ML20054K379

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Amend 71 to License DPR-53,authorizing Operation During Cycle 6
ML20054K379
Person / Time
Site: Calvert Cliffs 
Issue date: 06/24/1982
From: Clark R
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20054K380 List:
References
NUDOCS 8207010517
Download: ML20054K379 (55)
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{{#Wiki_filter:_ [> Aft (On,Io UNITED STATES l' 'g NUCLEAR REGULATORY COMMISSION - g ) l W ASHINGTON, D. C. 20555 gg Cortified B 3 BALTIM0RE GAS AND ELECTRIC COMPANY DOCKET N0. 50-317 CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 71 License No. OPR-53 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Baltimore Gas & Electric Company (the licensee) dated February 17, 1982, as supple-mented April 29, 1982, complies with the standards and require-ments 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 regula-tions; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public, and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations'and all applicable requirements have been satisfied. 0207010517 820624 DRADOCK05000g

s. l. 2. Accordingly, Facility License No. DPR-53 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2), is hereby amended to read as follows: i (2) Technical Specifications i The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 71, are hereby incorporated in the license. The licensee ^' shall operate the facility in accordance with the Technical Specifications. l o f.r. 3. The license amendment is effective as'of the date of its issuance."# FOR THE NUCLEAR REGULATORY COMMISSION ) ..{' ~ L, \\ - L m Robert A. Clark, Chief Operating Reactors Branch #3 i Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: June 24, 1982 ( i l l I i i + i

8 ATTACHMENT TO LICENSE AMENDMENT NO. 71 FACILITY OPERATING LICENSE N0. DPR-53 DOCKET.NO. 50-317 Replace the following pages of 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 main-tain document completeness, i Pages 2-2 3/4 3-4 2-3 3/4 3-6 2-9 3/4 3-15 2-10 3/4 3-17 2-11 3/4 3-21 i B 2-1 3/4 7-9 B 2-3 B 3/4 1-1 B 2-7 8 3/4 2-2 3/4 1-1 5-4 3/4 1-27 3/4 2 3/4 2-4 3/4 2-4a (new) 3/4 2-6 3/4 2-7 4 i 3/4 2-7a (new) 3/4 2-7b (new) 3/4 2-9 3/4 2-11 3/4 2-13 4 3/4 2-14 i j t I l 4 m- ~. ag.w y s,., .a m,-.- -w

!-i 2.0 SAFETY LIMITS AND LIMIlING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS 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 of two, three and four reactor coolant pump operation. ] APPLICABILITY: MODES 1 and 2.' ' ~ ACTION: ~ ~ P 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. APPLICABILITY: MODES 1, 2, 3, 4 and 5. l 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. l CALVERT C'LIFFS - UNIT 1 2-1 l

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VAllD FOR AXIAL SHAPES AND o INTEGRATED ROD RADIAL PEAKING FACTORS LESS THAN OR EQUAL TO o THOSE ON FIGURE B2.1-1 ',> d 520 N O g REACTOR OPERATION LIMITED TO LESS 3 THAN 580*F BY ACTUATION OF THE E SECONDARY SAFETY VALVES X g i il g 480 - ACCEPTABLE OPERATION e-N n e g ~$ 4,4 '+ g i i l l I I I I l p 0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 2.0 FRACTION OF RATED THERMAL POWER w*.' q e Fjgure 2.1-1 [ Reactor Core Thermal Margin Safety Limit Four Reactor Coolant Pumps Operating

l i This page left blank pending NRC approval of ECCS analysis for three pump operation. a I 4 i t t 4 i i Figure 2.1-2 Reactor Core Thermal Margin Safety Limit - l Three Reactor Coolant Pumps Operating -CALVERT CLIFFS - UNIT 1 2-3 ~ l .r.e- .-..e ,_..,.,m -.--,-e% ,--r-,_,,m.-f. .- - - - -, ~ - --m s, ,.my e3-< w - r-gm

{ t This page left blank pending NRC approval of ECCS analysis for two pump (same loop) opera-tion. ~ a 1 Figure 2.1-3 Reactor Core Thermal Margin Safety Limit - Two Reactor Coolant Pumps Operating - Same Loop CALVERT CLIFFS - UNIT 1 2-4

e o o TABLE 2.2-1 (Cont'd) E M REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS E FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 4 4. Pressurizer Pressure - Nigh < 2400 psia 1 2400 psia E 5. Containment Pressure - High < 4 psig < 4 psig Q 6. Steam Generator Pressure - Low (2) > 635 psia > 635 psia 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 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 Trin setpoint adjusted to Operating not exceed the limit lines ~be not less than th~e'la_rger of Figures 2.2-2 and 2.2-3., of (1) the value calculated from Figures 2.2-2 and 2.2-3 and (2) 1875.psig. b. Steam Generator Pressure < 135 psid < 135 psid Difference - High (1) 10. Loss of Turbine -- Ilydraulic > 1100 psig 1 1100 psig Fluid Pressure - Low (3) 5 11. Rate of Change of Power - High (4) 1 2.'6 decades per minute < 2.6 debades per minute 5 TABLE NOTATION -4 (1 ) Tripmaybebypassedbe}ow10 % of RATED THERMAL POWER; bypass shall be automatically removed when ~ TilERMAL POWER is > 10' % of RATED TilERNAL POWER. 4 ^~ e

  • O 9

k TABLE 2.2-1 (Cont'd)

u TABLE NOTATIONS (Cont'd)

