Text

Amend 61 to License DPR-69,authorizing Plant Operation During Cycle 5 at Rated Thermal Power of 2,700 Mwt
	ML20028D915
Person / Time
Site:	Calvert Cliffs
Issue date:	01/10/1983
From:	Lainas G; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20028D916	List: ML20028D915; ML20028D931; ML20028D938;
References
	NUDOCS 8301200085
	Download: ML20028D915 (63)
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~ BALTIM0RE GAS AND ELECTRIC COMPANY DOCKET NO. 50-318 CALVERT CLIFFS NUCLEAR POWER PLANT UNIT N0. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 61 License No. DPR-69 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Baltimore Gas & Electric Company (the licensee) dated October 15, 1982, as supplemented November 17, 1982, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the comnon defense and security or to the health and safety of the public; and' E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. 8301200005 830110 gDRADOCK 050003g8

. 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.2 of Facility Operating License No. DPR-69 is hereby amended to read as follows: 2. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 61, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. Within 30 days after the effective date of this amendment, or such other time as the Commission may specify. the licensee shall satisfy any applicable requirement of P.L. 97-425, dated January 7,1983. 4. This license amendment is effective as of the date of its issuance. FOR Tile NUCLEAR REGULATORY COMMISSION (. / ' Gus C. Lainas, Assistant Director for Operating Reactors Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: January 10, 1983 i

ATTACHMENT TO LICENSE AMENDMENT NO. 61 FACILITY OPERATING LICENSE N0. DPR-69 7 DOCKET N0. 50-318 Replace the following p, ages c.~ the Appendix A Technical Specifications with the enclosed pages as indicated.- 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 completeness. Page 2-2 3/4 3-17 2-9 3/4 3-19 2-10 3/4 3-20 2-11 3/4 3-21 B 2-1 3/4 3-23 B 2-3 3/4 3 B_2-4 3/4 3-42 B 2-5 B 3/4.1-1 B 2-6 B 3/4 1-2 B 2-7 -3/4 7-5 3/4 1-1 5-4 3/4 1-5 3/4 1-6 3/4 1-27 3/4 2-2 3/4 2-4 3/4 2-4a (new) 3/4 2-6 3/4 2-7 3/4 2-7a.(new) 3/4 2-8 3/4 2-8 (cont.) 3/4 2-9 3/4 2-10a (new) 3/4 2-11 3/4 2-13 3/4 2-14 3/4 3-4 3/4 3-6 3/4 0-14 3/4 3-15 O O

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s [ l 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS ~ 2.1 SAFETY LIMITS e REACTOR CORE 2.1.1 The combination of THERMAL POWER,-pressurizer pressure, and highest operating loop cold leg coolant temperature shall not exceed the limits shown in Figures 2.1-1, 2.1-2, 2.1-3 and 2.1-4 for the various combinations .l of two, three and,four reactor coolant pump operation. APPLICABILITY: MODES 1 and 2. ACTION: Whenever the point defined by the combination of the highest operating loop cold leg temperature 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. . t. APPLICABILIT'- 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 i'ts 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. l CALVERT CLIFFS-UNIT 2 2-1 Amendment No. 6

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N / oc FLUCTUATIONS 9 9 n. E s40 - VALID FOR AXIAL SHAPES AND c INTEGRATED ROD RADIAL PEAKING FACTORS LESS THAN OR EQUAL TO THOSE ON FIGURE B2.1-1 O Yo s20 ~ o 2 REACTOR OPERATION LIMITED TO LESS 3 THAN s80*F BY ACTUATION OF THE 2 SECONDARY SAFETY VALVES x y ~ II ACCEPTABLE g 480 OPERATION >2 s s 4 kk 4 =o g i I I I I I I I l y 0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 2.0 g FRACTION OF. RATED THERMAL POWER ~ on W Figure 2.1-1 Reactor Core Thermal Margin Safety Limit Four Reactor Coolant Pumps Operating o

TABLE _2.2-1 (Cont'd) n ?

f, REACTOR PROTECTIVE IllSTRUMENTATION TRIP SETPOINT LIMITS E

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES' a f 4. Pressurizer Pressure - High 1 2400 psia-1 2400 psia E 5. Containment. Pressure - High 1 4 psig 1 4 psig 6. Steam Generator Pressure - Low (2) > 685 psia > 685 Psia l ~ m 7. Steam Generator Water Level - Low > 10 inches below top > 10 inches, below top of feed ring. of feed ring. 8. Axial flux offset (3) Trip setpoint adjusted to Trip setpoint adjusted to ~i not exceed the limit lines not exceed the limit lines of Figure 2.2-1. of Figure 2.2-1. 9. Thermal Margin / Low Pressure (1) a. Four Reactor Coolant Pumps Trip setpoint adjusted to ' Trio setpoint adjusted to Operating .not exceed the limit lines be not less than the larger of Figures 2.2-2 and 2.2-3. .of (1) the value.calculatbd from Figures 2.2-2 and '2.2-3 and (2).1875 psig. b. Steam Generator Pressure -< 135 psid -< 135 psid Dif ference - High (1) F 10. Loss of Turbine -- Hydraulic > 1100 psig > 1100 psig Fluid Pressure - Low (3) S 11. Rate of Change of Power - High (4) 1 2.6' decades per minute 1 2.6 decades per minute 0 TABLE NOTATION m b (1) Trip may be bypassed be}ow 10 % of RATED THERMAL POWER; bypass shall be automatically remo'ved when ~4 THERMAL POWER is > 10" % of RATED THERf1AL P0HER. cre M .5 i x 1.j

. ; 3.. 'r 0 y se 1 .. G TABLE:2.2-1 (Cont'd) T .m:o TABLEIl0TATIONS (Cont'd) 9 G. (2) Trip may be manually bypassed. below 785 psia; bypass shall be' automatically removed 'at or above' 785 psia. l. ' D -(3) Trip:may be bypassed below 15% of RATED TilERitAL POWER; bypass shall.be automatidellyfremoved when - ci THERilAL POWER is > 15% of RATED TIIEIUIAL -I'0WER. = -4 ] (4) Trip may be bypassed below 10 % and above' 12% of ~ RATED THERttAL POWER.- v .M' 4'", b o s- [. iir g an .F g

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I I I I l (0,1.20) 1.20 UNACCEPTABLE UNACCEPTABLE OPERATION OPERATION 1.10 REGION REGION 5 mO 1.00 (-0.2.1.00) (0.2,1.00) $s .5 h 0.90 k =. u O ACCEPTABLE OP R lON 0.80 - (-0.4, 0.80) i > (0.4, 0.80) reg p 4 E ~ 0.70 0.60 .~ l 0.50 0.6 0.4 -0.2 0 0.2 0.4 0.6 i P'ERIPHERAL AXIAL SHAPE INDEX, Y, l Figure 2.2-1 Per!pheral Axial Shape Index, Y, vs Fraction of Rated Thermal Power CALVERT CLIFFS - VillT 2 2-11 A,aendment tio. 9. 31, 61 i l

