Text

Amend 31 to License DPR-69,authorizing Cycle 4 Operation W/ Increase in Fuel Enrichment,Modified Guide Tubes for Control Element Assemblies & Use of New Reactor Protection Sys Trip
	ML19341C319
Person / Time
Site:	Calvert Cliffs
Issue date:	02/10/1981
From:	Clark R; Office of Nuclear Reactor Regulation
To:	;
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ML19341C320	List: ML19341C319; ML19341C321; ML19341C322;
References
	NUDOCS 8103020732
	Download: ML19341C319 (85)
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{{#Wiki_filter:. l e o,, UNITED STATES NUCLEAR REGULATORY COMMISSION o g .a WASHINGTON, D. C. 20555 r ,W...../ BALTIMORE GAS AND ELECTRIC COMPANY DOCKET NO. 50-318 CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NO. 2 -AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 31 License No. DPR-69 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The applications for amendment by Baltimore Gas & Electric Company (the licensee) dated December 4 and 12,1980 and January 29, 1981, as supplemented, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the 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 satisified. 810s ou 73%

e o. 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the a'ttachment to this license amendment, and paragraph 2.C.2 of Facility Operating License No. DPR-69 is hereby amended to read as follows: 2. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 31, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of the date of its issuance. FOR THE NUCLEAR REGULATORY COMMISSION x Robert A. Clark, Chief Operating Reactors Branct #3 Division of Licensing

Attachment:

Changes to ths Technical Specifications Date of Issuance: February 10, 1981 L l' f G

r ATTACHMENT TO LICENSE AMENDMENT NO. 31 FACILITY OPERATING LICENSE NO. DPR-69 DOCKET NO. 50-318 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 maintain document contpleteness. - Page 2-2 3/4 2-5 3/4 9-1 2-9 3/4 2-6 3/4 9-8 2-10 3/4 2-8 3/4 10-1 2-11 3/4 2-9 B 3/4 1-1 2-12 3/4 2-11 B 3/4 1-2 B 2-1 3/4 3-2 B 3/4 1-3 B 2-3 3/4 3-4 B 3/4 2-2 B 2-5 3/4 3-6 B 3/4 9-1 B 2-6 3/4 3-7 5-4 8 2-7 3/4 3-15 3/4 1-1 3/4 3-17 3/4 1-3 3/4 3-20 3/4 1-5 3/4 4-2 3/4 1-9 3/45-1 3/4 1-11 3/4 5-2 3/4 1-13 3/4 5-5 3/4 1-14 3/4 5-7 3/4 1-15 3/4 7-45 3/4 1-16 3/4 7-46 3/4 2-4 3/4 7-53 0 6 6 1 1

e 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i 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 l of two, three and four reactor coolant pump operation. APPLICABILITY: MODES 1 and 2. ACTION: Whenever the point defined by the combination of the highest operating loop cold leg temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure 1ine, be in HOT STANDBY within 1 hour. REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia. APPLICABILITY: MODES 1, 2, 3, 4 and 5. ACTION: MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2750 psia, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour. MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to within its limit within 5 minutes. i CALVERT CLIFFS-UNIT 2 2-1 Amendment No. 6 ~ i

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-i FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE' VALUES 4. Pressurizer Pressure - High _ 2400 pe.ia 1 2400 psia E 5. Containment Pressure - High 4 psig 1 4 psig ro 6. Steam Generator Pressure. Low (2) 570 psia _1 570 psia l 7. Steam Generator Water Level - Low - 10 inches below top > 10 inches below top of feed ring, of feed ring. 8. Axial flux of fset (3) Trip setpoint adjusted to Trip setpoint adjusted to not exceed the limit lines not axceed the limit lines o f figure 2.2-1. of Figure 2.2-1. ? 9. Thermal Margin / Low Pressure (1) a. Four Reactor Coolant Pumps Trip setpoint adjusted to Trip setpoint adjusted to Operating not exceed the limit lines not exceed the limit lines of figures 2.2-2 and 2.2-3. of Figures 2.2-2 and 2.2-3. b., Steam Generator Pressure ~ 135 psid ~< 135 psid i Di f ference - High (l') F 10. Loss of Turbine -- Hydraulic _ 1100 psig 1 1100 psig Fluid Pressure - Low (3) 11. Rate of Change of Power - High (4) 1 2.6 decades per minute 1 2.6 decades per minute g E( TABLE NOTATION -4 3 (1 ) Tripmaybebypassedbe}ow10 % of RATED TilERMAL POWER; bypass shall be automatically removed where TilERMAL POWER is 1 10 % of RATED TilERiiAL POWER. =

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~ 9 s i 2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel-is. prevented by maintaining the steady state peak linear heat rate at or less than 21 kw/ft. Centerline fuel melting will not occur for this peak linear heat rate. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer. coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. ~ Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temper-ature and Pressure have been related to DNB through the CE-1 correlation. The CE-1 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distri-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 1.195 l This value corresponds to a 95 percent probability at a 95 percent con-fidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions. The curves of Figures 2.1-1, 2.1-2, 2.1-3 and 2.1-4 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperature of various pump combinations for which the minimum DNBR is no less than 1.195 for the family of axial shapes and l corresponding radial peaks shcwn in Figure B2.1-1. The limits in Figures 2.1-1, 2,1-2, 2.1-3 and.2.1-4 w,ere calculated for reactor coolant inlet temperatures less than or equal to 580 F. The dashed line at 580'F coolant inlet temperature is not a safety limit; however, operation above 580 F is not possible because of the actuation of the main steam line. safety valves which limit the maximum value of reactor inlet temperature. Reactor operation at THERMAL POWER levels higher than 112% of RATED THERMAL j POWER is prohibited by the high power level trip setpoint specified in CALVERT CLIFFS - UNIT 2 B 2-1 Amendment No.78, 31 b

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SAFETY LIMITS BASES Table 2.1-1. The area of safe operation is below and to the left of these lines. The conditions for the Thermal Margin Safety Limit curves in Figures 2.1-1, 2.1-2, 2.1-3 and 2.1-4 to be valid are shown on the figures. The reactor protective system in combination with the Limiting Conditions for Operation, is designed to prevent any anticipated combina-tion of transient conditions for reactor coolant system temperature, pressure, and THERMAL POWER level that would result in a DNBR of less than 1.195 and preclude the existence of flow instabilities. l 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. l i l I CALVERT CLIFFS - UNIT 2 B 2-3 Amendment No. 78,31 l

I 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2. 2.1 REACTOR TRIP SETPOINTS The Rcontor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Operation with a trip set less conservative than its Trip-Setpoint but within its speci-fied Allowable Value is acceptable on the basis that each Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses. Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability. 4 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 /Luw Pressure trip. The Power Level-High trip setpoint is operator adjust'able 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 rip setpoint is automatically decreased as THERMAL power decreases. The trip setpoint 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 TH,ERMAL 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 DNB in the event of a sudden significant decrease in reactor coolant flow. Provisions have been made in the reactor protective system to permit i CALVERT CLIFFS-UNIT 2 B 2 -4 Amendment No. 18

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 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 1.195 under normal operation I and expected transients. For reactor operation with only two or three reactor coolant pumps operating, the Reactor Coolant Flow-Low trip set-points, the Power Level-High trip setpoints, and the Thermal Margin / Low Pressure trip setpoints are automatically changed when the pump condition selector switch is manually set to the desired two-or three-pump p'osi tion. Changing these trip setpoints during two and three pump operation prevents the minimum value of DNBR from going below 1.195 during I normal operational transients and anticipated transients when only two or three reactor coolant pumps are operating. Pressurizer Pressure-High The Pressurizer Pressure-High trip, backed up by the pressurizer cede i safety valves and main steam line safety valves, provides reactor coolant system protection against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is 100 psi below the nominal lift setting (2500 psia) of the pressurizer code safety valves and its concurrent operation with the power-operated relief valves avoids the undesirable operation of the pressurizer code safety valves. Eontainment Pressure-High The Containment Pressure-High trip provides assurance that a reactor trip is initiated concurrently with a safety injection. The setpoint for this trip is identical to the safety injection setpoint. Steam Generator Pressure-Low The Steam Generator Pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the reactor coolant. The setting of 570 psia l is sufficiently below the full-load operating point of 850 psia so as not to interfere with normal operation, but still high enough to l 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. CALVERT CLIFFS - UNIT 2 B 2-5 Amendment No. Jp,31 e

