Text

Amend 52 to License DPR-65 Authorizing Cycle 3 Operation at 2,560 Mw W/Modified Guide Tubes for Control Element Assemblies
	ML19269E355
Person / Time
Site:	Millstone
Issue date:	05/12/1979
From:	Reid R; Office of Nuclear Reactor Regulation
To:	;
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ML19269E356	List: ML19269E355; ML19269E357; ML19269E358; ML19269E359;
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	NUDOCS 7906270258
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UNITED STATES NUCLEAR REGULATORY COMMISSION yg WASHINGTON, D. C. 20555 5 e. g

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/ g \\ *v 8 THE CONNECTICUT LIGHT AND POWER C0f1PANY, THE HARTFORD ELECTRIC LIGHT COMPANY, WESTERN ttASSACHUSETTS ELECTRIC COMPANY, AND NORTHEAST NUCLEAR ENERGY COMPANY DOCKET NO. 50-336 MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 52 1.icense No. DPR-65 The Nuclear Regulatory Commission (the Commission) has found that: 1. The applicatiors for amendment by The Connecticut Light and A. Power Company, The Hartford Electric Light Company, Western Massachusetts Electric Company, and Northeast Nuclear Energy Company (the licensees), dated December 16, 1977, December 15, 1978, February 12, 1979 and March 2,1979, as supplemented, comply 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 applications, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized 4' this amendbent 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; 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 require-ments have been satisfied. 7 2131 040 7906270#58'

. Accordingly, the licensc is amended by changes to the Technical 2. Specifications as indicated in the attachement to this license amendment and paragraph 2.C(2) of Facility Operating License No. DPR-65 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 52, are hereby incorporated in the If cense. The licensee shall operate the facility in accordance with the Technical Specifications, except that maximum authorized power level remains as 2560 Wt as specified by paragraph 2.C.(1),above. 3. This license amendment is effective as of the date of its issuance. FOR THE NUCLEAR REGULATORY COMMISSION h Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: May 12, 1979 E 2131 041 e

ATTACHMENT TO LICENSE AMENDMENT NO. 52 FACILITY OPERATING LICENSE NO. DPR-65 DOCKET N0.~50-336 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document comp 1steness. Pages Pages 1-1 2-2 3/4 4-5 2-4 3/4 4-7a 2-5 3/4 4-7b 2-7 3/4 4-7c 2-8 3/4 4-7d 2-9 3/4 4-7f B 2-1 3/4 4-79 B 2-2 3/4 4-7h B 2-3 3/4 5-3 B 2-5 3/4 5-4 B 2-6 3/4 5-5 B 2-8 (added) 3/4 5-6 3/4 1-26 3/4 7-1 3/4 1-30 3/4 7-2 3/4 2-1 3/4 7-3 3/4 2-2 3/4 7-21 3/4 2-4 3/4 7-22 3/4 2-5 3/4 7-35 3/4 2-6 3/4 10-1 3/4 2-8 3/4 10-2 3/4 2-9 B 3/4 2-1 3/4 2-10 B 3/4 2-2 3/4 2-12 B 3/4 4-2 3/4 2-14 B 3/4 4-2a 3/4 2-15 B 3/4 7-1 3/4 3-3 8 3/4 7-2 3/4 3-6 3/4 3-8 3/4 3-21 3/4 3-22 0 2131 042

' ;p 1.0 DEFINITIONS DEFINED TERMS 1.1 The DEFINED TERMS of this section appear in capitalized type and are applicable throughout these Technical Specifications. THERMAL POWER 1.2 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. RATED THERMAL PC'4ER 1.3 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2700 MWt. l OPERATIONAL MODE 1.4 An OPERATIONAL MODE shall correspond to any one inclusive com-bination of core reactivity condition, power. level and average reactor coolant temperature specified in Table 1.1. ACTION 1.5 ACTION shall be those additional requirements specified as corollary statements to each principle specification and shall be part of the specifications. OPERABLE - OPERABILITY 1.6 A system, s'ubsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of perfonning its specifted function (s). Implicit in this definition snall be the assumption that all necessary attendant instrumentation, controls, electric power, cooling or seal water, lubrication or other required auxiliary equipment is also OPERABLE. REPORTABLE OCCURRENCE 1.7 A REPORTABLE OCCURRENCE shall be any of those conditions specified as a reportable occurrence in Revision 4 of Regulatory Guide 1.16, " Reporting of Operating Information - Appendix "A" Technical Specifications." MILLSTONE - UNIT 2 1 -1 Amendment No. J,52 n itic 2131 043

>j DEFINITIONS CONTAINMENT INTEGRITY 1.8 CONTAINMENT INTEGRITY shall exist when: 1.8.1 All penetrations required to be closed during accident conditions are either: Capable of being closed by an OPERABLE containment a. automatic isolation valve system, or b. Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, 1.8.2 The equipment hatch is closed and sealed, and 1.8.3 The airlock is OPERABLE pursuant to Specification 3.6.1.3. CHANNEL CALIBRATION 1.9 A CHANNEL CALIBRATION shall be the ad:iustment, as necessary, of the channel output such that it responds with :he necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire cha1nel including the sensor and alam and/or tr p functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequen-tial, overlapping or total channel steps such that the entire channel is calibrated. CHANNEL CHECK 1.10 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter. CHANNEL FUNCTIONAL TEST 1.11 A CHANNEL FUNCTIONAL TEST shall be the injection of a. simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY including alam and/or trip functions. MILLSTONE - UNIT 2 1-2 f., : 2131 044

s 3 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and maxi-mum cold leg coolant temperature shall not exceed the limits shown on Figure 2.1-1. APPLICABILITY: MODES 1 and 2. ACTION: Whenever the poi.nt defined by the combination of maximum cold leg temper-ature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour. REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia. APPLICABILITY: MODES 1, 2, 3, 4 and 5. ACTION: MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2750 psia, be in HOT STANDBY with'the Reactor Coolant System pressure within its limit within 1 hour. MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to within its limit within 5 minutes. MILLSTONE - UNIT 2 2-1 2131 045 cy:.c;_

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c SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SETPOINTS 2.2.1 The reactor protective instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1. APPLICASILITY: AS SHOWN FOR EACH CHANNEL IN TABLE 3.3-1. ACTION: With a reactor protective instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value. MILLSTONE - UNIT 2 2-3 2131 047 sn l._,' 3 . CV

i5 lABLE 2.2-1 f$ REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS E 7 FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES l E 1. Manual Reactor Trip Not applicable Not Ar;plicable .a ro 2. Power Level - High [^ Four Reactor Coolant Pumps 1 9.88% above THERMAL POWER, with 1 9.88% above THERMAL POWER and Operating a min; mum ctpoint of s 15% of a minimum setpoint. of s 15% of l RATED THERMAL POWER, and a maximum RATED THERMAL POWER, and a maximum of 1 107% of RATED THERMAL POWER. of 1 107% of RATED THERMAL POWER. 3. Reactor Coolant Flow - Low (1) Four Reactor Coolant Pumps > 91.7% of design reactor coolant > 91.7% of design reactor coolant Operating fTow with 4 pumps operating

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[ 4. Pressurizer Pressure - High 1 2400 psia 1 2400 psia 5. Containment Pressure - High 1 4.75 psig 1 4.75 psig 6. Steam Generator Pressure - > 500 psia > 500 psia Low (2)(5) g Low (5) -> 36.0% Water Level - each -> 36.0% Water Level - each p 7. Steam Generator Water Level-steam generator steam generator o.l 8. Local Power Density - High (3) Trip setpoint adjusted to Trip setpoint adjusted to not exceed the limit lines not exceed the limic lines c+ g of Figures 2.2-1 and 2.2-2. of Figures 2.2-1 and 2.2-2. h9. Thermal Margin / Low Pressure (1) N t e4 m Four Reactor Coolant Pumps Trip setpoint adjusted to Trip setpoint adjusted to N Operating not exceed the limit lines not exceed the limit lines of Figures 2.2-3 and 2.2-4. of Figures 2.2-3 and 2.2-4.

Design reactor coolant flow with 4 pumps operating 1s 370,000 gpm

O!, TABLE 2.2-1 (Continued) ~' ~.. Y 3 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS ~. ~- p FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 2

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2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel pladding and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate at or less than 21 kw/f t. Centerline fuel melting will not occur for this peak linear heat rate. Overheating of the fuel cladding is operatica to within the nucleate boiling prevented by restricting F 1 regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturstion temperature. Operation abene the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB). and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measur-able parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the CE-1 correlation. The CE-1 DNB correlation has been developed to l predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB. The minimum value of the DNBR during steady state operation, nomal operational transients, and anticipated transients is limited to 1.19. This.value corresponds to a 95 percent probability at a 95 l percent confidence level that DNB will not occur and is chosen as as appropriate margin to DNB for all operating conditions. The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperature with four Reactor Coolant Pumps operating for whi:h the minimum DNBR is no less than 1.19 for the family of axial shapes and correspending i radial peaks shown in Figure B2.1-1. The limits in Figure 2.1-1 were calculated for reacter coolant inlet temperatures less than or equal to 580'F. The da,hed line at 580'F coolant inlet temperatures is not [ a safety limit; nowever, operation above 580*F is not possible because of the actuation of the main steam line safety valves which limit the maximum value of reactor inlet temperature. Reactor operation at THERMAL POWER levels higher than 112% of RATED THERMAL POWER is prohibited by the high power level trip set point specified in Table 2. 2-1. The area of safe operation is below and to the left of these lines. MILLSTONE - UNIT 2 8 2-1 Amendment No. 7,52' eu E 2131 054

,i. 3 F gg e rn b E \\ Q 1.6 3 l ~ 7 i z, W N 4 to T i-N Q A \\ ROD j= i x w a^o'ai =^* 7 7 x N b ^; / N ~ 2.00 N 1.76 d' S 0 ~ 0 10 20 30 40 50 60 70 80 90 100 c+ PERCENT OF ACTIVE CORE LENGTH FROM BOTTOM ? vi N FIGURE B2.1-1 Axial Power Distributions for Thermal Margin Safety Limits m u ~ O e L.n ~ LT1.

