Text

Proposed Tech Specs 3/4.4.3,replacing Pressurizer Max Water Inventory Requirement W/Pressurizer Level Requirement
	ML20216A890
Person / Time
Site:	Millstone
Issue date:	04/07/1998
From:	; NORTHEAST NUCLEAR ENERGY CO.
To:	;
Shared Package
ML20216A884	List: B17028, Application for Amend to License NPF-49,revising TS 3/4.4.3 by Replacing Pressurizer Max Water Inventory Requirements W/ Pressurizer Level Requirement; ML20216A890;
References
	NUDOCS 9804130316
	Download: ML20216A890 (29)
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,g,cc g. 4-r f F-//4, magy JAN 31 1986 IEEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.3-13 RADI0 ACTIVE GASE0US EFFLUENT fR*ITORING INSTRUNENTATION..........................................

3/4 3-75 TABLE 4.3-9 RADI0 ACTIVE GASE005 EFFLUENT IGNITORING INSTRUNENTATION SURVEILLANCE REQUIREDENTS...............

3/4 3-78 1/4. 3. 4 TbdINE OVERSPEED PROTECTION..............................

3/4 3-81 3/4.4 R'EACTOR COOLANT SYSTEN 3/4.4.1 REACTOR C0OLANT LOOPS AND COOLANT CIRCULATION Startup and Power 0peration..............................

3/4 4-1 Hot Standby..............................................

3/4 4-2 Hot Shutdown.............................................

3/4 4-3 Cold Shutdown - Leops F111ed.............................

3/4 4-5 Col d shutdown - Loops Not F111ed.........................

3/4 4-6 Iso]atedLoop............................................

3/4 4-7 Isolated Loop Startup....................................

3/4 4-8 3/4.4.2 SAFETY VALVES Shutdown.................................................

3/4 4-9 0perating................................................

3/4 4-10 3/4.4.3 PRESSURIZER..............................................

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v4.4.4 REu EF VALVES............................................

3/4 4-2 3/4.4.5 STFAN GENERATORS.........................................

3/4 4-14 TABLE 4.4-1 MINIMUN MitBER OF STEAM GENERATORS 10 SE IMSPECTED DURING INSERVICE INSPECTION.............................

3/4 4-19 TABLE 4.4-2 STEAM GENERA'f0R TLSE INSPECT 10N.......................

3/4 4-20 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................................

3/4 4-21 Operational Leakage......................................

3/4 4-22 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES......

3/4 4-24 i i 1

3/4.4.7 CHEMISTRY................................................

3/4 4-25

i TABLE 3.4-2 REACTOR C00LANT SYSTEM CHEMISTRY LIMIT 5...............

3/4 4-26 i

TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS.............................................

3/4 4-27 3/4.4.8 SPECIFIC ACTIVITY........................................

3/4 4-28 9804130316 980407 PDR ADOCK 05000423 P

pgg MILLSTONE - UNIT 3 v11

January 3,1995 LINITING CONDf710NS FOR OPERATION Am SURVEILLANCE REOUIRENENTS 4

EG10N 2AE TABLE 3.7 STEAM LINE SATETY VALVES PER LOOP 3/4 7-3 Auxiliary Feedwater System 3/4 7-4 buineralized Water Storage Tank.......... 3/4 7-6 Specific Activity

................. 3/47-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEN SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM................ 3/4 7-8

. Main Steam Line Isolation Valves 3/4 7-9

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3/4.7.2

$ TEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION 3/4 7-10 3/4.7.3 REACT 0P PLANT COMP 0NENT COOLING WATER SYSTEM 3/4 7-11

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3/4.7.4 SERVICE MATE SYSTEM................ 3/47-12 3/4.7.5 ULTIMATE HEAT SIIK................. 3/4 7-13 3/4.7.6 FLOOD PROTECTION.................. 3/47-14 3/4.7.7 CONTROL R00M EMERGENCY VENTILATION SYSTEM 3/4 7-15 3/4.7.8 CONTROL R00N ENVELOPE PRESSURIZATION SYSTEM 3/4 7-1B 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM 3/4 7-20 3/4.7.10 SNUBBERS