P (2) Trip may be manually bypassed below 710 psia; bypass shall be automatically removed at or above 710 psia. 3 (3) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatica11'y removed when TilERitAL POWER is > 15% of RATED THERMAL POWER. c= ~4 (4) Trip may be bypassed below 10 % and above 12% of RATED TifERMAL POWER. r 0 O s s 4 a W n 1 l 5 J g s O h 0 !.~,\\.'g k, /, y 4 g 4 i \\ ~ ,) i s e N ' 5., s,,i. i 5 ,[ e -s l g ~ ~J e r .7, u .k / l i p s T \\ 1 1 s U \\, s k,

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_ - _-_~_- o 2.1 SAFETY LIMITS BASES ] 2.1.1 REACTOR CORE lI The restrictions of this safety limit prevent dVerheat'ing'of the fuel cladding and possible cladding perforation which would result in the i 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 21.3 kw/ft. Centerline fuel melting.will not occur [ for this peak linear heat ra.te. 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 I 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 DNS through the CE-1 correlation. 1 The CE-l DNB correlation has been developed to predict the DNB flux and the location of DNB for axially unifonn and non-uniform heat flux distri-1 butions. The local DNB heat flux ratio, DNBR, defined as the ration 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, 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. i The curves of Figures 2.1-1, 2.1-2, 2.1-3 and 2.1-4 show the l loci of points'of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperature of various pump combinations for which-the minimum DNBR is no less than 1.23 for the family of axial shapes and l 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 3 i inlet temperatures less than or equal to 580 F. The dashed line at 580*F coolant inlet ^.emperature 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 l POWER is prohibited by the high power level trip setpoint specified in CALVERT CLIFFS - UNIT 1 B 2-1 Amendment No.38,pp, y i __.

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o. SAFETY LIMITS BASES Table 2.1-1. The area of safe operation is below and to the left of o these lines. The conditions for the Thermal Margin Safety Limit curves in Fi 2.1-1, 2.1-2, 2.1-3 and 2.1-4 to be valid are shown on the figures. gures The reactor protective system in combination with the Limiting Conditions for Opera ~ tion, is~ designed to prevent any anticipated combina-tion of transient conditions for reactor coolant system temperature. pressure, and THERMAL POWER level that woul'd result in'a DNBR of less than 1.23 and preclude the existence of flow instabilities. 2.1.2 REACTOR COOLANT SYSTEM PRESSURE 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 III, 1967 Edition, 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,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 associated code requirements. The entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to initial operation. CALVERT CLIFFS - UNIT 1 B 2-3 Amendment No. 33, Ef,/y171

e 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SETPOINTS 7 The' Reactor Trip Setpoints specified in Table 2.2-1 are tRe 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 acc.eptable on the basis that'the difference between the trip setpoint and 'the Allowable Value is equal to or hss than the drift allowance assumed for iach t, rip in the safety 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 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 The Reactor Coolant Flow-Low trip provides core protection to prevent l DNB in the event of a sudden significant decrease in reactor coolant flow. Provisions have been made in the reactor ptStective system to permit CALVERT CLIFFS - UNIT I B 2-4 Amendment No. AS,7 i A t

j 1 LIMITING SAFETY SYSTEM SETTINGS i BASES 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 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 and expected transients. l For reactor operation with only two or three reactor coolant pumps operating, the Reactor Coolant Flow-Low trip setpoints, the Power Level-High trip set-points, and th? 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 l l during normal operational transientr 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 opera-tion of the pressurizer code safety valves. 2 Containment Pressure-High The Containment Pressure-High. trip provides assurance that a reactor j trip is initiated concurrently with a safety injection. The setpoint for this trip is identical to the safety injection setpoint. i Steam Generator Pressure-Low The Steam Generator Pr. essure-Low trip provides protection against an excessive rate of heat extr' action from the steam generators and subs'equent cooldown of the reactor coolant. The setting of 635 psia is sufficiently I below the full-load operating point of 850 psia so as not to interfere with normal operation, but still high enough to provide the required protec-tion in the event of excessively high steam flow. This setting was used with an uncertainty factor of + 87 psi which was based on the main steam line break event. i CALVERT CLIFFS - UNIT 1 B 2-5 Amendment No. 33, fB,71 l ---y -+-w + --+ g. ~~ -?--- = ,.--+--q ri, e g -v ew-eey.- p"--9 r e--< --,yy, e-

} LIMITING SAFETY SYSTEM SET 14. BASES Steam Generator Water Level The ' Steam Generator Water Level-Low trip provides core prote.ction 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 setroint provides allowance that there will be sufficient water. inventor.y,in the steam generators at the time of trip to provide a margin of more than 13 minutes before auxiliary ~~ t feedwater is required. 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 detectors. The trip setpoints 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 ma1 distributions. These trip set-points were derived from an analysis of many axial power shapes with allowances for instrumentation inaccuracies and the uncertainty associated with the excore to incore axial flux offset relationship. I Themal 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 number of reactor coolant pumps operating. 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 genera-tion of this trip function. In addition, CEA group sequencing in accor-dance with Specifications 3.1.3.5 ano 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. CALVERT CLIFFS - UNIT 1 B 2-6 Amendment No. 33, W,M, 71

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 exhihitt the most rapid decr' ease 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.rqactor trip when the difference in pressure between the two steam generators exceeds the trip setooint. 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 tenperatures. The most limit'ing 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 wil' ensure a trip at or before reaching the analysis setpoint. 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 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 Protec-tion System. I 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 apability at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System. CALVERT CLIFFS - UNIT 1 B 2-7 Amendment No. 27, 32, 39, AB, 71 4

I 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL SHUTDOWN MARGIN - T > 200 F avo LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be > 5.3%* Ak/k. ~ l APPLICABILITY: MODES 1, 2**. 3 and 4. ACTION: With the SHUTDOWN MARGIN < 5$3%* ak[k, immediately initiate and continue I ~ ' ' 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 > 5.3%* 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 SHUTOOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable CEA(s). b. When in MODES 1 or 2, at least once per 12 hours by verifying that CEA group withdrawaJ is within the Transient Insertion i Limits of Specification 3.1.3.6. 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-l down margin to match the shutdown margin requirements of the safety analyses.

