERABLE and available for automatic transfer via static switches VS1 and VS2 to power busses "!AC 4 and VIAC,t, respectively.

APPLICABILITY: MODES 1, 2 & 3 With inverter 5 or 6 inopera51e, restore the inverter to 2

ACI1ON:

a.

OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Q b.

With inverter 5 or 6 unavailable for automatic transfer via static switch VS1 or VS2 to power bus 4 ACM or

-VIAC4 respectively, restore the automatic transfer t/Ad capabilitywithin7daysorbeinHOTSHUTDOWNwithinthe next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

g.g With inverters 5 and 6 fnopera le or unavailable for c.

automatic transfer /via static sw tches VS! and VS2 to power busses W AC-1 and M

, respectively, restore the inverters to OPERABLE status or restore their automatic transfer capability within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS Verify correct inverter voltage, frequency, and alignment 4.8.2.!A a.

for automatic transfer via static switches VS! and VS2 to power busses H AC and V!"C-2, respectively, at least once per 7 days.

b.

Verify that busses V-IAC4 and W A automatically transfer to their alternate power sources, inverters 5 and 6 respectively, at least once per refueling during shutdown.

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BASES 1/41.1MOVEARLECONTROLASSEMBLIES(Continued)

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safety analyses and LC0 and LSSS setpoints determination.

Therefore, time limits have been imposed on operation with inoperable CEAs to preclude such adverse conditions from develnping, i

Operability of the CEA position indicators (Specification 3.1.3.3) is required to determine CEA positions and thereby ensure compliance with the CEA alignment and insertion limits and ensures proper operation of the rod block circuit.

The CEA " Full In' and " Full Out" limits provide an additional independent means for determining the CEA positions when the CEAs are at either their fully inserted or fully withdrawn positions.

Therefore, the ACTION statements applicable to inoperable CEA position indicators permit continued operations when the positions of CEAs with inoperable position indicators can be verified by the " Full In" or " Full Out" limits.

CEA positions and OPERABILITY of the CEA position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with more frequent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCO's are satisfied.

The maximum CEA drop time permitted by Specification 3.1.3.4 is the assumed CEA drop time used in the accident analyses. Measurement with T 515'F and with all reactor coolant pumps operating ensures that UE measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.

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REACTIV!TY CONTET. SYSTEMS

" BASES

-C 3/4.1.3 MOVABLE CON'iROL ASSEMBLIES (Continued)

The LSSS setpoints and the power distribution LCOs were generated

! based upon a core burnup which would be achieved with the core operating

in an essentially unro::ed configuration. Therefore, the CEA insertion I limit specificatior.s require tnat during MODES 1 and 2, the full length CEAs be nearly fully withdrawn. The amount of CEA insertion permitted by the Long Term Steady State Insertion Limits of Specification 3.1.3.6 will not have a significant effect upon the unrodded burnup assumption but will still provide sufficient reactivity control. The Transient Insertion Limits of Specification 3.1.3.6 are provided to ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUIDOWN MARGIN is maintained, and (3) the potential effects of a CEA ejection accident are limited to acceptable levels; however, long term operation at these insertion limits could have adverse effects on core power distribution during subsequent operation in an unrodded configura-

*"* K Q @ h The control rod crive mec anism requirement of Specification 3.1.3.7 is provided to assure that the consequences of an uncontrolled CEA with-(

drawal from subcritical transient will stay within acceptable levels.

This specification assures that reactor coolant system conditions exist which are consistent with the plant safety analysis prior to energizing the control rod drive mechanisms. The accident is precluded when condi-tions exist which are inconsistent with the safety aralysis since de-energized drive mechanisms car:not withdraw a CEA. The drive mechanisms i

may be energized with the boron concentration greater than or equal to the refueling concentration since, under these conditions, adequate SHUT-DOWN MRGIN is maintained,even if all CEAs are fully withdrawn from the Core.

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EMERGENCY CORE COOLING SYSTEMS i

J BASES I

The' purpose of the ECCS throttle valve surveillance requirements is to provide assurance that proper ECCS flows will be maintained in the event of a LOCA.

Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to:

(1) prevent total pump flow from axceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or,z.abeus that assumed in the ECCS-LOCA analyses.

4 MEcr c) w Only one HP51 pu:; say be OPERABLE in MODE 4 with RCS temperatures less than or equal to 275'F due to the restricted relief capacity with Low-Temperature Overpressure Protection System.

To reduce shutdown risk by having additional pumping capacity readily available, a HPSI pump may be made inoperable but available at short notice by shutting its discharge valve with the key lock on the control panel.

3/4.5.4 REF0ELIM WsTER STORAGE TANK (RWST)

The OPERABILITY of the RWST as part of the ECCS ensures that a sufficient supply of boratee water is available for injection by the ECCS in the event of a LOCA. The Insits on RWST minimum volume and boron concentration ensure that

1) sufficient water is available within containment to persit recirculation cooling flow te tre core, and 2) the reactor will remain subcritical in the cold condition f c11o ing mixing of the RWST and the RCS water volumes with all control rods inse ted except for the most reactive control assembly. These assumptions are coaststent with the LOCA analyses.

i MILLSTONE - UNIT 2 8 3/4 5-2 AmendmentNo.JJJJ),g[

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[nsert C - Pace B 3/4 5-2 In MODE 4 the automatic safety injection signal generated by low pressurizer pressure and high containment pressure and the automatic sump recirculation actuation signal e

generation by low refueling wate' storage tank level are not required to be OPERABLE.

i Automatic actuation in MODE 4 is not required because adequate time is available for i

plant operators to evaluate plant conditions and respond by manually operating engineered safety features components. Since the manual actuation (trip pus'1 buttons) i portion of the safety injection and sump recirculation actuation signal generation is required to be OPERABLE in MODE 4, the plant operators can use the manual trip pushbuttons to rapidly position all components to the required accident position.

Therefore, the safety injection and sump recirculation actuation trip pushbuttons satisfy the requirement for generation of safety injection pnd sump recirculation actuation signals in MODE 4.

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CONTAINMDfT SYSTEMS

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RASES 3/4.6.3 CONTAINMENT ISOLATION VALVES (continued)

The bypasses around the main steam supplies to the turbine driven auxiliary feedwater cump opened to drain water from(2 MS-201 and 2 MS-202), 2-MS-458 and 2 MS-459, ari t

1 the steam supply lines. When either 2-MS 458 or 2-MS 459 is o control room,pened, a dedicated operator, in continuous comunication with the is required. Operation of these valves is expected during pla nt startup.

j The containment station air header isolation, 2-SA-19, is opened to supply station air to containment.

When 2-5A-19 is opened, a dedicated cperator, in continuous comunication with the control room, is required.

Operation of this valve is only expected for maintenance activities inside containment.

The backup air supply master stop, 2-IA-566, is opened to supply backup air to 2 CH-SIf, 2-CH-518, 2 CH-519, 2-EB 88, and 2-EB 89. When 2-IA 566 is opened, a dedicated operator, in continuous comunication with the control room, is requh td.

Operation of this valve is only expected in response f 3a 1

loss of the normal air supply to the valves listed.

J The nitrogen header drain valve, 2-51-045, is opened to depressurize the i

containment side of the nitrogen supply header stop valve 2-SI-312. When [

2-SI MS is opened, a dedicated operator, in continuous co,munication with the control room, is required. Operation of this valve is only expectaJ after using the high pressure nitrogen system to raise SIT nitrogen pressure.

The containment waste gas header test connection isolation valve, 2 GR 63, is opened to sample the primary drain tank for oxygen and nitroger When 2-GR-63 is opened, a dedicated operator, in continuous comunication with the control rooni, :s required. Operation of this valve is expected during plant startup and shutdown.

The determination of the appropriate administrative controls for these containment isolation valves included an evaluation of the expected environmental conditions.