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y 2.1 SAFETY LIMITS l BASES 2.1.1 REACTOR CORE. The restrictions of this safety. limit prevent overheating of the fuel cladding 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 22.0 kw/ft. Centerline fuel melting will not occur l for this peak linear heat rate. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. Operation above 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 measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temper-ature and Pressure have been related to DNB through the CE-1 correlation. The.CE-1 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distri-betions. The local DNB heat flux ratio, DNBR, defined as the ration of-c 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 mininun value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.23 l This value corresponds to a 95 percent probability at a 95 percent con-fidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions. The curves of Figures 2.1-1, 2.1-2, 2.1-3 and 2.1-4 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperature of varices pump combinations for which the minimum DNBR is no less than 1.23 for the family of axial shapes and l corresponding radial peaks shown in Figure B2.1-1. The limits in Figures 2.1-1, 2.1-2, 2.1-3 and 2.1-4 were calculated for reactor coolant inlet temperatures less than or equal to 580*F. The dashed line at 580*F coolant inlet temperature is not a safety limit; however, 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 110% of RATED THERMAL POWER is prohibited by the high power level trip setpoint specified in l l \\, CALVERT CLIFFS - UNIT 2 B 2-1 Amendment No. 78, 37,,61 I

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[L l SAFETY LIMITS BASES Table 2.1-1. The area of safe operation is below and to the left of these lines. The conditions for the Thermal Margin Safety Limit curves in Figures 2.1-1, 2.1-2, 2.1-3 and 2.1-4 to be valid are shown on the figures. The reactor protective system in combination with the Limiting Conditions for Operation, is designed to prevent any anticipated combina-tion of transient. conditions for reactor coolant system temperature, pressure, and THERMAL POWER level that would result in a DNBR of less than 1.23 and preclude the existence of flow instabilities. l 2.1.2 REACTOR COOLANT SYSTEM PRESSUP.E The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere. The reactor pressure vessel and pressurizer are designed to Section !!I, 1967 Edition, of the ASME Code for Nuclear Power Plant Com which pennits a maximum transient' pressure of 110% (2750 psia) ponents of design pressure. The Reactor Cnolant System piping, valves and fittings, are designed to ANSI B 31.7, Class I,1969 Edition, 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 associtted code requirements. ~ The entire Rea: tor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to initial operation. CALVERT CLIFFS - UNIT 2 B 2-3 Amendment No. 78,7I, 61 C .[ g-

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2. 2.1 REACTOR TRIP SETPOINTS 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'Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Operation with a trip set less conservative than its Trip Setpoint but within its speci-fied Allowable Value is acceptable en the basis that each Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses. Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels cnd 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. lhe Power Level-High trip setpoint is. operator adjustable and can be set no higher than 10% 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 107.0% of RATED THERMAL POWER and a minimum setpoint of 30% of RATED THERMAL POWER. Adding 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 would be actuated is 110% of RATED l THERMAL POWER, which is the value used in the safety analyses. Reactor Coolant Flow-Low j 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 (,- e CALVERT CLIFFS-UNIT 2 B 2-4 Amendment No. AB, g [ I l ~ _,

LIMITING SAFETY SYSTEM SETTINGS BASES operation of.the reattor at reduced power if one or two reacto'r coolant pumps are taken out of service. The low-flow trip setpoints and Allowable Values for the various reactor coolant pump combinations have been derived in consideration of instrument errors and response times of equipment involved to maintain the DNBR above 1.23 under normal 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 Thermal 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 1.23 during l normal operational 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 pressurizer code safety valves and its concurrent operation with the power-operated relief valves avoids the undesirable operation of the pressurizer code safety valves. Containment Pressure-High The Containment 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-Low 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 685 psia l is sufficiently below the full-load operating point of 850 psia so l as not to interfere with normal 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 + 85 psi in the accident analyses.which was based on the Main Steam Eine Break [ event. o CALVERT CLIFFS - UNIT 2 B 2-5 Amendment No. 4, af, '61 l C 4

LIMITING SAFETY SYSTEM SETTINGS BASES Steam Generator Water Level 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 pressure of the reactor coolant system will not exceed its Safety Limit. The specified setpoint in combination with the auxiliary _ feedwater actuation system ensures,that sufficient water inventory exists in both steam generators to remove decay heat following a loss.of main feedwater flow event. Axial Flux Offset The axial flux offset trip is provided to ensure that excessive axial peaking will not cause fuel damage. The axial flux offset is determined from the axially split excore ' etectors. The trip setpoints d ensure that neither a DNBR of less than 1.23 nor a peak linear heat rate which corresponds to the temperature for fuel centerline melting will exist as a consequence of axial power maldistributions. These trip set-points were derived from an analysis of many axial power shapes with .y ellowances for instrumentation inaccuracies and the uncertainty associated with the excore to incore axial flux offset relationship. Thermal Margin / Low Pressure The Thermal Margin / Low Pressure trip is provided to prevent operation when the DNBR is less than 1.23. The trip is initiated whenever the reactor coolant system pressure signal drops below either 1875 psia or a computed value as described l below, whichever is higher. The computed value is a function of the higher of AT power or neutron power, reactor inlet temperature, and the l number of reactor coolant pumps operating. The minimum value of reactor coolant flow rate, the maximum AZIMUTHAL POWER TILT and the maximum CEA l l' deviation permitted for continuous operation are assumed in the genera-t tion of this trip function. In addition, CEA group sequencing in accor-dance with 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 l operational occurrence prior to a Power Level-High trip is assumed. CALVERT CLIFFS - UNIT 2 B 2-6 Amendment No.JE,If,6 1 l l ^ l 6

i LIMITING SAFETY SYSTEM SETTINGS BASES The Thermal Margin / Low Pressure trip setpoints include allowances for equipment response time, measurement uncertainties,. processing error.- and a further' allowance of 40 psia to compensate for the time delay associated with providing effective termination of the occurrence that exhibits the most rapid decrease in margin to the safety limit. Asymmetric Steam ~ Generator Transient Protection Trip Function (ASGTPTF) The ASGTPTF utilizes steam generator pressure inputs to the TM/LP. calculator, which causes a reactor trip when the difference in pressure between the two steam generators exceeds the trip setpoint. The ASGTPTF is designed to provide a reactor trip for those Anticipated Operational Occurrences associated with secondary system malfunctions which result in asymmetric primary loop coolant temperatures. The most limiting event is the loss of load to one steam generator caused by a single Main Steam - Isolation Valve closure. The equipment trip setpoint and allowable values are calculated to account for instrument uncertainties, and will ensure a trip at or before reaching the anclysis setpoint. Loss of Turbine 9 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 transient, thus extending the service life of these valves. 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. Rate of Change of Power-High The Rate of Change of Power-High trip is provided to protect the core during startup operations and its use serves as a backup to the administra-tively enforced startup rate limit. Its trip setpoint does not correspond to a Safety Limit and 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. i, l l CALVERT CLIFFS - UNIT 2 B 2-7 Amendment No. 9, JS, 31,g 1 l L

l 3/4.1 REACTIVITY CONTROL SYSTE5-3/4.1.1 BORATION CONTROL SHUTDOWitl'tARGIN - T,yg > 200*F. LIMITING CONDITION FOR OPERATION ~ 3.1.1.1 The' SHUTDOWN MARGIN shall be > 5.2%* ak/k. l APPLICABILITY: MODES 1, 2**,-3 and 4. ACTION: With the SHUTDOWN. MARGIN < 5.2%* ak/k, immediately initiate and continue [l boration at > 40 gpm of 2300 ppm boric acid solution or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be > 5.2%* ak/k: l a. Within one hour after detection of an inoperable CEA(s) and at least once per 12 hours thereafter while the CEA(s) is inoperable. If the inoperable CEA is immovable or untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at N least equal to the withdrawn worth of the innovable or untrippable CEA(s), b. When in MODES 1 or 2", at least once per 12 hours by verifying that CEA group withdrawal is within the Transient Insertion Limits of Specification 3.1.3.6. N c. When in MODE 2 , within 4 hours prior to achieving. reactor criticality by verifying that the predicted critical CEA position is within the limits of Specification 3.1.3.6. d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below,- with the CEA groups at the Transient Insertion Limits of-Specification 3.1.3.6. Adherence to Technical Specification 3.1.3.6 as specified in Surveillance Requirements 4.1.1.1.1 assures that there is sufficient available shut-down IEargin to match the shutdown margin requirements of the safety analyses.