    • See Special Test Exception 3.10.1.

eff 1.0.

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CALVERT CLIFFS - UNIT 1 3/4 1-1 Amendment No. 27,74,AE,71

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) e. When in MODES 3 or 4, at least once per 24 hours by con-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 Cays after each fuel loading. f

  • CALVERT CLIFFS - UNIT 1 3/41-2

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  • 0.20 g

og LONG TERM SHORT TERM 4 STEADY STEADY STATE STATE INSERTION 0.10 INSERTION LIMIT g LIMIT m I I ,I I I I I I I I I I I I I 0 h ALLOWABLE BASSS e m OPERATING REGION 3 5 3 1 GROUPSl l l l l l l l l l l l l l l l l l g 0 20 40 60 80 100 0 20 40 60 80 100 0 20 40 60 80 100 (136)(108.8)(81.6)(54.41 (27.21 (0) 1136) (108.8)(81.61 (51.41 (27.21 (0) (136)(108.8)(81.61 (54 4) (2/ 2) (0) I I I I I I I I I I I I s O 20 40 60 80 100 0 20 40 60 80 100 I (136)(108.8)(81.61 (54.41 (27.21 (0) (136)(108.8)(81.61 (54.4)(27.21 (0) uw %CEA INSERTION 4 (INCllES CEA WITHDRAWN) s-s Figure 3.1-2 CEA insertion Limits vs Fraction of Allowable Thermal Power for Existing RCP Combination

T l 3/4.2 POWER DISTRIBUTION LIMITS LINEAR HEAT RATE LIMITING CONDITION FOR OPERATION e 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 b. Be in at least HOT STANDBY within the next 6 hours. SURVEILLANCE 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 distribution 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 1 3/4 2-1 Amendment No. 27,33

e POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) c. Verifying at least once per 31 days that the AXIAL SHAPE INDEX is m'aintained within the limits of Figure. 3.2-2, where 160 percent of the allowable power represents the maximum THERMAL POWER allowed by the following expression: MxN where: 1. M is the maxirum allowable TH'ERMAL POWER 1evel for the existing Reactor Coolant Pump combination. 2. N is the maximum allowpble fraction of RATED THERMAL POWER as determined by the F curve of Figure 3.2-3b. l xy 4.2.1.4 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 alarms: 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 1.070, 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'. I;ALVERT CLIFFS - UNIT 1 3/4 2-2 Amendment No. 27, 25, $7, 77, Af, 71

005 4 004 r e a .e A s p u n ru N S B Y s O N A v IT O D e 0 I A T 0 I t R A 3 R a E R E R P W 1 t E O P O 2 a e E O P 3 H E L L r E B L R a L A B U U e T A F G n i P T E F L I E P V k C E 0 a I i C C 0 T e d C 2 C P E N A F e l U F b E a w lo lA 00 1 0 6 5 4 3 1 1 1 1-Eop<Noo2+05 sw k~ h dF4m H b I g4w3' x d a. Jmqgo$< 9Gh c,r~g'cz*' R* N.w $aa.S g

  • o* U.nn f

c

O 1 1 I I I ~ ~ ~ (-0.M,1.0) - - (0.12,1.0) 1.00 e 5 6 c. 0.90 UNACCEPTABLE UNACCEPTABLE OPERATION OPERATION { REGION REGION w I H g 0.80 S i .s a< u. A PT (-0.3. 0.70) < 0.70 <, (0.3, 0.70) OPERAT REGION O< E 0.s0 I I I I I o,g -0.6 -0.4 -0.2 0 0.2 0.4 ,0.6 PERIPHERAL AXIAL SHAPE INDEX, Yg Figure 3.2 2 Linear Heat Rate Axial Flux Offset Control Units CALVERT CLIFFS 'JNIT 1 3/4 2-4 Amendment No. 39,JB,yg 1

~ )06 0 0. 3 7 e e I 1 1 ) i I G 8 I 8 0 4. 76 1 ( W l N 1 sv s ro tc b a F 3-g 2 2 Y n l l 6 I X 3i k 1 F e a ru e U igP E f F L

d A

a V R L l M la E l B l-to A T T P E CCA l l M. ) 1 00 1 0, 1 5 1 ( 05 l l 1 6 4 1 0 0 0 0 0 0 0 1 0 9 8 7 6 5 1 1-0 0 0 0 0 N n? <mp" nC Mw e C54 a Rb 7sn p@"u$ ~ _? e s

_i _._-2.; ____3_. l=__. J } _.. ^ 1.06 ____I.__._2l-.;; .a__.. -= -- - t= = =. -- .. _ __ _ x _- - 4;.. ; ;.___ t;- :.= L=F=S=&5JhEm=: _=f= =asiiEJEE= ' =.Tt=_=EMM134@ 1:0551 ? ____. ;a__-. , _.._._7_ -~---..Z=~t=: =.8 ,f_. _ _. _

u. __. _.. - _

2 ~~ ~ _ _ _.__.,_. _ _, (118.6,1.050) 72M - = ---~--"e._".-_".~_.._~.__a"__--~"__...._-_~~_"__~~_._._ 1.05 7 -- . =. = - _a._._. -... cr. _... g . ;y _ g _--_-t-- t----- . - - - __4- ..a__. f..,.....v._., + #--# (10_6.5,1.046) = =~ -._:t . _. - + - ", m -.-- r__y _-.; =: - _ f;. ,,r v.-: 'T- _-. _.., _r_ ;_z

=;=. a=_ ;e=. =__:.__,__.r-4

t

_.4_ 1_. _ _ H_ ;- - - --.* : t --. _. u_a -- E r y_: -~2[z (90.5.1.041P--- -s=:. ri. 3= *=2-E-.__ =r: ^~-~__-=c- '- 1 --==L====.3Ei--; c: o 1.04

.