This evaluation has concluded environmental conditions will not preclude access to close the valve, and this action will prevent the release of radioactivity outside of containment through the respective pe-t*ation.

sentel h The contaica ent pu beVclosedWalertrindresupplyandexhaustisolationvalvesarerequiredto y deactivated during plant operation since these valves have not been demonstrated capable of closing during a LOCA or steam line break accident.

Such a demonstration would require justification of the mechanical operability of the phege valves and consideration of the appropriateness of the electrical override circuits. Maintaining these valves closed during plant operations ensures that excessive quantities of radioactive materials will not be released via the containment purge system.

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The containmeni purge supply and exhaust isolation valves are sealed closed by j

removing power from the valves. This is accomplished by pulling the control power l

fuses for each of the valves. The associated fuse blocks are then locked. This is consistent with the guidance contained in NUREG-0737 Item II.E.4.2 and Standard Review Plan 6.2.4, " Containment isolation System," Item II.f.

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k 5.0 DESIGN FEATURES 5.1 -SITE e

EXCLUSION AREA 5.1.1 Tae exclusion area is shown on Figure 5.1-1.

LOW POPULATION ZONE 5.1.2 The low population zone is shown on Figure 5.1-2.

FLOOD CONTROL 5.1.3 The flood control provisions shall be designed and maintained in accordance with the rigin design provisions contained in Section 2.5.4.2 of the FSAR.

5.2 CONTAINMENT

,' i CONFIGURATION 5.2.1 The reactor containment building is a steel lined, reinforced concrete building of cylindrical shape, with a doce roof and having the following design features:

a.

Nominal inside diameter = 130 feet, b.

Nominal inside height = 175 feet.

c.

Minimum thickness of concrete walls = 3.75 feet.

d.

Minimum thickness of concrete dome =' 3.25 feet.

e.

Minimum thickness of concrete floor pad = 8.5 feet, f.

Nominal thickness of steel liner = 0.25 inches.

6 g.

Net free volume = 1.9 x 10 cubic feet.

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M5!aN FEATUP.ES i

DfE1CN PREISURE AND TEMPERATURf 5.2.2 The reactor containment building is designed and shall'be maintained i

for a maximum internal pressure of 54 psig and an equilibrium liner [

temperature of tal'F.

PENETRATIONI i

5.t.3 Penetrations through the reactor containmenij uildin re designed and shall be maintsined in accordance with the < r design provisions contained in Section 5.2.8 of the FSAR with allow [an_iginal ce or normal degradation pursuant to the appitcable Surveillance Requirements.

5.1 REACTOR CORf 4

FUEL-ASSEMBL1ft 5.3.1 The reactor core shall contain 217 fuel assemblies with each fuel assembly containing 176 rods. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 4.5 weight percent of U 235.

4 CONTROL ELEMENT ASSEMBL1ES 2

5.3.2 The reactor core shall contain 73 full length and no part length control element assemblies. The control element assemblies shall be designed and maintained in accordance with the 4 Fro ~f ign provisions contained in i

Section 3.0 of the FSAR with allowance for norma degradation pursuant to the appitcable Surveillance Requirements.

5.4 REatTOR COOLANT SYSTEM DESIGN PRES $URE AND TEMPERATURE I

5.4.1 The reactor coolant system is designed and shall be maintained:

a.

In accordance with tha code requirements specified in Section 4.2.2 of the FSAR with allowance for normal degradation pursuant of the applicable surveillance Requirements, b.

For a pressure of 2500 psia. and c.

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temperature of 650'F ex:ept for the pressurizer which is 700'F.

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-Juu. 2, 1007 DE51GN FEATURES 5.7 $E!$MIC CLA55!FICAT!0N 5.7.1 Those structures, systems and components identified as Categorf I Itoms in Section 5.1.1 of the FSAR shall be designed and maintained to the fr sin esign provisions contained in Section 5.8 of the F5R4 with allowance normal degradation pursuant to the applicable Surveillance Requiremen 5.8 METEOROLOGICAL TOWER LOCATION fi.8.1 The meteorological tower location shall be as shown on Figura

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5.9 SHORECM PROTECT!CV 5.