LIMITING SAFETY SYSTEM SETTINGS BASES Steam Generator Water Level The Steam Generator Water Level-Low trip provides core protection t/ 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 axial flux offset is determined from the axially split excore detectors. The trip setpoints ensure that neither a DNBR of less than 1.195 nor a peak linear heat rate which corresponds to the temperature for fuel centerline melting will exist as a consequence of axial power maldistributions. These trip set-points were derived from an analysis of many axial power shapes with allowances for instrumentation inaccuracies and the uncertainty associated with the excore to incore axial flux offset relationship. Thermal Margin / Low Pressure The Thermal Margin / Low Pressure trip is provided to prevent operation when the DNBR is less than 1.195 l The trip is initiated whenever the reactor coolant system pressure signal drops below either 1750 psia or a computed value as described below, whichever is higher. The computed value is a function of the higher of AT power or neutron power, reactor inlet 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 assume 4 in the genera-tion of this trip function. In addition, CEA group sequencing in accor-dance with Specifications 3.1.3.5 and 3.1.3.6 is assumed. Finally, the maximum insertion of CEA banks whic'n can occur during any anticipated operational occurrence prior to a Power Level-High trip is assumed. CALVERT CLIFFS - UNIT 2 B 2-6 Amendment No.Jg,31 ~' +

e LIMITING SAFETY SYSTEM SETTINGS BASES The Thermal Margin / Low Pressure. trip setpoints are derived from the core safety limits through application of appropriate allowances for equipment response time, measurement ur certainties and processing error. A safety margin is provided which includes: an allowance of 5% of RATED THERMAL POWER to compensate for potential power measurement error; an allowance of 2 F to compensate for potential temperature measurement uncertainty; and a further allowa..r: of 92 psia to compensate for l pressure measurement error, trip system processing trror, and time delay ussociated with providing effective termination of the occurrence that exhibits the most rapid decrease in margin to the safety limit. The 92 l . psia allowance is made up of a 22 psia pressure c.easurement allowance and a 70 psia time delay allowance. 1 Asymmetric Steam Generator Transient Protection Trip Function ( ASGTPTF) The ASGTPTF utilizes steam generator pressure inputs to the TM/LP calculator, which causes a reactor trip when the difference in pressure between the two steam generators exceeds the trip setpoint. The ASGTPTF is designed to provide a reactor trip for those Anticipated Operational Occurrences associated with secondary system malfunctions which result iri asymmetric primary loop coolant temperatures. The most limiting event is the loss of load to one steam generator caused by a single Main Steam Isolation Valve closure. The equipment trip setpoint and allowable values are calculated to account for instrument uncertainties, and will ensure a trip at or before reaching the 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 Protection System. Rate of Change of Power-High The Rate of Change of Power-High trip is provided to protect the core during startup operations and its use serves as a backup to the administra-tively enforced startup rate limit. Its trip setpoint does not correspond to a Safety Limit and no credit was taken in the accident analyses for operation of this trip. Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System. CALVERT CLIFFS - UNIT 2 B.2-7 Amendment No.9, 78, 31 4

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL SHUTDOWN MARGIN - T,yg > 200*F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be 1 4.3%* ak/k. APPLICABILITY: MODES 1, 2**, 3 and 4. ACTION: With the SHUTDOWN MARGIN < 4.3%* ak/k, immediately initiate and continue boration at 1 40 gpm of 2300 ppm boric acid solution or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.l.1.1.1 The SHUTDOWN MARGIN shall be determined to be 3 4.3%* 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 SHUTDOWN 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 withdrawal is within the Transient Insertion Limits of Specification 3.1.3.6. When in MODE 2 ', within 4 hours prior to achieving reactor c. 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. i Adherence to Technical Specification 3.1.3.6 as specified in Surveillance Requirements 4.1.1.1.1 assures that there is sufficient available shut-down margin to match the shutdown margin requirements of the safety analyses.

- See Special Test Exception 3.10.1.

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With K 1 1.0.

eff 7* With K < l.0. eff CALVERT CLIFFS - UNIT 2 3/4 1-1 Amendment No. 9, JS,31 l

4 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 thernal energy generation, 5. Xenon concentration, and 4 Samarium concentration.: 6. i 4.1.1.1.2 The overall core reactivity balance shall be compared to p'redicted 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 (nornalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading. 4 4 i l CALVERT CLIFFS-UNIT 2 3/4 1-2 9 --o = - - y-- -,. ~ -e --w_,. ~ -. -

REACTIVITY CONTROL SYSTEMS 0 SHUTDOWN MARGIN - T,yg 1 200 F LIMITIrlG CONDITION FOR OPERATION 3.1.1. 2 The SHUTDOWN MARGIN shall be > 3.0%' Ak/k. l APPLICABILITY: MODE 5 a. Pressurizer level > 90 inches from bottom of the pressurizer. b. Pressurizer level < 90 inches from bottom of the pressurizer and all sources of non-borated water 1 88 gpm. a ACTION: With the SHUTDOWN MARGIN < 3.0% Ak/k, immediately initiate and continue a. boration at > 40 gpm of 2300 ppm bo,ic acid solution or equivalent until the required SHUTDOWN MARGIN is restored. b. With the pressurizer drained'to < 90 inches and all sources of non-borated water > 88 gpm, immediately suspend all operations involvi'ng positive reactivity changes while the SHUTDOWN MARGIN is increased to compensate for the additional sources of non-borated water or reduce the sources of non-borated water to 1 88 gpm. SURVEILLAWCE REQUIREMENTS .i l4.1.1.2 The SHUTDOWN MARGIN shall be determined to be > 3.0% Ak/k: l ll a. Within one hour after detection of an inoperable CEA(s) and at j least once per 12 hours thereafter while the CEA(s) is inoperable. If the inoperable CEA is immovable or untrippable, the above required SHUTDOWN tiARGIN shall be increased by an amount at least i! equal to the withdrawn worth of the immovable or untrippable CEA(s). b. At least nnte per 24 hours by consideration 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. 4.1.1.2.2. With the piessurizer drained to 1 90 inches determine: a. Within one hour and every 12 hours thereafter that the level in the reactor coolant system is above the bottom of the hot leg nozzles, and I i b. Within one hour and ~every 12 hours thereafter that the sources l of non-borated water are 1 88 gpm or the shutdown margin has compensated for the additional sources. CALVERT CLIFFS' - UNIT 2 3/4 1-3 Amendment No. 31

REACTIVITY CONTROL SYSTEMS BORON DILUTION LIMITING CONDITION FOR OPERATION 3.1.1.3 The flow rate of reactor coolant through the reactor coolant system shall be 3,3000 gpm whenever a reduction in Reactor Coolant System boron concentration is being made. APPLICABILITY: ALL MODES. ACTION: With the flow rate of' reactor coolant through the reactor coolant system < 3000 gpm, i mediately suspend all ope ~ rations involving a reduction in boron cor -tration of the Reactor Coolant System. SUT.'EILLANCE REQUIREMENTS 4.1.1.3 The flow rate of reactor coolant through the reactor coolant system shall'be determined to be 2,3000 gpm within one hour prior to the start of and at least once per hour during a reduction in the Reactor Coolant System boron concentration by either: a. Verifying at least one reactor coolant pump is in operation, or b. Verifying that at least one low pressure safety injection pump is in operation and supplying y_3000 gpm through the reactor coolant system. i l i CALVERT CLIFFS-UNIT 2 3/4 1-4

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

With Keff 1

See Special Test Exception 3.10.2.