- See Special Test Exception 3.10.1.

With.K,ff 1 1.0.

H With K,ff < l.0. s CALVERT CLIFFS - UNIT 2 3/4 1-1 Amendment No. ), M,37.61

N REACTIVITY CONTROL SYSTEMS ') SURVEILLANCE REQUIREMENTS (Continued) WheA in_ MODES 3 or 4, at least once per 24 ho'urs by c'on-e. sideration of the following factors: '1. Reactor coolant system boron concentration, ~ 2. CEA position, 3. Reactor coolant system average temperature, 4. Fuel burnup based on gross thermal energy generation, 5. Xenon concentration, and 6. Samarium concentration. 4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within + 1.0% ak/k at least ' once per 31 Effective Full Power Days (EFPD). ThTs comparison shall consider at least those factors stated in Specification 4.1.1.1.1.e, above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading. c t i \\.* e l CALVERT CLIFFS-UNIT 2 3/4 1-2 l l ~ F

1 REACTIVITY CONTROL SYSTEMS - 30DERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.4 The moderator temperature coefficient (MTC) shall be: Less positive than 0.5 x 10-4 ok/k/*F whenever THERMAL a. POWER is 1 70% of RATED THERMAL POWER, b. Less positive than 0.2.x 10-4 Ak/k/*F whenever THERMAL POWER is > 70% of RATED THERMAL POWER, and Less negative than -2.2 x 10~4 Ak/k/*F at RATED THERMAL l c. POWER. APPLICABILITY: MODES 1 and 2*# ACTION: With the moderator temperature coefficient outside any one of the above limits, be in at least HOT STANDBY within 6 hours. ( SURVEILLANCE REQUIREMENTS 4.1.1.4.1 The MTC shall be determined to be within its limits by confirmatory measurements. MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the above limits.

With K,ff > l.0.

See Special Test Exception 3.10.2.

CALVERT CLIFFS - UNIT 2 3/4 1-5 Amendment No. 78, 31 E 6 _v-, n - - - r

i REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.1.1.4.2 The MTC shall txt determined at the following frequen'cies and THERMAL POWER conditions during each fuel cycle: a. Prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading. b. At any THERMAL POWER; above 90% of RATED THERMAL POWER,'within 7 EFPD af'ter initially reaching'an equilibrium condition at or above 90% of RATED THERMAL POWER after each fuel loading. At any THERMAL Powe','within 7 EFPD after reaching a RATED c. r THERMAL POWER equilibrium boron concentration of 300 ppm. l v 1 CALVERT CLIFFS-UNIT 2 3/4 1-6 Amendment No.6 1 ,n

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_. - 8g = Ipasut %gg 0 g dno.19 owe m m g5Edi-p - g* 'E z (uopiesul %sg Oyyz-3 = @ g dnoJD '0'll m2 tuopassui %gg @ g dnoJo co g) g9 - R "g

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e i o 40 0 i I l I i i l l I ~WE_o% E 8 R E R S S g R R "o E E s6 6 o 6 o o e 6 s =E a 34 h 35 @ NOl1VN18WO3 d3W DN11SIX3 W03 W3 mod lVWW3H13'l0VM011V 30 NOll3VWd o o a D. <O CALVERT CLIFFS - UNIT 2 3/4 1-27 Amenhent No. Si 18,6 1 i .. l

O 3/4.2 POWER DISTRIBUTI0ft 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: With the linear heat rate exceeding its limits, as indicated by four or more coincident incore channels or by the AXIAL SHAPE INDEX outside of the power dependent control limits of Figure 3.2-2, within 15 minutes initiate corrective action to reduce the linear heat rate to within the limits and either: a. Restore the linear heat rate to within its limits within one hour, or i b. -Be in at least HOT STANDBY within the next 6 hours. s SURVEILLAfiCE REQUIREMENTS 4.2.1.1 The provisions of Specification 4.0.4 are not applicable. 4.2.1.2 The linear heat rate shall be determined to be within its limits by continuously monitoring the core power distribution with either the excore detector monitoring system or with the incore detector monitoring ~ system. 4.2.1.3 Excore Detector Monitoring System - The excore detector moni-toring system may be used for monitoring the core power distritution by: a. Verifying at least once per 12 hours that the full length CEAs are withdrawn to and maintained at or beyond the Long Term Steady State Insertion Limit 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. CALVERT CLIFFS-UNIT 2 3/42-1 Amendment No. 9, 18 r

d p0tlER DISTRIBUTION LIMITS S_URVEILLANCE REQUIREMENTS (Continued) ~ c. Verifying at least once per 31 days that the AXIAL SHAPE INDEX is maintained within the 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 maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination. 2. N is the maximum allow 9ble fraction of RATED THERMAL POWER as determined by the F curve of Figure 3.2-3b. l xy 4.2.1.4 Incore Detector Monitorina 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 alarms: e a. Are adjusted to satisfy the requirements of the core power distribution map which shall be updated at least once per 31 days of accumulated operation in MODE 1. 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, 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. v CALVERT CLIFFS-UNIT 2 3/4 2-2 AmendmentNo.5,9,76,JS,h4,;y

8 m. 3 E D m m z a o z p a O S <- -g gn m = c w E E C 5 E. % w o c. o - e w o a. M r w a a m w a c: c: e a m o e o u e5 w t" c U 8 P E x 'u-'u a o c. u m 2 i 3 o 2 w = .2 E 8 o s W 3 ? (Ho1V83aoW + av73 -land 14/MM '31V8 IV3H UV3 Nil MV3d 318VMO'llV ~ CALVERT cliff 5 - UNIT 2 3/* 2-3 Amendment No.5. 18

' 10 l l 1, I ,I (-0.Os,1.0) - - (0.12,1.0) 1.00 n0 0.90 UNACCEPTABLE UNACCEPTABLE 4 OPERATION OPERATION ] REGION REGION w 5 i 5 g 0.80 3 S. u. O R opg ggLE ACC g 0.70 ( 0.3. 0.701 < . (0.3, 0.70) i g REGION {< E i 0.60 i I l i I I o.So ,0.6 -0.6 -0.4 -0.2 0 0.2 0.4 l PERIPHERAL AXIAL SHAPE INDEX, Yg Figure 3.2 2 Linear Heat Rate Axial Flux Offset Control limits C.1LVERT CLIFFS 'JNIT 2 3/4 2-4 Amendinent No. 9, JE, 24, 3J, 61 i l

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POWER DISTRIBUTION LIMITS TOTAL' PLANAR RADIAL PEAKING FACTOR Ffy LIMITING CONDITI0'N FOR OPERATION T T xy(1+T ), shall be l 3.2.2.1The calculated value of F , defined as F ,p limited to < 1. 70. xy xy q 1 APPLICABILITY:, MODE 1*.. ACTION: With Ff, >.1.70 within 6 hours either: l Reducy THERMAL POWER to bring the canbination of THERMAL POWER a. and F to within the limits of Figure 3.2-3a and withdraw the full ilngth CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6; or b. Be in at least HOT STANDBY. u SURVEILLANCE REQUIREMENTS 4.2.2.1.lThe provisions of Specification 4.0.4 are not applicable.