The provision r shoreline rotection desc i ed in Amendmen JE and 36 to thef1AR shall be c latedbyJune) 1976. /

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J MILLSTONE - UNIT 2 5-6 Amendment No. f.777.

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a Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Compliance issues Number 2 Retyped Pages t

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DEFINITIONS i

AZIMUTHAL POWER TILT - Tq 1.18 AZIMUTHAL POWER TILT shall be the difference between the maximum l

power generated in any core quadrant (upper or lower) and the average power of all quadrants in that half (upper or lower) of the core divided by the average power of all quadrants in that half (upper or lower) of the core.

Maximum oower in any core auadrant (unoer or lower)

-1 AZIMUTHAL POWER TILT Average power of all quadrants (upper or lower)

DOSE E0VIVALENT I-131 1.19 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (micro-curie / gram) which alone would produce the same th.yroid dose as the quantity and isotopic mixture of I-131,1-132,1-133,1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Regulatory Guide 1.109 REv.1, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50 Appendix 1."

E-AVERAGE DISINTEGRATION ENERGY 1.20 E shall be the average sum of the beta and lamma energies per dis-integration (in MF.V) for isotopes, other than iodines, with half lives greater than 15 m.nutes, making up at least 95% of the total noniodine activity in the coolant.

STAGGERED TEST BASIS 1.21 A STAGGERED TEST BASIS shall consist of:

a.

A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval-into n equal subinterval, and b.

The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.

FRE0VENCY NOTATION i

1.22 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

MIgLSTONE-UNIT 2 1-4 Amendment No. JPJ,

O 9

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) d.

When in MODES 3 or 4, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consider-ation of the following factors:

1.

Reactor coolant system boron concentration, 2.

CEA position, 3.

Reactor coolant temperature, 4.

Fuel burnup based on gross thermal energy generation, 5.

Xenon concentration, and 6.

Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement with i 1.0% Ak/k at least once per 31 Effective Full Power Days.

This comparison shall consider at least those factors stated in Specification 4.1.1.1.1.d, above.

The predicted reactivity values may be adjusted (normalized) to correspond l

to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each refueling.

'l NILLSTONE UNIT 2 3/4 1-2 Amendment No.

0336

REACTIVITY CONTROL SYSTEMS 4

CEA DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual ful' length (shutdown and control) CEA drop time, from a fully withdraw.. >osition, shall be s 2.75 seconds from when electrical power is inti.crupted to the CEA drive mechanism until the CEA reaches its 90 percent insertion position with:

a.

T., 2 515'F, and b.

All reactor coolant pumps operating.

APPLICABIllTY:

MODES 1 and 2.

ACTION:

a.

With the drop time of any full length CEA determined to exceed the above limit, restore the CEA drop time to within the above limit prior to proceeding to MODE 1 or 2.

b.

With the CEA drop times within limits but determined at less than full reactor coolant flow, operation may proceed arovided THERMAL POWER is restricted to less than or equal to tie maximum THERMAL POWER level allowable for the reactor coolant pump combination operation at the time of CEA drop time determination.

SURVEILLANCE REQUIRENENTS 4.1.3.4 The CEA drop time of full length shall be demonstrated through measurement prior to reactor criticality:

a.

For all CEAs following each removal of the reactor vessel

head, b.

For specifically affected individual CEAs following any main-tenance on or modification to the CEA drive system which could affect the drop time of t50se specific CEAs, and c.

At least once per 18 months.

MILLSTONE - UNIT 2 3/4 1-26 Amendment No. 77, 77, pp.

0337

__~

-L.

O REACTIVITY CONTROL SYSTEMS REGULATING CEA INSERTION LIMITS (Cor.tingtdl

$URVEILLANCE REQUIRENENTS c.

With the regulating CEA grou)s inserted between the Long Term Steady State Insertion Limits and tie Transient Insertion Limits specified in the CORE OPERATING LIMITS REPORT for intervals > 5 EFPD per 30 EFPD interval or > 14 EFPD per calendar year, except during operations pursuant to the provisions of ACTION items c. and d. of Specification 3.1.3.1, either:

1.

