ML19241B992

Amend to License DPR-53,authorizing Operation W/Modified Guide Tubes for Control Element Assemblies W/High Burnup Demonstration Fuel Assembly Installed in Core & Revising Tech Specs Based on Analysis of Cycle 4 Reload Fuel

ML19241B992
Person / Time
Site:	Calvert Cliffs
Issue date:	06/14/1979
From:	Reid R Office of Nuclear Reactor Regulation
To:
Shared Package
ML19241B993	List: ML19241B992 ML19241B994 ML19241B995
References
NUDOCS 7907260092
Download: ML19241B992 (39)

Text

{{#Wiki_filter:. ./#"% UNITED STATES y7 t NUCLEAR REGULATORY COMMISSION i 1(l j j $m a f r WASHINGTON, 0. C. 20585 %,'.v ,/ BALTIMORE GAS & ELECTRIC COMPANY DOCKET NO. 50-317 CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Anendment f!o. 39 License No. OPR 53 1. The Nuclear Regulatory Comission (the Comission) has found that: A. The application for amendment by Baltirore Gas and Electric Company (the licensee) dated 'ebruary 23, 1979 as supplemented, complies with r the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; 8. The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Comission; C. There is reasonabla assurance (1) that the activities authorized by this amendment.2n be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D. The issuance of this amendment will not be inimical to the comon 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 Comission's regulations and all applicable requirements have been satisfied. 126 0o ib c 190 483 263

' i' 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-33 is hereby amended to read as follows: (2) Techr.ical Specifications The Technical Specifications contained in Appendices / A and B, as revised through Amendment No. 39, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the l Technical Specifications. i 3. This license amendment is effective as of the date of its issuance. FOR THE NUCLEAR REGULATORY COMMISSICN l' ly'Y - s Robert W. Reid', Chief Operating Reactors Branch #4 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: June 14, 1979 t l

ATTACHMENT TO LICENSE AMENDMENT N0._30 FACILITY OPERATING LICENSE NO. ".dR-53 DOCKET NO. 50-317 Replace the following pages of the Appendix "A" Technical Specifications The revised pages are identified by Amendment with the enclosed pages. The number and contain vertical lines. indicating the area of change. corresponding overleaf pages are also provided to maintain document completeness. Paaes IV l-3 2-7 -2 12 2-13 B 2-3 B 2-4 B 2-6 B 2-7 3/4 1-23 3/4 2-2 3/4 2-4 3/4 2-5 3/4 2-8 3/4 2-9 3/4 2-11 3/4 2-13 3/4 2-15 3/4 3-6 B 3/4 2-1 B 3/4 2-2 483 265

.a INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICAP'LITY.................... 3/4 0-1 3/4.1 REACTIVITY CONTRPL SYSTEMS 3/4.1.1 B0 RATION CONTROL Shutdown Margin - T > 200 F................ 3/4 1-1 avg Shutdown Margin - T 1 200 F....................... 3/4 1-3 avg Bo.on Dilution.............................. 3/4 1,-4 Moderator Tempera ture Coef ficient................... 3/4 1-5 Minimum Temperature for Criticality.................. 3/4 1-7 3/4.1.2 B0 RATION SYSTEMS Flow Paths - Shutdown................................ 3/4 1-8 Fl ow Pa ths - 0pera ti ng............................... 3/4 1-9 Charging Pump - Shutdown............................. 3/4 1-10 Charging Pumps ~0perating........................... 3/4 1-11 -1 B]ric Acid Pumps - Shutdown.......................... 3/4 1-12 Eoric Acid Pumps - Operating......................... 3/4 1-13 30 rated Water Sources - Shutdown..................... 3/4 1-14 Borated Water Sources - Opera ting.................... 3/4 1-16 3/4.1.3 MOVABLE CONTROL ASSEMBLIES 3/4 1-17 Full Length CEA Position Position Indicator Channels.......................... 3/4 1-21 CEA Drop Time..................... 3/4 1-23 Shutdown CEA Insertion Limits..................... 3/4 1-24 Regulating CEA Insertion Limits.............. 3/4 1-25 CALVERT CLIFFS - UNIT 1 III Amendment No. 32 483 266