4. 2.2.1.2 Ffy shall be calculated by the expression F,p T

xy(1+T)andFfy y q shall be determined to be within its limit at the following intervals: a. Prior to operation above 70 percent of RATED THERMAL POWER after each fuel loading, b. At least once per 31 days of accumulated operation in MODE 1, and c. Within four hours if the AZIMUTHAL POWER TILT (T ) is > 0.030. q

See Special Test Exception 3.10.2.

~~ CALVERT CLIFFS - UNIT 2 3/4 2-6 Amendment No.9, 78,/),81 O v

POWER DISTRISUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) T 4.2.2.1.3 F,y shall be determined each time a calculo6 ton of F is required l xy by using the incore detectors to obtain a power distribution map with, all full length CEAs at or above the Long Term Steady State Insertion Limit for the existing Reactor. Coolant Pump combination. This determina-tior shall be limited to core planes between 15% and 85% of full core hei!J t inclusive and shall exclude regione influer.ced by grid effects. h shall be determined each time a calculation of Ffy is required l 4.7.2.1.4 Tq and the value of T used to determine F shall be the measured value of q xy 9' J s .~ CALVERT CLIFFS-UNIT 2 Amendment No. 9,- 18, 61 3/42-7 as

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.e POWER DISTRIBUTI0fl LIMITS TOTALPLANARRADIALPEAKINGFACTOR-F[y LIMITING CONDITION FOR OPERATION 3.2.2.2 The value of N presently used in specification 4.2.1.3 shall l be in accordance with Figure 3.2-3b. APPLICABILITY: MODE 1 when operating in accordance w Mh specification 4.2.1.3. l ACTION: With the value of N.p.sently used in specification 4.2.1.3 exceeding the' l limit shown in Figure 3.2-3b, within 6 hours either: a. Reduce the value of N used in specification 4.2.1.3 to within the limits of Figure 3.2-3b;-or b. Be in at least HOT STANDBY. SURVEILLANCE REQUIREMENTS i 4.2.2.2.1 The provisions of Specification 4.0.4 are oct applicable. l FlyshallbecalculatedbytheexpressionF[y=Fxy(1+T)andN 4.2.2.2.2 q T shall be determined to be within its limit by monitoring F at the following y intervals: a. Prior to operation above 70 percent of RATED THERMAL POWER after each fuel loading, b. At'least once per 3 days of accumulated operation in MODE 1. l CALVERT CLIFFS - UNIT 2 3/4 2-8 Amendment rio. 9, JS, 31,g 1 l +

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) shall be determined each time a calculation of Ffy is required l 4.2.2. 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 Pump combination. This determina-tion shall be limited to core planes between 15% and 85% of full core height inclusive and shall exclude regions influenced by grid effects. 1 ime a calculation of Ffy is required l 4.2.2. 2.4 T shall be determined each q and the value of T used to determine F shall be the measured value of q xy 4 q. 1 i t;, f .~ i CALVERT CLIFFS-UNIT 2 Amendment No. 9, IB g y 3/4 2-8 (cont.) r ~S-

t POWER D STRIBUTION LIMITS T T0TAL INTEGRATED RADIAL PEAKING FACTOR - F r LIMITING CONDITION FOR OPERATION T 3.2.3 The' calculated value of F, daf;ned as FT = F (1+T ), shall be 9 limited to < l.650. l APPLICABILITY: MODE l*. ACTION:' With FT > 1.650, within 6 hours either: g r a. Be in at least HOT STANDBY, or b. Reduce THERMAL POWER to bring the combination of THERMAL POWER and Ff to within the limits of Figure 3.2-3c, withdraw the full length CEAs to or beyond the Long Term Steady State Limits-ofSpecification3.1.3.6,andinsertnewvalueofFfinBASSS;or c. Reducy THERMAL POWER to bring the combination of THERMAL POWER and F to within the limits of Figure 3.2-3c and withdrawn.the full l lengt) CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6. The THERMAL POWER limit determined from Figure 3.2-3c shall then be used to establish a l y revised upper THERMAL POWER level limit on Figure 3.2-4 (truncate-Figure 3.2-4 at the allowable fraction of RATED THERMAL POWER determined by figure 3.2-3c) and subsequent operation shall be l maintained within the reduced acceptable operation region of Figure 3.2-4. SURVEILLANCE REQUIREMENTS 4.2.3.1 The' provisions of Specification 4.0.4 are not applicable. T T shall be calculated by the expression F,, e (1+T ) and F 1.2.3.2 F r r q r shall be detennined to be within its limit at the following intervals: a. Prior to operation above 70 percent of RATED THERMAL POWER af ter each fuel loading, b. 'At least once per 31 days of accumulated operation in MODE 1, and c. Within four hours if the AZIMUTHAL POWER TILT (T ) is > 0.030. q iSee Special Test Exception 3.10.2. CALVERT CLIFFS - UNIT 2 3/4 2-9 Amendment No. 9, 16, 78,31,0 1

SURVEILLAtiCE REQUIREMENTS (Continued) shall be determined each time a calculation of Ff is required 4.2.3.3 Fr by using the.incore detectors to obtain a' power distribution map with all. full length CEAs at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump combination. T 4.2.3.4 T shall be determined each time a calculation of F is required r q and the value of T used to determine F shall be the measured value of q r q' e CALVERT CLIFFS - UNIT 2 3/4 2-10 Amendment No. 9, 76,18

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1.2 l 1 1 I I 1.1 x $z 1.0 (-0.15,1.0) - - (0.15.1.0) o9

c. r tlNACC EPTAtsLE UNACCEPTABLE miii OPERATION OPERATION

{0 g REGION REGION 3 0.9 ' O. 40 2* ae EE Et; 4-ACCEPTABLE Ey 0.8 ( 0.30. 0.60) ' OPERATION (0.30. 0.80) 'c 0: REGION 0 zw 9d tiB <a 5 0.7 0.6 ~ I I I I I 0.5 0.6 -0.4 0.2 0 0.2 0.4 0.6 PERIPHERAL AXIAL SHAPE INDEX. Y, t. Figure 3.2-4 DNB Axial Flux Offset Control Limits CALVERT CLIFFS 'illT 2 3/4 2-11 .bendment flo. 9, Z8, 31,$ 1 J i i +,

POWER DISTRIBUTION LIMITS AZIMUTHAL POWER TILT - T q LIMITING CONDITION FOR OPERATION 3.2.4 The AZIMUTHAL POWER TILT'(T ) shall not exceed 0.030. l q APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER.* i ACTION: a. With the indicated AZIMUTHAL POWER TILT determined to be > 0.030 but < 0.10, either correct the power tilt within two hours or determine within the next 2 hours and at least once per subsequent 8 hours, that the TOTAL PLANAR RADIAL PEAKING FACTOR (Ffy)andtheTOTALINTEGRATEDRADIALPEAKINGFACTOR (F ) ar ' within the limits of Specifications 3.2.2 and 3.2.3. r b. With the indicated AZIMUTHAL POWER TILT determined to be > 0.10, operation may proceed for up to 2 hours provided that T the TOTAL INTEGRATED RADIAL PEAKING FACTOR (F ) and TOTAL 7 PLANAR RADIAL PEAKING FACTOR (F y) 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 allowable provided the THERMAL POWER level is restricted to < 20% of the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination. SURVEILLANCE REQUIREMENTS 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 12 hours, and b. Using the incore detectors to determine the AZIMUTHAL POWER TILT at least once per 12 hours when one excore channel is l inoperable and THERMAL POWER is > 75% of RATED THERMAL POWER.

See Special Test Exception 3.10.2.