XG.t-..i. s==1=r-n = =3 witn =.=:_.== : =-

==_ _. t-

== =v-4 === - - r f _._..,__,.=._r___. 0 .a.- _ m. _ _.. _...==-3=.l_==r.._._. _ _2_s .__'- t___n _ .___ _EE____ +---+---j--- E!- :T-55-~~ ~ "i-5 E "; =.=iM".=-5ET:=i"[ii _ 4:_1.035,) =3[ gJ=q3-i; jisp._ 5_ l74 Q 1.03.5iE"$iE:7 ! IJEH=. d-E.4EEE 4-'

1!=t--i=45s[e.i:d.f:_T._EdMEL f Eili~ r

'9

(62.3,1.030) - _

=_ _i - =+:=I=u i. -.=:= : ru. t= - :.== ---=.:.==-a. - =r- ..__ r t_.__.__ g { ~~5NE~ --Nf = C- ~ = :--[ -- i --.. _ -_-f---- -d-n- - - - - ' _4 1;... = _ :;_=. ;.. -. - / --

==_._--

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2

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  • = * * =

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2. :..._-__4 g;;, ;

m ( 1.02 ... ;;=4.;_;=..._;; = . _ _, _.,.g_, _ _ . __a__ _. -~ . _.. _ _. _ = _ -.: (30.2,1.017) . __ ~ ___ = - - -, - - - -. -r- - r =.t= n - mf._=_ ;,... --. r..__._. _ _. - ={= .=. t- -2 = - t= =iu _..._2

=1 r =

=: r :f._. _- _

a..

=. _ ._ _;_ _w =.- =_,3 2.. _;; ;;. -- :. _f --........ __ -2r-= (18.1.1.011)==-t-~== :t==2 --. 1.01 ---;-$. _. _,., _. _ _ _ _ _ _, _. - __ _ L_ ~ - -- g. en. %Eiis Eist==.=e

_.,_ _.. _.m.i = e d = a = ~ E = = G.==r =_..=._-teswemE

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< (2_.0_, 1. 002. ),-... _ _...... _. _.,"_ _ :... g".. _.__ _ m-g.__... ._4 _ _'r..- + +- 1.00 0 20 40 60 80 100 120 140 CORE HEIGHT, INCHES t t l l FIGURE 4.2-1 ' Augrnentation Factor vs Distance from Bottom of Core CALVERT CLIFFS - UNIT 1 3/4 2-5 Amendment No.48 l l l I

POWER DISTRIBUTION LIMITS TOTALPLANARRADIALPEAKINGFACTOR-Fh LIMITING CONDITION FOR OPERATION T T 3.2.2.1 The calculated value of F*Y, defined as F =F*Y(1+T ), shall be 7 1imited to < 1.65.

  • Y..

8 APPLICABILITY: MODE 1*. ACTION: With Ff, > 1.65, within 6 hotjes either: [- a. Reduce THERMAL POWER to bring the tombination of THERMAL POWER andFfy to within the limits of Figure 3.2-3a and withdraw the l 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 STANDBY. SURVEILLANCE REQUIREMENTS

4. 2. 2.1.1 The provisions of Specification 4.0.4 are not applicable.

l Ffy shall be calculated by the expression F =F 42.2.1.2 xy(1+T ) and l q F shall be determined to be within its limit at the following intervals: xy 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 Within four hours if the AZIfiUTHAL POWER TILT (T ) is > 0.030. c. q

  • See Special Test Exception 3.10.2.

CALVERT CLIFFS - UNIT 1 3/4 2-6 Amendment No. 32, 33, 48,7 1

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.2.1.3 F shall be. determined each time a calculation of FT is required I o . XY .xy by using the incore detectors to obtain a power.distfibution 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. I 4.2.2.1.4 T shallbedeterkinedea'c.h ime a calculation of Ffy is required. l ~ ', ~ q and the value of T used to determine F shall be the measured value of q xy 2-L CALVERT' CLIFFS-UNIT 1 Amendment No. 27, 77, 7 j j 3/42-7 I I

57 ~ 1 )06 0 03 m 7 1 ( 3 E 7 1, V 1 R U C T E I r L M e w BN IL o AO N 1 TI Y P i P TO

  1. x 7

l l E AI F 1 a G CR 6 m E CE

  1. r r

R e P A F h .O O T N / s U de ta 9 R 6 l I f 1 o no i tca r aF 3 T. r - le

2. b 7

6 'Y 3 1 I a 1 T X ew ruo l gl iA F )0 sv 1 s r 0 5 o 5 6 c t 6 - l I 1 a 1 E ( LN F BO N g AITO n i T AI k P G a ER e E C E P R P C .l O A 3 ia 6 d I I 1 a R la to T 1 6 I l 1 9 5 1 0 0 0 0 0 0 0 1 0 9 8 7 6 5 1 1 0 0 0 0 0 xyOa. J4E$E O 4[oz o4x$,m43,a< 9rMp* hymm e _I Z " R4 m& 6."& 3" =0-E