INDEX_ LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE..................................... 3/4 2-1 l 3/4.2.2 TOTAL PLANAR RADIAL PEAKING FACT 0R................... 3/4 2-6 1/4.2.3 10TAL INTEGRATED RADIAL PEAKING FACTOR............I.. 3/4 2-9 3/4.2.4 AZIMUTHAL POWER TILT.............._....,.............. 3/4 2-12 3/4 2-13 l 3/4.2.5 DEI ""............................................. 3/4.2.6 rARAMETERS....................................... 3/4 2-14 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION................... 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION......................,............. 3/4 3-10 3/4.3.3, MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation................. 3/4 3-25 Incore Detectors............................,........ 3/4 3-29 Seismic Instrumentation.............................. 3/4 3-31 Meteorological Instrumentation....................... 3/4 3-34 Remote Shu tdown Ins trumentation...................... 3/4 3-37 Post-Accident Instrumentation........................ 3/4 3-40 Fire Detection Instrumentation....................... 3/4 3-43 3 /4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT L00PS................................ 3/4 4-1 3/4.4.2 SAFETY VALVES - SHUTCCWN............................. 3/4 4-3 3/4.4.3 SA FET V AL V ES - 0 P ER ATI NG............................ 3/4 4-4 CALVERT CLI""S - UNIT 1 IV Amendment No. 27, 75 19 48[ /Ul

DEFINITIONS CHANNEL CHECK 1.10 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter. CHANNEL FUNCTIONAL TEST 1.11 A CHANNEL FUNCTIONAL TEST shall be: a. Analog channels - the injection af a simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions. b. Bistable channels - the injection o# e simulated signal into the channel sensor to verify OPERABILITY including alarm and/or trip functions. CORE ALTERATION 1.12 CORE ALTERATION shall be the movement or manipulation of any component pithin,the reactor pressure vessel with the vessel head removed and fuel in 'the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a -component to a safe conservative position. SHUTDCWN MARGIN 1.13 SHUTCCWN "ARGIN shall be the instantaneous amount of reactivity by which tho reactor is subcritical or would be subcritical irem its present conditi;.. assuming all full length control element assemblies (shutdown and regulating) are fully insarted except for the single assembly of highest reactivity worth which is assumed to be fully withdrawn. 0 CALVERT CLIFFS-UNIT I 1-3 Amendment No. 483 260

DEFINITIONS IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be: Leakage (except CONTROLLED LEAKAGE) into closed systems, such a. as pump seal or valve packing leaks that are captured, and conducted te a sump or collecting tank, or Leakage into the containment atmosphere from sources that are b. both specifical, ~.ocated and known either not to interfere with the opera +'an of leakage detection systems or not to be PRESSURE B0UNDARY LEAKAGE, or Reactor coolant system leakage through a steam generator to the c. secondary system. UNIDENTIFIED LFAKAGE UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED 1.15 LEAKAGE or CONTROLLED LEAKAGE. PRESSURE BOUNDARY lFAKAGE PRESSURE EOUNDARY LEAKAGE shall be leakage (except steam generator 1.16 tube leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall. CONTROLLED LEAKAGE CONTROLLEC LEAKAGE shall be the water flow from the reactor coolant 1.17 pump seals. AZIMUTHAL POWER TILT - Tq AZIMUTHAL PCWER TILT shall be the maximum difference between the 1.18 pcwer generated in any core quadrant (upper or lower) and the average power of all quadrants in that half (upper or lower) of the core divided by the average power of all quadrants in that half (upper or lower) of the core. DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of 1-131 (uCi/ gram) 1.19 which alone would produce the same thyroid dose as the quantity and isotcpic mixture of I-131, I-132,1-133, I-134 and I-135 actually prese nt. The thyroid dose conversion factors used for this calculation sha*1 be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites." CALVERT CLIFFS - UNIT 1 1-4 40) 2b9