.................. 3/4 7-22 TABLE 4.7-2 SNUBBER VISUAL INSPECTION INTERVAL 3/4 7-27 1

FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST..... 3/4 7-29 1

3/4.7.11 SEALED SOURCE CONTAMINATION 3/4 7-30 3/4.7.12 DELETED Table 3.7-4 DELETED Table 3.7-5 DELETED 3/4.7.13 DELETED 3/4.7.14 AREA TEMPERATURE MONITORING 3/4 7-32 TABLE 3.7-6 AREA TEMPERATURE ETORING 3/4 7-33 y;yy ccartrre&

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1 RILLSTONE - ISIIT 3 x

Amendment No. R. M.100 eum

IgL1 03/24/94

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RAsrs EGII2f f.EE 2/4.0 APPLICABIL1TY...............................................

B 3/4 0-1 3/.,4.1 REAtTIv1TY CONTROL SYSTEMS 3/4.1.1 BORATION CONTR0L.......................................... B3/41-1 3/4.1.1 BORATION SYSTEMS.......................................... B3/41-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES................................

B 3/4 1-3a l

3/4.2 POWER DI STRI BUT ION L IMITS................................... B 3/4 2-1 3/4.2.1 AXIAL FLUX DIFFERENCE.....................................

B 3/4 2-1 l

3/4.2.2 and3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE NOT CHANNEL FACTOR.........

B 3/4 2 3 3/4.2.4 QUADRANT POWER TILT RATI0.................................

B 3/4 2-5

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I 3/4.1.5 DNB PARAMETERS............................................

B 3/4 2-5 3/4.3 INSTRUMENTATION

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3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUKENTATION...........................................

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 l.

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............................................. B3/44-2 3/4.4.3 PRESSURIZER...............................................

B3/44-2 B3/44-2h 3/4.4.4 RELIEF VALVES.............................................

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.T m m,A m S..........................................

B 3,4 4 3 3/4.4.s REACTOR COOLANT SYSTEM LEAKAGE............................

B3/44-4 3/4.4.7 CHEMISTRY.................................................

B 3/4 4-5

~ ~3/4.4.5 SPECIFIC ACTIVITY.........................................

B 3/4 4-5 3/4.4.9 PRES SURE/ TEMPERATURE LIMITS............................... B 3/4 4-7 MILLSTONE - UNIT 3 xiii Amendment No. S. 89 9141 m

January 3,1995 REACTOR C00LANL SYSTEN 3/4.4.3 PRESSURIZER frw m Wu pr7M LINITING CONDITION FOR OPERATION e,,

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dqRABLE with 4 water [el ef ler:

3.4.3 The pressurizer shall be OP th: er supplied by emergency power ea)ch having a capacity of at least 175 at least two groups of pressurizer heaters APPLICABILITY: MODES ACTION:

a.

With only one group of pressurizer heaters supplied by emergency power 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 ::d in '

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"^T "**^.' within the fell;.;;;g S i::n.

4-r With the pressurizer otherwise 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 :.ad ta liOT 4

6"m"" eithin the fell =ini5 i::r:.

l SURVEILLANCE REQUIREMENTS 5g e

4.4.3.

I The pressurizer waterJ.

shall bef:5!.i';td to be withinh\\ ggg.

If=it arleast once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

h 4.4.

The capacity of each of the above required grodps of pressurizer-heaters supplied by emergency power shall be verified by energizing the heaters and measuring circuit current at least once each refueling interval.

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MILLSTONE - LMIT 3 3/4 4-11 Amendment No. 100 4347

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l REACTOR COOLANT SYSTEM January 3,1995 3/S.4.3""rEE""!!""Q LINITING CONDITION FOR OPERATION

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,1 J s k a lu a/ h r % k 'ldO pys q ;yi ws 3.4.3 The pressurizer shall be OPERABLE wit enerMT ~ e? a+F+"a

^^" *^ "'" (1555 rdt: f::t), i:D wo groups of pressurize heaters suppliedbyemergencypowereachha$atleast a-i ing a ~ capacity of at least 175 k

ud APPLICABILITY: MODk2,__

3.