4 e [ f POWER DISTRIBUTION LIMITS TOTAL PLANAR RADIAL PEAKING FACTOR'- F i LIMITING CONDITION FOR OPERATION c \\ 3.2.2.2 The value of N presently used in Specification 4.2.1.3 shall be in i j accordance with Figure 3.2-3b. i a APPLICABILITY: MODE.1 when operating in accordance with Specification 4.2.1.3. l ACTION: r With the value of N presently used in Specification 4.2.1.3 exceeding the 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 i b. Be in at least HOT STANDBY. SURVEILLANCE REQUIREMENTS 1 6 4.2.2.2.1 The provisions of Specification 4.'O.4 are not applicable. i Ffy shall be calculated by the expression F F (1+T ) and N 4.2.2.2.2 q l shall be determined to be within its limit by monitoring F at the following i intervals: U a. Prior to operation above 70 parcent of RATED THERMAL POWER - af+er each fuel loading, j i b. At least once per 3 days of accumulated operation in MODE 1. 3 T 4.2.2.3 F shall be determined each time a calculation of F is required [ xj xy by using the incore detectors to obtain a power distribution map with all full f length CEAs at or above the Long Term Steady State Insertion Limit for the l ~ existing Reactor Coolant Pump combination. This determination shall be limited i to core planes between 15% and 85% of full core height inclusive and shall exclude regions influenced by grid effects. l ime a calculation of Ffy is required j 4.2.2.2.4 T shall be determined each q and the value of T used to determine F shall be the meas'ured value of T. q q CALVERT CLIFFS - UNIT 1 3/4 2-8 Amendment No. 71 l

~ POWER DISTRIBUTION LIMITS TOTALINTEGRATEDRADIALPEAKINGFACTOR-~F[' LIMITING CONDITION FOR OPERATION T 3.2.3 The calculated value of F, defined as FT = F (1+T ), shall be limited to < l.650. r r r q g APPLICABILITY: MODE 1*. CTION: WithFf>1.650within6ho.urs,either: 8 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 3.2.3a, withdraw the full length CEAs to or beyond the Long Term Steady State Limits of Specification 3.1.3.6, and insert new value of Ff in BASSS; or Reduce THERMAL POWER to bring the combination of THERMAL POWER and c. TF to within the limits of Figure 3.2-3a and withdraw the full length I p CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6. The THERMAL POWER limit determined from Figure 3.2-3a shall then be used to establish a revised upper THERMAL I POWER level limit on Figure 3.2-4 (truncate Figure 3.2-4 at the allowable fraction of RATED THERMAL POWER determined by Figure 3.2-3a) and subsequent operation shall be maintained within the reduced acceptable operation region of Figure 3.2-4. I SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. Ff shall be calculated by the expression Ff = F (1+T ) and Ff 4.2.3.2 r q shall be determined 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 j and l c. Within four hours if the AZIMUTHAL POWER TILT (T ) is > 0.030. q

  • See Special Test Exception 3.10.2.

CALVERT CLIFFS - UNIT 1 3/4 2-9 Amendment No. 37,33,39, M,7l

- i f SURVEILLANCE REQUIREMENTS (Continued) 4.2.3.3 F shall be determined each time a calculation of FT is required r 7 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 combinati,on. 4.2.3.4 T shallbedeterminedeachtimeacalculationofF[isrequired q T and the value of T used to determine F shall be the measured value of T. q 7 q r J CALVERT CLIFFS - UNIT 1 3/4 2-10 Amendment No. 27, 32 i

1.2 I I I i l 1.1 l l l a: yg 1.0 (.0.15,1.0) - - (0.T5,1.0) o9

a. r UNACCEPTABLE UNACCEPTABLE miii OPERATION OPERATION jg REGION REGION 30 0.9

'4 a 0 2"

s c 2E R V' ACCEPTABLE Ed 0.8

( 0.30, 0.80) i OPERATION (0.30, 0.80) u-e REGION Oo z u-9d 6> Nd E 0.7 1 0.6 1 I l l I I 0.5 -0.6 -0.4 0.2 0 0.2 0.4 0.6 PERIPHERAL AXlAL SHAPE INDEX, Y, Figure 3.2-4 DNB Axial Flux Offset Control Limits CALVERT CLIFFS - UNIT 1 3/4 2-11 ,bendment lio. AB, 71

7 POWER DISTRIBUTION LIMITS AZIMUTHAL POWER TILT - Tq LIMITING CONDITION FOR OPERATION ~ 3.2.4 The AZIMUTHAL POWER TILT (T ) shall not exceed 0.030. q APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER.* ACTION: a. With the indicated AZIMUTHAL POWER TILT determined to be > 0.030 l but < 0.10, either correct the power tilt within two hours or determine within the next 2 hours and at least once per subse-quent8 hours,thattheTOTALPLANARRADIALPEAKINGFACTOR(Ffy) 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 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 REQUIREMENT 4.2.4.1 The p'rovisions.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 hcurs when one excore channel is inoperable and THERMAL POWER IS > 75% of RATED THERMAL POWER.

  • See Special Test Exception 3.10.2.

