Text

Amend 160 to License NPF-49,replacing Pressurizer Max Water Inventory Requirement W/Pressurizer Max Indicated Level Requirement
	ML20248A812
Person / Time
Site:	Millstone
Issue date:	05/27/1998
From:	Mckee P; NRC (Affiliation Not Assigned)
To:	;
Shared Package
ML20248A817	List: ML20248A812; ML20248A829;
References
	NUDOCS 9806010047
	Download: ML20248A812 (12)
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UNITED STATES g

g NUCLEAR REGULATORY COMMISSION e

WASHINGTON. D.C. - anni s.,...../

NORTHEAST NUCLEAR JNERGY COMPANY. ET AL.

DOCKET NO. 50-423 MILLSTONE NUCLEAR POWER STATION. UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 160 License No. NPF-49 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 April 7,1998, 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; 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 Commission's regulations; i

D.

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.

9806010047 980527 PDR ADOCK 05000423 P

PM U _ _ __-_ _____ _ ___ ___________- __-_ -__ _____

'. 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-49 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.160, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of issuance, to be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

-T y

Phillip F. McKee Deputy Director for Licensing Special Projects Office Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance: May 27,1998 l

4 ATTACHMENTTO LICENSE AMENDMENT NO. inn FACILITY OPERATING LICENSE NO. NPF-49 DOCKET NO. 50-423 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.

Etm2Ya insert vil vii x

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xiii xiii 3/4 4-11 3/4 4-11 3/4 4-11a 3/4 4-11b B 3/4 4-2 B 3/4 4-2 B 3/4 4-2a B 3/4 4-2a B 3/4 4-2b l

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18D11 a turvina enuntrinut rno nornartnu aun sinnvrti n auer nrnisterururt SECTION ggg TABLE 3.3-13 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION.....................

3/4 3-75 TABLE 4.3-9 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS........

3/4 3-78 3/4.3.4 TURBINE OVERSPEED PROTECTION..............

3/4 3-81 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation ;. J.~.... i. ~...... 3/4 4-l~

Hot Standby........................

3/4 4-2 Hot Shutdown 3/44-3 Cold Shutdown - Loops Filled 3/44-5 Cold Shutdown - Loops Not Filled 3/4 4-6 Isolated Loop....................... 3/44-7 Isol ated Loop Startup...................

3/4 4-8 3/4.4.2 SAFETY VALVES Shutdown 3/4 4-9 Operating........................ 3/44-10 3/4.4.3 PRESSURIZER Startup and Power Operation............... 3/44-11 FIGURE 3.4-5 PRESSURIZER LEVEL CONTROL.............. 3/4 4-lla Hot Standby....................... 3/4'4-Ilb 3/4.4.4 RELIEF VALVES...................... 3/4 4-12 3/4.4.5 STEAM GENERATORS 3/4 4-14 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION............... 3/44-19 l

TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION........... 3/44-20 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE b,

L'eakage' Debetion $ystems j ;.'..... c N. '.'. '.'... 3/4 4-21,';,

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Operational Leahge..... ;.......:...... 3/44-22 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES

.. 3/4 4-24 3/4.4.7 CHEMISTRY........................ 3/44-25 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS...

3/44-26 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS 3/44-27 3/4.4.8 SPECIFIC ACTIVITY...............'...... 3/44-28 NILLSTONE - UNIT 3 vii Amendment No. '160 om

IlEEK LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION Egg TABLE 3.7-3 STEAM LINE SAFETY VALVES PER LOOP 3/4 7-3 Auxiliary Feedwater System 3/4 7-4 Dominera11 zed Water Storage Tank 3/47-6

-Specific Activity -... r.~... r.-....... 3/47-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM 3/47-8 l