ACTION:

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

OPERABLE, restore at least two groups to OPERABLE status within Q* Uggj 72 hou or be in et lert "a? n?"a"Y titM: 'l: as;t 6 ;,w. ; r.d in '

' HOT SH within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ingaa 6 k.

5# M b.

With the pressurizer otherwise ino t:ith tt: ";;.;t:r Trip Sy:t= bre"perable, be in at leirt HOT STANBBV-

"; L*"; within the f:ll::ing S 5;;r:. ?'m within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ::d i: "^T tre r 3*

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SURVEILLANCE REhu!REMENTS cutc 3A

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. The pressurizer water vohns shall be determined to behi+lin itt

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least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.3 The capacity of each of the above required grotips of pressurizer-heaters supplied by emergency power shall be verified by energizing the heaters and measuring circuit current at least once each refueling interval.

l 4

MILLSTONE - INIT 3 3/44-iip Amendment No. 100 0247

9 REACTOR COOLANT SYSTEM JAN 31 1986 '

BASES 3/4.4.2 SAFETY VALVES j

.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 stesis 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 i

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 cgainst RCS overpressurf:ation at low temperaturt

- 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 cperation of the power-operated relief valves or steam dump valves.

ochnr ilh dur1Nh W I"

' 3 Ji c Demonstration.ofthe:safetyvalves'-lift.httings i

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??"'%ffthe: ASME Soller and Pressure Code-' ' shutdown.and will: be performedlin accordance U

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0 3/4.4.3 PRESSURIZER 7_ ___

The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the normal steady-state envelope of operation assumed in the SAR. The Itmit is consistent with the initial SAR a,ssumptions. '

The 12-hour periodic surveillance is ' sufficient to ensure that--therparameter m

is restored to within its limit following expected transient operation. The maximum water volume also ensures th t steam bubble is formed and thus the RCS is not a hydraulically solid systra. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant t control Reactor Coolant System pressure and establish natural circulation.,

n a 3/4.4.4 RELIE VAL ES

.:r>v' M (c The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the dasign step load decrease with steam dump. Operation of the PORVs minialzes the undesirable opening of the spring-loaded pressurizer Code safety valves.

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

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

MILLSTONE - Oh!T 3 B 3/4 4-2

1 l..

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

insert B 4

b.

With pressurizer water level not maintained within programmed level, +/-

6% of full scale, 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 .

Insert C 3/4.4.3 PRESSURIZER The pressurizer provides a point in the RCS where liquid and vapor are maintained in equilibrium under saturated conditions for pressure control 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.

1 MODES 1 AND 2 l

The requirement for the pressurizer to be OPERABLE, with pressurizer level l

maintained at programmed level within +/- 6% of full scale is consistent with the accident analysis in Chapter 15 of the FSAR. The accident analysis assumes that pressurizer ievelis being maintained at the programmed level by the automatic control system and when in manual control similar limits are established. The programmed level ensures the capability to establish and maintain pressure 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

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into account for those transients where pressurizer overfill is a concern (e.g., loss of feedwater, feedwater line break and inadvertent ECCS actuation at power).

When in manual control, the goal is to maintain pressurizer level at the program level value.The 16 % of full scale acceptance criterion in the Technical Specification establishes a band for operation to accomodate variations between level measurements. This value is bounded by the margin applied to the pressurizer overfill events.

l The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance requires that pressurizer level be maintained at j

programmed level within +/- 6% of full scale. The surveillance is performed by observing the indicated level. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown oy operating practice to be sufficient to regularly assess level for any deviation and to ensure J

l 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 . 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 reactor coolant. Unless I

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 j

l power source or the emergency power supply. The minimum heater capacity I

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 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 l

within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is reasonable considering that a demand caused by loss of offsite power would be unlikely in this time period. Pressure control may be maintained during this time using normal station powered heaters.

l I

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% level preserves the steam space for pressure control. The 89% level has been established to ensure the capability to establish and maintain pressure control for MODE 3 and to ensure a bubble is present in the pressurizer. Initial pressurizer level is not significant for those events analyzed for MODE 3 in Chapter 15 of the FSAR.