CALVERT CLIFFS-UNIT 2 3/4 2-12 Amendment No. 9 t._

I POWER DISTRIBUTION LIMITS DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters'shall be maintained within the limits shown on Table 3.2-1: a. Cold Leg Temperature b. Pressurizer Pressure c. Reactor Coolant System Total Flow Rate d. AXIAL SHAPE INDEX, Core pawer APPLICABILITY: MODE 1. ACTION: With any of the above parameters exceeding its limit, restore the parameter ) to within its limit within 2 hours or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours. -s SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters of Table 3.2-1 shall be verified to be within their limits at least once per 12 hours. 4.2.5.2 The Reactor Coolant System total flow rate shall be ditermined to be within its limit by measurement at least once per 18 months. .~ CALVERT CLIFFS-UNIT & Amendment No. 9,18,61 3/4 2-13

e TABLE 3.2-1 n [" DNB PARAMETERS 9 -n pr LIMITS 5 Four Reactor Three Reactor Two Reactor Two Reactor l_- Coolant Pumps Coolant Pumps Coolant Pumps - Coolant Pumps g Parameter Operating Operating Operating-Same Loop Operating-Opposite Loop -n Cold Leg Temperature < 548*F ra l Pressurizer Pressure > 2200 psia

Reactor Coolant System Total Flow Rate

> 370,000 gpm l AXIAL SHAPE INDEY, Core *** }{ Power 7*

Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% of RMTED THERMAL POWER per minute or a THERMAL POWER step increase of greater than 10% of RATED THERMAL POWER.
- These values left blank pending NRC approval of ECCS analyses for operation with less than four _ reactor coolant pumps operating.
  - The AXIAL SHAPE INDEX, Core Power shall be maintained within the limits established by the~Better Axial Shape Selection System (BASSS) for CEA insertions of the lead bank of 1 55% when BASSS is l},

OPERABLE, or within the limits of FIGURE 3.2-4 for CEA insertions specified by FIGURE 3.1-2, 7 il l n if

o s,

n..

TABLE 3.3-1 (Continued) c3 ): 4 Ts REACTOR PROTECTIVE INSTRUMENTATION ?l c, 1: MINIMUM

R TOTAL NO.

CHANNELS CHANNELS APPLICABLE y' FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE-MODES ACTION Ei

11. Wide Range Logarithmic Neutron

1 Flux Monitor n3 a.

Startup and Operating--Rate of Change of Power - High 4 2(d) 3(f) 1,'2 and * - 2# b. Shutdown 4-0 2 3,4,5 3-t

12. Reactor Protection System 6

1 6-1,. ;2* 4 i* Logic Matrices w

13. Reactor Protection System 4/ Matrix 3/ Matrix-4/ Matrix 1, 2*

4 Logic Matrix Relays

14. Reactor Trip Breakers 8

6 8 1, 2* '4 e D e m-9 O

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.

The provisions of Specification '3.0.4 are not applicable.

(a) Trip may be bypassed below 10-4be automatically removed wh of RATED THERMAL POW THERMAL POWER. of RATED (b) Trip may be' manually bypassed below 785 psia; bypass shall be automatically removed at or above' 785 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) Trip may be bypassed below 10-4% and above 12% of RATED THERMAL POWER. (e) Trip may be bypassed during testing pursuant to Special Test Excep-tion 3.10.3. (f) There shall be at least two decades of overlap between the Wide e Range Logarithmic Neutron Flux Monitoring (,hannels and the Power Range Neutron Flux Monitoring Channels. ACTION STATEMENTS With the number of channels OPERABLE one less than ACTION 1 required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE stattas within 48 hours or be in HOT STANDBY within the next 6 hours and/or open the protective system trip breakers. With the number of OPERABLE channels one less than the ACTION 2 Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied: a. The inoperable channel is placed in either the bypassed or tripped condition within 1 hour. For the purposes of testing and maintenance, the inoperable channel may be bypassed for up to 48 hours from time of initial loss of OPERABILITY; however, the inoperable channel shall then be either restored to OPERABLE status or placed in the tripped condition. CALVERT CLIFF.S - UNIT 2 3/4 3-4 Amendment No.JJ, G 1

IT TABLE 3.3-1 (Continued) ACTION' STATEMENTS b. Within one hour, all functional units receiving an input from the inoperable channel are also placed in the same condition (either bypassed or tripped, as applicable) as that required by a. above for the inoperable channel. c. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for ,up to 48 hours while performing tests and maintenance on that channel provided the other inoperable channel is placed in the tripped condition. With the number of channels OPERABLE.one less than required ACTION 3 by the Minimum Channels OPERABLE requirement, verify compli-ance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, within 1 hour and at least once per 12 hours thereafter. With the number of channels OPERABLE one less than required ACTION 4 by the Minimum Channels OPERABLE requirement, be in HOT STANDBY within G haurs; however, one channel may be bypassed for up to 1 hour for surveillance testing per Specification 4.3.1.1. me a f CALVERT CLIFFS - UNIT 2 3/4 3-5 t-

TABLE 3.3-2 REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES { FUNCTIONAL UNIT RESDONSE TIME. 1. Manual Reactor Trip Not Applicable e g 2. Power Level - High < 0.40 seconds *# and < 12.0 seconds ## [ 3. Reactor Coolant Flow - Low 1 0.50 seconds 4. Pressurizer Pressure - High 1 0.90 seconds 5. Containment Pressure - High 1 0.90 seconds 6. Steam Generator Pressure - Low 1 0.90 seconds 7. Steam Generator Water Level - Low 1 0.90 seconds y 8. Axial Flux Offset < 0.40 seconds *# and < 12.0 seconds ## [ 9.a. Thermal.'targin/ Low Pressure < 0.90 seconds *# and < 12.0 seconds n b. Steam Generator Pressure Difference - High .,0.90 seconds ~

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

11. Wide Range Logarithmic Neutron Flux Monitor Not Applicable E.

h

Neutron detectors are exempt from response time cesting.

Response time of the neutron flux signal portion of the channel shall be measurcd from detector output or inp,ut of first electronic component in channel. O;

Response time does not include contribution of RTDs.

m

1. RTD response time only. This value is equivalent to the time interval required for the RTDs output D

to achieve 63.2% of its total change when subjected to a step change in RTD temperature.

e _

m O m.

? TABLE 3.3-3 (Continued) l ENGINEERED SAFETY FEATURE-ACTUATION SYSTEM INSTRUMENTATION MINIM 0M c, p: TOTAL NO. CHANNELS CHANNELS APPLICABLE' FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES

ACTION 4

6. C0flTAINMENT FURGE n r-VALVES ISOLATION ## n UI a. Manual (Purge Valve Control Switches) 2/ Penetration 1/ Penetration-2/ Penetration' 5, ~ 6 8 8' =:; b. Containment Radiation - g High j n, Area Monitor 4 2 3 6 8. J 7. LOSS OF POWER a. 4.16 kv Emergency Bus Undervoltage (Loss 's of Voltage 4/ Bus 2/ Bus 3/ Bus 1,~2, 3 7*- b. 4.16 kv Emergency Bus -Undervoltage (Degraded Voltage) 4/ Bus 2/ Bus 3/ Bus 1, 2, 3 7* Pl