Main Steam Lina Isolation Valves 3/47-9 Steam Generator Atmospheric Relief Bypass Lines 3/4 7-9a l 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION 3/4 7-10 3/4.7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM 3/4 7-11 3/4.7,4 SERVICE WATER SYSTEM 3/4 7-12 3/4.7.5 ULTIMATE HEAT SINK 3/4 7-13 3/4.7.6 FLOOD PROTECTION.................._ 3/4 7-14 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM 3/4 7-15 3/4.7.8 CONTROL ROOM ENVELOPE PRESSURIZATION SYSTEM 3/4 7-18 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM 3/4 7-20 i

3/4.7.10 SNUBBERS 3/4 7-22 j

TABLE 4.7-2 SNUBBER VISUAL INSPECTION INTERVAL 3/47-27 FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST 3/47-29 3/4.7.11 SEALED SOURCE CONTAMINATION 3/47-30 3/4.7.12 DELETED

- Table 3.7-4 DELETED Table 3.7-5

"... DEL. ET.ED 3/4'.7.13

. DELETED.

I 3/4.7.14' AREA TEMPER'ATURE MONITORING '............ 3/47-32 TABLE 3.7-6 AREA TEMPERATURE MONITORING 3/47-33

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NILLSTONE - UNIT 3 x

Amendment No. 71, 77, JPP.160 ossa

g BASES SECTION E&gE 3/4.0 APPLICABILITY

....................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL..................... B 3/4 1-1 3/4.1.2 B0 RATION SYSTEMS..................... B 3/4 1-2 3/,4.1.3 MOVABLE, CONTROL ASSEMBLIES.............,..........B 3/.4_l-3a 3/4.2' POWER DISTRIBUTION LIMITS

................. B 3/4 2-1 3/4.2.1 AXIAL FLUX DIFFERENCE B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR B 3/4 2-3 3/4.2.4 QUADRANT POWER TILT RATIO B 3/4 2-5 3/4.2.5 DNB PARAMETERS...................... B 3/4 2-5 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION

.......... B 3/4 3-1

-3/4.3.3 MONITORING INSTRUMENTATION................ B 3/4 3-3 3/4.3.4 TURBINE OVERSPEED PROTECTION............... B 3/4 3-6 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION B 3/4 4-1 3/4.4.2 SAFETY VALVES B 3/4 4-2 3/4.4.3 PRESSURIZER B3/44-2 3/4.4.4 RELIEF VALVES B3/44-2bl 3/4.'4.5 STEAM GENERATORS..................... B 3/4 4-3 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE..... '........ B 3/4 4-4 l

3/4.4.7 CHEMISTRY........................ B3/44-5 3/4.4.8 SPECIFIC ACTIVITY B3/44-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS............... B 3/4 4-7 MILLSTONE - UNIT 3 xiii Amendment No. pp, pp.160 one j

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REACTOR COOLANT SYSTEM 3/4.4.3 PRES $URIZER STARTUP AND POWER OPERATION LIN! TING CONDITION FOR OPERATION 3.4.3.1 The pressurizer shall be OPERABLE with:

at least two groups of pressurizer heaters supplied by emergency a.

-power,- each having a capacity of at ^1 east 175-kW; and b.

water level maintained at programmed level +/-6% of full scale (Figure 3.4-5).

APPLICABILITY: N0 DES I and 2.

l ACTION:

With only one group of pressurizer heaters supplied by emergency power a.

OPERABLE, restore at least two groups to OPERABLE status 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 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With pressurizer water level outside the parameters described in Figure 3.4-5, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months restore programmed level to within +/- 6%

of full scale, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With the pressurizer othenvise inoperable, be in at least HOT STANDBY c.

with the Reactor Trip System breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS.

4.4.3.1.1 The pressurizer water level shall be verified to be within programmed i.

level +/- 6% of full scale at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.3.1.2 The capacity of each of the above required groups of pressurizer heaters supplied by emergency power shall be verified by energizing the heaters l

and measuring circuit current at least once each refueling interval.