I 1

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 abnormallevelindications.

l The basis for the pressurizer heater requirements is identical to Modes 1 and 2.

l l

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Docket No. 50-423 817028 5

l Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specification Pressurizer Level (TSCR 3-10-98)

Retvoed Paaes I

1 April 1998

U.S. Nuclear Regulttory Commission B17028%ttachment 3\\Page1 RETYPE OF PROPOSED REVISION l

Refer to the attached retype of the proposed revision to the Technical Specifications.

The attached retype reflects the currently issued version of the Technical l

Specifications. Pending Technical Specification revisions or Technical Specification revisions issued subsequent to this submittal are not reflected in the enclosed retype.

The enclosed retype should be checked for continuity with Technical Specifications l

prior to issuance.

I l

INDEX s turTrue enuntTinut rna nornartnu aun sinnvrri t auer nrnaiteruruTt SECTION EaEE TABLE 3.3-13 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION.....................

3/4 3-75 TABLE 4.3-9 RADI0 ACTIVE GASE0US 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................

3/4 4-1 Hot Standby........................

3/4 4-2 Hot Shutdown 3/4 4-3 Cold Shutdown - Loops Filled 3/4 4-5 Cold Shutdown - Loops Not Filled 3/4 4-6 Isolated Loop.......................

3/4 4-7 Isolated Loop Startup................... 3/4 4-8 3/4.4.2 SAFETY VALVES Shutdown 3/4 4-9 Operating........................

3/4 4-10 3/4.4.3 PRESSURIZER Startup and Power Operation...............

3/4 4-11 FIGURE 3.4-5 PRESSURIZER LEVEL CONTROL..............

3/4 4-11a Hot Standby.......................

3/4 4-11b 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............... 3/4 4-19 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION 3/4 4-20 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................ 3/4 4-21 Operational Leakage...................

3/4 4-22 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES 3/4 4-24 3/4.4.7 CHEMIS1RY........................ 3/4 4-25 i

TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS 3/4 4-26 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS 3/4 4-27 3/4.4.8 SPECIFIC ACTIVITY.................... 3/4 4-28 MILLSTONE - UNIT 3 vii Amendment No.

0502 l

4 IlE.El LIMITING C00EITIONS FOR OPERATION Als SURVEILLANCE REQUIRENENTS SECTION MGE TABLE 3.7-3 STEAM LINE SAFETY VALVES PER LOOP 3/4 7-3 Auxiliary Feedwater System 3/4 7-4 Domineralized Water Storage Tank 3/4 7-6 Specific Activity 3/4 7-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM 3/4 7-8 Main Steam Line Isolation Valves 3/4 7-9 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 3/4.7.10 SNUBBERS 3/4 7-22 TABLE 4.7-2 SNUBBER VISUAL INSPECTION INTERVAL 3/4 7-27 FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST 3/4 7-29 3/4.7.11 SEALED SOURCE CONTAMINATION 3/4 7-30 3/4.7.12 DELETED Table 3.7-4 DELETED Table 3.7-5 DELETED 3/4.7.13 DELETED 3/4.7.14 AREA TEMPERATURE MONITORING 3/4 7-32 TABLE 3.7-6 AREA TEMPERATURE MONITORING 3/4 7-33 NILLSTONE - UNIT 3 x

Amendment No. JJ, pp, MP, 0603

INDEX BASES SECTION E6GE 3/4.0 APPLICABILITY B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL.....................

B 3/4 1-1 3/4.1.2 BORATION SYSTEMS.....................

B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES................ B 3/4 1-3a 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 D 3/4 4-1 3/4.4.2 SAFETY VALVES B 3/4 4-2 3/4.4.3 PRESSURIZER B 3/4 4-2 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 3/4.4.7 CHEMISTRY B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS B 3/4 4-7 I

MILLSTONE - UNIT 3 xiii Amendment No. pp, pp, 0584

REACIPR_C00LAML1Y31EM 3/4.4.3 PRESSUR11EB STARTUP AND POWER OPERATION LINITING 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 least 175 kW; and b.

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

APPLICABILITY: MODES 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 , c.