CALVERT CLIFFS - UNIT 1 3/4 2-12 Amendment No. ZJ, 32

t 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 power 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. 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 determined to be within its limit by measurement at least once per 18 months. CALVERT CLIFFS-UNIT 1 Amendment No. 39. Si,71

s, a TABLE 3.2-1 m( DNB PARAMETERS 9 -i p3 LIMITS ~ 4 Four Reactor Three Reactor Two Reactor Two Reactor [ Coolant ~ Pumps Coolant Pumps Coolant Pumps Coolant Pumps z Parameter Operating Operating Operating-Same Loop Operating-Opposite Loop _ _4 i Cold Leg Temperature < 548'F Pressurizer Pressure 3,2225 psia

  • t Reactor Coolant System l

i Total Flow Rate 3,370,000 gpm AXIAL SHAPE INDEX l', ^ c 2 t S*

  • Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% of RATED THERMAL POWER !

II ~ per minute or a THERMAL POWER step increase of greater than 10% qf 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 < 55% when BASSS is

( inoperable, or within the limits of FIGURE 3.2-4 for CEA insertions specifie3' by FIGURE '3.1-2. 1 2 a 5 ~ m.-

i, ~ g TABLE 3.3-1 (Continued) G g_ REACTOR PROTECTIVE INSTRUMENTATION 4 i q MINIMUM g TOTAL N0. . CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION C3

11. Wide Range Logarithmic Neutron H

Flux Monitor 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

12. Reactor Protection System 6

1 6 1, 2

  • 4 Logic Matrices
13. Reactor ~ Protection System 4/ Matrix 3/ Matrix 4/ Matrix 1, 2*

4 R Logic Matrix Relays u y

14. Reactor Trip Breakers 8

6 8 1, 2

  • 4 w

L I 4 4 y -..r,- .-...,,mm -v y 7- --y

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.
  1. The provisions of Specification 3.0.4 are not applicable.

Trip may be bypassed below 10'4be automatically removed whe of RATED THERMAL POW (a) THERMAL POWER. of RATED (b) Trip may be manually bypassed below 710 psia; bypass shall be automatically removed at 'o'r abov'e 710 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 Range Logarithmic Neutron Flux Monitoring Channels and the Power j Range Neutron Flux Monitoring Channels. I 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 status 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 ei~ her 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 CLIFFS - UNIT 1 3/4 3-4 Amendment No. AS, 71 1 i

e 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 I i up to 48 hobrs while performing tests and maintenan' e c on that channel provided the other inoperable channel is placed in the tripped' condition. ' ACTION 3 With the number of channels OPERABLE one less than required 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 thereaf'er. ACTION 4 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in HOT STANDBY within 6 hours; however, one channel may be 4 bypassed for up to 1 hour for surveillance testing per Specification 4.3.1.1. i I e - e I I P l' CALVERT CLIFFS - UNIT 1 3/4 3-5 l

n _ TABLE 3.3-2 .md REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES b. FUNCTIONAL UNIT Y RESPONSE TIME 7 1. Manual Reactor Trip Not Applicable 2. Power Level - High < 0.40 seconds *# and < 12.0 seconds ## l 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 l'evel - Lo'w < 0.90 seconds 8. Axial Flux Offset ,1 0.40 seconds *# and _< 12.0 seconds ## l 9.a. Thermal fiargin/ Low Pressure O'.90 seconds *# and < 12.0 seconds ## b. Steam Generator Pressure Difference - High 1 0.90 seconds a-.g 10. Loss of Turbine--Hydraulic I Fluid Pressure - Low Not' Applicable

11. Wide Range Logarithmic Neutron Flux Monitor Not Applicable a

4 P

  • Neutron detectors are exempt from response time testing.

e of the channel shall be measured from detector output or input of first electronic component

  1. Response time does not include contribution of RTDs.

P

    1. RTD response time only.

to achieve 63.2% of its total change when subjected to a step change in RTD tem ?

TABLE 3.3-3 (Continued) TABLE NOTATION (a) Trip function may be bypassed in this fiODE when pressurizer pressure is < 1800 psia; bypass shall be automatically removed when pressurizer pressure is 11800 psia. I (c) Trip function may be bypassed in this MODE below-710 psfa;' bypass shall be automatically removed at or above 710 psia. The provisions of Specification 3.0.4'are not applicable. . ACTION STATEMENTS With the number of OPERABLE channels one less than the ~ ACTION 6 ~ 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 Total Number of Channels, 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. 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. 2 i CALVERT CLIFFS - UNIT 1 3/4 3-15 Amendment No. AS, 71

TABLE 3.3-3 (Continued) ACTION 8 With less than the Minimum Channels OPERABLE, operation 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. l L CALVERT CLIFFS - UNIT 1 3/4 3-16

e, TABLE 3'.3-4 9G ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUE_S_ w ALLOWABLE r- _FU_NCTIONAL UNIT TRIP SETPOINT ' VALUES 1. SAFETY INJECTION (SIAS) a. Manual (Trip Buttons) Not Applicable Not Applicable g b. Containment Pressure - High 1 4.75 psig 1 4.75 psig c. Pressurizer Pressure - Low > 1725 psia g 1725 psia l 2. CONTAINMENT SPPAY (CSAS) a. Manual (Trip Buttons) Not Applicable Not Applicable b. Containment Pressure -- High 1 4.75 psig < 4.75 psig 3. CONTAINMENT ISOLATION (CIS) # { a. Manual CIS (Trip Buttons) Not Applicable Not Applicable b. Containment Pressure - liigh 1 4.75 psig

4.75 psig 4.