NILL 5 TONE."- UNI 3

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'3/4 '4 AInend nt No. h97.16d MM

PRESSURIZER LEVEL CONTROL 70

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. MILL 5 TONE'- UNIT 3 S

3/4 4,11a

.Amehdient..No.160

, ma

REACTOR C0OLANT SYSTEN H0T STAMBY LIMITING CONDITION FOR OPERATION i

3.4.3.2 The pressurizer shall be OPERABLE with:

l at least two groups of pressurizer heaters supplied by emergency a.

power, each having a capacity of at least 175 kW; and b.-

water level-less than or' equal to 89% of full scale. -

APPLICABILITY: NODE 3 ACIIDli:

With only one group of pressurizer heaters supplied by emergency power a.

OPERABLE, restore at least two groups to OPERABLE status within i

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of being declared inoperable, or be in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

I b.

With the pressurizer otherwise inoperable, be in H0T SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

I i

SURVEILLANCE REQUIREMENTS 4.4.3.2.1 The pressurizer water level shall be determined to be less than or I

equal' to 89% of full scale at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

.4.4.3.2.2'The capacity of nach of.the above. requiredhroups of pressur.izer. -

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heaters supplied by emergency power shall'.be'-verified by inergizing l

' and measuring circuit current.at least once each refueling interval.the heaters..-

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NILI.$51'0NE[ 'tiNIT 3-

'3/4 4-11'b Amendsent No.160,

. 0586

REACTOR C00UMT $YSTEM BASES 3/4.4.2 SAFETY VALVES The pressurizer Code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia. Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam at the valve Setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.

In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. -In addition-the Cold Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures.

During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2750 psia. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss-of-load assuming no Reactor trip until the first Reactor Trip System Trip Setpoint is reached (i.e., no credit is taken for a direct Reactor trip on the loss-of-load) and also assuming no operation of the power-operated relief valves or steam dump valves.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.

3/4.4.3 PRESSURIZER The pressurizer provides a point in the RCS when liquid and vapor are maintained in equilibrium under saturated conditions for pressure contro1 purposes to prevent bulk boiling in the remainder of the RCS. Key functions include maintaining required primary system pressure during steady state operation and limiting the pressure changes caused by reactor coolant thermal expansion and contraction during load transients.

MODES 1 AND 2 The requirement for the pressurizer to be OPERABLE, with pressurizer level maintained at programmed level.withip i 6% of full scale is.consis. tent with the accident analysis in. Chapter.15 of the.FSAR..The accident analysis assumes that

.- ' l... ~ pressurizer. level is' being maintained at the.>rogrammed.levil by the'.abtioimatic J contr'olf system, and when in manual control similar~ 1imits are established. The programmed level ensures the capability to establish and ' aintain pressure m

control for steady state operation and to minimize the consequences of potential overpressure and pressurizer overfill transients. A pressurizer level control error based upon automatic level control has been taken into account for those transients where pressurizer overfill is a concern (e.g., loss of feedwater, z

feedwater line break, and inadvertent ECCS actuation.at powar. When in manual control, the goal is to maintain pressurizer level at the prog) ram level value.

The 16 % of full scale acceptance criterion in the Techn.ical Specification establishes a band for operation to accommodate variations between level l

measurements /, This v.alue. is bounded,by.t.he mapgin app, lied to the pressurizer

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MILLSTONE.-UNIT 3 B3/4'4-2 I Amendment No.160

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O REACTOR C0OLANT SYSTEN BASES 3/4.4.3 PRESSURIZER fcont'd.)

The 12-hour periodic surveillance require that pressurizer level be maintained at programmed level within i 6% of full scale. The surveillance is performed by observing the indicated level. The 12-hour interval has been shown by operating practice to be sufficient to regularly assess level for any deviation and to ensure that the appropriate level exists in the pressurizer.

During transitory conditions, i.e., power changes, the operators will maintain programmed level, and deviations greater than 6% will be corrected within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Two hours has been selected for pressurizer level restoration after a transient to avoid an unnecessary downpower with pressurizer' level outside~ the~

operating band. Normally, alarms are also available for early detection of abnormal level indications.

Electrical immersion heaters, located in the lower section of the pressurizer vessel, keep the water in the pressurizer at saturation temperature and maintain a constant operating pressure. A minimum required available capacity of pressurizer heaters ensures that the RCS pressure can be maintained.