With the pressurizer otherwise inoperable, be in at least HOT STANDBY 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 REQUIRENENTS 4.4.3.1.1 The pressurizer water level shall be verified to be within programmed 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.

MILLSTONE - UNIT 3 3/4 4-11 Amendment No. Jpp, 06M

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PRESSURIZER LEVEL CONTROL 70

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0 551 557 562 567 572 577 582 587.1 591.1 T (AVG) l l

l FIGURE 3.4-5 l

NILLSTONE - UNIT 3 3/4 4-11a Amendment No.

oses l

l l

l

REACTOR C0OLANT SYSTEN HOT STAMBY LIMITING COMITION FOR OPERATION 3.4.3.2 The pressurizer shall be OPERABLE with:

a.

at least two groups of pressurizer heaters supplied by emergency power, each having a capacity of at least 175 kW; and b.

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

APPLICABILITY: MODE 3 ACTION:

a.

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

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

l SURVEILLANCE REQUIRENENTS 1

4.4.3.2.1 The pressurizer water level shall be determined to be less than or 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 each of the above required groups of pressurizer l

heaters supplied by emergency power shall be verified by energizing the heaters and measuring circuit current at least once each refueling interval.

NILLSTONE - UNIT 3 3/4 4-11b Amendment No.

l om i

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4 REACTOR C00LANT SYSTEN 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 s' Nown and will be performed in accordance with the provisions of Section XI

o. 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 control purposes to prevent bulk boiling in the remainder of the RCS. Key functions y

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 i

The requirement for the pressurizer to be OPERABLE, with pressurizer level i

maintained at programmed level within 6% of full scale is consistent with the accident analysis in Chapter 15 of the FSAR. The accident analysis assumes that i

pressurizer level is being maintained at the programmed level by the automatic control system, and when in manual control, similar limits are established. The i

programmed level ensures the capability to establish and maintain pressure control for steady state operation and to minimize the consequences of potential

- overpressure and pressur*zer 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, feedwater line break, and inadvertent ECCS actuation at power). When in manual control, the goal is to maintain pressurizer level at-the program level value.

The i6 % of full scale acceptance criterion in the Technical Specification establishes a band for operation to accommodate variations between level measurements. This value is bounded by the margin applied to the pressurizer overfill events.

NILLSTONE - L5IIT 3 53/44-2 Amendment No.

osa

REACTOR COOLANT SYSTEM BASES 3/4.4.3 PRESSURIZER fcont'd.)

The 12-hour periodic surveillances require that pressurizer level be maintained at programmed level within 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 i

programmed level, and deviations greater than 6% will be corrected within i

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 I

abnormal level indications.

Electrical immersion heaters, located in the lower section of the t

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 ubcooled conditions in the RCS and ensuring the capability to remove core may heat by either forced or natural circulation of the reactor coolant. L ss adequate heate; capacity is available, the hot high-pressure condition ca..10t be maintained indefinitely and still provide the required I

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 LC0 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 l

the pressure near the operating conditions, a wide margin to subcooling can be

{

l obtair.ed in the loops.

The emergency power supply requirements for the heaters J

provides assurance that the heaters can be energized during a loss of offsite power condition to maintain natural circulation at HOT STANDBY.

I 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 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is reasonable considering that a demand caused he loss of offsite power would be unlikely in this time period.

Pressure contro' aay be maintained during this time using 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% level preserves the steam space for pressure control. The 89% level has been established to ensure the capability to establish and maintain pressure control for MODE 3 and to ensure a brbble is present in the pressurizer.

Initial pressurizer level is not significant for those events analyzed for MODE 3 in Chapter 15 of the FSAR.

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

0586

REACTOR COOLANT SYSTEN

-BASES 3/4.4.3 PRESSURIZER fcont'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 1 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.

1 1

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NILLSTONE - UNIT 3 B 3/4 4-2b Amendment No.

osee

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Docket No. 50-423 I

B17028 l

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i Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specification l

Pressurizer Level (TSCR 3-10-98) l Backaround and Safety Assessment l

l l

April 1998

... +

U.S. Nuclear Regul: tory Commission B17028\\ Attachment 4\\Page1, 1

Backcround The proposed change to Technical Specircation 3/4.4.3, Pressurizer, replaces the pressurizer maamum water inventory requirement with a pressurizer level requirement. The Bases section is also being modified.