A k e a 3 SAFETY LIMITS BASES The conditions for the Thermal Margin Safety Limit curves in Figure 2.1-1 to be valid are shown on the figure. The retctor protective system in combination with the Limit 1ng Condittons for Operation, is designed to prevent any anticipated com-bination of transient cenditions for reactor coolant system temperature, pressure, and thermal power level that would result in a DNBR of less than 1.19 and preclude the existence of flow instabtitties. l 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity uf the Reactor Coolant System from overpressurization and thereby praents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere. The reactor pressure vessel and pressurizer are uesigned to Section III of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110% (2750 psia) of design pressure. The Reactor Coolant System piping, valves and fittings, are designed to ANSI B 31.7, Class I which permits a maximum transient pressure of 110% (2750 psia) of component design pressure. The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code require-ments. The entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to initial operation. MILLSTONE - UNIT 2 B 2-3 Amendment No. 7,52' N iti 2131 056

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SET POINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each parameter. The Trip Values have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Operation with a Trip Setpoirt less conservative than its setpoint but within its specified Allowable Value is acceptable on the basis that each Allowable Value is equal to or less than the drift allowance assumed to occur for each trip used in the accident analyses. Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability. Power Level-High The Power Level-High trip provides reactor core protection against ' reactivity excursions which are too rapid to be protected by a Pressurizer Pressure-High or Thermal Margin / Low Pressure trip. The Power Level-High trip setpoint is operator adjustable and can be 7 set no higher than 9.88% above the indicated THERMAL POWER level. Operator action is required to increase the trip setpoint as THERMAL POWER is increased. The trip setpoint is automatically decreased as THERMAL power decreases. The trip setpoint has a maximum value of 107% of RATED THERMAL POWER and a minimum setpoint of 15% of RATED THERMAL POWER. Adding l to this maximum value the possible variation in trip point due to calibration and instrument errors, the maximum actual steady-state THERMAL POWER level at which a trip wou1d be actuated is 112% of RATED THERMAL POWER, which is the value used in the accident analyses. Reactor Coolant Flow-Low The Reactor Coolant Flow-Low trip provides core protection to prevent DNB in the event of a sudden significant decrease in reactor coolant flow. Provisions have been made in the reactor protective system to permit Change No. 4 MILLSTONE - UNIT 2 B24 September 26, 1975 .E7. 2131 057

LIMITING SAFETY SYSTEM SETTINGS BASES Reactor Coolant Flow-Low (Continued) operation of the reactor at reduced power if one or two reactor coolant pumps are taken out of service. The low-flow trip setpoints and Allowable Values for the various reactor coelant pump combinations have bean derived in consideration of instrument errors and response times of equipment involved to maintain the DNBR above 1.19 under nonnal operation l and expected transients. For reactor operation with only two or three reactor coolant pumps operating, the Reactor Coolant Flow-Low trip set-points, the Power Level-High trip setpoints, and the Themal Margin / Low Pressure trip setpoints are automatically changed when the pump condition selector switch is manually set to the desired two-or three-pump position. Changing these trip setpoints during two and three pump operation prevents the minimum value of DNBR from going below 1.19 during nomal operational l transients and anticipated transients when only two or three reactor coolant pumps are operating. Pressurizer Pressure-High The Pressurizer Pressure-High trip, backed up by the pressurizer code safety valves and main steam line safety valves, provides reactor coolant system protection against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is 100 psi below the nominal lift setting (2500 psia) of the pressurtzer code safety valves and its concurrent operation with the power-operated relief valves avoids the unde'sirable operation of the pressurtzer code safety valves. Containment Pressure-High The Contaiment Pressure-High trip provides assurance that a reactor trip is initiated concurrently with a safety injection. The setpoint for this trip is identical to the safety injection setpoint. Steam Generator Pressure-tow The Steam Generator Pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the reactor coolant.. The setting of 500 psia is sufficiently below the full-load operating point of 815 psia so as not to interfere with nomal operation, but still high enough to provide the required protection in the event of excessively high steam flow. This setting was used with an uncertainty factor of + 22 pst in the accident analyses. MILLSTONE - UNIT 2 B 2-5 Amendment No. 521 s. r. ic, i m M 1 058 o

o LIMITING SAFETY SYSTEM SETTINGS BASES Steam Generator Water Level - Low The Steam Generator Water Level-Low Trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity and assures that the design pressure of the reactor coolant system will not be exceeded. The specified setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of trip to provide a margin of more than 10 minutes before auxiliary feedwater is required. Local Power Density-N gh The Local Power Density-High trip, functioning from AXIAL SHAPE INDEX monitoring, is provided to ensure that the peak local power density in the fuel which corresponds to fuel centerline melting will not occur as a consequence of axial power maldistributions. A reactor trip is initiated whenever the AXIAL SHAPE INDEX exceeds the allowable limits of Figure 2.2-2. The AXIAL SHAPE INDEX is calculated from the upper and lower ex-core neutron detector channels. The calculated setpoints are generated as a function of THERMAL POWER level with the allowed CEA group position being inferred from the THERMAL POWER level. The trip is automatically bypassed belw 15 percent power. The maximum AZIMUTHAL POWER TILT and maximum CEA misalignment permitted for continuous operation are assumed in generation of the setpoints. In addition, CEA group sequencing in accordance with the Specifications 3.1.3.5 and 3.1.3.6 is assumed. Finally, the maximum insertion of CEA banks which can occur during any anticipated opera-tional occurrence prior to a Power Level-High trip is assumed. Thermal Margin / Low Pressure The Thermal Margin / Low Pressure trip is provided to prevent l operation when the DNBR is less than 1.19. MILLSTONE - UNIT 2 B 2-6 Amendment No. S. 6,52' w 2131 059 e

3, LIMITING SAFETY SYSTEM SETTINGS BASES Thermal Margin / Low Pressure (Continuedl The trip is initiated whenever the reactor coolant system pressure signal drops below either 1750 psia or a computed value as described below, whichever is higher. The computed value is a function of the higher of AT power or neutron power, reactor inlet temaerature, the number of reactor coolant pumps operating and the AXIA'. SHAPE INDEX. The minimum value of reactor coolant flow rate, the maximum AZIMUTHAL POWER TILT and the maximum CEA deviation permitted for continuous operation are assumed in the generation of this trip function. In addition, CEA group sequencing in accordance with Specificattons 3.1.3.5 and 3.1.3.6 is assumed. Finally, the maximum insertion of CEA banks which can occur during any anticipated operational occurrence prior to a Power Level-High trip is assumed. The Thermal Margin / Low Pressure trip setpoints are derived from the core safety limits through application of appropriate allowances for equipment response time measurement uncertainties and processing error. A safety margin is provided which includes: an allowance of 5% of RATED THERMAL POWER to cmpensate for potential power measurement error; an allowance of 2*F to compensate for potential temperature measurement uncertainty; and a further allowance of 67 psi to compen-l sate for pressure measurement error, trip system processing error, and time delay associated with providing effective termination of the occurrence that exhibits the most rapid decrease in margin to the safety limit. The 67 psi allowance is made up of a 22 psi pressure measurement allowance and a 45 psi time delay allowance. Loss of Turbine A Loss of Turbine trip causes a direct reactor trip when operating above 15% of RATED THERMAL POWER. This trip provides turbine protection, reduces the severity of the ensuing transient and helps avoid the lifting of the main steam line safety valves during the ensuing transtent, thus extending the service life of these values. No credit was taken in the accident analyses for operation of this trip. Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System. MILLSTONE - UNIT 2 B 2-7 Amendment No. M, 52' idU I M 2131 060

,i ( LIMITING SAFETY SYSTEM SETTINGS BASES Underspeed - Reactor Coolant Pumps The Underspeed - Reactor Coolant Pumps trip provides core protection to prevent DNB in the event of a sudden significant decrease in reactor coolant pump speed (with resulting decrease in flow) on all four reactor coolant pumps. The trip setpoint ensures that a reactor trip will be generated, considering instrument errors and response times, in sufficient time to allow the DNBR to be maintained above 1.19 following a 4 pump loss of flow event. t MILLSTONE - UNIT 2 B 2-8 Amendment No. 52' 2131 061

REACTIVITY CONTROL SYSTEMS POSITION INDICATOR CHANNELS (Continued) LIMITING CONDITION FOR OPERATION b) The CEA group (s) with the inoperable position indi-cator is fully inserted, and subsequently maintained fully inserted, while maintaining the withdrawal sequence and THERMAL POWER level required by Speci-fication 3.1.3.6 and when this CEA group reaches its fully inserted position, the " Full In" limit of the CEA with the inoperable position indi-cator is actuated and verifies this CEA to be fully inserted. Subsequent operation shall be within the limits of Specification 3.1.3.6. c. With a maximum of one reed switch position indicator channel per group or one pulse counting position indicator channel per group inoperable and the CEA(s) with the inoperable position indicator channel at either its fully inserted position or fully withdrawn position, operation may continue provided: 1. The position of this CEA is verified immediately and at least once per 12 hours thereafter by its " Full In" or " Full Out" limit (as applicable), 2. The fully inserted CEA group (s) containing the inoperable position indicator channel is subsequently maintained fully inserted, and 3. Subsequent operation is within the limits of Specifica-tion 3.1.3.6 d. With one or more pulse counting position indicator channels inoperable, operation in MODES 1 and 2 may continue for up to 24 hours provided all of the reed switch position indicator channels are OPERABLE. SURVEILLANCE REQUIREMENTS 4.1.3.3 Each position indicator channel shall be determined to be OPERABLE by verifying the pulse counting positior. indicator channels and the reed switch position indicator channels agree within 6 steps at l least once per 12 hours except during time intervals when the Deviation circuit is inoperable, then compare the pulse counting position indicator and reed switch position indicator channels at least once per 4 hours. Change No. 4' MILLSTONE - UNIT 2 3/4 1-25 September 26, 1975 m N. 2131 062