i CALVERT CLIFFS - UNIT 2 3/4 1-5 Amendment No. 7$, 31

L REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.1.1.4.2 The MTC shall be determined at the following frequencies and THERMAL POWER conditions during each fuel cycle: a. Prior.to initial operation above 5% of RATED THERMAL POWER, after each fuel loading. b. At any THERMAL' POWER, within 7 EFPD after reaching a RATED THERMAL POWER equilibrium boron concentration of 900 ppm. c. At any THERMAL POWER, within 7 EFPD after reaching a RATED THERMAL POWER equilibrium boron concentration of 300 ppm. 3 O CALVERT CLIFFS-UNIT 2 3/4 1-6 M

REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION ~FOR OPERATION 3.1.2.2 At least two of the following tnree boron injection flow paths and one associated heat tracing circuit shall be OPERABLE: a. Two flow paths from the boric acid storage tanks via either a boric acid pump or a gravity feed connection, and a charging pump to the Reactor Coolant System, and 4 b. The flow path from the refueling water tank via a charging pump to the Reactor Coolant System. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 3; ak/k at 200*F within the next 6 hours; restore at l least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REQUIREMENTS 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE: a. At least once per 7 days by verifying that the temperature of the heat traced portion of the flow path from the concentrated boric acid tanks'is above the temperature limit line shown on Figure 3.1-1. b. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position. c. At least once per 18 months during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on a SIAS test signal. CALVERT CLIFFS - UNIT 2 3/4 1-9 Amendment No. 31

REACTIVITY CONTROL SYSTEMS . CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 At least one charging pump or one high pressure safety injection pump in the boron injection flow path required OPERABLE pursuant to Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency bus. APPLICABILITY: MODES 5 and 6. ACTION: With no charging pump or high pressure safety injection pump OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one of the required pumps is restored to OPERABLE status. SURVEILLANCE

REQUIREMENTS 4.1.2.3 No additional Surveillance Requirements other than those required by Specification 4.0.5.

i l t e CALVERT CLIFFS-UNIT 2 3/4 1-10

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING e i LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE. APPLICABILITY: ' MODES 1, 2, 3 and 4. ACTION: With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated.to a SHUTDOWN MARGIN equivalent to at least 3 % ak/k at l 200*F within the next 6 hours; restore at least two charging pumps to OPERABLE status within the next'7 days or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REQUIREMENTS 4.1.2.4 No additional Surveillance Requirements other than those required by Specification 4.0.5. a e i l CALVERT CLIFFS - UNIT 2 3/4 1-11 Amendment No. 31 e t-rw- = ~x a

REACTIVITY CONTROL SYSTEMS BORIC ACID PUMPS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 At least one boric acid pump shall be OPERABLE and capable of being powered from an OPERABLE emergency. bus if only the flow path through the boric acid pump in Specification 3.1.2.la above, is OPERABLE. APPLICABILITY: MODES 5 and 6. ACTION: With no boric acid pump OPERABLE as required to complete the flow path of Specification 3.1.2.la, suspend all operations involving CORE ALTERA-TIONS or positive reactivity changes until at least one boric acid pump is restored to OPERABLE status. SURVEILLANCE REQUIREMENTS 4.1.2.5 No additiona1' Surveillance Requirements other than those required by Specification 4.0.5. t CALVERT CLIFFS-UNIT 2 3/4 1-12 l l --=,r e-

I REACTIVITY CONTROL SYSTEMS, BORIC ACID PUMPS - OPERATING LIMITING CONDITION FOR OPERATION ^ 3.1.2.6 At least the boric acid pump (s) in the boron injection flow path (s) required OPERABLE pursuant to Specification 3.1.2.2a shall be 0PERABLE and capable of being powered from an OPERABLE emergency bus if the flow path through the boric acid pump (s) in Specification 3.1.2.2a is OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one boric acid pump required for the boron injection flow path (s) pursuant to Specification 3.1.2.2a inoperable, restore the boric acid pump to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and borated to a SHUTDOWN MARGIN equivalent to at least 3% ak/k at 200*F; restore the above required boric acid pump (S) l to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REQUIREMENTS d i t 4.1.2.6 No additional Surveillance Requirements other than those required by Specification 4.0.5. F i l ~ CALVERT CLIFFS - UNIT 2 3/4 1-13 Amendment No. 31 L

t REACTIVITY CONTROL SYSTEMS B0 RATED WATER SOURCES - SHUTDOWN I LIMITING CONDITION FOR OPERATION 3.1.2.7 As a minimum, one of the following borated water sources shall be OPERABLE: a. One boric acid storage tank and one associated heat tracing circuit with the tank contents in accordance with Figure 3.1-1. b. The refueling water tank with: 1. A minimum contained bor'ated water volume of 9,844 gallons, l 2. A minimum boron concentration of 2300 ppm, and l 3. A minimum solution temperature of 35"F. ~ APPLICABILITY: MODES 5 and 6. AC :0": With no borated water sources OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one borated water source is restored to OPERABLE status. j SURVEILLANCE REQUIREMENTS 4.1 2.7 The above required borated water source shall be demonstrated OPERABLE: 3 a. At least once per 7 days by: i 1. Verifying the boron concentration of the water,. 2. Verifying the contained borated water volume of the tank, and 3. Verifying the boric acid storage tank solution temperature when it is the source of borated water. b. At least once per 24 hours by verifying the RWT temperature when it is the source of borated water and the outside air temperature is < 35'F. 4 CALVERT CLIFFS - UNIT 2 3/4 1-14 Amendment No.5,31 w m- .m y ,w

e e 170 ..._..f._*.I*;.~.'.

l-

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= =1Er=

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0 t.._.._ i._.... l. _g .,. }....,.. ..L... ~...... 6 7 8 9 10 11 12 STORED BORIC ACID CONCENTRATION (WT%) { FIGURE 3.11 1 Minimum Boric Acid Storage Tank Volume and Temperature as a Function of Stored Boric Acid Concentration CALVERT CLIFFS - UNIT 2 3/4 1-15 Amendment Noe 6. 3r

_R_EACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.8 At least two of the following three~ borated water sources shall 'I be OPERABLE: 'a. Two boric acid storage tank (s) and one associated heat tracing circuit per tank with the contents of the tanks in accordance with Figure 3.1-1 and the boron concentration limited to < 8%, and b. The refueling water tank wit.h: 1. A minimum contained borated water volume of 400,000

gallons, 2.

A beron concentration of betwaen 2300 and 2800 ppm, l 3. A minimum solution temperature of 40 F, and 4. A maximum solution temperature of 100 F in MODE 1. APPLICAEILITY: MODES 1, 2. 3 and 4. ACTION: With only one borated water source OPERABLE, restore at least two borated water sources to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and borated to a SHUTDOWN MARGIN equivalent to at least 3',, ak/k at 200 F; restore at least two borated water sources l to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REQUIREMENTS 4.1.2.8 At least two borated water sources shall be demonstrated OPERABLE: a. At least once'per 7 days by: 1. Verifying the boron concentration in each water source, 2. Verifying the contained borated water volume in each water source, and 3. Verifying the boric acid storage tank :,olution temperature. b. At least once per 24 hourt by verifying the RWT temperature when the outside air temperature is < 40 F. l CALVERT CLIFFS - UNIT 2 3/4 1-16 Amendment Nc. 6, 8, II,31

8 b V C 3 C M 3 2 o d o o a >- -d N g m m m m m I L ~ O O y @ m m = m o m P % N r C C m m m oF H y m 3e m x 'g-8 9 ? y o m z m r C k E' 8 l G O' I )ROTAREDOM + D ALC -LEUF( TF/WK.ETAR TAEH RAENIL KAEP ELBAWOLLA 3Vl'A381 3lIJdS - ON11 2 C/7 2-E Vwaupuauq NO'g' L9' i

1.0 ; . _ 7... UNACCEPTA BLE'.f.. ~J,. .i.. ~2.2._,. :_ UNACCEPTA BLE l-OP ERATION ='4 27f = ~9" F ~ W - OPERATION REGION !=:MifaM# - REGION i = { j n.r:.. .n j t

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i: s.i==si+ o i =:

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O.5 "= ACCEPTABLE ~ .. - ::= = 2: -= opgggylon i~l + =~- JM5BEr REGION f =._. 2. i..Ji.E=? ._:.... c u.