Restore the regulating groups to within the Long Term Steady State Insertion Limits provided in the CORE OPERATING LIMITS REPORT within two hours, or 2.

Be in HOT STANDBY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIRENENTS 4.1.3.6 The position of each regulating CEA group shall be determined to be within the Transient Insertion Limits provided in the CORE OPERATING LIMITS REPORT at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the PDIL alarm is inoperable, then verify the individual CEA positions at least once l

per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The accumulated times during which the regulator.< CEA groups are inserted between the Long Term Steady State Insertion Limits ar.d the Transient Insertion Limits specified in the CORE OPERATING LIMITS REPORT shall be determined at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months specified.

l i

I MILLSTONE - UNIT 2 3/4 1-29 AmendmentJJJ,JJJ.

0338

EMERGENCY _ CORE.C00 LING.5YSTEMS ECCS SUBSYSTEMS - T,,_

< 300*F LIMITING CONDITION FOR OPERATION i

3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a.

One # OPERABLE high-pressure safety injection pump, and b.

An OPERABLE flow path capable of taking suction from the refuel-ing water storage tank on a safety injection actuation signal and automatically transferring suction to the containment sump on a sump recirculation actuation signal.***

l APPLICABillTY: MODES 3* and 4.

ACTION:

a.

With no ECCS subsystem OPERABLE, restore at least one ECCS subsystem to OPERABLE status within one hour or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

c.

With two or more high pressure safety injection pumps OPERABLE and the temperature of one or more of the RCS cold legs s 275'F take immediate action to have a maximum of one high pressure safety injection pump OPERABLE.

SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2.

With pressurizer pressure < 1750 psia.

A maximum of one high-pressure safety injection pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is s 275'F.

In H0DE 4, the requirement for OPERABLE safety injection and sump recirculation actuation signals is satisfied by use of the safety injection and sump recirculation trip pushbuttons.

MILLSTONE - UNIT 2 3/4 5-7 Aw ndment No. 77. Jp),

0339

i CONTAll0ENT SYSTEMS CONTAllMENT VENTILATION SYSTEM t

LIMITING CONDITION FOR OPERATION 3.6.3.2 The containment purge supply and exhaust isolation valves shall be i

sealed closed.

l APPLICABillII:

MODES 1, 2, 3 and 4.

ACTION:

With one containment purge sup91y and/or one exhaust isolation valve open, close the o)en valve (s) within one hour or be in at least HOT STANDBY within the next 61ours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.3.2 The containment purge supply and exhaust isolation valves shall be determined sealed closed at least once per 31 days.

I MILLSTONE - UNIT 2 3/4 6-19 Amendment No. JJ, 0340 m

ELECTRICAL POWER SYSTENS 3/_4 8.2 ONSITE POWER DISTRIBUTION SYSTEM A.C. DISTRIBUTION - OPER&I186 LINITING CONDITION FOR OPERATION 3.8.2.1 The following A.C. electrical busses shall be OPERABLE and energized from sources of power other than the diesel generators with tie breakers open f

between redundant busses:

4160 volt Emergency Bus # 24 C 4160 volt Emergency Bus #24 0 480 volt Emergency Load Center #22 E 480 volt Emergency Load Center #22 F 120 volt A.C. Vital Bus # VA-10 120 volt A.C. Vital Bus # VA-20 120 volt A.C. Vital Bus # VA-30 120 volt A.C. Vital Bus # VA-40 APPLICABILITY: H0 DES 1, 2, 3 and 4.

ACTION:

With less than the above complement of A.C. busses OPERABLE, restore the inoperable bus and/or associated load center to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

SURVEILLANCE REQUIRENENTS 4.8.2.1 The specified A.C. busses shall be determined OPERABLE and energized from normal A.C. sources with tie breakers open between redundant busses at least once per 7 days by verifying correct breaker alignment and indicated power availability.