1 REACTIVIf( CONTROL SYSTEMS CEA DROP TIME 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 4 3.1 seconds from I when electrical power is interrupted to the CEA ifrive mechanism until the CEA reaches its 90 percent insertion position with: 1,yg >_515 F, and a. b. All reactor coolant pumps operating. APPLICABILITY: MODE 3. 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 provided THEs, MAL POWER is restricted to less than or equal to tie maximum THERMAL POWER level allowable for the reactor coolant pump combination operating at the time of CEA drop time determination. SURVEILLANCE REQUIREMENTS 4.1.3.4 The CEA drop time of full length CEAs 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 modiftcation to the CEA drive system which could affect the drop time of those specific CEAs, and c. At least once per 18 months. MILLSTONE - UNIT 2 3/4 1-26 Amendment No. SE, 52' ~ ?'s 2131 063

-t REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) c. With the regulating CEA groups inserted between the Lcng Term Steady State Insertion Limits and the Transient Insertion Limits for intervals > 5 EFPD per 30 EFPD interval or > 14 EFPD per calendar year, except during operations pursuant to the provi-sions 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 within two hours, or 2. Be in HOT STANDBY within 4 hours. SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each regulating CEA group shall be detennined to be within the Transient Insertion Limits at least once per 12 hours except during time intervals when the PDIL Auctioneer Alann Circuit is inoperable, then verify the individual CEA positions at least once per 4 hours. The accumulated times during which the regulating CEA groups are inserted between the Long Term Steady State Insertion Limits and the Transient Insertion Limits shall be detennined at least once per 24 hours. MILLSTONE - UNIT 2 3/4 1-29 .c '. 2131 064

990 lEl2 e. 1. - e. ,c, 3k'd 'ON quampuaW OE-Lt/C 2 IIN1 - 3N01S17IW = FRACTION OF RATED THERMAL POWER 5 9 9 o p a o o o o e o s M 8 8 8 8 a 8 8 8 o-i I i .i g _._I . _ __f u m ---(-----,--, LONG TERM m STE ADY STATE [ m 3

t-INSERTION

- ~ * * '

P i -.1 LIMIT J - {...-- t-4 o .d _ _,_ _ _. - b .i. o g I o ^ = i m ?' ^i ~z l l a._._-_..-.._- l M g g SHORT TERM 3 g o 5, +- STEADY STATE - l-g.. i m h INSERTION d ~- ~ gr - -*- 8 uutT 4 3 o zi F i j i e gj 3j g' n' s I w Q o_ .. ~ T"~'*~ ~~~ 'f~'~ t" ao 3 i m, y, .I i 3 4 l } >l -g: z i m g g -.---, --+- -. *,: - ~. g; - - f - ---.. - - z 3 . ~4 s N 3 g 3}! '._ d i ' '" I i g I I q gg_ g i 'a i s I. jz. o i m o_ g _ _.._ l ! "g !.__.__p._._.. m g 4 i. m !s = i i I -e 2 3 1' 8 l _4 - i @g 3 3' _j g _. r.. _4 z _.a.. ._+.._p.__-._-____. g y .I .j c.. I I z si .i x 6 g l 3 !m i _.i s_ i a g 's i i n o } f - { l l .t 4 i Q p g i n o ._..____.._.._4_a.-.;_.- --M_--..]._._.7._... g g i i. j. j g m i 1 i y g g g.._.L.-.._.,. _.._ _. _ _l_ _ _. 9 4 p. t t i 3 x .l. 1.- }.. a. g 4 i i 6 g . g -h ^ e i 4-__- E o i l g 1... ___,l.. i.. _ l. i = g in 7 i i i ..._t.. .I 4 g l 1 I e i ..}. j j i t, i 1 e f {, e j. f e

3/4.2 POWER DISTRIBUTION LIMITS LINEAR HEAT RATE LIMITING CONDITION FOR OPERATION 3.2.1 The linear heat rate shall not exceed the limits shown on Figure 3.2-1. APPLICABILITY: MODE 1. ACTION: During operation with the linear heat rate being nonitored by the Incore Detector Monitoring System, comply with the following ACTION: With the linear heat rate exceeding its limit, as indicated by four or more coincident incore channels, within 15 minutes initiate corrective action to reduce the linear heat rate to within the limits and either: a. Re: ore the linear heat rate to within its limits within one hour, or b. Be in at least HOT STANDBY within the next 6 hours. During operation with the linear heat rate being monitored by the Excore Detector Monitoring System, comply with the following ACTIONS: With the linear heat rate exceeding its limit, as indicated by the AXIAL SHAPE INDEX being outside of the power dependent limits on the Power Ratio Recorder and with the THERMAL POWER: a. Above 100% of the allowable power level determined by the expression (M) (N) in Specification 4.2.1.2.c, within 15 l minutes either restore the AXIAL SRAPE INDEX to within the limits of Figure 3.2-2 or reduce THERMAL POWER to < 100% of the allowable power level determined by the express 3on (M)(N) in Specification 4.2.1.2.c. b. < 100% of the allowable power level determined by the expres-i' ion (M)(N) in Specification 4.2.1.2.c, either restore the l AXIAL-SHAPE INDEX to within the limits of Figure 3.2-2 within 1 hour from initially exceeding the linear heat rate limit or be in HOT STANDBY within the next 4 hours. SURVEILLANCE REQUIREMENTS 4.2.1.1 The linear heat rate shall be determined tc be within its limits by continuously monitoring the core power distribution with either the excore detector monitoring system or with the incore detector moni,toring system. MILLSTONE - UNIT 2 3/42-1 'AmendmentNo.77,AE,52f ~i 2131 066 ema , e s.

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.1.2 Excore Detector Monitoring Systen - The excore detector monitoring system may be used for monitoring the core power distribution by: Verifying at least once per 12 hours that the f01 length CEAs a. are withdrawn to and maintained at or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6. b. Verifying at least once per 31 days that the AXIAL SHAPE INDEX alarm setpoints are adjusted to within the limits shown on Figure 3.2-2. Verifying at least once per 31 days that the AXIAL SHAPE INDEX is I c. maintained within the allowable limits of Figure 3.2-2, where 100 percent of the allowable power represents the maximum THERMAL POWER allowed by the following expression: MxN where: 1. M is the maximm allowable THERMAL POWER level for the existing Reactor Coolant Pump combination. N is the maximum allow 9 le fraction of RATED THERMAL POWER b 2. as detemined by the F curve shown on Figure 3.2-3 of M Specification 3.2.2. 4.2.1.3 Incore Detector Monitoring System - The incore detector moni-toring system may be used for monitoring the core power distribution by verifying that the incore detector Local Power Density alams: Are adjusted to satisfy the requirements of the core power dis-a. tribution map which shall be updated at least once per 31 days. b. Have their alarm setpoint adjusted to less than or equal to the limits shown on Figure 3.2-1 when the following factors are appropriately included in the setting of these alarms: 1. Flux peaking augmentation factors as shown in Figure 4.2-1, 2. A measurement-calculational uncertainty factor of 1.07, l 3. An engineering uncertainty factor of 1.03, 4. A linear heat rate uncertainty factor of 1.01 due to axial fuel densification and thermal expansion, and 5. A THERMAL POWER measurement uncertainty factor of 1.02. MILLSTONE - UNIT 2 3/42-2 Amendment No. p, pp,52' f b le i 2131 067

2 r-O# c'., 17.C ~% m rw '~e C

o. -

3 g -4 x Jg N EO =Q >- m dg 16.0 IO m2 6+ UNACCEPTAPLE OPERATION '5 " ? w gu ACCEPTABLE OPERATION N (+ N E'd N wo dD U 15.0 'i:o CD _l co CD EI 14.0' n. 3 0 100 200 300 400 500 to Dr EFFECTIVE FULL POWER DAYS

=

P Figure 3.21 Allowable Peak Linear Heat Rate vs Burnup w N w b 00 v <r.;

1.10 J' .. f.._ 1.00 UNACCEPTABLE D UNACCEPTABLE OPERATION OPERATION REGION ^ REGION -l l: 0.90 . (-0.06, 0.89), (0.15, 0.89) { i' /i -\\ l _ i i J 1 p i L_. uj [ 9 l m 0.80 h- ~~ ' li-~ l i -) {/ \\ l { i9 - it: 0.70 ~ ~ 4 (-0.3, 0.65) (0.3, 0.65) ' ACCEPTABLE

j-.

y f... z l l l-OPERATION t: .i o 0.60 REGION

ig..

b i E I 'l I .i 0.50 .g.. i; I .g 0.40 I ..[.1 _!.- .I l-l i~~ l.' .l -i-q .i 0.30 -0.6 -0.4 -0.2 0 0.2 0.4 0.6 AX1 AL SHAPE INDEX FIGURE 3.2-2 AX1 AL SH APE INDEX vs Fraction of Allowable Power Level per Specification 4.2.1.2c MILLSTONE - UNIT 2 3/4 2-4 Amendment NO. 77, EE, 52' ?.13\\ 069 r;;;,;

1.060 . il, j! D. ih.I. :.i.n.. i.:. i.l.'.. .i l.. '} h l! a'lUg.' _' l. '.' a: ' L' u'l l }. q .c u. .q; p. a L' 'E ni i l! lI h, l t i i l i'. - I!ll it .! ji r--

i I

h 'i. t, /),[ .i! h ni i .,J. L, i.n { . a'a .y! my /; ev. -l \\ o 1.050 2 i, '..1, ".i "j) /. .1'1.. +i I,-- d-J! ~. t' ',t -- / m a p t-- r l ,:l 'l i j, I / i i-b '!: l i' c i' i '~ 2 i i, + - - t -- E 1.040 ,[i.. [ j j s .l'i i H .; }. . 1, ;j N O R1 .1 >-o o q u. ~ __i_.M.. z 9 / i,o3o Q /_ ___. __J _ _ i i s i 'l l / F- ,z ..i .J._: / N a / r w < 1.020 /... g,, ) w s . f m J CD F CD . l/ y i.oio _._. [ l a. 9. .. u i o r. e 1.000 0 20 40 60 80 100 120 140 DISTANCE FROM BOTTOM OF CORE, INCHES coar+ 2 O FIGURE 4.2-1 Augmentation Factors vs Distance From Bottom of Core wW N. 2 .,mI4 *)

q POWER DISTRIBUTION LIMITS TOTAL PLANAR RADIAL PEAKING FACTOR - F LIMITING CONDITION FOR OPERATION ThecalculatedvalueofF[y,definedasF =F (1+T ), shall 3.2.2 q l be limited to < l.615. APPLICABILITY: MODE 1*. ACTION: WithFfy>1.615,within6nourseither: l Reduce THERMAL POWER to bring the combination of THERMAL POWER a. andFfy to within the limits of Figure 3.2-3 and withdraw the full length CEAs to or beyond the Long Term Steady State Insertion Limit of Specification 3.1.3.6; or b. Be in at least HOT STANDBY. SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. xy(1+T ) and Ffy shall be calculated by the expression F =F 4 2.2.2 q F shall be determined to be within its limit at the following intervals: xy Prior to operation above 70 percent of RATED THERMAL POWER after a. each fuel loading, b. At least once per 31 days of accumulated operation in MODE 1, and Within four hours if the AZIMUTHAL POWER TILT (T ) is > 0.02. c. q

See Special Test Exception 3.10.2.