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r- : " '_.._:r - " :n - ;_ _ R = - = _ - - rar--1:.9 .- f r- - - _i__ E:EEi El[(h,[*~1.011j9f-MESI E:##i~t sin ' Et 9 E M i 9 : W 2 1.01 _. _.. _ h_. _..... f~..=4_,- - :.r. = : =-- - : = _. ---.l ---. . r:z ;_ :- - -3==== t==-

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.: =

.-r 2 _. ...q=.,... 4-- . _.=t =; ' .= i_2__5._=~ 1... _=._i==. _+ =_i..= L_ - "_=l:.=..fE::.q:. f 5-Ll=r_i. il:u. = = E }:E=.

== =2..'._a=..u ! _.:z. E"f=..=._=. .. _.-.. _.... J =.u.. : : :=...____ n._.

== 1.00 ~ ~ ~ ~ . p, =}isi.= ~.. - =ij=EE p-.; :- i; r= =iz !-~ wia."=:4:---- . (2.0,1.002) 0 20 40 60 80 100 120 140 CORE HEIGHT, INCHES o FIGURE 4.2-1 ' Augmentation Factor vs Distance from Bottom of Core CALVERT CLIFFS - UNIT 2 3/4.2-5 Amendment No. 9.18, 31 i i j

POWER DISTRIBUTION LIMITS TOTALPLANARRADIALPEAKINGFACTOR-F[y LIMITING CONDITION FOR OPERATION T T xY(1+T ), shall be 3.2.2 The calculated value of F*Y, defined as F =F 9 XY l l limited to 1 1.620. APPLICABILITY: MODE 1*. I j ACTION: } With.F > 1.620, within 6 hours either: l ReducyTHERMALPOWERtobringthecombinationofTHERMALPOWER a. l and F to within the limits of Figure 3.2-3 and withdraw the full Mngth CEAs' to or oeyond the Long Term Steady State Insertion Limits of Specift.ation 3.1.3.6; or b. Be in at least HOT STANDBY. SURVEILLANCE REQUIREMENTS 1 I 4.2.2.1 The provisions of' Specification 4.0.4 are not applicable. Ffy shall be calculated by the expression Ffy = Fxy(1+T)andFfy 1 4.2.2.2 q shall be determined to be within its limit at the following intervals: a. Prior to operation above 70 percent of RATED THERMAL POWER after each fuel loading, b. At least once per 31 days of accumulated operation in MODE 1, and c. Within four Hours if the AZIMUTHAL POWER TILT (T ) is > 0.030. q

See Special Test Exception 3.10.2.

] l CALVtRT CLIFFS - UNIT 2 3/4 2-6 Amendment No.P. 78,31

,m

..,4

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) T 4.2.2.3 F shall be determined each time a calculation of F is required xy by using the incore detectors to obtain a power distribution map with all full length CEAs at or above the Long Term Steady State Insertion L.imit for the existing Reactor Coolant Pump combination. This determina-tion shall be 1imited to core planes between 15% and 85% of full core ~ height inclusive and shall exclude regions. influeaced by grid effects. T shall be determined each time a calculation of F, is required 4.2.2.4 T q and the value of T used to determine F shall be the measured value of q xy q. 4 i e CALVERT CLIFFS-UNIT 2 3/42-7 Amendment No. 9, 18 l

Tr ~ l ll!Il1lll1I11llllllll F i T G I _)%.k l*62, 1 00) UNACCEPTABLE 'O 1.00 I ll-i it-OPERATION n 9 x REGION s !i i 4 J,,*I i'd,,[ ij r- !!ilizittliffittitittiltitfit n k I , Fhy E 2 LIMIT CURVE s, s < 0. 90 -:"s a a ls l [] ,,' ll .s, T 1 3 F LIMIT CURVE ! I;- t ll s s 5 E Y ~ 'il 'Y N's (1.70,.85) ^ H ~ s s O 's is 0.80 N l a,b n j (1.695,.775)j m ACCEPTABLE O ..;i OPERATION 0.70 REGION p a, M l' em 3 l i; . j! l!! ll q; j Q 0.60 i .l- !, l'.i n! j!l !l I

'l l!

3 jl! 3 [: i

l l

lj, [ l j i I p' il it I u I;i i i i i t i. i l 11! a l i l 0.50 l. I I II l l i il!l l l i i i i ! ![!Illi lll l-I i A 1.56 1.58 1.60 1.62 1.64 1.66 1.68 1.70 1.72 ~ F T T IXY f b 4 r 5 Figure 3.2-3 g TOTAL RADIAL PEAKIt1G FACTORS vs ALL0HABLE FRACTION OF RATED TilERMAL POWER b

v POWER DISTRIBUTION LIMITS T TOTAL INTEGRATED RADIAL PEAKING FACTOR - Fr LIMITING CONDITION FOR OPERATION a 3.2.3 The calculated value of F, defined as FT = F (1+T ), shall be 9 limited to < 1.620. l APPLICABILITY: MODE 1*. ACTION: With F > 1.620, within 6 hours either: g 7 a. Be in at least HOT STANDBY, or b. Reducy THERMAL POWER to. bring the combination of THERMAL POWER and F to within the limits of Figurc 3.2-3 and withdraw the full lengtb CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6. The THERMAL POWER limit i determined from Figure 3.2-3 shall then be used to establish a revised upper THERMAL POWER level limit on Figure 3.2-4 (truncate Figure 3.2-4 at the allowable fraction of RATED THERMAL POWER determined by Figure 3.2-3) and subsequent operation shall be maintained within the reduced acceptable operation region of Figure 3.2-4 SURVEILLANCE REQUIREMENTS

.2.3.1 The provisions of Specification 4.0.4 are not applicable.
.2.3.2 F

shall be calculated by the expression FT = f (1+T ) and FT r r r q r shall be detennined to be within its limit at the following intervals: l' 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. qALVERT CLIFFS - UNIT 2 3/4 2-9 Amendment No. 9, 76, 78,31 l c.

.= i SURVEILLANCE REQUIREMENTS (Continued) shall be determined ea:h time a calculation of Ff is required 4.2.3.3 F r by using the incore detectors to obtain.a power distribution map with all i full length CEAs at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump combination. T 4.2.3.4 T shall be determined each time a calculation of F is required q usedtodetermineFfshallbethemeasuredvalueof and the value of T q 9-4 I l 4 7 f e i i ~ CALVERT CLIFFS - UNIT 2 3/4 2-10 Asendment 'No. 9, 76,18 9 + .,c ,,.-.r. -p ~ -.m ,-...,,,n,e n.

j i i i zo i-<z 1A* } CO

r..:;

r. un s.

..= r- ~ UN.ACCEPTA BLE zd=~.=E===q=. 2"" = =~g"=~j?" -~j = 2 .; a :.r.nn. ^ o

.=--

OP ERATION OPERATION ' = o REGION REGION sd= =s=e =sd=E:- -!. c- = o d - (.15, 1.0) 7 - ~.. 7 " 1.0 - . P:10,1.0) /r~ .. '. ;. ;. =;.= - -- i ~E.2 L- \\ .. =i1 '. n ' ' : _. I\\ .J- ,i;

"~

. k-~ 5-]_.[3.* f.i_*'.~ 3 7 .:.1;.n =:.1 :C-.=-- / y-n n:r -1 =- ~ .. \\ 7. ..... ?!.....; -..... .i 1'

! : := = : =.- _- - 17 /;

-- -. *. x 1 ; .. \\ - : l " C "'_ U :C'. ":- "-. X - r /:_n .2 w I- , * " ;* /.:.n;y. r ; - 1.- . x L. 2:. 'l r.-- ce 0.9 a. i.. - _ =i/ = = 1u..--. . : :.==..,. = o ^ __..n. : u. e w i .0. r/...__._+..............u =:.--u = =: :.:- -- - :: \\ :.. C=.. ~..] : s .i.. -. t .......=......=....__4....._,......b_...__......_........g-.._.=... = =.. 13J

T. ". r_. '. -..- !

J.. r.

y. _ _-t. __"..2 7

. /..: 1. _.T.. ~:._. 7..-._.: =._. _.. J. _: * 'dd .1,! _. :,=-. ,.- - e : /m =.===- - E=n==n I r._ - \\ - ~ - - 50.8 .3.. y...=_=_ g. - _,._.g. g. g. p ...._ \\ g _.=. y g ;......., g CL ..-- - (. 3 0,. 7 5 ) J =_..=_..._ _ =p= ' =_ = =... '..= =.. ' =. '.. hu ---- - - -(. 30,. 75 a p..