TABLE 2.2-1 s2 REACTOR PROTECTIVE INSTRUMENTA110N TRIP SETPOINT LIMITS r 9d [2 FUNCTIONAL UNIT TRIP SETPOINT_ ALLOWABLE VALUES Maa 1. Manual Reactor Trip Not Applicable Not Applicable 2 2. Power Level - High a. Four Reactor Coolant Pumps < 10% above THERMAL POWER, with 5,10% above THERMAL POWER, and Operating a minimum setpoint of 30% of RATED a minimum setpoint. of 30% of RATED THERMAL POWER, and a maximum of THERMAL POWER and a maximum of < 107.0% of RATED THERMAL PUJER. 1 107.0% of RATED THERMAL POWER. l b. Three Reactor Coolant Pumps s 10% above THERMAL POWER, with 1 10% above THERMAL POWER, and Operating a minimum setpoint of 30% of RATED a minimum setpoint of 30% of RATED THERMAL POWER, and a maximum of THERMAL POWER and a maximum of n, L < 80% of RATED THERMAL POWER. 1 80% of RATED THERMAL POWER. c. Two Reactor Coolant Pumps 1 10% above THERMAL POWER, with s 10% above THERMAL POWER, and Operating - Same Loop a minimum setpoint of 30% of RATED a minimum setpoint of 30% of RATED ff THERMAL POWER, and a maximum of THERMAL POWER and a maximum of 5 ~< 46.8% of RATED THERMAL POWER. 1 46.8% of RATED THERMAL POWER. 3 k d. Two Reactor Coolant Pumps < 10% above THERMAL POWER, with s 10% above THERMAL POWER, and Operating - Opposite Loops a minimum setpoint of 30% of RATED a minimum setpoint of 30% of RATED THERMAL POWER, and a maximum of THERMAL POWER and a maximum of o < 51.1% of RATED THERMAL POWER. 1 51.1% of RATED THERMAL POWER. e (c 45m C 'A k N si CD

n# TABLE 2.2-1 (Cont'd) i t 9] -REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS a P;; FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES U 3. Reactor Coolant Flow - Low (1) c:5 a. Four Reactor Coolant Pumps > 95% design reactor coolant > 95% of design reactor coolant Operating flow w ch 4 pumps operating

• flow with 4 pumps operating
• b.

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• flow with 4 pumps operating
• c.

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• flow with 4 pumps operating
• 7 ca d.

Two reactor Coolant Pumps > 50% of design reactor coolant 1 50% of design reactor coolant Operating - Opposite Loops Tlow with 4 pumps operating

• flow with 4 pumps operating *
• Design reactor coolant flow with 4 pumps operating is 370,000 gpm.

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This page left blank pending NRC approval of ECCS analysis for three pumo operation. Figure 2.2-4 Thermal Margin / Low Pressure Trip Setpoint-Part 1 Three Reactor Ccolant Pumps Operating CALVERT CLIFFS - UNIT 1 2-14 Amendment No. 21 I0b ?l?b t

B g SAFETY 1.IMITS 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 conditicns for reactor coolant system temperature, pressure, and THERMAL POWER level that would result in a ONBR of less than 1.19 ar 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 '"ereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere. The reactor pressure vessel and oressurizer 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 ccnsistent with the design criteria and associated code requirements. The entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to initial cperation. CALVERT CLIFFS - UNIT 1 B 2-3 Amendment No. 33, 3 9 483 2?6

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SETPOI qS. 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 frcm exceeding their safety limits. Operation with a trip se' less conservative than its Trip Setpoint but within its speci-fied Allowable Value is acceptable on the basis that ti.e difference between the trip setpoint and the 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 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 pcwer decreases. The trip set; int has a maximum value of 107.0% of RATED l 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 112% of RATED THERMAL POWER, which is the value used in the safety analyses. Reactor Coolant Flow-Low The Reactor Coolant Ficw-Low trip provides core protecticn to prevent DNB in the event of a sudden,ignificant decrease in reactor coolant Provisions have been made in the reactor protective system to permit fl ow. CALVERT CLIFFS - UNIT 1 B 2-4 Amendment No. 3 3 4d) 2/!