MILLSTONE - UNIT 2 3/4 2-6 Amendment No. 35,52Y 2131 071

s. F* POWER DISTRIBUTION LIMITS SURVEILLANCE RE0UIREMENTS (Continued) shallbedeterminedeachtimeacalculationofFfy is required 4.2.2.3 Fxy by using the incore detectors to obtain a power distribution map with all full length CEAs at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Famp combination. This detennination shall be limited to core planes between 15% and 85% of full core height inclusive and shall exclude regions influenced by grid effects. shallbedeterminedeachtimeacalculationofFfy 4.2.2.4 T is required q and the value of T used to determine F shall be measured value of T. q xy q MILLSTONE - UNIT 2 ~3/4 2-7 Amendment No. 38 (Io l. 2131 072

Iii l' -b UNACCEPTABLE l 2 1 p j. OPERATION r-l T(1.6 5,~ ~1.0) k650* 1. F[y ~ LIMIT CURVE ~

q g

w 1.0 s-- (1.656,0.975): \\ j-- - ~ ]-- T -'- 8 E D M 0 + (1.644,0.90) _{ ~~ j 5' _ ' (1.689, 0.80). y ~ I m =* (1.776,0.8025) to w 0.8 I (1.717,0.755) - 8 t. .- ;.q_ .._ [ _. _.f _ _ O l-g I-

y.._

H _.7__ p - [_._ _._ .\\ -l (1.776,0.675) E [ N T WRVE l 0.6 r O 8 _ _- {__ a I-Q _g OA ACCEPTABLE i .9-OPERATION ~~ ~ ~ -~ -~ REGION .j. jl k i 3: .. p__ g ( 0.2 i .l- ._i ...j i i i t f. 0.0 g 1.60 1.62 1.64 1.66 1.68 1.70 1.72 1.74 1.76 1.78 1.80 g FT.F IF, x U + T ): F xU+TH y q xy g M zo b FIGUR E 3.2-3 Total Radial Peaking Factor Versus Allowable Fraction of R ATED THERMAL POWER tm Ut l\\). e

POWER DISTRIBUTION LIMITS TOTAL INTEGRATED RADI AL PEAKING FACTOR - Ff L_IMITING CONDITION FOR OPERATION 3.2.3 The calculated value of F, defined as FT = F (1+T ), shall be T 7 7 r q l limited to < 1.630. APPLICABILITY _: MODE 1*. ACTION: With FT > 1.630, within 6 hours either: l 7 Reduce THERMAL POWER to bring the combination cf THERMAL POWER a. T and F to within the limits of Figure 3.2-3 and withdraw the 7 full length CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6; or b. Be in at least HOT STANCBY. SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.Q.4 are not applicable. T 4.2.3.2 F shall be calculated by the expression FT = F (1+T ) and F T 7 q y 7 shall be detennined to be within its limit at the following intervals: Prior to operation above 70 percent of RATED THERMAL POWER a. after each fuel loading, b. At least once per 31 days of accumulated operation in MODE 1, and Within four hours if the AZIMUTHAL POWER TILT (T ). is > 0.020. c. q T 4.2.3.3 F shall be detennined each time a calculation of F is required 7 r by using the incore detectors to obtain a power distribution map with all full length CEAs at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Ptnp Combination. T 4.2.3.4 T shall be determined each time a calculation of F is required q and the value of T used to detennine F shall be the measured value of q r q.

See Special Test Exception 3.10.2.

MILLSTONE - UNIT 2 3/42-9 Amendment No. pg,52' d I "'. i d a 2131 074

POWER DISTRIBUTION LIMITS AZIMUTHAL POWER TILT - T q LIMITING CONDITION FOR OPERATION 3.2.4 The AZIMUTHAL POWER TILT (T ) shall not exceed 0.02. q APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER *. l ACTION: a. With the indicated AZIMITHAL POWER TILT determined to be > 0.02 but < 0.10, either correct the power tilt within two hours or determine within the next 2 hours and at least once per sub-sequent 8 hours, that the TOTAL PLANAR RADIAL PEAKING FACTOR (F and the TOTAL INTEGRATED RADIAL PEAKING FACTOR (F ) are within r the limits of Specifications 3.2.2 and 3.2.3. b. With the indicated AZIMUTHAL POWER TILT determined to be > 0.10, operation may. proceed for up to 2 hours provided that the TOTAL T INTEGRATED RADIAL PEAKING FACTOR.(F ) and TOTAL PLANAR RADIAL 7 PEAKING FACTOR (Ffy). are within the limits of Specifications 3.2.2 and 3.2.3. Subsequent operation for the purpose of measurement and to identify the cause of the tilt is allowabM provided the THERMAL POWER level ts restricted to x 20% of tLe maximum allowable THERMAL POWER level for the exisOng Reactor Coolant Pump combination. SURVEILLANCE REQUIREMENT 4.2.4.1 The provisions of Specification 4.0.4 are not applicable. 4.2.4.2 The AZIMUTHAL POWER TILT shall be determined to be within the limit by: a. Calculating the tilt at least once per 7 days when the Channel High Deviation Alann is OPERABLE,

See Special Test Exception 3.10.2.

MILLSTONE - UNIT 2 3/4 2-1.0 Amendment No. n, 52' 2131 075

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) b. Calculating the tilt at les:t once per 12 hours when the Channel High Deviation Alarm is inoperable, and Using the incore detectors to determine the AZIfiUTHAL POWER c. TILT at least once per 12 hours when one excore channel is inoperable and THERMAL POWER is > 75% of RATED THERHAL POWER. filLLSTONE - UNIT 2 3/4 2-11 Amendment flo. 38 l \\i, if: 2131 076

This.page intentionally left blank. a MILLSTONE - UNIT 2 3/4 2-12 Amendment No j!E,, 33, $$,52' '. ' ' ~ 21'31 077

i? POWER DISTRIBUTION LIMITS DNB MARGIN LIMITING CONDITION FOR OPERATION I 3.J.6 The DNB margin shall be preserved by maintaining the cold leg temperature, pressurizer pressure, reactor coolant flow rate, and AXIAL SHAPE INDEX within the limits specified in Table 3.2-1 and Figure 3.2-4. APPLICABILITY: MODE 1. ACTION: With any of the above parameters exceeding its specified limits, restore the parameter to within its above specified limits within 2 hours or reduce THERMAL POWER to < 5% of RATED THERMAL POWER within the next 4 hours. ~ SURVEILLANCE REQUIREMENTS 4.2.6 The cold leg temperature, pressurizer pressure, reactor coolant l flow rate, and AXIAL SHAPE INDEX shall be determined to be within the limits of Table 3.2-1 and Figure 3.2-4 at least once per 12 hours. 2 MILLSTONE - UNIT 2 .3/4~2-13, Amendment tio. 38 1 7h ry 'd ILg 3 2131 078

ik y TABLE 3.2-1 DNB MARGIN LIMITS Four Reactor Coolant Pumps Parameter Operating Cold Leg Temperature 5,549'F l Pressurizer Pressure t,2225 psia" Reactor Coolant Flow Rate t,370,000 gpm AXIAL SHAPE INDEX Figure 3.2-4

Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step increase of greater than 10% of RATED THERMAL POWER.

MILLSTONE - UNIT 2 3/4 2-14 Amendment No. EE, 52' 2131 079 . i'

l

(,t 6 V t s

a -g i 1.2 _{..... ..l-M __J L 1.1 l 'l' i .i i i I 2.f._._.f' ~ -(0.15,1.0) j 1.0 (-0.10,1.0) _.}. i m I i-i -- OPE R ATION ) UN ACCEPTABLE - w ~~ UNACCEPTABLE - OPERATION 0.9 REGION 'i-I. REGION _j__ 1 i-Y _ _L_ 1 i iE L -t i _L _.;_ J w t-I. I i - (0.30, 0.80) o m 0.8 (-0.30, 0.80) ~ ACCEPTABLE - - - -- OPER ATION y -} REGION m _j__ _ _, _ _ O l l h 0.7 { o . __.}. ..l... g. _. ...g _. _.i j. 4 m I_ j i .+. 0.6 l i i i I I 1 t- -- i i ...I _ ...g 0.5 I '~i i ]-- '! j j j j ._j _ l.! I I I . _ f.._ I ~l .I r l j 0.4-0.6 -0.4 -0.2 0 0.2 0.4 0.6 Y g FIGURE 3.2-4 AXI AL SHAPE INDEX Operating Limits with Four Reactor Coolant Pumps Operating MILLSTONE - UNIT 2 3/4 2'-15 Amendment No. 5, M,52 e 2131 080

~ d'> (12 C::a TABLE 3.3-1 0 Continued) [1

5 REACTOR PROTECTIVE INSTRUMENTATION p

u ( v3jl MINIMUM n' TOTAL NO. CHANNELS CHANNELS APPLICABLE i FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION e []

11. Wide Range Logarithmic Neutron Flux Monitor - Shutdown 4

0 2 3,4,5 4

12. Underspeed - Reactor 4

2(a). 3 1,2(e) 2 Coolant Pumps ps>ae tsa e La. - G) O CO W = ~ IV ao D ch D\\{ s ..\\

..j TABLE 3.3-1(Continued [ TABLE NOTATION

With the protective system trip breakers in the closed position and the CEA drive system capable of CEA withdrawal.