NILLSTONE - UNIT 2 3/4 8-6 Amendment No.

0341 I

I 1

ELECTRICAL!2ffA _411TIH1 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. DISTRIBUTION - OPERATING LIMITING C0feITION FOR OPERATION (Continued) 3.8.2.lA Inverters 5 and 6 shall be OPERABLE and available for eutomatic transfer via static switches VS1 and VS2 to power busses VA-10 and VA-20, l

l respectively.

APPLICABILITY: MODES 1, 2 & 3 AGIl0!i:

a.

With inverter 5 or 6 inoperable, restore the inverter to OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

With inverter 5 or 6 unavailable for automatic transfer via static switch VS1 or VS2 to power bus VA-10 or VA-20, l

respectively, restore the automatic transfer capability within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c.

With inverters 5 and 6 inoperable or unavailable for automatic transfer via static switches VS1 ano '<S2 to power busses VA-10 and VA-20, respectively, restore the l

inverters to OPERABLE status or restore their automatic transfer capability within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.8.2.lA a.

Verify corred:, inverter voltage, frquency, and alignment for auton tic transfer via static switches VS1 and VS2 to power busses VA-10 and VA-20, respectively, at least once l

per 7 days, b.

Verify that busses VA-10 and VA-20 automatically transfer l to their alternate power sources, inverters 5 and 6, respectively, at least once per refueling during shutdown.

NILLSTONE - UNIT 2 3/4 8-6a Amendment No. JJF om

o o

SA$ES 3/4.1.3 MOVEABLE CONTROL ASSEMBLIES (Continued) safety analyses and LCO and LSSS setpoints determination.

Therefore, time limits have been imposed on operation with inoperable CEAs to preclude such adverse conditions from developing.

Operability of the CEA position indicators (Specification 3.1.3.3) is required to determine CEA positions and thereby ensere compliance with the CEA alignment and insertion limits and ensures proper operation of the rod block circuit.

The CEA " Full In" and " Full Out" limits provide an additional independent means for determining the CEA positions when the CEAs are at either their fully inserted or fully withdrawn positions.

Therefore, the ACTION statements applicable to inoperable CEA position indicators permit continued operations when the positions of CEAs with inoperable position indicators can be verified by the " Full In" or " Full Out" limits.

CEA positions and OPERABILITY of the CEA position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with more frequent verifications required if an automatic monitortig channel is inoperable.

These verification frequencies are adequate fo" assuring that the applicable LCO's are satisfied.

The maximum CEA drop time permitted by Specification 3.1.3.4 is the assumed CEA drop time used in the accident analyses.

Measurement with Tav 2 515'F and with all reactor coolant pumps operating ensures that the measure 0 l drop times will be representative of insertion times experienced during a reactor trip at operating conditions.

l-MILLSTONE - UNIT 2 8 3/4 1-4a Amendment No. JJJ, 0342

REACTIVITY CONTROL SYSTEMS nun 3/4.1.3 MOVABLE CONTROL ASSEMBLIES (Continued)

The LSSS setpoints and the power distribution LCOs were generated based upon a core burnup which would be achieved with the core operating in an essentially unrodded configuration.

Therefore, the CEA insertion limit specifications require that during MODES I and 2, the full iength CEAs be nearly fully withdrawn. The amount of CEA insertion permitted by the long Term Steady State Insertion Limits of Specification 3.1.3.6 will not have a significant effect upon the unrodded burnup assumption but will still provide sufficient reactivity control.

The Transient Insertion Limits of Specification 3.1.3.6 are provided to ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of a CEA ejection accident are limited to acceptable levels; however, long term operatira at these insertion limits could have adverse effects on core power distribution during subsequent operation in an unrodded configura-tion.

The PDIL alarm is provided by the CEAPDS computer.

l The control rod drive mechanism requirement of specification 3.1.3.7 is provided to assure that the consequences of an uncontrolled CEA withdrawal from subcritical transient will stay within acceptable levels.