- -. - - - - - _-:=---..;:

_=..

w .... _... _... _ _. _. _ _.. - -- =:. =_r_t_ =.._._ : r_;=.. _.::. _.._: :...._ _._ :... _ _._...... = _, - n.._ -J ...'...__=.._r.-: =. _=.. _=. =:. r..... ; ;.:_.u.n.,. m_u :._. : n... _. _ ;.._.. p-- .=...2.,__=_.._t.- C g7 < v.I g 2.:== 2 c r.

:_..- }

= ..n ;;2..;; 7 o .a. _a -J

nr:= r=:: :.

.=. 1.

==d =.~.h i '=.1.= := i i ;. :iz. _ =. =. - - ~ =c .. = mur=J- = : ::- - .r .c

==.tu n a r P l A D Lt "----+ m-n. r-

_r

c:.x :.;: c

,3 ().O m, -- -.------- A L L t ... - - -..::. _.-:c_ '.._=_..:_:. =.-

== .:.= :.. OPMAIN-- g 1 y l

==. :.:

n: =._ -.

n t e s rw =:. :,; ----- -- - _. :. u.

c. = r --- -

IR.U 1UIN r,_.... ;. t m - 4 g [. _- ::=_=nn ----- -" - - ;- n :.:.u j... :.;. l..n.;,.-...... :_.; :_ = _ -. -. - -- =.;;-" - g t_....... -. _. _. _.. - 5 '5r= 0 = = r= = = = -, z

====a -0.6 -0.4 -0.2 0

0. 2

0. 4 0.6 C

PERIPHERAL AXIAL SHAPE INDEX> YI o< ce w Figure 3.2-4 i DHB AXIAL FLUX OFFSET CONTROL LIMITS t t l CALVERT CLIFFS - UNIT 2 3/4 2-11 Amendment No.9, 78, 31

POWER DISTRIBUTION LIMITS AZIMUTHAL POWER TILT - Tq LIMITING CONDITION FOR OPERATION 3.2.4 The AZIMUTHAL POWER TILT (T ) shall not exceed 0.030. l q APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER.* ACTION: a. With the indicated AZIMUTHAL POWER TILT determined to be > 0.030 but < 0.10, either correct the power tilt within two hours or determine within the next 2 hours and at least once per subsequent 8 hours, that the TOTAL PLANAR RADIAL PEAKING FACTOR (F ) and the TOTAL INTEGRATED RADIAL PEAKING FACTOR (Ff) are within 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 INTEGRATED RADIAL PEAKING FACTOR (Ff) and TOT PLANAR RADIAL PEAKING FACTOR (F ) are within the limits of Specifications 3.2.2 and 3.2.3. Subsequent operation for the purpose of measurement and to identify the cause of the tilt is allowable provided the THERMAL POWER level is restricted to < 20% of the maximum allowable THERMAL POWER level for the existing Reactor Coolant Pump combination. SURVEILLANCE REQUIREMENTS 4.2.4.1 The provisions of Specification 4.0.4 are not applicabl.e. 4.2.4.2 The AZIMUTHAL F3WER tit'T 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 POWER TILT at least once per 12 hours 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 2 3/4 2-12 Amendment No. 9 I

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION LIMITING CONDITION FOR OPERATION

3. 3.1.1 As a minimum, the reactor protective instrumentation channels and bypasses of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2.

APPLICABILITY: As shown in Table 3.3-1. ACTION: As shown in Table 3.3-1. SURVEILLANCE REQUIREMENTS 4.3.1.1.1 Each reactor protective instrumentation channel shall be demonstrated OPERABLE by the. performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the MODES and at the frequencies shown in Table 4.3-1. 4. 3.1.1. 2 The logic for the bypasses shall be demonstrated OPERABLE prior to each reactor startup unless performed during the preceding 92 days. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation. 4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit at.least once per 18 months. Each test shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the " Total No. of Channels" column of Table

3. 3-1.

CALVERT CLIFFS - UNIT 2 3/4 3-1

I l 1^P!r.3.3-1 REACTOR PROTECTIVE INSTRUMENTATION ES MINIMUM ~* TOTAL NO. CHANNELS CHANNELS APPLICABLE l p2 FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION ul 1. Manual Reactor Trip 2 1 2 .1, 2 and

1 j

2. Power level - liigh 4 2 3(f) 1, 2' 2# 7 3. Reactor Coo,lant Flow - Low 4/SG 2(a)/SG 3/SG 1, 2 (e) 2# 4. Pressurizer Pressure - liigh 4 2 3 1, 2 2# 5. Containment Pressure - liigh 4 2 3 1, 2 2# 6. Steam Generator Pressure - Low 4/SG 2(b)/SG 3/SG 1, 2 '2# 7. Steam Generator Water Id Level - Low 4/SG 2/SG 3/SG 1, 2 2# Y' 8. Axial flux Offset 4 ?(c)

I 1

2# h) 9.a. iheriul Margin / Low Pressure 4 2(a) 3 1, 2 (e) 2# b. Steam Generator Pressure Dif ference - liigh 4 2(a) 3 1,2 (e) 2# f[ 10. Loss of Turbine--Hydraulic Fluid Pressure - Low 4 2(c) 3 1 2# a = ? S e e

TABLE 3.3-1 (Continued) c3 .N si REACTOR PROTECTIVE INSTRU!'.ENTATION E n [~ MINIMUM TOTAL NO. CHANNELS CHANNELS. APPLICABLE %l OF CHANNELS TO TRIP OPERABLE MODES ACTIOh f FUNCTIONAL UNIT Ei

11. Wide Range Logaritnmic Neutron El Flux Monitor m

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 R

12. Reactor Protection System 6

1 6 1, 2* '4 i' Logic Matrices w

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

4 Logic Matrix Relays

14. Reactor Trip Breakers 8

6 8 1,l2* 4

TABLE 3.3-1 (Continued) TABLE NOTATION

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

The provisions of Specification 3.0.4 are not applicable.

Trip may be bypassed below 10'be automatically removed when THE of RATED THERMAL POW (a) of RATED THERMAL POWER. (b) Trip may be manually bypassed below 685 psia; bypass shall be automatically removed at or above 685 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 Range fleutron Flux Monitoring Channels. ACTION STATEMENTS ACTION 1 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in HOT STANDBY within the next 6 hours and/or open the protective system trip breakers. ACTION 2 With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied: a. The inoperable channel is placed in either the bypassed or tripped condition within 1 hour. For the purposes of testing and maintenance, the inoperable channel may be bypassed for up to 48 hours from time of initial loss of OPERABILITY; however, the inoperable channel shall then be either restored to OPERABLE status or placed in the tripped condition. CALVERT CLIFFS - UNIT 2 3/4 3-4 Amendment No. 31 l l

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 net; however, one additional channel may be bypassed for up to 48 hours while performing tests and rmaintenance on that channel provided the other inoperable channel is placed in the tripped condition. With the number of channels OPERABLE one less tham required ACTION 3 by the Minimum Channels OPERABLE requirement, verify compli-ance with the SHUTDOWN MARGIN requirements of Spe:cification i 3.1.1.1 or 3.1.1.2, as applicable, within 1 hour and at least once per 12 hours thereafter. With the number of channels OPERABLE one less tha-n required ACTION 4 by the Minimum Channels OPERABLE requirenent, be in HOT STANDBY within 6 hours; however, one channel may.be bypassed for up to 1 hour for surveillance testing per Specification 4.3.1.1. h e CALVERT CLIFFS - UNIT 2 3/4 3-5

TABLE 3.3-2 l n REACTOR PROTECTIVE INSTRUMENTATION RESPONSE' TIMES 4 { FUNCTIONAL UNIT RESPONSE TIME (n 1. Manual Reactor Trip Not Applic6ble g 2. Power Level - High < 0.40 seconds *# and < 8.0 seconds ## [ 3. Reactor Coolant Flow - Low 1 0.50 seconds 4. Pressurizer-Pressure - High 1 0.90 seconds 5. Containment Pressure - High 1 0.90 seconds 6. Steam Generat'or Pressure - Low 1 0.90 seconds 7. Steam Generator Water Level - Low 1 0.90 seconds { 8.~ Axial Flux Offset < 0.40 seconds *# and < 8.0 seconds ## [ 9.a. Thermal 1argin/ Low Pressure < 0.90 seconds *# and < 8.0 seconds ## b. Steam Generator Pressure Di f ference - High 1 0.90 seconds

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

11. Wide Range Logarithmic Neutron Flux Monitor Not Applicable E

R

Neutron detectors are exempt from response time testing.