(a) Trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is >_ 5% of RATED THERMAL POWER. (b) Trip may be manually bypassed below 600 psia; bypass shall be automatically removed at or above 600 psia. (c) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > 15% of RATED THERMAL POWER. (d) Deleted. (e) Trip may be bypassed during testing pursuant to Special Test Excep-tion 3.10.3. (f) AT Power input to trip may be passed below 5% of RAVED THERMAL Power; bypass shall be automatically removed when THERMAL POWER is > 5% of RATED THERMAL POWER. ACTION STATEMENTS ACTION l With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in HOT STANDBY within the next 4 hours and/or open the protective system trip breakers. With the number of OPERABLE channels one icss than the ACTION 2 Total Number of Channels and with the THERMAL POWER level: a. < 5% of RATED THERMAL POWER, imediately place the Inoperable channel in the bypassed condition; restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 5% of RATED THERMAL POWER. b. > 5% of RATED THERMAL POWER, operation may continue with the inoperable channel in the bypassed condi-tion, provided the following conditions are satisfied: MILLSTONE - UNIT 2 IC!.3/43-4 Amendment No. 9, 38 ~ ~2131 082

TABLE 3.3-1 (Continued) ACTION STATEMENTS All functional units receiving an 'nput from the bypassed channel are also placed in the bypassed condition. 2. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service for up to 2 hours for surveillance testing per Specification 4.3.1.1 provided one of the inoperable channels is placed in the tripped condition. ACTION 3 With the number cf OPERABLE channels one less than the Total Number of Channels and with the THERMAL POWER level: a. < 5% of RATED THERMAL POWER, imediately place the inoperable channel in~the bypassed condition, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 5% of RATED THERMAL POWER. b. > 5% of RATED THERMAL POWER, power operation may continue. ACTION 4 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, imediately verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, and at least once per 4 hours thereafter. MILLSTONE - UNIT 2 3/4.3-5 2131'083 ..s M.. R ! s

i TABLE 3.3-2 z REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES F r-FUNCTIONAL UNIT RESPONSE TIME M 1. Manual Reactor Trip 2 1 0 seconds 5 2. Power Level - High < 0.40 seconds *# and < 8.0 seconds ## l 3. Reactor Coolant flow - Low < 0.65 seconds m ~ s 4. Pressurizer Pressure - High. < 0.90 seconds 5. Contaiment Pressure' 'High ' Not Applicable 6. Steam Generator Pressure - Low < 0.90 seconds 7. Steam Generator Water Level - Low < 0.90 seconds R 8. Local Power Density - High < 0.40 seconds *# and < 8.0 seconds ## a 9. Thennal Margin / Low Pressure < 0.90 secondr*# and < 8.0 seconds ## w b

10. Loss of Turbine--Hydraulic Fluid Pressure - Low Not Applicable

11. Underspeed - Reactor Coolant Pumps 1

45 seconds, j 0 h

Neutron actectors are exenpt from response time testing. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel, E

1 Response time does not include contribution of RTDs. D N ffRTD response time only. This value is equivalent to the time interval required for the RTD's t.ra M output to achieve 63,2% of its total chan'ge when subjected to a step change in RTD temperature. co =. CD =. A 7-w M

,'[ TABLE 4.3-1 z f REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS o5 CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE g FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED [ 1. Manual Reactor Trip N.A. N.A. S/II(1) N.A. 2. Power Level - High a. Nuclear Power S D(2),M(3),Q M 1, 2 b. AT Power S D(4),Q M 1 3. Reactor Coolant Flow - Low S R M 1, 2 [- 4. Pressurizer Pressure - High S R M 1, 2 3 5. Containment Pressure - High S R M 1, 2 C) Y' 6. Steam Generator Pressure - Low S R M 1, 2 co W 7. Steam Generator Water Level - Low S R M 1, 2 8. Local Power Density - High S R M 1 9. Thermal Margin / Low Pressure S R M 1, 2

10. Loss of Turbine--Hydraulic Fluid Pressure - Low N.A.

N.A. S/U(1) N.A.

TABLE 4.3-1_{ Continued) REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS 3 "r-r- CHANNEL MODES IN WHICH E CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE '" FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED E 11. Wide Range Logarithmic Neutron 5 N.A. S/U(1) 3, 4, 5 and

Z Flux Monitor ro

12. Underspeed - Reactor S

R M 1, 2 Coolant Pumps

13. Reactor Protection System s

g,. Logic N.A. N.A. M and S/U(1) 1, 2 '14. Reactor Trip Breakers N.A, N.A. M 1, 2 and

e.

W 3 I i 9l N e Oco e M Wm

TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 1. Manual a. SIAS Safety Injection (ECCS) Not Applicable Containment Isolation Not Applicable Enclosure BeiIding Filtration System Not Applicable b. CSAS Containment Spray Not Applicable c. CIAS Containment Isolation Not Applicable d. SRAS Containment Sump Recirculation Not Applicable e. EBFAS Enclosure Cuilding Filtration System Not Applicable 2. Pressurizer Pressure-Low Safety Injection (ECCS) a. 1) High Pressure Safety Injection 1 30.0*/5.0** 2) Low Pressure Safety Injection 1 50.0*/5.0** 3) Charging Pumps 1 40.0*/40.0** 4) Containment Air Recirculation System 1 31. 0 * / 31. 0** b. Containment Isolation 1 7.5 Enclosure Building Filtration System 1 50.0*/50.0** l c. MILLSTONE - UNIT 2 3/4 3-21 endment No. ZE, (5,57 ~

TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 3. Containment Pressure-High a. Safety Injection (ECCS) 1) High Pressure Safety Injection 1 30.0*/5.0** 2) Low Pressure Safety Injection 1 50.0*/5.0** 3) Charging Pumps 1 40.0*/40.0** 4) Containment Air Recirculation System 1 31. 0*/ 31. 0** b. C ntainment Isolation 1 7.5 c. En?.losure Building Filtration System 1 50.0*/50.0** l 4. Containment Pressure--High-High 35.6*( )/35.6**(I) a. Containment Spray 1 5. Containment Radiation-High a. Containment Purge Valves Isolation 1 ounting period C plus 7.5 6. Steam Generator Pressure-Low a. Main Steam Isolation 1 6.9 b. Feedwater Isolation 1 60 7. Refueling Water Storage Tank-Low a. Containment Sump Recirculation i 120 TABLE NOTATION Diesel generator starting and sequence loading delays induded. Diesel generator starting and sequence loading delays not included. Offsite power available. (I) Header fill time not included. MILLSTONE - UNIT 2 3/4 3-22 Amendment No. O, 52 l .1 2131 088

i REACTOR COOLANT SYSTEM STEAM GENERATORS LIMITING CONDITION FN OPERATION 3.4.5 Each steam generator shall be OPERABLE. APPLICABILITY: MODES 1, 2 and 3. ACTION: With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing T,yg above 200 F. SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following Augmented Inservice Inspection Program. 4.4.5.1 Augmented Inservice Inspection Program 4.4.5.1.1 Steam Generator Sanple Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-5. 4.4.5.1.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-6. The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.5.1.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.1.4. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except: a. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas. b. The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include: MILLSTONE - UNIT 2 3/4 4-5 AmendmentNo.22,#,h1 h;s 2131 089 .a

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 1. All nonplugged tubes thai; previously had detectable wall penetrations (>20%). 2. Tubes in those areas where experience has indicated potential problems. 3. A tube inspection (pursuant to Specification 4.4.5.1.4.a.8) l shall be performed on each selected tube. If.any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection. The tubes sel a l as the sCr.nd and third samples (if c. required by Table 4.4-6) during each inservice inspection may be subjected to a partial tube inspection provided: 1. The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found. 2. The inspection include those portions of the tubes where imperfections were previously found. The results of each sample inspection shall be classified into one of the following three categories: Category Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective. C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes. ii C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective. Note: In all inspections, previously degraded tubes must exhibit significant (>10%) further wall penetrations to be included in the above percentage calculations. MILLSTONE - UNIT 2 3/4 4-6 Amendment No. 72,37 M!5 2131'090

= i REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.1.3 Inspectio~n Frequencies - The above required inservice inspections l of steam generator tubes shall be performed at the following frequencies: The first inservice 17spection shall be perfomed after 6 a. Effective Full Power Months but within 24 calendar months of initial criticality. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections following service under AVT conditions, not including the preservice inspection, result in all inspec-tion results falling into the C-1 category or if two consecu-tive inspections demonstrate that previously observed degrada-tion has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months. b. If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-6 at 40 month intervals fall into Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspec-l tions satisfy the criteria of Specification 4.4.5.1.3.a; the interval may then be extended to a maximum of once per 40 months. Additional, unscheduled inservice inspections shall be perfomed c. on each steam generator in accordance with the -first sample inspection specified in Table 4.4-6 during the shutdowrt subse-quent to any of the following conditions: 1. Primary-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6~.2. 2. A seismic occurrence greater than the Operating Basis Earthquake. 3. A loss-of-coolant accident requiring actuation of the engineered safeguards; 4. A main steam line or feedwater line break. MILLSTONE - UNIT 2 3/4 4-7. Amendment No. 72,37 2131'091 3:

- 5 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4. 5.1. 4 Acceptance Criteria As used in this Specification a. Imperfection means an exception to the dimensions, finish or 1. contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections. Degradation means a service-induced cracking, wastage, 2. wear or general corrosion occurring on either inside or outside of a tube. 20% of 3. Degraded Tube _ means a tube containing imperfections t the nominal wall thickness caused by degradation. 4. % Degradation means the percentage of the tuoe wall thick-ness affected or removed by degradation. Defect means an imperfection of such severity that it 5. exceeds the plugging limit. A tube containing a defect is defective. Plugging Limit means the imperfection depth at or beyond 6. which the tube shall be removed from service because it may become unserviceable prior to the next inspection and is equal to 40% of the nominal tube wall thickness. Unserviceable describes the condition of a tube if it leaks 7. or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.1.3.c, above. Tube Inspection means an inspection of the steam generator 8. tube from the point of entry (hot leg side) completely around the U - Bend to the top support of the cold leg. The steam generator shall be determined OPERABLE after completing b. the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-6. Amendme.nt Ng. y, 77,5h ~ MILLSTONE - UNIT 2 3/4 4-7a 2131 092

2 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.1.5 Reports a. Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Comission within 15 days. b. The complete results of the steam generator tube inservice inspection shall be included in the Annual Operating Report for the period in which this inspection was completed. This report shall include: 1. Number and extent of tubes inspected. '2. Location and percent of wall-thickness penetration for each indication of an imperfection. 3. Identification of tubes plugged. c. Results of steam generator tube inspections which fall into Category C-3 and require prompt notification of the Comission shall be reported pursuant to Specification 6.9.1 prior to resumption of plant operation. The written followup of this report shall provide a description of investigations conducted to detemine cause of the tube degradation and corrective measures taken to prevent recurrence. MILLSTONE - UNIT 2 3/4.4-7b. AmendmentNo.77,37,hh' 2131 093

m. !