This specification assures that reactor coolant system conditions exist which are consistent with the plant safety analysis prior to energizing the control rod drive mechanisms.

The accident is i

precluded when conditions exist which are inconsistent with the safety analysis since deenergized drive mechanisms cannot withdraw a CEA.

The drive mechanisms may be energized with the boron concentration greater than or equal to the refueling concentration since, under these conditions, adequate SHUTDOWN MARGIN is maintained, even if all CEAs are fully withdrawn from the core.

l l

l oMLSTONE-UNIT 2 8 3/4 1-5 Amendment No. 77,JJJ M

EMERGENCY CORE COOLING SYSTEMS BASES l

The purpose of the ECCS throttle valve surveillance requirements is to provide assurence that proper ECCS flows will be maintained in the event of a LOCA.

Maintenance of propar flow resistance and pressure drop in the piping system to each injection point is necessary to:

(1) provent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECC3-LOCA analyses.

In MODE 4 the automatic safety injection signal generated by low pressurizer pressure and high containment pressure and the automatic sump recirculation actuation signal generation by low refueling water storage tank level are not required to be OPERABLE.

Automatic actuation in MODE 4 is not required because adequate time is available for plant operators to evaluate plant conditions and respond by manually opera;ing engineered safety features components.

Since the manual actuation (trip pushbuttons) portion of the safety injection and sump recirculation actuation signal generation is required to be OPERARLE in MODE 4, the plant operators can use the manual trip pushbuttons to rapidly position all components to the required accident position.

Therefore, the safety injection and sump recirculation actuation trip pushbuttons satisfy the requirement for generation of safety injection an : ump recirculation actuation signals in MODE 4.

Only one HPSI puwp may be OPERABLE in MODE 4 with RCS temperatures less than or equal to 275'F due to the restricted relief capacity with Low-Temperature Overpressure Protection System.

To reduce shutdown risk by having additional pumping capacity readily available, a HPSI pump may be made inoperable but available at short notice by shutting its discharge valve with the key lock on the control panel.

3/4.5.4 REFUELING WATER STORAGE TANK (RWST)

The OPERABILITY of the RWST as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits en RWST minimum volume and boron concentration ensure that 1) sufficient water is available within containment to permit recirculation cooling flow to the core, and 2) the reactor will remain subcritical in the cold condition following mixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses.

HILLSTONE - UNIT 2 B 3/4 5-2 Amendment No. J), JJ7, JP),

0343

w CONTAIMENT SYSTEMS BASES 3/4.6.3 CONTAINMENT IS01ATION VALVES (continued)

The bypasses around the main steam supplies te the turbine driven auxiliary feedwater pump (2-MS-201 and 2-MS-202), 2-MS-458 and.2-MS-459, are opened to drain water from.the steam supply lines. When either 2-MS-458 or 2-MS-459 is opened, a dedicated operator, in continuous communicetion with the control rooin, is required. Operation of these valves is expected during plant startup.

The containment station air header isolation, 2-SA-19, is openec t.,

supply. station air to containment. When 2-SA-19 is opened, a dedicated operator, in continuous cor..munication with the control room, is required.

Operation of this valve is only expected for maintenance activities inside containment.

The backup air supply master stop, 2-IA-566, is opened to supply t,ackup air to 2-CH-517, 2-CH-518, 2-CH-519, 2-EB-88, and 2-EB-89. When 2-IA-566 is opened, a dedicated operator, in continuous communication with the control room, is required. Operation of this valve is only expected in response to a loss of the normal air supply to the valves listed.

The nitrogen header drain valve, 2-51-045, is opened to depressurize the containment side of the nitrogen supply header stop valve, 2-SI-312.

When 2-51-045 is opened, a dedicated operator, in continuous communication with the control room, is required. Operation of this valve is only expected after using the high pressure nitrogen system to raise SIT nitrogen pressure.

The containment waste gas header test connection isolation valve, 2-GR-63, is opened to sample the primary drain tank for oxygen and nitrogen.

When 2-GR-63 is opened, a dedicated operator, in continuous communication with the control room, is required. Operation of this valve is expected during i

plant startup and shutdown.