The exishng Technical Speedication has been revised to reflect the values required to support operation in Modes 1, 2, and 3.

Since the purpose and requirements for pressurizer level are signdicantly ddlerent between Modes 1/2 and Mode 3, the Technical Speedication has been segmented to reflect this fact and make them easier to use.

The change to the basis clanfies the pressurizer level requirement for the various modes of plant operation. In Modes 1 and 2 the values presented represent the values used in the accident analysis whk:h are maintained either by automatic or procedural controls. These controls maintan pressurizer level consistent with the Safety Analysis Report The value shown for Mode 3 is nearelatad with confirming the existence of a bubble in the pressunzer. FSAR Sechon 15.5.1 Inadvertent Operation of the Emergency Core Cooling System During Power Operation is also being revised to clearly identify that this event is not analyzed for Mode 3.

The proposed Technmal Specification change is necessary for the Technical Specircations to reflect the. FSAR Chapter 15 initial condition assumptions. The basis for the current Technical Speedication is to assure the presence of a steam bubble in Modes 1,2 and 3 and is not directly tied to the initial conditions assumed in the accident analysis presented in Chapter 15. The current Technical Specifcation is only related to the accident analysis assumphons in that by assuming the programmed level as an initial condition, it also obviously assumes the presence of a steam bubble. However, the current wording in the basis of this Technical Specification is very misleading in that it states "the limit is consisterd with the initial SAR assumptions."

in order to correct the Todmical Specircation basis, NNECO has submitted a proposed Techncal Specircation change to the NRC via a "B16624, Application for Amend to License NPF-49,proposing Changes That Will Affect Nominal Trip Setpoints & Allowable Values. Proprietary Rev 5 to WCAP-10991 & non-proprietary Rev 5 to WCAP-10992,encl.Proprietary Rept Withheld,Per [[CFR" contains a listed "[" character as part of the property label and has therefore been classified as invalid..790|letter dated October 15,1997]], and clarified the submittal via a letter of December 17,1997. The NRC's letter of February 13, 1998, indcated that the proposed change was unacceptable because it did not address Criterion 2 of 10CFR50.36(c)(2)(ii)(B), in that what was proposed was not a process vanable that is an initial condition of a design basis accident or transient analysis._ The NRC requested NNECO to resubmit an LCO for pressurizer level that meets the above stated regulation. Thus, the Technical Specification is being changed to accurately reflect the Chapter 15 accident analysis initial condition assumptions with respect to pressurizer level.

l

U.S. Nucl:ar Regulatory Commission

. B17028\\ Attachment 4\\Page 2 SAFETYASSESSMENT The proposed changes are being made to ensure that the Technical Specification for pressunzer level is consistent with the initial condition assumed in the accident analysis. The Chapter 15 FSAR accident analysis assumes that pressurizer level is being maintained at the programmed level. For most of the accident analysis, pressunzer level is assumed to be at 61.5% for power conditions and 28% for hot zero power.

For events where pressurizer level overfill is a concem, initial pressunzer level is assumed to be 6% above the nominal value of 61.5% at full power. This bounds the automatic control system uncertainty as documented in WCAP 14353 Thus, the proposed Technical SpedfE4;en LCO for Modes 1 and 2 Is consistent with the Chapter 15 FSAR eMant analysis. When pressurizer level is j

being maintained by manual operator action, the 6% band is an operating band This band is mnsistent with the 6% error assumed for the pressurizer overfill events, but it does not take into acmunt instrument uncertainty.

Current operating procedures require that automatic pressurizer level control be in service in Modes 1 and 2 and manual control is used only when automatic control is not available Because of the infrequent use of manual operation, combined with i

the multiple main board indications and the randomness associated with instrumentation uncertainty, it is unnecessary to apply instrument uncertainty effects

)

on top of the operating band As such, the 6% band is bounded by the current i

Chapter 15 FSAR analysis. Thus, it is concluded that the proposed Technical j

Specification is consistent with the analysis assumptions.