- s

This page intentionally left blank. I t l l Y MILLSTONE - UNIT 2 3/4 4-7c Amendment No. 7/ 52 S s s 2, 31 094

\\ This page intentionally left blank. I~ MILLSTONE - UNIT 2 3/4 4-7d Amendment No. 77, 55'

$e;o

!e i. 2131 095

TABLE 4.4-5 3 MINIMUM NUMBER OF STEAM GENERATORS TO BE { INSPECTED DURING INSERVICE INSPECTION 3 [ Preservice Inspection Yes 5 a No. of Str am Generators per Unit Two to ,[ ' First Inservice Inspection One l Second & Subsequent Inservice Inspections One m Table Notation: t,~ ,s [ 1. The inservice inspection may be limited to one steam generator on a rotating schedule 4 encompassing 3 N % of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous inspections indicate that all steam generators o are performing in a like manner. Note that under some circumstances, the operating con-ditions in one or more steam generators may be found to be more severe than those in other steam generators. Under such circumstances the sample sequence shall be mdified to inspect the most severe conditions. N ~ u co a. m z Os ? M. O e e

i 2

,1

~ gp TABLE 4.4-6 r-4N -4@ STEAM GENERATOR TUBE INSPECTION r, 1ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION C$ Sample Size Result Action Required Result Action Required Result Action Pequired -4 ro A minimum of C-1 None N/A N/A .7m N/A S Tubes per S. G. C-2 Plug defective tubes C-1 None N/A N/A and inspect additional Plug defective tubes C-1 None 2S tubes in this S. G. C-2 and inspect additional C-2 Pieg defective tunes 4S tubes in this S. G. Perform action for ca y ) C-3 ' C-3 result of first sample A Perform action for U-C-3 C-3 result of first N/A N/A sample C-3 Inspect all tubes in All other O D, this S. G., plug de-S. G.s are None N/A N/A fective tubes and C-1 N inspect 2S tubes in Some S. G.s each other S. G. Perform action for N/A N/A C-2 but no C-2 result of second g additional sample m Prompt notification S. G. are g to NRC pursuant C-3 a to specification Additional inspect all tubes in 6.9.1 S. G. is C-3 each S. G. and plug r+ defective tubes. y Prompt notification N/A. N/A to NRC pursuant m to specification .N 6.9.1 wa." S=3 % Where N is the number of steam generators in the unit, and n is the number of steam generators inspected ut n during an inspection N.

This page intentionally ' Oft blank. MILLSTONE - 11 NIT 2 3/4 4 79 AmendmentNo.77,431 2131 098

This page lef t blank intentionally, / t Amendment.No.37,Sh MILLSTONE - UNIT 2 3/4 4-7h '~, h,g -(' 2131 099 ~ ~

REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS 1.IMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System leakage detection systems shall be OPERABLE: a. A containment atmosphere particulate radioactivity monitoring

system, b.

The containment sump level monitoring system, and c. A containment atmosphere gaseous radioactivity monitoring system. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: a. With one of the above radioactivity monitoring leakage detection systems inoperable, operations may continue for up to 30 days provided: 1. The other two above required leakage detection systems are OPERABLE, and 2. Appropriate grab samples are obtained and analyzed at least once per 24 hours; otherwise, be in COLD SHUTDOWN within the next 36 hours. b. With the containment sump level monitoring system inoperable, restore the inoperable system to OPERABLE status within 7 days or be in COLD SHUTDOWN within the next 36 hours. SURVEILLANCE REQUIREMENTS 4.4.6.1 The leakage detection systems shall be demonstrated OPERABLE by: a. Containment atmosphere gaseous and particulate monitoring systems-performance of CHANNEL CHECK, CHANNEL LALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencier, specified in Table 4.3-3, and b. Containment sump level monitoring jystem-performance of CHANNEL CALIBRATION TEST at least once per 18 months. MILLSTONE - UNIT 2 3/44-8 100 ~.

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T t 300 F avg LIMITING CONDITION FOR OPERATION 3.5.2 Two separate and independent ECCS subsystems shall be OPERABLE with each subsystem comprised of: a. One OPERABLE high-pressure safety injection pump, b. One OPERABLE low-pressure safety injection pump, c. A separate and independent OPERABLE flow path capable of taking suction from the refueling water storage tank on a safety injection actuation signal and automatically trans-ferring suction to the containment sump on a sump recircu-lation actuation signal, and d. One OPERABLE charging pump with a separate and independent OPERABLE flow path from an OPERABLE Boric Acid Storage Tank via either sn OPERABLE Boric Acid Pump or a gravity feed connection. APPLICABILITY: MODES 1, 2 and 3*. ACTION: a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 48 hours or be in HOT SHUTDOWN within the next 12 hours. 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 I and the total accumulated actuation cycles to date.

With pressurizer pressure 1 1750 psia.

MILLSTONE - UNIT 2 3/4 5-3 Amendment No. 52' 2131 101 4;v '- + t.., e.,

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE: a. At least once per 31 days on a STAGGERED TEST BASIS by: 1. Verifying that each high-pressure safety injection pump: a) Starts automatically on a test signal. b) Develops a discharge pressure of > 1125 psig on recirculation flow. c) Operates for at least 15 minutes. 2. Verifying that each low-pressure safety injection pump: a) Starts automatically on a test signal. b) Develops a discharge pressure of > 162 psig on recirculation flow, c) Operates for at least 15 minutes. 3. Verifying that each charging pump: a) Starts automatically on a test signal. b) Operates for at least 15 minutes. 4. Verifying that each boric acid pump (when required OPERABLE per Specification 3.5.2.d): a) Starts automatically on a test signal. b) Develops a discharge pressure of > 98 psig on re;irculation flow. c) Operates for at least 15 minutes. 5. Verifying that upon a sump recirculation actuation signal, the containment sump isolation valves open. 6. Cycling each testable, automatically operated valve through at least one complete cycle. 7. Verifying the correct position for each manual valve not locked, sealed or otherwise secured in position. 8. Verifying the correct position' for each remote or auto-matica11y operated valve. 9. Verifying that each ECCS subsystem is aligned to receive electrical power from separate OPERABLE emergency busses. MILLSTONE - UNIT 2 3/4 5-4 AmendmentNo.h 2131 102

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

10. Verifying that the following valves are in the indicated l

position v'th power to the valve operator removed: Valve Famber Valve Function Valve Position 2-SI 306 Shutdown Cooling Open Flow Control 2-SI-660 SRAS Recirc. Open* 2-SI-660 SRAS Recirc. Open* 2-CH-434 Thennal Bypass Closed ** b. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sLmp and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed: 1. For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and 2. Of the areas affected within containment at the completion of containment entry when CONTAINMENT INTEGIRTY IS established. c. At least once per 18 months by: 1. Verifying automatic isolation of the shutdown cooling system from the Reactor Coolant System when the Reactor Coolant System pressure is above 300 psia. 2. A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or corrosion. 3. Verifying that a minimum total of 110 cubic feet of solid granular trisodium ph%phate dodecahydrate (TSP) is contained with the TSP storage baskets.. 4. Verifying'that when a representative sample of 0.35 1 0.05 lbs of TSP from a TSP storage basket is sutxaerged, without agitation, in 50 1 5 gallons of 180 1 10'F borated water from the RWST, the pH of the mixed solution is raised to > 6 within 4 hours.

To be closed prior to recirculation following LOCA.
- 2-CH-434, a manual valve, shall be locked closed.

Amendment No. 7, M,5k MILLSTONE - UNIT 2 3/4 5-5 2131 103

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) d. At least once per 18 nionths, during shutdown, by cycling each power operated valve in the subsystem flow path not testable during plant operation through one complete cycle of full

travel, e.

By a visual verification that each of the throttle valves in Table 4.5-1 will open to the correct position. This verifica-tion shall be perfonneo: Within 4 hours following the completion of each valve i. stroking operation, 2. Immediately prior to returning the valve to service after maintenance, repair, or replacement work is performed on the valve or its associated actuat.. or its control circuit, or 3. At least once per 18 months. f. By conducting a flow balance verification innediately prior to returning to service any portion of a subsystem after the comple-tion of a modification that could alter system flow characteris-tics. The injection leg flow rate shall be as follows: 1. HPSI Headers - the sum of the thrca lowest injection flows must be > 471 gpm. The sum of.the four injection flows must be < 675 gpm. 2. LPSI Header - the sum of the three lowest injection flows must be > 2370 gpm. The sum of the four injection flows must be - 4500 + ; RWST level N - 10(O x 200 ; 90% g. At least once per 18 months, during shutdown, by verifying that on a Safety Injection Actuation test signal: 1. The valves in the boron injection flow path from the boric acid storage tank via the boric acid pump and charging pump actuate to their required positions, and 2. The charging pump and boric acid pump start automatically. MILLSTONE - UNIT 2 3/4 5-6 Amendment No. 6,521 ). 2131 164 m

3/4.7 PLANT SYSTEMS 3.4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam'line code safety valves shall be OPERABLE. APPLICABILITY: MODES 1, 2 and 3. ACTION: a. With both reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2 and 3 may proceed provided, that within 4 hours, either the inoperable valve is restoced to OPERABLE status or the Power Level-High trip setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours, b. The provisions of Specification 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.7.1.1 Each main steam line code safety valve shall be demonstrated OPERABLE, with lift settings and orifice sizes as shown in Table 4.7-1, in accordance with Subsection IWV-3510 of Section XI of the ASME Boiler and Pressure Vessel Code, dated July 1,1974. MILLSTONE - UNIT 2 3/4 7-1 Amendment No. 529 2131 105 30i 'c:

TABLE 3.7-1 MAXIMUM ALLOWABLE ')0WER LEVEL-HIGH TRIP SET >0 INT WITH IN0PERABLE z f STEAM LINE SAFETY VA_VE5 DURING OPERATION WITi BOTH STEAM GENERATORS W oM Maxtmum Allowable Power Maximum Number of Inoperable Safety Level-High Trip Setpoint g Valves on Any Operating Steam Generator (Percent of RATED THERMAL POWER) Z 1 93.6 2 80.2 3 66.8 [ .. :~ [ T N o = ct M 2 Ccn mN.