~ LIMITING SAFETY SYSTEM SETTINGS BASES operation of the reactor at reduced power if one or two reactor coolant pumps are taken out of service. The low-flow trip set;~ oints 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 1.19 under nomal operation l and expected transients. For reactor operation with only two or three reactor coolant pumps operating, the Reactor Coc' ant Flow-Low trip set-points, the Power Level-High trip setpoints, and the Thermal Margin / tow 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 D 3 h from going below 1.19 during l normal operational transients and anticipated transients when only two or three reactcr coolant pumps are operatir.g. 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 thc 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 fo: this trip is identical to the safety injection setpoint. S_ team 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 500 psia is sufficiently 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 protection in the event of excessively high steam flow. This setting was used with an uncertainty factor of + 22 psi in the accident analyses. .ll CALVERT CLIFFS - UNIT 1 8 2-5 Amen nt No.

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 provides allowance that there will be sufficient water inventory in the steam generators at the time of trip to provide a margin of more than 13 minutes before auxiliary 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 axiel flux offset is determined from the axially split excore detectors. The trip setpoints ensure that neither a ONBR of less than 1.19 nor a peak linear heat rate which corresponds to the temperature for fuel centerline melting will exist as a consequence of axial pcwer maldistributions. 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 axia'. flux offset relationship. Thermal Margin / Low Pressure The Thermal Margin / Low Pressure trip is provided to prevent operation when the DNBR is less than 1.19. The trip is initiated whenever the reactor coolant system pressure signal drops below either 1750 psia or a computed value as described The computed value is a function of the below, whichever is higher. higher of ST power cr neutron power, reactor inlet temperature, ar.d the The minimum value of reactor number of reactor coolant pumps operating. 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-Finally, the dance with Specifications 3.1.3.5 and 3.1.3.6 is assumed. maximum insertion of CEA banks which can occur during any anticipated operational occurrt:,ve prior to a Power Level-High trip is assumed. Amendment No. 33,39 CALVERT CLIFFS - UNIT 1 B 2-6 483 D9

LIMITfNG SAFETY SYSTEM SETTINGS BASES The Therma'. Margin / Low P. essure trip setpoints are derived from the core safety limits through application of appropriate allowances for equipment response time, measurement uncertainties and procer, sing error. an allowance of 5% of A safety margir, is provided which includes: RATED THERMAL POWER to compensate for potential power measurement error; an allowance of 2 F to compensate for potential temperature measurement l uncertainty; and a f.rti.ar allowance of 84 psia to compensate for pressure ;neasurement errc r, trip system processing error, and time delay associated with providinc, effective termination of the occurrence ; hat The 84 I exhibits the most rapid cecrease in margin to the safety limit. psia allowance is made up of a 22 psia pressure measurement allowance l end a 62 psia time delay allowance. L_oss of Turbine A Loss of Turbine trip Lauses a direct reactor trip when operating This trip provides turbine protection, above 15% of RATED THERMAL POWER. reduces the severity of the ensuing tri.nsient and helps avoid the lifting thus of the main steam line safety valves during the ensuing transient, No credit was taken in the extending the service life of these valves. Its functional capability accident analyses for operation of this trip. 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 Its functional capability at the specified trip operation of this trip. setting is required to enhance the overall reliability of the Reactor Protection System. Amendment No. 27, 32,i3 B 2-7 CALVERT CLIFFS - UNIT 1 48:

280,

REACTIVITY CONTROL SYSTEMS CEA DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full length (shutdown and control) CEA drop time, from a fully withdrawn position, shall be 1 3.1 seconds from when the l electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90 percent insertion position with: avg _ 515 F, and e. T b. All reactor coolant pumps operating. APPLICABILITY: MODES 1 and 2. ACTION: With the drop time of any full length CEA determined to exceed a. the above limit, restore the CEA drop time to within the above limit prior to proceeding to MODE 1 or 2. b. Witn the CEA drop times within limits but detennined at less than full reactor coolant flow, operation may pr oceed provided THERMAL POWER is restricted to less than o equal to the. maximum THERMAL POWER level allowable for t~.e reactor coolant pump combination operating at the time of CEA drop time detennination. SURVEILLANCE REQUIREMENTS 4.1.3.4 The CEA drop time of full length CEAs shall be demonstrated through measurement prior to reactor criticality: For all CEAs following each removal of the reactor vessel head, a. For specifically affected individual CEAs folicwing any main-b. tenance on or modification to the CEA drive system which could affect the drop time of those specific CEAs, and At least once per 18 months. c. CALVERT CLIFFS-UNIT 1 3/4 1-23 /cendment No. 32, 3 9 ~ 483 281

REACTIVITY CONTROL SYSTEMS SHUTDOWN CEA INSERTION LIMIT LIMITING CONDITION FOR OPERATION l All shutdown CEAs shall be withdrawn to at least 129.0 inches. 3.1.3.5 APPLICABILITY: MODES 1 and 2*#. ACTION: With a maximum of one shutdown CEA withdrawn, except for surveillance I testing pursuar.t to Specification 4.1.3.1.2, to less than 129.0 inches, within one hour either: l Withdraw the CEA to at least 129.0 inches, or a. Declare the CEA inoperable and apply Specification 3.'. 3.1. b. SURVEILLANCE REQUIREMENTS Each shutdown CEA shall be determined to be withdrawn to at I 4.1.3.5 least 129.0 inches: Within 15 minutes prior to withdrawal of any CEAs in regulat-ing groups during an approach to reactor criticality, and a. At least once per l'2 hours thereafter. b. See Special Test Exception 3.10.2.

1. With Keff > 1.0.

CALVERT CLIFFS-UNIT 1 3/4 1-24 Amendment No. 28 483 282

3/4.2 POWER DISTRIBUTION LIMITS LINEAR HEAT RATE LIMITING CONDITION FOR OPERATION 3.2.1 The linear heat rate shall not exceed the limits shown on Figure 3.2-1. APPLICABILITY: MODE 1. ACTION: 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 STANOBY 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 determ'ined 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: Verifying at least once per 12 hours that the full length CEAs a. 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. 27,33 383 483 c

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) Verifying at least once per 31 days tha. ine AXIAL SHAPE INDEX is c. 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 expressicn: MxN where: 1. M is the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination. 2. N is the maximum alltwable fraction of RATED THERMAL POWER as detemined by the Fly curve of Figure 3.2-3. 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 incere 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 nperation 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 apprcpriately included in the setting of these alarms: 1. Flux peaking augmentation facters as shown in Figure 4.2-1, 2. A measurement-calculational uncertainty factor of 1.070, l 3. An engineering uncertainty factor of 1.03, 4. A linear heat rate uncertainty factor of 1.01 due to axial fuel densification and themal expansion, and S. A THERMAL POWER measurement uncertainty factor of 1.02. l I CALVERT CLIFFS - UNIT 1 3/4 2-2 Amendment No. 27, 25, 32, 37,3 9 [} 8 [. 20k

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_s d POWER DISTRIBUTI0t' LIMITS TOTALPLANARRADIALPEAKINGFACTOR-Ffy LIMITING CONDITION FOR OPERATION T T xY(1+T ), shall be =F 3.2.2 The calculated value of F*Y, defined as F 9 l

• 7 limited to 1 1.660.