Response time of the neutron flux signal portion

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

Response time does not include contribution of RTDs.

E

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

to achieve 63.2% of its total change when subjected to a step change in RTD temperature. 26 M a

_ TABLE 4.3-1 1 9_ RIACIOlt_j'It0iLCTIVE INS 1RUMENTATION SURVEILLANCE REQUIREMENTS ~ ,g 4 P CHANNEL IODES IN WHICH CilANNEL CilANNEL FUNCTIONAL SURVEILLANCE 5 v> FUNCTIONAL UNIT CHECL CALIBRATION TEST REQUIRED i g 1. Manual Reactor Trip N.A. N.A. S/U(1) N.A. 2. Power Level - High a. Nuclear Power S 0(2),tf(3),Q(5) M 1, 2 .m b. AT Power S D(4),R M 1 3. Reactor Coolant Flow - Low S R 'M 1, 2 4. Pressurizer Pressure - liigh 5 R M 1, 2 j 5. Containment Pressure - liigh S R M 1, 2 [ wO 6. Steam Generator Pressure - Low S R M 1, 2 t 7. Steam Generator Water R M 1, 2 Level - Low t

8., Axial Flux Offset S

R M 1 9.a. Thermal Margin / Low I'ree.sure S it I, 2 l F l b. Steam Generator Pressure Difference - S R 1, 2 liigh 10. Loss of Turbine--Ilydraulic Fluid W Pressure - Low N.A. N.A. S/U(1) N.A. S t i

g TABLE 4.3-1_(Continued) G g REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS P CHANNEi. MODES IN WHICH qg CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT C' HECK CALIBRATION ~ TEST REQUIRED'

11. Wide Range Logarithmic Neutron S

R(5) S/U(1) 1,2,3,4, H Flux Monitor 5 and

to 12.

Reactor Protection System Logic Matr' ices N.A. N.A. M and S/U(1) 1, 2 13. Reactor Protection System Logic Matrix Relays N.A. N. A. M and S/U (1) 1, 2 14. Reactor Trip Breakers N.A. N.A. M 1, 2 and

Y=

4 m

TABLE 3.3-3 (Continued) TABLE NO.TATION (a) Trip function may be bypassed in this t10DE when pressurizer pressure is < 1700 psia; bypass shall be automatically removed when pressurizer pressure is > 1700 psia. (c) Trip function may be bypassed in this MODE below 685 psia; bypass shall be automatically removed at on above 685 psia. The provisions of Specification 3.0.4 are not applicable. ACTION STATEMENTS With the number nf 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. ACTION 7 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the following condit4rns 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 OPERAELE 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 Min. mum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 48 hours while performing tests and maintenance on that channel provided the other inoperable channel is placed in the tripped condition. l l l CALVERT CLIFFS - UNIT 2 3/4 3-15 Amendment No.31

TABLE 3.3-3 (Continued) With less than the Minimum Channels OPERABLE, operation ACTION 8 may continue provide the containment purge valves are maintained closed. ACTION 11 - With the number of OPERABLE Channels one less than the Total Number of Channels, operation may prc eed 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. 1 CALVERT CLIFFS - UNIT 2 3/4 3-16 Amendment No. 3 ~ l -_y_

TAP.If 3.3-4 ? G EttGINEERED SAFETY TEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES m A P ALLOWABLE FUNCTIONAL UNIT TRIP SETPOINT VALUES ~n A 1. SAFETY INJECTION (SIAS) e a. tianual (Trip Buttons) Not Applicable Not Applicable _z b. Containment Pressure - liiqh < 4.75 psig 1 4.75 psig c. Pressurizer Pressure - Low > 1578 psia > 1578 psia l 2. CONTAINMENT SPRAY (CSAS) a. Manual (Trip Buttons) Not Applicable Not Applicable i b. Containment Pressure -- liiqh

4.75 psig 1 4.75 psig 3.

CONTAINMENT ISOLATION (CIS) a. Manual CIS (Trip Buttons) Not Applicable Not Applicable j } Y b. Containment Pressure - liigh 1 4.75 psig 1 4.75 psig C 4. MAIN STEAM LINE ISOLATION a. Manual (MSIV iland Switches and Feed Head Isolation h" Hand Switches) Not Applicable Not Applicable g b. Steam Generator Pressure - Low

570 psia

> 570 psia l n p i S a w r

TABLE 3.3-4 (Continued) 9 G ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES "x w ' ALLOWABLE P FUNCTIONAL UNIT TRIP VALUE VALUES ~n 5 5. CONTAINMENT SUMP RECIRCULATION (RAS) a. Manual RAS (Trip Buttons) Not Applicable Not Applicable [5 b. Refueling Water Tank - Low > 24 inches above > 24 inches above tank bottom tank bottom 6. CONTAINMENT PU,RGE VALVES ISOLATION ~ Manual (Purge Valve Control Switches) Not Applicable Not Applicable a. b. Containment Radiation - High Area Monitor 1 229 mr/hr 1 220 mr/hr g a 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) 210.2 second time delay 2[0.2 second time delay \\ l b. 4.16 kv Emergency Bus Under-3628125 volts with a 3628125 volts with a i voltage (Degraded Voltage) 810.4 second time delay 810.4 second time delay i i E i l 8" M i

O y TABLE 3.3-4 (Continued) EtlGINECRED SAFETY FEATURE ACTUATION SYSTEtt IrlSTRUMEtlTATI0fl TRIP VALUES P G ALLOWADLE 5 FUNCTIONAL UNIT TRIP VALUE VALUES l g 8. CVCS ISOLATION [ 11est Penetration Room / < 0.5 psig < 0.5 psig Letdown Heat Exchanger Room Pressure - High R = G h a R 8n E

TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 1. Manual a. SIAS Safety Injection (ECCS) Not Applicable b. CSAS Containment Spray Not Applicable c. CIS Containment Isolation Not Applicable d. RAS Containment Sump Recirculation Not Applicable 2. Pressurizer Pressure-Low a. Safety Injection (ECCS) 1 30*/30** 3. Containment Pressure-High a. Saf'ety Injection (ECCS) < 30*/30** b. Containment Isolation 1 30 c. Containment Fan Coolers 1 35*/10** 4. Containment Pressure--High a. Containment Spray 1 60*/60** l 5. Containment Radiation-Hioh a. Containment Purge Valves Isolation 15 l Amendment No.6. 31 CALVERT CLIFFS - UNIT 2 3/4 3-20

3/4.4 REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3.4.1 Both reactor coolant loops and both reactor coolant pumps in each loop shall be in operation. APPLICABILITY: As noted below, but excluding MODE 6*. ACTION: MODES 1 and 2: a. With one reactor coolant pump not in operation, STARTUP and/or continued POWER OPERATION may proceed provided THERMAL POWER is restricted to < 80% of RATED THERMAL POWER and the setpoints for the followTng trips have been reduced to the values specified in Specification 2.2.1 for operation with three reactor coolant pumps operating: 1. Power Level-High 2. Reactor Coolant Flow-Low 3. Thermal Margin / Low Pressure ~' 4. Axial Flux Offset b. With two reactor coolant pumps in opposite loops not in opera-tion, STARTUP and/or continued POWER OPERATION may proceed pro-vided THERMAL POWER is restricted to < 51.1% of RATED THERMAL POWER and the setpoints for the following trips have been reduced to the values specified in Specification 2.2.1 for operation with two reactor coolant pumps operating in opposite loops: 1. Power Level-High 2. Reactor Coolant Flow-Low 3. Thermal Margin / Low Pressure 4. Axial Flux Offset c. With two reactor coolant pumps in the same loop not in opera-tion, STARTUP and/or continued POWER OPERATION may proceed provided the water level in both steam generators is maintained above the Steam Generator Water Level-Low trip setpoint, the THERMAL POWER is restricted to < 46.8% of BATED THERMAL POWER,

See Special Test Exception 3.10.3.