TABLE 4.7-1 ii E STEAM LINE SAFETY VALVES 'l .:l 8 m VALVE NUMBERS LIFT SETTING (t.1%)_ ORIFICE SIZE l e en G 4.515 in,g

1 a.

2-MS-246 & 2-MS-247 1000 psia m 4.515 in.2 b. 2-MS-242 & 2-MS-254 1005 psia c. 2-MS-245 & 2-MS-249 1015 psia 4.515 in. N ,u d. 2-MS-241 & 2-MS-252 1025 psia 4.515 in.2 ~ w s N e. 2-MS-244 & 2-MS-251 1035 psia 4.515 in.2 'o""* N w f. 2-MS-240 & 2-MS-250 1045 psia 4.515 in.2 g. 2-MS-239, 2-MS-243, 1050 psia 4.515 in.2 2-M 248 & 2-MS-253 'E 8n N

PLANT SYSTEMS AUXILIARY FEE 0 WATER PUMPS LIMITING CONDITION FOR CPERATION 3.7.1.2 At least three steam generator auxiliary feedwater pumps shall be OPERABLE with: a. Two feedwater pumps capable of being powered from separate OPERABLE emergency busses, and b. One feedwater pump capable of being powered from an OPERABLE steam supply system. APPLICABILITY: MODES 1, 2 and 3. ACTION: With one auxiliary feedwater pump inoperable, restore at least three auxiliary feedwater pumps (two capabic of being powered from separate OPERABLE emergency busses and one capable of being powered by an OPER-ABLE steam supply system) to OPERABLE status within 48 hours or be in HOT SHUTDOWN within the next 12 hours. SURVEILLANCE REQUIREMENTS 4.7.1.2 Each auxiliary feedwater pump shall be demonstrated OPERABLE at least once per 31 days by: a. Starting each pump from the control room, b. Verifying that: 1. Each motor triven pump develops a discharge pressure of 3.1070 psig on recirculation flow, and 2. The steam turbine driven pump develops a discharge pres-sure of 2.1080 psig on recirculation flow. / o MILLSTONE - UNIT 2 3/4 7-4 I 2131 108 r ,\\ 4 =

PLANT SYSTEMS 3/4.7.8 HYDRAULIC SNUBBERS LIMITING CONDITION FOR OPERATION 3.7.8.1 All hydraulic snubbers listed in Table 3.7-1 shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one or more hydraulic snubbers inoperable, restore the inoperable snubber (s) to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.7.8.1 Hydraulic snubbers shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program: Each hydraulic snubber with seal matertal fabricated from a. ethylene propylene or other materials demonstrated compatible with the operating envirorment and approved as such by the NRC, shall be determined OPERABLE in accordance with the inspection -schedule of Table 4.7-3, by a visual inspection of the snubber. Visua1' inspections of the snubbers shall include, but are not necessarily limited to, inspection of the hydraulic fluid reservoirs, fluid connections, and linkage connections to the pising and anchors. Initiation of the Table 4.7-3 inspection ~scledule shall be made assuming the unit.was previously at the 6 Sonth inspection interval, b. Each hydraulic snubber with seal material not fabricated from l ethylene propyl ~ene or other materials demonstrated com3atible with the operating environment shall be determined OPERABLE at least once per 31 days by a visual inspection of the snubber. Visual inspections of the snubbers shall include, but are not necessarily limited to, inspection of the hydraulic fluid r reservoirs, fluid connections, and linkage connections to the piping and anchors. MIL,L. STONE. - UNIT 2 3/4 7-21 AmendmentNo.7J,(J,h 0il 7:>1 2131 109

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) c. At least once per 18 months during shutdown, a representative sample of at least 10 hydraulic snubbers or at least 10% of all snubbers listed in Table 3.7-1, whichever is less, shall be selected and fur.ctionally tested to verify correct pisten movement, lock up and bleed. Snubbers greater than 50,000 lb. capacity may be excluded from functional testing requirements. Snubbers selected for functional testing shall be selected on a rotating basis. Snubbers identified as either "Especially Difficult to Remove" or in "High Radiation Zones"1nay be exempted from functional testing provided these snubbers were demonstrated OPERABLE during previous functional tests. Snubbers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation. For each snubber found inoperable during these functional tests, an additional minimum of 10% of all snubbers or 10 snubbers, whichever is less, shall also be functionally tested until no more failures are found or all snubbers have been functionally tested. AmendmentNo.JJ,A75b MILLSTONE - UNIT 2 3/4 7-22 4!- 2131 110

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) e. At least once per 18 months by performing a system functional test which includes simulated automatic actuation of the system throughout its operating sequence, and: 1. Verifying that each pump develops at least 1800 gpm at a system head of 100 psig, 2. Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel, and 3. Verifying that each high pressure pump starts (sequentially) to maintain the fire suppression water system pressure > 75 psig. f. At least once per 3 years by performing a flow test of the sys+.em in accordance with Chapter b Section 11 of the Fire Protection Handbook, 14th Edition, published by the National Fire Protection Association. 4.7.9.1.2 The fire pump diesel engine shall be demonstrated OPERABLE: a. At least once per 31 days by verifying; 1. The fuel storage tank contains at least 125 gallons of fuel, and 2. The diesel starts from ambient conditions and operates for at least 20 mi.nutes. b. At least once per 92 days by verifying that a sample of diesel fuel from the fuel storage tank, obtained in accordance with ASTM-D270-65, is within the acceptable limits specified in Table 1 of ASTM D975-74 when checked for viscosity, water and sediment. c. At least once per 18 months by: l 1. Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for the class of service, and 2. Verifying the diesel starts from ambient conditions on the auto-start signal and operates for > 20 minutes while loaded with the fire pump. , MILLSTONE - UNIT 2 3/4 7-35 Amendment No. M, n 5 2 t 2131 111

PLANT SYSTEMS, SURVEILLANCE REQUIREMENTS (Continued) 4.7.9.1.3 The fire pump diesel starting 24-volt battery bank and charger shall be demonstrated OPERABLE: a. At least once per 7 days by verifying that: 1. The electrolyte level of each battery is above the plates, and 2. The overall battery voltage is > 24 volts. b. At least once per 92 days by verifying that the specific gravity is appropriate for continued service of ~he battery. c. At least once per 18 months by verifying that: 1. The batteries, cell plates and battery racks show no visual indication of physical damage or abnormal deterio-ration, and 2. The battery-to-battery and terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material. MILLSTONE - UNIT 2 3/4 7-36 AmendmentNo./ji,43 2131 112 i

L j 3/4.10 SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of CEA worth and shutdown margin provided reactivity equivalent to at least the highest estimated CEA xorth is available for trip insertion from OPERABLE CEA(s). APPLICABILITY : MODES 2 and 3. ACTION: a. With any full length CEA not fully inserted and with less than the above reactivity equivalent available for trip insertion, imediately initiate and continue boration at > 44 gpm of > 1720 ppm boric acid solution or its equivalent until the 1 511UTDOWN MARGIN required by Specification 3.1.1.1 is restorc.i. b. With all full length CEAs inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at > 44 gpm of > 1720 ppm boric acid I solution or its equivalent until the 511UTDOWi MARGIN required by Specification 3.1.1.1 is restored. SURVEILLANCE REQUIREML:?S 4.10.1.1 The positien of each full length CEA required either partially or fully withdrawn shall be determined at least once per 2 hours. 4.10.1.2 Each CEA not fully inserted shall be demonstrated capable of full insertion when tripped frun at least the 50% withdrawn position within 24 hours prior to reducing the SHUTDOWN MARGIN to less than the limits of Speci fication 3.1.1.1. MILLSTONE - UNIT 2 3/4 10-1 Amendment No. /E, ag,52 2131 113

SPECIAL TEST EXCEPTIONS GROUP HEIGHT AND INSERTION LIMITS . LIMITING CONDITION FOR OPERATION 3.10.2 The requirements of Specifications 3.1.1.4.3.1.3.1,3.1.3.2,3.1.3.5,l 3.1.3.6, 3.2.2, 3.2.3 and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided: The THERMAL POWER is restricted to the test power plateau which a. shall not exceed 85% of RATED THERMAL POWER, and b. The limits of Specification 3.2.1 are maintained and detennined as specified in Specification 4.10.2 below. APPLICABILITY: H0 DES 1 and 2. ACTION: With any of the limits of Specification 3.2.1, being exceeded while the requirements of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.2, 3.1.3.5, 3.1.3.6, l 3.2.2, 3.2.3 and 3.2.4 are suspended, immediately: Reduce THERMAL POWER sufficiently to satisfy the requirements a. of Specification 3.2.1 or b. Be in HOT STANDBY within 2 hours. SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be detarmined at least once per hour during PHYSICS TESTS in which the requirements of Specifications 3.1.1.4, l 3.1.7.1, 3.1.3.2, 3.1.3.5, 3.1.3.6, 3.2.2, 3.2.3 or 3.2.4 are suspended and shall be verified to be within the test power plateau. 4.10.2.2 The linear heat rate shall be determined to be within the limits of Specification 3.2.1 by monitoring it continuously with the Incore Detector Monitoring System pursuant to the requirements of Specifi-cations 4.2.1.3 and 3.3.3.2 during PHYSICS TESTS above 5% of RATED THEMAL POWER in which the requirements of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.2, l 3.1.3.5, 3.1.3.6, 3.2.2, 3.2.3 or 3.2.4 are suspended. MILLSTONE - UNIT 2 3/4 10-2 Amendment No. E,6'd a l 2131 114 f