The capability to maintain and control system pressure is important for maintaining subcooled conditions in the RCS and ensuring the capability to remove core decay heat by either forced or natural circulation of the reactor coolant. Unless adequate heater capacity is available, the hot high-pressure condition cannot be maintained indefinitely and still provide the required subcooling margin in the primary system.

Inability to control the system pressure and maintain subcooling under conditions of natural circulation flow in the primary system could lead to a loss of single-phase natural circulation and decreased capability to remove core decay heat.

The LCO requires two groups of OPERABLE pressurizer heaters, each with a capacity of at least 175 kW, capable of being powered from either the offsite power source or the emergency power supply. The minimum heater capacity required is sufficient to maintain the RCS near normal operating pressure when accounting for heat losses through the pressurizer insulation. By maintaining the pressure near the operating conditions, a wide margin to subcooling can be l

obtained in the loops. The emergency power supply requirements for the heaters provides assurance that the heaters can be energized during a loss of offsite power condition to maintain natural circulation at HOT STANDBY.

. If one required group of pressurizer. heaters is inoperable, restoration is

' required within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .. The Completion. Time of'72. hours is reasonable.

'considering tha't a demand. caused:6y lo'ss of.c'ffsite power wo.uld be unlikel'y in

. this time period.

Pressure control may be maintained during this time using

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normal station powered heaters.

MODE 3 The requirement for the pressurizer to be OPERABLE, with a level less than or equal to 89%, ensures that a steam bubble exists. The 89% lavel preserves the steam space for pressure control. The 89% level has been established to ensure the capability to establish and maintain pressure tontrol for MODE 3 and to ensure a bubble is present in the pressurizer.

Initial pressurizer level is.

. f...

not.signif,1 cant.for those. events. analyzed for: MODE 3 in. Chapter 15.cf.tha:ESAR.

s MILLSTONE - UNIT 3 8 3/4 4:2a Amendment No.160 osos

REACTOR COOLANT SYSTEN BASES 3/4.4.3 PRESSURIZER (cont'd.)

The 12-hour periodic surveillance requires that during MODE 3 operation, pressurizer level is maintained below the nominal upper limit to provide a minimum space for a steam bubble. The surveillance is performed by observing the indicated level. The 12-hour interval has been shown by operating practice to be sufficient to regularly assess level for any deviation and to ensure that a steam bubble exists in the pressurizer. Alarms are also available for early detection of abnormal level indications.

The basis for the pressurizer heater requirements is identical to MODES I and 2.

3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer Code safety valves.

Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable.

Requiring the PORVs to be OPERABLE ensures that the capability for depressurization during safety grade cold shutdown is met.

Action statements a, b, and c distinguishes the inoperability of the power operated relief valves (PORV).

Specifically, a PORV may be designated inoperable but it may be able to manually open and close and therefore, able to perform its function.

PORV inoperability may be due to seat leakage, instrumentation problems, automatic control problems, or other causes that do not prevent manual use and do not create a possibility for a small-break LOCA.

For these reasons, the block valve may be closed but the action requires power to be maintained to the valve. This allows quick access to the PORV for pressure control. On the other hand if a PORV is inoperable and not capable' of being manually cycled, it must be either restored or isolated by closing the associated block valve and removing power.-

The prime importance for the capability to close the block valve is to isolate a stuck-open PORV. Therefore, if the block valve (s) cannot be restored to operable status within I hour, the remedial action is to place the PORV in manual control (i.e. the control switch in the "CLOSE" position) to preclude its automatic opening for an overpressure event and to avoid the potential of a stuck-open PORV at a time that the block valve is inoperable. The time allowed to restore the block valve (s) to operable status is based upon the remedial action time limits for inoperable PORV per ACTION requirements b and c.

These actions do not specify closure of the block valves because such action would not likely be possible when the block valve is inoperable.

NILLSTONE - UNIT 3 8 3/4 4-2b Amendment No.160 osos

ML20248A812

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