MAIN STEAM LINE ISOLATION a. Manual (MSIV Hand Switches and feed Head Isolation y Hand Switches) Not Applicable Not Applicable E !} b. Steam Generator Pressure - Low 1 635 psia 3,635 psia l 3" l e t-2 P

  1. Containment isolation of non-essential penetrations is also initiated by SIAS (functional units 1.a and 1.c).

D = 9

TABLE-3.3-4 (Continued) (( ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES Bifi ALLOWABLE PP-FUNCTIONAL UNIT TRIP VALUE VALUES MM ?? 5. CONTAINMENT SUMP RECIRCULATION (RAS) 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 #f l a. Manual (Purge Valve Control Switches) Not Applicable Not Applicable b. Containment Radiation - High Area Monitor < 220 mr/hr < 220 mr/hr y 7. . LOSS OF POWER a. 4.16 kv Emergency Bus Under-2450+105 volts'with a 2450+105 volts with a voltage (Loss of Voltage) 2 0.2 second tirne delay 2[0.2 second time delay t b. 4.16 kv Emergency Bus Under-. 3628+25 volts with a 3628+25 volts with a voltage (Degraded Voltage) 810.4 second time delay 810.4 second time delay NN JT

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

s #. .sW 5:0 9 r ~ ~

1 i TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 6. Steam Generator Pressure-Low

  • 's 12, g'

] a. Main Steam Isolation b. Feedwater Isolation 1 80 i 7. Refuelina Water Tank-Low, a. Containment Sump Recirculation < 80 8. Reactor Trip a. Feedwater Flow Reduction to 5% 1 20 9. Loss of Power a. 4.16 kv Emergency Bus 1 2.2 Undervoltage (Loss of Voltage) i I l b. 4.16 kv Emergency Bus 1 8.4 l Undervoltage (Degraded i Voltage) } 10. Steam Generator Level - Low a. Auxiliary Feedwater System 1 360*/360** (2) TABLE NOTATION

  • 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 t

to the diesel generator start signal. (1 ) Header fill time not included. (2) Includes time delay of 3 to 5 minutes. l I CALVERT CLIFFS - UNIT 1 Amendment No. 4D. 44, 71

TABLE 4.3-2 hh ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTAION SURVEILLANCE REQUIREMENTS

  • M

$g CHANNEL MODES IN WHICH ng CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE Cg FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED 5 "m m 1. SAFETY INJECTION (SIAS) i . i a. Manual (Trip B'uttons) N.A. N.A. R N.A. E c;; b. Containment Pressure - liigh S R M 1, 2, 3 qO c. Pressurizer Pressure - Low S R M 1,2,3 m] d. Automatic Actuation Logic R.A. N.A. M(1)(3) 1, 2, 3 2. CONTAINMENT SPRAY -(CSAS) a. Manual (Trip Buttons) N.A. N.A. R N.A. b. Containment Pressure -- High S R M 1, 2, 3 c. Automatic Actuation Logic N.A. N.A. M(1) 1, 2, 3 M 3. CONTAINMENT ISOLATION (CIS) # l a. Manual CIS (Trip Buttons) N.A. N.A. R N.A. w y b. Containment Pressure - High S R M 1,2,3 c. Automatic Actuation Logic N.A. N.A. M(1)(4) 1, 2, 3 i t 4. MAIN STEAM LINE ISOLATION (SGIS) a. Manual SGIS (MSIV Hand EE Switches and Feed Head Isolation Hand Switches) N.A. N.A. R N.A. 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 EE

  1. Containment isolation of non-essential penetrations is also initiated by SIAS (functional units wm 1.a and 1.c).

t - I.

PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve shsll be.0PERABLEr APPLICABILITY: MODES 1, 2 and 3. ACTION: With one main steam line isolation valve inoperable, MODE 1 POWER OPERATION may continue provided the inoperable valve is either restored to OPERABLE status or closed within 4 hours; otherwise, be in HOT SHUTDOWN within the next 12 hours. MODES 2 With one main steam line isolation valve inoperable, sub-and 3 sequent operation in MODES 1, 2 or 3 may proceed provided: a. The isolation valve is maintained closed. b. The provisions of Specification 3.0.4 are not applicable. l Otherwise, be in HOT SHUTDOWN within the next 12 hours. SURVEILLANCE REQUIREMENTS 4.7.1.5 Each main steam line isolation valve shall be demonstrated l OPERABLE by verifying full closure within 4.0 seconds when tested l l pursuant to Spe.cification 4.0.5. l CALVERT CLIFFS-UNIT 1 3/4 7-9 Amendment No. 71 1 l

e l ~ Deleted CALVERT CLIFFS - UNIT 1 3/4 7-10 Amendment No. H. 59

t 3/4.1 REACTIVITY CONTROL SYSTEMS BASES F 3/4.1.1 B0 RATION 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 reactivit/ trinsients 1 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 cor'e life as a function of fuel depletion, RCS boron con' centration and RCS T The minimum available~ ~ ~~ SHUTDOWN MARGIN for no load operating' conditions $l9beginning of life is 4.5% - ak/k and at end of life is 5.3% ak/k. The SHUTDOWN MARDIN 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 5.3% Ak/k. Accordingly, theSHUTDOWNMARGINrequirementisbaseduponthislimitingconditfonandis consistent with FSAR safety analysis assumptions. With T 200 F, the reactivity transients resulting from any postulated accid $N <re minimal and a 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 horon concen-tration reducti-ons 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 that 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 pcoefficient changes slowly due principally to the reduction in RCS boron . concentration associated 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 1 B 3/4 1-1 Amendment No. 27, 22,4/. 7, t L