3/4.2 POWER DISTRIBUTION LIMITS BASES 3/4.2.1 LINEAR HEAT RATE The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200"F. Either of the two core power distribution monitoring systems, the Excore Detector Monitoring System and the Incore Detector Monitoring System, provide adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its limits. The Excore Detector Monitoring System perfonns this function by continu-ously monitoring the AXIAL SHAPE INDEX with two OPERABLE excore neutron flux detectors and verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits of Figure 3.2-2. In conjunction with the use of the excore monitoring system and in establishing the AXIAL SHAPE INDEX limits, tha following assumptions are made: 1) the CEA insertion limits of Specifications 3.1.3.2, 3.1.3.5 and 3.1.3.6 are satisfied, 2) the flux peaking augmentation factors are as shown in Figure 4.2-1, 3) the AZIMUTHAL POWER TILT restrictions of Specification 3.2.4 are satisfied, and 4) the TOTAL PLANAR RADIAL PEAKING FACTOR does not exceed the limits of Speci-fication 3.2.2. The Incore Detector Monitoring System continuously provides a direct measure of the peaking factors and the alarms which have been established for the individual incore detector segments ensure that the peak linear heat rates will be maintained V. thin the allowable limits of Figure 3.2-1 The setpoints for these alarv, d.nclude allowances, set in the conservative directions, for 1) flux peabr.g augmentation factors as shown in Figure 4.2-1, 2) a measurement-calc :lational uncertainty factor of 1.07, 3) an l engineering uncertainty facts r.' 1.03, 4) an allowance of 1.01 for axial fuel densification and thermal expansion, and 5) a THERMAL POWER measur-ement uncertainty factor of 1.02, 3/4.2.2, 3/4.2.3 and 3/4.2.4 TOTAL PLANAR AND INTEGRATED RADIAL PEAKING FACTORS-FhANDF T AND AZIMUTHAL POWER TILT - T r q The limitiations on F snd T are provided to ensure that the q assumptions used in the analyst's for establishing the Linear Heat Rate and Local power Density - High LCOs and LSSS setpoints remain valid during operation at the various allowable CEA group insertion limits. T The limitations on F and T are provided to ensure that the assumptions used in the analysis establ9shing the DNS Margin LCO, and Thermal Margin / r Low Pressure LSSS setp ints remain valid during operation at the various o MILLSTONE - UNIT 2 B 3/4 2-1 Amendment No. #,52' 215\\ \\\\5

POWER DISTRIBUTION LIMITS BASES allowable CEA group insertion limits. IfFfy,F T or T exceed their 7 q basic limitations, operation may continue under the additional restrictions imposed by the ACTION statements since these additional restrictions provide adequate provisions to assure that the assumptions used in establishing the Linear Heat Rate, Thermal krgin/ Low Pressure and Local Power Density - High LCOs and LSSS setpoints remain valid. An AZIMUTHAL POWER TILT > 0.10 is not expected and if it should occur, subsequent operation would be restricted to only those operations required to identify the cause of this unexpected tilt. he value of T that must be uecd in the equation F =Fy (1 + T ) q q and F =Fr (1 + T ) is the measured tilt. r q T The surveillance requirements for verifying that F ,F and Tq are within their limits provide assurance that the actual values of Fy, T VertfyingFhandF T F and T do not exceed the assumed values. y 7 q after each fuel loading prior to exceeding 75% of RATED THERMAL POWER provides additional assurance that the core was properly loaded. 3/4.2.6 DNB MARGIN The limitations provided in this specification ensure that the assumed margins to DNB are maintained. The limiting values of the parameters in this specification are those assumed as the initial conditions in the accident and transient analyses; therefore, operation must be maintained within the specified limits for the accident and transient analyses to remain valid. MILLSTONE - UNIT 2 B 3/4 2-2 Amendment No. 5,521 2131 116

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above 1.30 during all normal operations and anticipated transients. STARTUP and POWER OPERATION may be initiated and may proceed with one or two reactor coolant pumps not in operation after the setpoints for the Power Level-High, Reactor Coolant Flow-Low, and Thermal Margin / Low Pressure trips have been reduced to their specified values. Reducing these trip setpoints ensures that the DNBR will be maintained above 1.30 during three pump operation and that during two pump operation the core void fraction will be limited to ensure parallel channel flow stability within the core and thereby prevent premature DNB. A single reactor coolant loop with its steam generator filled above the low level trip setpoint provides sufficient heat removal capability for core cooling while in MODES 2 and 3; however, single failure consi-derations require plant cooldown if component repairs and/or corrective actions cannot be made within'the allowable out-of-service time. The restrictions on starting a Reactor Coolant Pump during MODES 4 and 5 with one or more RCS cold legs < 275'F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 43 F (31*F when measured by a surface contact instrument) above the coolant temperature in the reactor vessel, 3/4.4.2 and 3/4.4.3 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 295,000 lbs per hour of saturated steam at the valve setpoint. The relief capacity of a single safety valve is adequate to relieve any overpresst. condition which could occur during shutdown. In the event that no safeb.alves are OPERABLE, an operating shutdown cooling loop, connected to the RCS, provides overpressure relief capa-bility and will prevent RCS overpressurization. MILLSTONE - UNIT 2 B 3/4 4-1 Amendment No. 50 lt :::: 2\\51 \\\\1

REACTOR COOLA!J SYSTEM BASES 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 these valves is sufficient to limit the Reactor Coolant System pressure to within its Safety Limit of 2750 psia following a complete loss of turbine generator load while operating at RATED THERMAL POWER and assuming no reactor trip until the first Reactor Protective System trip setpoint (Pressurizer Pressure-High) is reached (i.e., no credit is taken for a direct reactor trip on the loss of turbine) and also assuming no operation of the pressurizer power operated relief valve or steam dump valves. 3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and is capable of accommodating pressure surges during operation. The steam btbble also protects the pressurizer code safety valves and power operated relief valve against water relief. The power operateil reitef valve and steam bubble function to relieve RCS pressure during ail design transients. Operation of the power operated relief valve in ccnjunction with a reactor trip on a Pressurizer-- Pressure-High signal, minimizes the undesirable opening of the spring-loaded pressurizer code safety valves. 3/4.4.5 STEAM GENEMTORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservire inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential l in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice taspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can b1 tak~en. The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible c6rrosion of the steam generator tubes. If the secondary coolant chemigtry is not mai'ntained within these limits, localized corrosion may likely result in stress corrosion cracking. MILLSTONE - UNIT 2 B 3/4 4-2 Amendment No. Eg,5,52' 2}hl c j

REACTOR COOLANT SYSTEM BASES The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage a 0.5 GPM,per steam generator). Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal oleration and by postulated acetdents. Operating plants have demonstrated tTat primary-to-secondary leakage of 0.5 gallon per minute can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged. Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging will be required for all tubes with imperfec-tions exceeding the plugging limit of 40% of the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness. Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to Specification 6.R.1 prior to resumption of plant operation. Such cases will be considered by the Coninission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary. i t , MILLSTONE - UNIT 2 B 3/4.4-2a Amendment No. 77, y,3 1 2131 119

3/4,7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limtted to within 110% (1100 psig) of its design pressure of 1000 psig during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).. The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code,1971 Edition. Thetotalrelgevingcapacityforall valves on all of the steam lines is 12.7x10 1bs/hr which is 108 percent 6 of the total secondary steam flow of ll.8x10 lbs/hr at 100% RATED THERMAL POWER. STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Level-High channels. The reactor trip setpoint reductions are derived on the following bases: for two loop operation SP = IX) ~ III (Y) X 107 X where: reduced reactor trip sett,oint in percent of RATED SP a THERMAL POWER maximum number of inoperable safety valves per steam V a line MILLSTONE - UNIT 2 B 3/4 7-1 Amendment No.52' n..im 2131 120

~ lPLANTSYSTEMS BASES Power Level-High Trip Setpoint for two loop operation 107 = Total relkr.ig capacity of all safegy valves per X = steam line in 1bs/ hour = 6.35 x 10 lbs/ hour Maximum relieving capacity of any one safety valve Y a 7.94210 lbs/ hour in 1bs/ hour = 3/4.7.1.2 AUXILIARY FEEDWATER PUMPS The OPERABILITY of the auxiliary feedwater pumps ensures that the Reactor Coolant System can be cooled down to less than 300 F from normal operating conditions in the event of a total loss of off-site power. Either two motor driven pumps or the steam driven pump have the requited capacity to provide sufficient feedwater flow to remove reactor decay heat and reduce the RCS tenperature to < 300*F where the shutdown cooling system may be placed into operation f6r continued cooldown. 3/4.7.1,3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available for cooldown of the Reactor Coolant System to less than 300*F in the event of a total loss of off-site power, The minimum water volume is sufficient to maintain the RCS at HOT STANDBY conditions for 10 hours with steam discharge to atmosphere. 3/4.7,1,4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction 1 MILLSTONE - UNIT 2 B 3/4 7-2 knendment No.52 ~' - ., c. y}}

v t e Letter Sequence Request
TAC:46174, (Closed, Approved)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance... Supplement Administration Meeting, Meeting, Meeting, Meeting Questions 14 March 1979 Answers Results Approval Other: ML19207C172, ML19209B403, ML19256A848, ML19259A812, ML19263B926, ML19270G086, ML19273B549, ML19274D206, ML19276E491, ML19276F808, ML19281A096, ML19282C563, ML19282C601, ML19282C615, ML19289F472, ML19317H313, ML20037A204, ML20037A229, ML20049A130, ML20062G381, ML20062G387, ML20064C234, ML20064D492, ML20064E453, ML20064H429, ML20076A802, ML20076A836, ML20076A843, ML20076A904, ML20076A952, ML20076B087, ML20076B088, ML20076B089, ML20244A548, ML20244A585
MONTHYEAR1979-03-02 [Table View]

ML19269E355

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