.CALVERT CLIFFS-UNIT 2 3/4 4-1 l I

REACTOR C00LAliT SYSTEft LIMIT!!iG C0flDITION FOR OPERATION (Continued) and the setpoints for. th,e following trips have' been reduced to the values specified in Specification 2.2.1 for operation with two reactor coolant pumps ope. rating in the same loop:

1..

Power Level-High 2. Reactor Coolant Flow-Low 3. Thermal fiargin/ Low Pressure 4. Axial Flux Offset MODE 3: Operation may proceed provided at least one reactor coolant pump i is in operation in each reactor coolant loop. !!0 DES 4#** and 5#**:.0peration may proceed provided at least one reactor l coolant loop is in, operation with an associated reactor coolant j pump o. shutdown cooling pump.* The provisions of Specifications t 3.0.3 and 3.0.4 are not applicable. s j All reactor coolant pumps and shutdown cooling pumps may be de-energized for up to 1 hour to accommodate transition between shutdown cooling pump i

[

and reactor coolant pump operation, provided no operations are permitted which could cause dilution of the reactor coolant system boron concentration. p i** A reactor coolant pump shall not be started with one or more of the RCS cold leg. temperatures 1 275 F unless 1) the pressurizer water volume is less than 600 cubic feet or 2) ghe segondary water temperature of each steam generator is less tMa 46 F (34 F when measured by a surface contact instrument) above the coolant temperature in the reactor vessel. i l SURVEILLAtlCE REQUIREMEr4_TS i i 4 L4.4.1 The Reactor Protective Instrumentation channels specified in the applicable ACTI0t1 statement above shall be verified to have had their trip setpoints changed to the values specified in Specification 2.2.1 for the applicable nunber of reactor coolant pumps operating either: a. Within 4 hours.after switching to a different pump combination l if switch -is made while operating, or b. Prior 'to reactor criticality if switch is made while shutdown.

See Special Test Exception 3.10.5.

CALVERT CLIFFS - UfJIT 2 3/4 4-2 Amendment flo. 5,16, 31 i i

.--~ _- 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), SAFETY INJECTION TANKS i LIMITING CONDITION FOR OPERATION 3.5.1 Each reactor coolant system safety injection tank shall be OPERABLE with: a. The isolation valve open, b. A contained borated water volume of between 1113 and 1179 cubic feet of borated water (equivalent to tank levels of between 187 and 199 inches, respectively), l c. A boron concentration of between 2300 and 2800 ppm, and l d. A nitrogen cover-pressure of between 200 and 250 psig. I APPLICABILITY: MODES 1, 2 and 3.* ACTION: a. With one safety injection tank inoperable, except as a result i of a closed isolation valve, restore the inoperable tank to OPERABLE status within one hour or be in HOT SHUTDOWN within the next 12 hours. l b. With one safety injection tank inoperable due to the isolation valve being closed, either immediately open the isolation 1 valve or be in ' HOT STANDBY within.one hour and be in HOT I SHUTDOWN within the next 12 hours. SURVEILLANCE REQUIREMENTS 4.5.1 Each safety injection tank shall be demonstrated OPERABLE: a. At least once per 12 hours by: 1. Verifying the contained borated water volume and nitrogen cover-pressure in the tanks, and 2. Verifying that each safety injection tank isolation valve i is open. t

With pressurizer pressure > 1750 psia.

CALVERT CLIFFS - UNIT 2 3/4 5-1 Amendment No. gy,31 =

EMERGENCY CORE C00LI"G SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b. At least once per 31 days by verifying _the_boton_ concentration of'the" safety injection tank solution. c. At least once per 31 days when the RCS pressure is above 2000 psig, by verifying that power to the isolation valve operator is removed by maintaining the feeder breaker open under administrative control. d. Within 4 hours prior to increasing the RCS pressure above 1750 psia by verifying, via local indication at the valve, that the tank isolation valve is open. e.- At least once per 18 months by verifying that each safety injection tank isolation valve opens automatically under each of the following conditions: t 1. When the RCS pressure exceeds 300 psia, and 2. Upon receipt of a safety injection test signal. f. Within one hour prior to each increase in solution volume of > 1% of normal tank volume by verifying the boron roncentration at the operating high pressure safety injection pump discharge is between 2300 and 2800 ppm. l i l I l 1 CALVERT CLIFFS - UNIT 2~ 3/4 5-2 Amendment No. 27, 31

4 EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) e. At least once per 18 months by: i ~ 1. Verifying automatic isolation and interlock action of the shutdown cooling system from the Reactor Coolant System .when the Reactor Coolant System pressure is above 300 psia. 2. A visual inspection of the containment sump and verifying' that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, l etc.) show no evidence of structural distress or corrosion. 3. Verifying that a minimum total of 100 cubic feet of l solid granular trisodium phosphate dodecahydrate (TSP) is contained within the TSP storage baskets. 4 Verifying that when a representative sample of 4.0 + 0.1 grams of TSP from a TSP storage basket is submerged-without 0 agitation, in 3.5 t 0.1 liters of 77 10 F borated water i ' from the RWT, the pH of the mixed solution is raised to > 6 within 4 hours. t. At least once per 18' months, during shutdown, by: i 1. Verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection Actuation test signal. 2. Verifying that each of the following pumps start auto-matically upon receipt of a Safety Injection Actuation Test Signal: a. High-Pressure Safety Injection pump. i b. Low-Pressure Safety Injection pump. g. By verifying the correct position of each electrical position stop for the following Emer gency Core Cooling System throttle i valves: I 1. During each performance of valve cycling required by Specification 4.0.5 by observation of valve position on the control boards. C ALVERT CLIFFS - UNIT 2 3/4 5-5 Amendment No. M, 31

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2. Within 4 hours following completion of maintenance on the valve or its operator by measurement of stem travel when the ECCS subsystems are required to be OPERABLE. HPSI SYSTEM Valve Number Valve Number MOV-616. MOV-617 MOV-626 MOV-627 MOV-636 MOV-637 MOV-646 MOV-646 b. By performing a flow balance test during shutdown following completion of HPSI system modifications that alter system flow characteristics and verifying the following flow rates: HPSI System Single Pump 170 + 5 gpm to each injection leg. e i CALVERT CLIFFS - UNIT 2 3/4 5-5a Amendment No.16 I

1 ' EMERGENCY CORE C0OLING SYSTEMS i - REFUELING WATER TANK LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water tank shall be OPERABLE with: s a. A minimum contained borated water volume of 400,000 gallons, b. A boron concentration of between 2300 and 2800 ppm, l c. A minimum water temperature of 40 F, and d. A maximum solution temperature of 100*F in MODE 1. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the refueling water tank inoperable, restore the tank to OPERABLE status within I hour or be in at least HOT STANDBY within 6 hours and in j COLD SHUTDOWN within the following 30 hours. 4 lSURVEILLANCEREQUIREMENTS 14.5.4 The RWT shall be demonstrated-0PERABLE: 1 I a. At least once per 7 days by: 1 i 1. Verifying the contained barated water volume in the tank, and 2. Verifying the boron concentration of the water. I b. At-least once per 24 hours by verifying the RWT temperature when the outside air temperature is < 40 F. 4 4 ) CALVERT CLIFFS - UNIT 2-3/4 5-7 - Amendment No. 6, 9, 27,31 t 4 ,,-n, ,e .,+w, -n nn,--r r -- -, e... ur ~~,----m,- ---r,--v, -, + -.,. - --,e-+-

9 ' { TABLE 3.7,4 SAFETY RELATED llYDRAUIIC SNUBBERS

C.