In addition, an evaluation has been performed for those events analyzed in Chapter 15 for Mode 3. The only accident analysis provided in Chapter 15 of the FSAR for Mode 3 is the boron dilution event. Pressurizer level has no impact on the results.

As stated in the evaluation, the other events either would not occur, or the plant response would be extremely slow or not meaningful without power generation. For the inadvertent ECCS actuation event, it is evaluated at power only and an inadvertent ECCS actuation in Mode 3 is not considered as part of the licensing basis. Thus, the current specification which assures that a steam bubble exists in Mode 3 is sufficient to ensure consistency with the accident analysis assumptions.

The proposed changes to the Technical Specification provide added assurance that pressunzer level will be maintained at the required value. Since the automatic control system maintains pressunzer level within 6% of the programmed level, the change in the Technical Specification just reflects current operating practice. In manual operation, the proposed Technical Specification requires the operator maintain pressunzer level within a 6% band around programmed level. This is more restrictive than the current Technical Specification upper limit of 92% by volume. A two hour restriction on operation with pressurizer level not within +/- 6% (full scale) of programmed level has been added. This provides added assurance that

U.S. Nuclear Regul tory Commission B17028%ttachment 4\\Page 3 the required pressurizer level assumed in the accident analysis is maintained. Thus, the changes reduce the likelihood of inadequate level control.

I i

.o Docket No. 50-423 B17028 1

l Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specification Pressurizer Level (TSCR 3-10-98)

Sionificant Hazards Consideration and Environmental Considerations l.

l April 1998 L

U.S. Nuclear Regul tory Commission f

B17028\\ Attachment 5\\Page 1 Sionificant Hazards Consideration i

Northeast Nuclear Energy Company (NNECO) has reviewed the proposed revision in accordance with 10CFR50.92 and has concluded that the revision does not involve a significant hazards consideration (SHC). The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not satisfied. The proposed revision does not involve a SHC because the revision would not:

1.

Involve a significant increase in the probability or consequence of an accident

{

j previously evaluated.

l l

The proposed Tedinical Speedication provides added restrictions on pressunzer l

level to ensure that the pressunzer will not overfill or empty in a transient and that l

RCS pressure control will be maintained The proposed Technical Specification l

requires pressunzer level to be maintained at the programmed level.

The programmed level is a curve that varies linearly from 28% at no load T to 61.5% at full power T This is more restrictive than the current upper limit of 92% of volume and provides added assurance that pressurizer overfill will not occur for those events where prevention of overfill is a criterion and that the pressurizer would not empty due to a transient in addition, it assures that there is enough steam spece available to prevent RCS overpressunzation in a transient This requirement also applies to manual opershon to ensure that pressurizer level is maintained in a band around the programmed level of +/- 6 % of full scale. A two hour restriction on operation with pressurizer level not within programmed level +/- 6 % of full scale has been added. - This will provide added assurance that operator error in pressunzer level control will not result in a transient Based on the above, the changes do not negatively impact the probability of occurrence of the previously evaluated accidents.

For Modes 1 and 2, the Chapter 15 FSAR accident analysis assumes that pressunzer level is being maintained by the automatic control system at the programmed level. For most of the accident analysis, pressurizer level is assumed j

to be at 61.5% for power conditions and 28% for hot zero power. For events where pressunzer level overfill is a concem, initial pressunzer level is assumed to be 6%

above the nominal value of 61.5% at full power. This bounds the automatic control system uncertainty as documented in WCAP 14353. Thus, the proposed Technical Spec;fication LCO for Modes 1 and 2 is consistent with the Chapter 15 FSAR l

accident analysis. When pressurizer level is being maintained by manual operator j

action, a 6% operating band is specified. This band is consistent with the 6% error j

assumed for the pressurizer overfill events, but it does not take into account instrument uncertainty. Because of the infrequent use of manual operation combined with the multiple main board indications and the randomness associated with instrumentation uncertainty, it is unnecessary to apply instrument uncertainty effects on top of the operating band As such, the 6% band is bounded by the I

U.S. Nuclear Regul: tory Commission B17028\\ Attachment 5\\Page 2 current Chapter 15 FSAR analysis.. Thus, it is concluded that the proposed Technical Speedicahon is consistent with analysis assumptions.