The determination of the appropriate administrative controls for ihese containment isolation valves included an evaluation of the expected environmental conditions. This evaluation has concluded environmental conditions will not preclude access to close the valve, and this action will prevent the release of radioactivity outside of containment through the respective penetration.

The containment purge supply and exhaust isolation valves are required to be sea 4H closed during plant operation since +hese valves have not been I

demonstated capable of closing during a LOCA or steam line break accident.

Such a demonstration would require justification of the mechanical operability of the purge valves and consideration of the appropriateness of the electrical override circuits. Maintaining these valves closed during plant operations ensures that excessive quantities of radioactive materials will not be released via the containment purge system. The containment purge supply and exhaust isolation valves are sealed closed by reeving power from the valves.

This is accomplished by pulling the control power fuses for each of the valves. The associated fuse blocks are then locked.

This is consistent with the guidance contained in NUREG-0737 Item II.E.4.2 and Standard Review Plan 6.2.4, " Containment Isolation System," Item II.f.

MILLSTONE - UNIT 2 B 3/4 6-3d Amendment No. US 0300

n..,

5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area is shown on Figure 5.1-1.

LOW POPULATION ZONE 5.1.2 The low population zone is shown on Figure 5.1-2.

FLOOD CONTROL 5.1.3 The flood control provisions shall be designed and maintained in accordance with the design provisions contained in Section 2.5.4.2 l

of the FSAR.

5.2 CQNTAINNENT CONFIGURATION 5.2.1 The reactor containment building is a steel lined, reinforced concrete building of cylindrical shape, with a dome roaf and having the following design features:

a.

Nominal inside diameter - 130 feet.

b.

Nominal inside height - 175 feet.

c.

Minimum thickness of concrete walls - 3.75 feet, d.

Minimum thickness of concrete dome - 3.25 feet, e.

Minimum thickness of concrete floor pad - 8.5 feet, f.

Nominal thickness of steel liner - 0.25 inches, g.

Net free volume - 1.9 x 10' cubic feet.

NILLSTONE'- IINIT 2 5-1 Amendment No.

0345

DE51GN FEATURES 4

DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be maintained for a maximum internal pres. ore of 54 psig~ and an equilibrium liner temperature of 289'F.

PENEIRATIONS 5.2.3 Penetrations through the reactor containment building are designed and shall be maintained in accordance with the design provisions contained in l Section 5.2.8 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.

5.3 REACTOR CORE FUEL ASSEMBLIES j

5.3.1 The reactor core shall contain 217 fuel assemblies with each fuel assembly containing 176 rods. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 4.5 weight j

percent of U-235.

CONTROL ELEMENT ASSEMBLIES 1

5.3.2 The reactor core shall contain 73 full length and no part length control element assemblies.

The control element assemblies shall be designed and maintained in accordance witt the design provisions contained inl Section 3.0-of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.

l 5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE i

5.4.1 The reactor coolant system is designed and shall be maintained:

a.

In accordance with the code requirements specified in Section 4.2.2 of the FSAR with allowance for normal degradation pursuant of the applicable Surveillance Requirements, b.

For a pressure of 2500 psia, and c.

For a

temperature of 650*F except for the pressurizer which is 700*F.

i i

MILLSTONE - UNIT 2 5-4 Amendment No. 77, J77, JJJ 7pf,

o

(

nFt10.M FFATilart 5.7 SEISMIC CLASSIFICATION 5.7.1 Those structures, systems and components identified as Category I Items in Section 5.1.1 of the FSAR shall be designed and maintained to the design provisions contained in Section 5.8 of the FSAR with allowance for l normal degradation pursuant to the applicable Surveillance Requirements.

5.8 METEOROLOGICAL TOWER LOCATION 5.8.1 The meteorological tower location shall be as shown on Figure 5.1-1.

k MILLSTONE - UNIT 2 5-6 Amendment No. 7, 197, JJ7, 0347

-

ML20203D103

Text

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