1. Contaimnent purge valve isolation is also initiated by SIAS (functional units 1.a, l.b, and l'.c).

g n O M. Go b -a h i ~ ~ m

h$ TABLE 3.3-3 (Continued) Ej ENGINEERED SAFETY FEATURE. ACTUATION SYSTEM INSTRUMENTATION [2 MINIMUM Gl TOTAL N0. CHANNELS CHANNELS APPLICABLE 5 FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION s. c; 8. CVCS ISOLATION a. Manual (CVCS Isolation Valve Control Switches) 1/ Valve 1/ Valve .1/ Valve .1, 2,'3, 4 6 b. West Penetration Room / Letdown Heat 4 Exchanger Room Pressure - High 4 2 3 1,2,3,4 f* w 3 9. AUXILIARY FEEDWATER wj; a. Manual 2 sets of 2 1 set of 2 2 sets of 2 1,2,3 6 o per S/G .per S/G per S/G 1 b. Steam Generator Level - Low 4/SG 2/Sa 3/SG 1, 2, 3 7 EL c. Steam Generator AP High 4/SG 2/SG 3/SG l', 2, 3 7- .Er l 30: t c) b4 i I l s l I L d -o

1 TABLE 3.3-3 (Continued)' ~ r- . TABLE NO,TATION Y -(a) Trip function may be' bypassed in this MODE when pressurizer pressure is <~1800 psia; bypass shallLbe automatically r,emoved-when pressurizer pressure is _> 1800 psia. (c) -Trip fur.ction may be bypassed in this MODE below'785 psia; bypa'ss shall be automatically removed at or above 785 psia. The provisions of Specification 3.0.4 are not applicable. ~ ACTION STATEMENTS ACTION'6 With the number of OPERABLE channels.one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within-the next 6 hours and in COLD SHUTDOWN within the following 30 hours. With the number of OPERABLE channels one -less than the ACTION 7 N Total-Number of Channels, operation may proceed provided the following conditions are satisfied: f a. The inoperable channel is placed.in either the bypassed-or tripped condition within 1 hour. For the purposes 4 of testing and maintenance, the inoperable channel may be bypassed for up to 48 hours fron time of initial loss of OPERABILITY; however, the inoperable channel shall then be either restored to OPERABLE status or placed in the tripped condition. 3 5 b. Within one hour, all functional units rec,eiving an input from the inoperable channel are also placed in the same condition (either bypassed or tripped, as applicable) as that required by a. above for the inoperable channel. c. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 48 hours while performing tests and maintenance on that channel provided the other inoperable channel is placed in the tripped condition. ~ l i \\ L CALVERT CLIEFS - UNIT 2 3/4 3-15 Amendment No.31', 61 i

TABLE 3.3-3 (Continued) With less than the Minimum Channels OPERABLE, operation ACTION 8 may continue provide the containment purge valves are maintained closed. ACTION 11 - With the number of OPERABLE-Channels one less than the Total. Number of Channels,. operation may proceed provided the inoperable-channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is demonstrated within 1 hour; one additional channel may be bypassed for up to 2 hours for surveillance testing per Specification-

4. 3. 2.1. -

C (' CALVERT CLIFFS - UNIT 2 3/4 3-16 Amendment No. 3 t

TABLE 3.3-4 9- { ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP YALUES 5 P ALLOWABLE 4 FUNCTIONAL UNIT TRIP SETPOINT-VALUES -i vi 1. SAFETY INJECTION (SIAS) i e a. Manual (Trip Buttons) Not Applicable Not_ Applicable 1 i 5 b. Containment Pressure - High 1 4.75 psig 1 4.75 psig ) ] l c. Pressurizer Pressure - Low > 1725 psia > 1725 psia l j j 2. CONTAINMENT SPRAY (CSAS) J a. Manual (Trip Buttons) Not Applicable . Not Applicable' i b. Containment Pressure -- High 1 4.75 psh 1 4.75 psig 3. CONTAINMENT ISOLATION (CIS) # ] y a. Manual CIS (Trip Buttons) Not Applicable Not Applicable b. Containment Pressure - High

1 4.75 psig 1 4.75 psig 4.

MAIN STEAM LINE ISOLATION a. Manual (MSIV Hand Switches and Feed Head Isolation l' Hand Switches) Not Applicable Not Applicable g-b. Steam Generator Pressure - Low > 685 psia > 685 psia l s

Containment isolation of non-essential penetrations is also initiated by SIAS (functional g

units 1.a and 1.c). H +

e TABLE 3.3-4 (Continued) (( ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES

x n ALLOWABLE

(( FUNCTIONAL UNIT TRIP VALUE VALUES nn CONTAldMENTSUMPRECIRCU,LATION(RAS) SU 5. a. Manual RAS (Trip Buttons) Not Applicable Not Applicable 55 b. Refueling Water Tank - Low > 24 inches above > 24 inches above tank bottom tank bottom 6. CONTAINMENT PURGE VALVES ISOLATION ## g a. Manual (Purge Valve Control Switches) Not Applicable Not Applicable b. Containment Radiation'- High Area Monitor 1 220 mr/hr 1 220 mr/hr-y 7. LOSS OF POWER a. 4.16 ky Emergency Bus Under-2450+105 volts with a 2450+105 volts with a voltage (Loss of Voltage) 210.2 second time delay 210.2 second tinie delay b. 4.16 kv Emergency Bus Under-3628125 volts with a 3628125 volts with a voltage (Degraded Voltage) 810.4 second time delay 810.4 second time delay NE - 8&$

1. Containment purge valve. isolation is also initiated by SIAS (functional units 1.a.1.b. and 1.c).

p P E. - g gs 9 n.

k . TABLE 3.3-4 (Continued) ENGINEERED SAFETY FEATURE ACTUATIO*1 SYSTEM INSTRUMENTATION TRIP VALUES-n. ALLOWABLE 3 FUNCTIONAL UNIT TRIP VALUE-VALUES- -g 8. CVCS ISOLATION [ Uest Penetration Room /.. 1 5 psig 1 0.5.psig: 0 Letdown Heat Exchanger Room Pressure - High 9. AUXILIARY FEE 0 WATER ACTUATION SYSTEM a. Manual (trip Buttons) -Not Applicable Not Applicable 7 b. Steam Generator (A or B) Level - Low -194"?to -149" -194" to -149" y (inclusive) (inclusive)- m a c. Steam Generator AP-High < 130.0 psid < 130.0 psid s-(SG-A > SG-B) d. Steam Generator AP-High -< 130.0 psid ~ < 130.0 psid (SG-B > SG-A) iir i a 8n e p e .>4 n-- ^% e--

TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 'l.. Manual a. SIAS Safety Injection (ECCS) Not Applicable. b. CSAS Containment Spray Not Applicable ( c. CIS Containment Isolation Not Applicable d. RAS Containment Sump Recirculation Not Applicable e. AFAS Auxiliary Feedwater Actuation Not Applicable 2. Pressurizer Pressure-Low C a. Safety Injection (ECCS) 1 30*/30** 3. Containment Pressure-High a. Safety Injection (ECCS) 1 30*/30** b. Containment Isolation 1 30 c. Containment' Fan Coolers 1 35*/10** 4. Containment Pressure--High a. Containment Spray 1 60*/60** 5. Containment Radiation-Hioh a. Containment Purge Valves Isolation 15 o (.- CALVERT CLIFFS - UNIT 2 3/4 3-20 Amendment No. 6, 3Z,g 1

e TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES p INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS. ' 6. Steam Generator Pressure-Low a. Main Steam Isolation '< 6.9 b. Feedwater Isolation 3:_ 80 7. Refueling Water Tank-Low a. Containment Sump Recirculation

80 8.