W:th regard to Mode 3 operation, an evaluation has been performed for those events analyzed in Chapter 15 for Mode 3. The only acculent analysis provided in Chapter 15 of the FSAR for Mode 3 is the boron dilubon event Pressurizer level has no irnpact on the results. As stated in the evaluation, the other events either would not occur, or the plant response would be extremely slow or not meaningful WWWt Power generation For inadwtont Operation of ECCS that increases Reactor Coolant inventory, the MP3 FSAR Sechon 15.5.1 clearly identifies this transient as an event evaluated at Power Operabon This is consistent with SRP Section 15.5.1-15.5.2 where the initial power condition is specified as the licensed core thermal power with allowance for measurement uncertainty.

Thus, the current licensing basis does not require analysis of this event for the shutdown modes, including Modes 3 and 4.

Thus, the current specification which assures that a steam bubble exists in Mode 3 is sufficient is to ensure consistency with the accident analysis assumptions.

Therefore, the proposed revision does not involve a significant increase in the probability or consequence of an accident previously evaluated.

2.

Create the possibility of a new or different kind of accident from any accident previously evaluated.

The Technical Specification changes provide tighter restrictions on pressurizer level to ensure that pressurizer level will be controlled as intended The Bases change J

better reflects what assures the validity of the accident analyses assumptions and the bases for the maximum level.

A two hour restriction on operation with pressurizer level not within +/- 6 % (full scale) has been added. This provides added assurance that pressurizer level will be maintained consistent with the acx:ident analysis initial condition assumption.

The changes provide added assurance that RCS pressure control will be maintained and reduces the likelihood of pressunzer emptying or overfill.

These changes modify neither accident mitigation nor system response post-accident.

Therefore, the proposed revision does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.

Involve a significant reduction in a margin of safety.

The Technical Specification changes provided are consistent with the initial condition assumed in the Chapter 15 accident analysis by placing tighter i

restnctions on pressunzer level. The Chapter 15 FSAR accident analysis assumes

(

a l

,s U.S. Nuclear Regulatory Commission B17028\\ Attachment 5\\Page 3 l

that pressunzer level is being maintained by the automatic control system at the programmed level. For most of the accident analysis, pressurizer level is assumed to be at 61.5% for power conditions and 28% for hot zero power. For events where pressunzer overfill is a concom, instial pressurizer level is assumed to be 6% above the nominal value of 61.5% at full power. This bounds the automatic control system control system uncertainty as documented in WCAP 14353. Thus, the proposed Technical Specification LCO for Modes 1 and 2 is consistent with the Chapter 15 FSAR accident analysis. When pressunzer level is being maintained by manual operator action, a 6% operating band is specifed. This band is consistent with the l

6% error assumed for the pressunzer overfill events, but it does not take into l

account instrument uncertainty. Because of the infrequent use of manual operation combined with the multiple main board indications and the rendciveess associated l

with wistrumentation uncertainty, it is unnecessary to apply instrument uncertainty effects on top of the operating band As such, the 6% band is bounded by the current Chapter 15 FSAR analysis. For Mode 3, the current specification which assures that a steam bubble exists in Mode 3 is sufficient to assure consistency with the accident analysis assumptions. The Bases are modified to reflect the proposed l

dianges and define the consistency with the Chapter 15 accident analysis.

l Therefore, the change does not reduce the margin of safety.

Therefore, the proposed revision does not involve a significant reduction in a margin of safety.

In conclusion, based on the information provided, it is determined that the proposed revision does not involve an SHC.

l Environmental Considerations NNECO has reviewed the proposed license amendment against the criteria of 10CFR51.22 for environmental considerations. The proposed revision does not involve a SHC, does not significantly increase the type and amounts of effluents that may be released offsite, nor significantly increase individual or cumulative occupational radiation exposures. Based on the foregoing, NNECO concludes that the proposed revision meets the criteria delineated in 10CFR51.22(c)(9) for categorical exclusion from the requirements for environmental review.

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ML20216A890

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