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1.5 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made crifical with the Reactor Coolant System average temaerature-less than 515 F. This limitation is required to ensure 1) the moderator-temperature coefficient is within its analyzed temperature range, 2) the protective instrumentation is within its normal operating range, 3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and 4) the reactor f issure vessel is above its minimum RT temperature. NDT 3/4.1.2 B0 RATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility. operation. The components required to perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow paths, 4) boric acid pumps, 5) associated heat tracing systems, and 6) an emergency power supply from OPERABLE diesel generators. With the RCS average temperature above 200 F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the systems inoperable. Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period. The boration capability of either system'is sufficient to provide a SHUTDOWN MARGIN from all gperating conditions of 3.0% ak/k after xenon decay and cooldown to 200 F. The maximum boration capability requirement occurs at E0L from full power equilibrium xenon conditions and requires 6500 gallons of 7.25% boric acid solution from the boric acid tanks or 55,627 gallons of 2300 ppm borated water from the refueling water tank. However, to be consistent with the ECCS requirements, the RWT is required to have a minimum contained volume of 400,000 gallons during MODES 1, 2, 3 and 4 The maximum baron concentration of the refueling water tank shall be limited to 2700 ppm and the maximum boron concentra-tion of the boric acid storage tanks shall be limited to 8% to preclude the possibility of boron precipitation in the core during long term ECCS cooling. With the RCS temperature below 200 F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restric-tions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable. CALVERT CLIFFS - UNIT 1 ' Amendment No. 27, 4, b5 CALVERT CLIFFS - UNIT 2 B 3/4 1-2 Amendment No. 31

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 ra,te does not. exceed its limits. The Excore Detector Monitoring System perfoms. this function by continu-ously monitoring the AXIAL SHAPE INDEX with the OPERABLE quadrant symmetric 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, the following assumptions are made:

1) the CEA insertion limits of Specifications 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 Specification 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 within the allowable limits of Figure 3.2-1. The setpoints for these alarms include allowances, set in the conservative directions, for 1) flux peaking augmentation factors as shown in Figure 4.2-1, 2) a measurement-calculational uncertainty factor of 1.070, 3) an l engineering uncertainty factor o' l.03, 4) an allowance of 1.01-for axial fuel densification and thermal expansion, and 5) a THERMAL POWER measurement uncertainty factor of 1.02. 3/4.2.2. 3/4.2.3 and 3/4.2.4 TOTAL PLANAR AND INTEGRATED RADIAL PEAKING AND Ff AND AZIMUTHAL POWER TILT - T FACTORS - F q T The limitations on F and T are provided to ensure that the assump-Y tions used in the analysis for es@ablishing the Linear Heat Rate and Local Power Density - High LCOs and LSSS setpoints remain valid during operation y the various allowable CEA group insertion limits. The limitations on F and T are provided to ensure that the assumptions used in r q

ALVERT CLIFFS - UNIT 1 B 3/4 2-1 Amendment No. 33, 39
ALVERT CLIFFS - LCIT 2

. 1c.n'acnt..o. 78, 24

POWER DISTRIBUTION LIMITS BASES the analysis establishing the DNB Margin LCO, and Thermal Margin / Low PressureLSSSsetpointsremainvalidduringToper9tionatthevarious allowable CEA grouc insertion limits. If F F or T exceed their basic limitations, operation may continue ufider Ehe id81tioilal'restric-tiens 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 ti,lt. T and.F{hevalueofT r(1+T)isthemeasuredtilt.that must be used in the equation F*7 = 9 =F r q T T The surveillance requirements for verifying that F p within their limits provide assurance that the actual v Yu,es}and T ye T 4 fF T and T do not exceed the assumed values. Verifying F and F afNr r each fuel loading prior to exceeding 75% of RATED THEMAL POWER provides 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 the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient varification of flow rate on a 12 hour basis. CALVERT CLIFFS - UNIT 1 B 3/4 2-2 Amendment No, y, gg, /p 7 I

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.._.. p.3,j,. g !py %...~ bi~.->;,,x.ggiY.7[Q y{U( q_ 1. ' G.:73 J.. ,~% 3 ) y te( . pp.o.. - c 7 w. -6 vw av .ges. jg; . - t : $. _: 2.. .s.,,. ._v , 9, c... z ; - u c.,r. j 'g .j.*' _',y; .2 ,T- , s t j. ~l. p d * - ' ' l 4 . (0- ,,f* x I N ,- G..** W G. ; "s S 3 J b&, h~.%: l FW'/m. E,,.d'r'~M x. s.k,h ? &;za.5'9yX ??:- c, x d',. o:~ N. :_. LOW POPULATION ZONE F I G. 5.1 - 2 CALVERT CLIFFS - UNIT 1 CALVERT CLIFFS - UNIT 2 5-3 1

i DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 tained for a maximum internal pressure of 50 psig and 276*F. ~ ~ 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 217 fuel assemblie~s with ea assembly containing a maximum of 176 fuel rods clad with Zircaloy-4. fuel rod shall have a nominal active fuel length of 136.7 inches and Each contain a maximum total weight of 3000 grams uranium. loading shall have a maximum enrichment of 2.99 weight percent U-235 The initial core 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 ELEMENT ASSEMBLIES 5.3.3 control element assemblies.The reactor core shall contain 77 full length and no pa 5.4 REACTOR COOLANT SYSTEM i e

  • DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

i In accordance with the code requirements specified in Section a. 4.2 of the FSAR with allowance for normal degradation pursuant of the applicable Surveillance Requirements, b. For a pressure of 2500 psia, and For a temperature of 650 F, except for the pressurizer which [ c. is 700 F. ) f l CALVERT CLIFFS - UNIT 1 Amendment No. 32,#2 71 5-4 __. _.}}

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