Reactor Trip a. Feedwater Flow Reduction to 5% < 20 9. Loss of Power a. 4.16 kv Emergency Bus < 2.2 Undervoltage (Loss of q Voltage) b. 4.16 kv Emergency Bus -< 8.4 Undervoltage (Degraded Voltage) 10. Steam Generator Level - Low a. Motor Dr.iven AFW Pump _

_ 54.5* / 14.5**

b. Steam Driven AFW Pump <: 54.5 11. Steam Generator AP-High a. Auxiliary Feedwater Isolation < 20.0 TABLE !!0TATION

Diesel generator starting and sequence loading delays included.
- Diesel generator starting and sequence loading delays not included.

Offsite power available.

- - Response time measured from the incidence of the undervoltage condition to the diesel generator start signal.

CALVERT CLIFFS - UNIT 2 3/4 3-21 Amendment No. 22, 37,6 1 4

e e TABLE 4.3-2 h$ h$ ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTAION SURVEILLANCE REQUIREMENTS. ?? si Ej 33 CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE c3 c, [: FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED r-n U3 1. SAFETY INJECTION (SIAS) '- e a. Manual (Trip Buttons) N.A. N.A. R N.A. Ei e b. Containment Pressure - High S. R M-1,2,3

q ::[

c. Pressurizer Pressure - Low S R M-1, 2, 3 - [ d. Automatic Actuation Logic H.A. N.A. M(1)(3) '1, 2, 3 n3 E. CONTAINMENT SPRAY (CSAS) a. Manual (Trip Buttons) N.A. N.A. R H.A. b. Containment Prcssure -- High S R M 1,2,3 c. Automatic Actuation Logic H.A. N.A. M(1) - 1,. 2, 3 ' 3. CONTAINMENT ISOLATION (CIS) # l ja a. Manual CIS (Trip Buttons) N.A. N.A. R N.A. R; b. Containment Pressure - High S R M 1., 2, 3 c. Automatic Actuation Logic N.A. N.A. M(1)(4) 1, 2,-3 4. MAIN STEAM LINE ISOLATION (SGIS) a. Manual SGIS (MSIV Hand 3I ET Switches and Feed Head Isolation Hand Switches) N.A. N.A. R N.A. 8-Et b. Steam Generator Pressure - Low S R M ' 1, 2, 3 c. Automatic Actuation Logic N. A '. N.A. M(1)(5) -1,2,3 r r F. F

Containment isolation of non-essential penetrations is also initiated by.SIAS (functio'nal units us ui 1.a and 1.c).

G O

a 9 TABLE 4.3-2 (Continued) k ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS n C. CHANNEL MODES IN WHICH CilANNEL CHANNEL FUNCTIONAL SURVEILLANCE v5 FUNCTIONAL UtlIT CilECK CALIBRATION TEST REQUIRED' i Yq 5 CONTAltiMENT SUMP RECIRCULATION (RAS) a. Manual RAS (Trip Buttons) N.A. N.A. R N.A. m b. Refueling Water Tank - Low N.A. R M 1, 2, 3 c. Automatic Actuation Logic N.A. N.A. M(1)' 1, 2, 3 6. CONTAINMENT PURGE VALVES ISOLATION w a. Manual (Purge Valve Control 2 Switches) N.A. N.A. R N.A. b. Containment Radiation - High w da Area Monitor S R M 6 w s 7. LOSS OF POWER a. 4.16 kv Emergency Bus Undervoltage (Loss of Voltage) N.A. R M 1, 2,'3 b. 4.16 ky Emergency Bus k . Voltage) N.A. R M .1, 2, 3 Undervoltage (Degraded g. @g 8. CVCS ISOLATION N.A. R M 1, 2, 3, 4 West Penetration Room / 2 P Letdown Heat Exchanger Room Pressure - High w N 9. AUXILIARY FEEDWATER a. Manual (Trip Buttons) N.A. N.A. R N.A. w* b. Steam Generator Level - Low S R M 1, 2, 3 w c. Steam Generator AP-High S R M 1, 2,'3 d. Automatic Actuation Logic N.A. N.A. M(1) 1, 2, 3 o s

1. Containment purge valve isolation is also initiated by SIAS (functional units 1.a,1.b and 1.c).

g.

^ TABLE 4.3-2 (Continued) TABLE NOTATION (1) - The logic circuits'shall be tested' manually at least once per 31-days. (3) SIAS logic circuits A-5, B-5,LA-10 and B-10 may be exempted from f testing during o~peration; however, these logic circuits shall be-tested at least once per 18 months during shutdown. (4) CIS logic' circuits A-5 and B-5 may be exempted from testing during operation; however, these logic circuits shall be tested at least once per 18 months during shutdown. (5) SGIS logic. circuits A-1 and B-1 may be exempted from testing during operation; however, these logic ' circuits shall be tested at least once per 18 months during shutdown. w C-CALVERT CLIFFS - UNIT 1 CALVERT CLIFFS - UNIT 2 3/4 3-24 Amendment No. 3 g 4-, ,,---e p w ..m

TACLE 3.3-10 y POST-ACCIDErlT' MONITORING IriSTRUMENTATION M MINIMUM CHANNELS OMAM P INSTURMENT m 7) 1. Containment Pressure 2 h 2. Wide Range Logarithmic Neutron Flux Monitor 2 = l [ 3. Reactor Coolant Outlet Temperature 2 ~ 'l: 4. Pressurizer Pressure 2 5. Pressurizer Level 2 6. Steam Generator Pressure 2/ steam generator 7. Steam Generator Level (Wide Range) 2/ steam generator M 8. Auxiliary Feedwater Flow Rate 2/ steam generator Y' H f_: 9. RCS Subcooled Margin Monitor 1 10. PORV/ Safety Valve Acoustic Flow Monitoring 1/ valve-F 11. PORV Solenoid Power Indication 1/ valve @~ y 12. Feedwater Flow 2 @n s .M e .) bb E

~*- ~ TABLE 4.3-10 POST-ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS', 9 r-f5i CHANNEL CHANNEL E INSTRUMENT CHECK CALIBRATION- 'l n 1. Containment Pressure M R [ 2. Wide Range Logarithmic Neutron Flux Monitor M N.A. E 3. Reactor Coolant Outlet Temperature M R. Z m 4. Pressurizer Pressure M R 5. Pressurizer Level M R ~ 6. Steam Generator Pressure M R 7. SteamGeneratorLevel(ilideRange) M R s* 8. Auxiliary Feedwater Flow Rate M R s b 9. RCS Subcooled Margin Monitor M R 10. PORV/ Safety Valve Acoustic Monitor N.A. R 11. PORY Solenoid Power Indication N.A. N.A. F 12. Feedwater Flow M R E s c, b6 s n.