APPLICABILITY: MODE 1*. ACTION: T within 6 hours either: l WithF;j>l.660 Reduc 9 THLR4AL POWER to bring the combinat 'n of THERMAL POWER a. and F to within the limits of Figure 3.2. and withdraw 11e full Mngth CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6; or b. Be in a'. least HOT STANDBY. \\ SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. xy(1+T)andFfy Ffy shall be calculated by the expression F F = 4.2.2.2 q shall be detennined to be within its limit at the following intervals: Prior to operation above 70 percent of RATd? THERMAL POWER a. after each fuel loading, b. At least once per 31 days of accumulated operation in MODE 1, and Within four hours if the AZIMUTHAL POWER TILT (T ) is > 0.030. c. q 5ee Special Test Exception 3.10.2. CALVERT CLIFFS - UNIT 1 3/4 2-6 Amendment No. 27, 24, 32, 33 l} 0 ?, kk0

POWER DISTRIBUTION LIMITS SlJRVEILLANCE REQUIREMENTS (Continued) shall be determined each time a calculation of Ffy is required 4.2.2.3 F by using the incore detectors to obtain a power distribution cap with all xy full length CEAs at or above the Long Tern Steady State Insertion Limit for the existing Reactor Coolant Pump combination. This determination shall be limited to core planes between 15% and 85% of full core height inclusive and shall exclude regions influenced by grid effects. T 4.2.2.4 T shall be determined each time a calculation of F*Y is equired q T and the value of T used to determine F, shall be the measured value of T. q q Miendment No. 27, 32 CALVERT CLIFFS - UNIT 1 3/4 2-7 483 pgo -

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r. E:l,is

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3 2 J o e. o. o o 8 4 d c5 d d d W3 Mod 1vnu3H1031VU do NOllovW3 378YM011V CALVERT CLIFT 5 - UMT 1 3/4 2-8 Amendment No. 27,24,32, I 33, 3 9 ... r.. n,. - d h3k m$$NMp k bg ,J L/

POWER DISTRIBUTION LIMITS T TOTAL INTEGRATED RADIAL PEAKIt'G FACTOR - F r LIMITING CONDITION FOR OPERATION The calculated value of F, deff ned as Ff = F (1+T ), shall be 3.2.3 r r q ligiited to < 1.571. l APPLICABILITY: MODE 1*. ACTION: With F > 1.571, within 6 hours either: l r a. Be in at least HOT STANDBY, or b. Reducy THERMAL POWER to bring the combination of TMERMAL POWER and F to within the limits of Figure 3.2-3 and withdraw the full lengtb CEAs to or beyond the Long Tenn Steady State Insertion Limits of Specification 3.1.3.6. The THERMAL POWER limit determined from Figure 3.2-3 shall then be used to establish a revised upper THERFAL 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 -3) and subsequent operation shall be 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. FfshallbecalculatedbytheexpressionF =F,.(1+T)&ndFf T 4.2.3.2 p q shall be detennined 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. ALVERT CLIFFS - UNIT 1 3/4 2-9 Amendment No. 27, 23, 32, 33, 3 1 l} 0 b l]

SURVEILLANCE REQUIREMENTS (Continued) T 4.2.3.3 F shall be determined each time a calculation of F is required 7 r by using the incore detectors to obtain a power distribution map with all full length CEAs at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump combination. T 4.2.3.4 T shall be determined each time a calculation of F is requir,ed r and the value of T used to determine Ff shall be the measured value of T. q q q CALVERT CLIFFS - UNIT 1 3/4 2-10 Amendment No. 27, 32 483 .292

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l REGION

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l i + {. !' l i -i l t-i i l -0.6 -0.4 -0.2 0 0.2 0.4 0.6 PERIPHER AL AX1AL SHAPE INDEX (Y ) g FIGURE 3.2-4 DNB Axial Flux Offset Control Limits CALVERT CLIFFS - UNIT 1 3/4 2-11 Amendment No. 27, 2f, 37, 33, 3 9 48} 2_ f) 3 e a