3/4.1 REACTIVITY CONTROL-SYSTEMS ~ BASES 3/4.1.1 BORATION CONTROL ~ 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions, 2) the reactivity transients-associated with postulated accident conditions are controllable within acceptable. limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition. ' SHUTDOWN MARGIN requirements vary throughout core life as a. function of fue1 ~ depletion, RCS br.ron concentration and RCS T The minimum available SHUTDOWN MARGIN for no load operating conditions.$E9beginning of life is 4.5% ak/k and at end of life is 5.2% ak/k. The SHUTDOWN MARGIN is based on the safety analyses performed for a steam line rupture event initiated at no load conditions. The most restrictive steam line rupture event occurs at EOC conditions. For the steam line rupture event at beginning of cycle conditions, a minimum SHUTDOWN MARGIN of less than 4.5% ak/k is required to control the reactivity transient, and end of cycle conditions require 5.2% ak/k. Accordingly, theSHUTDOWNMARGINrequirementisbaseduponthislimitingconditfonandis consistent with FSAR safety analysis assumptions. With T 200 F, the reactivity transients resulting from any postulated accid 8U <are minimal and a 3% ak/k shutdown margin provides adequate protection. With the pressurizer level less than 90 inches, the sources of non-borated water are restricted to increase the time to criticality during a boron dilution event. 3/4.1.1.3 BORON DILUTION A minimum flow rate of at least 3000 GPM provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration reductions in the Reactor Coolant System. A flow rate of at least 3000 GPM will circulate an equivalent Reactor Coolant System volume of 9,601 cubic feet in approximately 24 minutes..The reactivity change rate associated with boron concen-tration reductions will therefore be within the capability of operator recognition and control. 3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC) The limitations on MTC are provided to ensure.tfiat the assumptions used in the accident and transient analyses remain valid through each fuel cycle. The surveillance requirements for measurement of the MTC during each fuel cycle are adequate to confirm the MTC value since this . coefficient changes slow 1v due principally to the reduction in RCS boron . concentration ass tiated with fuel burnup. The confirmation that the measured MTC value is within its limit provides assurances that the coefficient will be maintained within acceptable values throughout each fuel cycle. CALVERT CLIFFS' - UNIT 2 B 3/4 1-1 Amendment No. 28, 3I,61 i

'l ' POWER __ DISTRIBUTION _ LIMITS BASES the analysis establi,shing the DNB Margin LCO, and Thermal Margin / Low-Pressure LSSS setpoints remain valid duringT per9 tion at the various o allowable CEA~ group insertion limits.. If F F or T exceed their basiclimitations,.operationmaycontinueuMerEhead81tionalrestric-tions imposed by the ACTION statements since these additional restric-tions provide adequate. provisions to assure that the assumptions.used in establishing the Linear Heat Rate. Thermal-Margin / Low Pressure and Local Power Density - High LCOs and LSSS setpoints remain valid. An AZIMUTHAL POWER TILT > 0.10. is 'not expected and if it should occur, sub-sequent operation would be restricted to only those operations required to _ identify the cause of this unexpected tilt. T he value of T that must be used in the equation F*Y = F*# (1.+ T ) andF}r=Fr(1+T)Tsthemeasuredtilt. 9 q T The surveillance requirements for verifying that F F and T aye 7 T 4 F within their limits provide assurance-that the actual v Yues yf F T r and T do not exceed the assumed values. Verifying F and F afNr each 9uel loading prior to exceeding 75% of. RATED theft!lAL POWER provides Y additional assurance that the core was properly loaded. 3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters.are maintained within the nomal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the safety analyses assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.23 throughout each l analyzed transient. The 12 hour periodic surveillance of these parameters through instru-ment readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The 18 month periodic measurement of t'he RCS total flow rate is adequate to detect flow degrac'ation and ensure correlation of the flow indication tnannels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour basis. (.- CALVERT CLIFFS - UNIT 2 B 3/4 2-2 Amendment No. 18,31,38,01 k

PLfd SYSTEMS' AUXILIARY FEEDWATER SYSTEM F. LIMITING CONDITION FOR OPERATION

3. 7.1. 2 At least two steam turbine driven steam generator auxiliary feedwater pumps and associated flu paths shall be OPERABLE and capable of automatically initiating flow, within' the area of acceptable operation. of Figure 3.7-1, to each steam generator *.

l APPLICABILITY: MODES 1, 2 and 3**. l ACTION: a. With one auxili,ary-feedwater pump inoperable, restore at least two auxiliary feedwater pumps to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours. b. Whenever a subsystem is inoperable for the performance of periodic testing (i.e., manual discharge valve closed for pump discharge head test) a dedicated operator will be stationed at the local station (i.e., closed va've), with direct communication to the Control Room, to return the subsystem to normal upon instruction from the Control Room. Upon completion of any testing, the subsystem (valve) will be returned to its proper position and verified in its proper position by an independent operator check. SURVEILLANCE REQUIREMENTS

4. 7.1. 2 Each auxiliary feedwater pump shall be demonstrated OPERABLE:

y a. At least once per 31 days by: 1. Verifying that each steam turbine driven pump develops a Total Dynamic Head of,> 2800 ft. on recirculation flow when the secondary steam supply pressure is greater than 800 psig. 2. Cycling each testable, remote operated valve that is not in its operating position through at least one complete cycie. 3. Verifying that each valve (manual, power operated or automatic) in'the direct flow path is in its correct position, b. Before entering MODE 3 after a COLD SHUTDOWN of at least 14 days by completing a flow test that verifies the flow path from the condensate storage tank to the steam generators. c. At least once per 18 months by: 1. Verifying that each automatic valve in the flow path actuates, to its correct position upon receipt of each auxiliary feedwater actuation ' test signal. For a period of up to 30 days following the entering into Mode 2 from Cycle 5 startup, at power levels not to exceed 55% of rated thermal power, the auxiliary feedwater system may be inoperable except with regard to the capability for manually initiating and manually controlling flow to either steam generator with the two steam driven auxiliary feedwater trains.

- Automatic flow initiation need not be OPERABLE during MODE 3.

CALVERT CLIFFS I UNIT 2 3/4 7-5 Amendment No. 37, /[ ; y

e PLANT' SYSTEMS AUXILIARY FEEDWATER SYSTEli SURVEILLANCE REQUIREMENTS (Continued) 2. Verifying that each auxiliary feedwater pump' starts 'as designed-automatically upon receipt of. each auxiliary feedwater actuation test signal. 3. Verifying, upon automatic initiation of auxiliary feedwater, a flow within the acceptable operating limits of Figure 3.7-1, Steam Generator Pressure Versus Auxiliary Feedwater Flow.

d. - Upon repositioning of 1/2-CV-4511 and/or 1/2-CV-4512 the valve shall be realigned to provide flow consistent with Figure 3.7-1.
Only applicable during 110 DES 1 and 2.

a t, CALVERT CLIFFS - UNIT 2 3/4 7-Sa Amendment No. 49 S

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n & Q _ \\ / h i C b l h b5* Q,. $. Y,'T.'~ &. &: ; jY ? ! ~ f .$P N b : hm f..h -. h\\Eh ' kl?' p[h& LOW POPULATION ZONE F I G. 5.1 - 2 CALVERT CLIFFS - UNIT 1 CALVERT CLIFFS - UNIT 2 5-3 ( t

DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be main-tained for a maximum internal pressure of 50 psig and a _ temperature of 276*F. 4 5.3 REACTOR CORE FUEL ASSEMBLIES 5. 3.1 The reactor core shall contain 217 fuel a'semblies with each fuel ^ s assembly containing a maximum of 176 fuel rods clad with Zircaloy-4. Each fuel rod shall have a nominal active fuel length of 136.7 inches and contain a maximum total weight of 3000 grams uranium. The initial core loading shall have a maximum enrichment of 2.99 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 4.1 weight percent U-235. l 5.3.2" Except for special test as authorized by the NRC, all fuel assemblies under control element assemblies shall be sleeved with a sleeve design previously approved by the NRC. CONTROL T'.EMENT ASSEMBLIES 5.3.3 The reactor core shall contain 77 full length and no part length

ontrol element assemblies.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be naintained: a. In accordance with the code requirements specified in Section 4.2 of the FSAR with allowance for normal degradation pursuant of the applicable Surveillance Requirements", b. For a pressure of 2500 psia, and c. For a temperature of 650 F, except for the pressurizer which is 700*F. CALVERT CLIFFS - UNIT 2 Amendment No. JS, 31, 6 1 5-4 - -}}

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