POWER DISTRIBUTION LIMITS AZIMUTHAL POWER TILT - Tq LIMITING CONDITION FOR OPERATIO_"_ 3.2.4 The AZIMUTHAL POWER TILT (T ) shall not exceed 0.030. q APPLICABILITY: MODE 1 above 50% c' RATED THERMAL POWER.* ACTION: With the indicated AZIMUTHAL POWER TILT determined to be > 0.030 l a. but < 0.10, either correct tM power tilt within two hours or dete7mine within the next 2 hours and at least once per subse-quent 8 hcurs, that the TOTAL PLANAR RADIAL PEAKDG FACTOR (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 AZ' < THAL POWER TILT 4 termined to be > 0.10, operation may proceeg. x up to 2 hours provided that the TOTAL INTEGRATED RADIAL PEAKINQ FACTOR (Ff) and TOTAL PLAN PEAKING FACTOR (F ) are within the limitsi 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 FlWER % vel is restricted ta < 20% of the maximum allcwable IHERMAL POWER level for the existing Reactor Coolant Pump combination. SURVEILLANCE REQUIREMENT 4.2.4.1 The provisions of Specification 4.0.4 are not applicable. 4.2.4.2 The AZIMUTHAL POWER TILT shall be determined to be within the limit by: Calculating the tilt at least once per 12 hours, and a. b. Using the incore detectors to determine the AZIMUTHAL F0WER TILT at least once per 12 hours when ena excore chann'el is inoperable and THERMAL POWER IS > 75% of RATED THERMA:. POWER.

• See Special Test Exception 3.10.2.

CALVERT CLIFFS - UNIT 1 3/4 2-12 Amendment No. 2J, 32 4 fj[ 29k

.a POWER DISTRIBUTION LIMITS FUEL RESIDENCE TIME LIMITING CONDITION FOR__0PERATION 3.2.5 This specification deleted. {JRVEILLANCE REQUIREMENTS 4.2.5 This specification deleted. 5, f3 '5 2)b CALVERT CLIFFS - UNIT 1 3/4 2-13 Amendment No. 27, 37, 3 9

POUER DISTRIBUTION LIf1ITS DGB PARAf1ETERS LIillTING CONDITION FOR OPERATION I 3.2.6 The following DNB related parameters shall be maintained within the limits shown on Table 2.2-1: a. Cold Leg Temperature b. Pressurizer Pressure Reactor Coolant System Total Flow Rate c. d. AXIAL SHAPE INDEX APPLICACILITY: 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 4.2.6.1 Each of the parwr.. s of Table 3.2-1 shall be verified to be I within their limits at least once per 12 hours. I 4.2.6.2 The Reactor Coolant ' stem total flow rate shall be determined to be within its limit by measu ement at least once per 18 months. CALVERT CLIFFS - UNIT 1 3/4 2-14 Arendment .h. 21 I k O[. h2h b

( TABLE 3.2-1 52 t-DNB PARAMETERS Q} w P LIMITS Th I Four Reactor Three Reactor Two Reactor Two Reactor Coolant Pumps Coolant Pumps Co9lant Pumps Coolant Pumps = Parameter Operating Operating Operating-Same Loop Operating-Opposite Loop c: Cold Leg Te.operature < 548 F Pressurizer Pressure > 2225 psia

• Reactor Coolant System Total Flow Rate

> 370,000 gpm AXIAL SHAPE INDEX Figure 3.2-4 p> 5; E ~ 4s, a [f$ ~* Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% of RATED THERMAL POWER ] per udsote or a THERMAL POWER step increase of greater than 10% of RATED THERMAL POWER. P

• • Trese value; lef t blank pending NRC approval of ECCS analyses for operation with less than four reactor coolant pumps operating.

(jj 73 sj IU Cc

TABLE 3.3-1 (Continued) ACTION STATL"_NTS 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. ACTICN 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 I hour and at least once per 12 hours thereafter. With the number of channels OPERABLE cne less than required ACTION 4 by the Minimum Channels OPERABLE requirement, be in H0T STANDBY within 6 hours; however, one channel may be bypassed for up to I hour for surveillance testing per Specification 4.3.1.1. CALVERT Cl.IFFS - UNIT 1 3/4 3-5 483 ?98

n TABLE 3.3-2 M REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES A FUNCTIONAL UNIT RESPONSE TIME 1. Manual Reactor Trip Not Applicable 2. Power Level - liigh < 0.40 seconds *# and < 8.0 seconds ## l 3. Reactor Coolant Flow - Low 1 0.50 seconds 4. Pressurizer Pressure - liigh 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 8. Axial Flux Offset 1 0.40 seconds *# and 1 8.0 seconds ## 9. Thermal Margin / Low Pressure < 0.90 seconds *# and < 8.0 secondsi#

10. Loss of Turbine--Ilydraulic Fluid Pressure - Low Not Applicable y

11. Wide Range Logarithmic Neutron Flux Monitor Not Applicable

$n

• Neutron detectors are exempt from response time testing.

Pesponse time of the neutron flux signal portion

c P

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

1. Response time does not include contribution of RTDs.

w e,-a

1. 1. RTD response time only. This value is equivalent to the time interval required for the RTDs output to achieve 63.2% of its total change when subjected to a step change in RTD temperature.

w N ~D ~C

l3/4.2 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 er the fuel cladding will not exceed 2200 F. Either of the two core pcwer distribution monitoring systems, the Excore Detector Monitcring System and the Incore Detector Monitoring System, provide adeqi. ate monitoring of the core powcr distributicn and are capable of verifying that the linear heat rate does not exceed its limits. The Excore Detector Monitoring System perfoms this functicn by continu-ously monitoring the AXIAL SHAPE INDEX with the OPERABLE quadrant symmetric excore neutron flux deteccors and vertfying 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 peakinj augmentation factors are as shown in Figure 4.2-1, 3l the AZIMUTHAL POWER TILT restrictions of Specification 3.2.4 are satisiied, 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 of 1.03, 4) an allowance of 1.01 for axial fuel densification and themal 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 T FACTORS - F AND F AND AZIMUTHAL POWER TILT - T 7 q T tions used in the analysis # and Tfores2ablishingthelinearHeatRateandLocal The limitations on F are provided to ensure that the assump-Power Density - High LCOs and LSSS setpoints remain valid during operation ay the various allowable CEA group insertion limits. The limitations on and T are provided to ensure that the assumptions used in r q

ALVERT CLIFFS - UNIT 1 B 3/4 2-1 Amendment No. 27, 32, 33, 3 9 483.300

POWER DISTRIBUTION LIMITS BASES the analysis establishing the DNB Margin LCO, and Thermal Margin / Low Pressure LSSS setpoints remain valid duringT per9 tion at the various o If F F or T, exceed their allowable CEA group insertion limits. basic limitations, operation may continue uMe,r [he Gtional restric-tions imposed by the ACTION statements since these additional restric-tions provide adequate provisions to assure that the assumptions used in establishing the Linear H<,, 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

1. 1. (1 + 1 9) he value of T that must be used in the equation F*Y =F andF}r"Fr (1+T ) is the measured tilt.

q T The surveillance requirements for verifying tc.st F F and T aye T within their limits provide assurance that the actual vNu,es 9fF 3F T r and T do not exceed the assumed values. Verifying F and F afNr each 9uel loading prior to exceeding 75% of RATED THENL POWER provides additional assurance that the core was properly loaded. 3/4.2.4 FUEL R:SIDENCE TIME _ This specification deleted. 3/4.2.5 DNB PARAMETERS The limits on the DN8 related parameters assure that each of the parameters are maintained within the nonnal steady state envelope of The limits are operation assumed in the transient and accident analyses. consistent with the safety analyses assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of ;.19 throughout each cnalyzed transient. The 12 hour periodic surveillarw of these parameters through instru-ment readout is sufficient to ensure t1at the parameters are restored within their limits following load chanyes and other expected transient The 18 month periodic n'easurement of the RCS total flow rate operation. is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated d flow rate on a percent flow will provide sufficient verification o 12 hour basis. 30 CALVERT CLIFFS - UNIT 1 8 3/4 2-2 Amendment No. 21, 32, )), 483 501 .}}