SNOBBER SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGil RADIATION ESPECIALLY DIFFICULT E-N0. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE ** TO REMOVE y (A or I) _(Yes or No) (Yes or No) 2-61-19 CONT. SPRAY HDR FOR SPRAY RING

22 39' I

Yes Yes 2-63-1 S/G #22 BLOWDOWN LINE 34' 11' A flo No 2-63-2 S/G #22 IiLOWDOWN LINE 27' 10' A No No 2-63-3 NITROGEN LINE TO S/G #22 77'6" I Yes No 2-63-4 NITR0 GEN LINE TO S/G #22 77'6" I Yes No w3 2-63-5 S/G #21 SURFACE BLOWDOWN LINE 76'9" I Yes No' .y L 2-63-6 S/G #21 SURFACE BLOWDOWN LINE 76'9" I Yes No 2-63-11 STEAM GENERATOR #21 75' *** I Yes Yes 2-63-12 STEAM GENERATOR #21 75' I Yes Yes 2-63-13 STEAM GENERATOR #21 75' I Yes Yes 2-63-14 STEAM GENERATO? #21 75' I Yes Yes 2-63-15 STEAM GENERATOR #21 75' I. Yes Yes b' l 2-63-16 STEAM GENERATOR #21 75' I Yes Yes 2-63-17 STEAM GENERATOR #21 75' I Yes Yes g 5 e TB S

T ABl.E ' 3. 7-4 h5' SAf t:TY RELATED liYDRAULIC SNUBBERS *

a B

n SNVBBER SYSTEM SNUB 0ER INSTALLED ACCESSIBLE.0R llIGil RADIATION ESPECIALLY DIFFICULT C NO. ON LOCATI0ri AND ELEVATION INACCESSIBLE ZONE ** TO REMOVE g. ~~ ~~ --(A or T)" (Yes or No) (Yes or No) h 2-63-18 STEAM GENERATOR #21 75' I I Yes Yes

1 2-63-19 STEAM GENERATOR #22 75'***

I Yes Yes m 2-63-20 STEAM GENERATOR #22 75' I Yes Yes 2-63-21 STEAM GONERATOR #22 75' I Yes Yes J, 2-63-22 STEAM GENERATOR #22 75' I Yes Yes ' 2 2-63-23 STEAM GENERATOR #22 75' I Yes - Yes y E 2-63-24 STEAM GENERATOR #22 75' I Yes Yes 2-63-25 STEAM GENERATOR #22 75' I Yes Yes 2-63-26 , STEAM GENERATOR #22 75' I Yes Yes E 2-64-1 PRESSURIZER REL PIPING UPSTREAM MOV 403 81'6" I Yes No 2-64-2 PRESSURIZER REL PIPING TO RV 200 f 79'11" I Yes No Si 2-64-3 PRESSURIZER REL PIPING DOWNSTREAM MOV 405 84'3" I .Yes No i e G e

_.m l l l-1 i n{ TABLE,3,.7-4 SAFETY RELATED HYDRAULI_C SNUBBERS

P

?! SNUBBER' SYSTEM SNUBBER INSTALLED ACCESSIBLE OR HIGH RADIATION ESPECIALLY' DIFFICULT T NO. ON, LOCATION AND ELEVATION INACCESSIBLE ZONE ** TO REMOVE jE (A or I) (Yes or No) (Yes or No) -t m 2-83B-2 MSIV #21 HYDRAULIC SUPPLY 27' A No No 2-83B-3 MSIV #21 HYDRAULIC RETURN 27' A No - No . j{

5nubbers may be added to safety related systems without prior License Amendment to Table 3.7-4 provided that a revision to Table 3.7-4 is included with the next License Amendment request.

y [n

- Modifications to this table due to changes in high radiation areas shall be submitted to the NRC as part of the next License Amendment request.
  - Snubters. served by a ccmmon hydraulic reservoir are indicated by a bracket.

All reservoirs servinq gp more than one snubber shall be inspected to ensure adequate hydraulic level. m E a ' Mthin 7 da ys after reactor startup fo.llowing a major outage or following any maintenance in the S immediate vicinity of these snubbers, reservoirs or associated hydraulic piping; and ,+ b. Every 31 da ys 25 percent. S m-

I 9. TABLE 4.7-4 { i. l N t [.. ilYDRAULIC-SNUBBER INSPECTION SCHEDULE r, - n NEXT REQUIRED 5 NUMBER OF SNUBBERS FOUND.IN0PERABLE INSPECTION INTERVAL ** l-DURING INSPECTION-0R DURING INSPECTION INTERVAL

5g 18 months + 25%

0 12 months T 25% 1 6 months T 25% 2 . 3 or 4 124 days T 25% ' 5, 6, or 7 62 days 'T 25% 31 days + 25% -.8 k R.. -? - E

Snubbers may be categorized into two groups, " accessible" and " inaccessible". This categorization shall be based upon the snubber's accessibility for inspection during reactor operation. These two groups may be l

inspected independently according to the-above schedule.

- The required inspection interval shall not be lengthened more than one step at a time.

L I

4 3/4.9 REFUELING OPERATIONS BORON CONCENTRATION LIMITING CONDITION FOR OPERATION ' 3.9.1 With the reactor vessel head unbolted or removed, the boron concentration of all filled portions of the Reactor Coolant System and the refueling pool shall be maintained uniform and sufficient to ensure that the more restrictive of following reactivity conditions is met: a. Either a K of 0.95 or less, which includes a 1% Ak/k conser-vativeall8bknceforuncertainties,or b. A boron concentration of > 2300 ppm, which includes a 50 ppm l conservative allowance for uncertainties. APPLICABILITY: MODE 6*. ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity c' ances and initiate and continue boration at > 40 gpm of 2300 ppm l r or tne baron concentration is restored to >23Db' ppm, educed to < 0.95 std: acid solution or its equivalent until K is r j whichever is the more restrictive. The provisions of Specification 3.0.3 are not apolicable. SURVEILLANCE REQUIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to: a. Removing or unbolting the reactor vessel head, ar.d b. Withdrawal of any full. length CEA in excess of 3 feet from its fully inserted position. 4.9.1.2 The baron concentration of the reactor coolant system and the refueling pool shall be determined by chemical analysis at least 3 times'per 7 days with a maximum time interval between samples of 72 hours.

The reactor shall be maintained in MODE 6 when the reactor vessel head is unbolted or removed.

CALVERT CLIFFS - UNIT 2 3/4 9-1 Amendment No. 31 l l

REFUELING OPERATIONS l INSTRUMENTATION LIMITING COND! TION FOR OPERATION 3.9.2 As a minimum, two source range neutron flux monitors shall be operating, each with continuous visual indication in the control room and one,with audible indication in the containment and control room. APPLICABILITY: MODE 6. ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes. The provisions of Specification 3.0.3 are not applicable. i i SURVEILLANCE REQUIREMENTS l 4.9.2 Each source range neutron flux monitor shall be demonstrated i OPERABLE by performance of: a. A CHANNEL FUNCTIONAL TEST at least once per 7 days. b. - A CHANNEL FUNCTIONAL TEST within 8 hours prior to the initial l start of CORE ALTERATIONS, and e c. A CHANNEL CHECK at least once per 12 hours during CORE ALTERATIONS. 4 i CALVERT CLIFFS-UNIT 2 ' 3/4 9-2 I l-

REFUELING OPERATIONS CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING c LIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of 1600 pounds shall be prohibited from travel over fuel assemblies in the storage pool. APPLICABILITY: - With fuel assemblies in the storage pool. ACTION: With the requirements'of the above specification not satisfied, place the crane load in a safe condition. The provisions of Specification

3. 0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.7 The weight of each load, other than a fuel assembly and CEA, shall be verified to be < 1600 pounds prior to moving it over fuel assemblies. l CALVERT CLIFFS UNIT 2 3/4 9-7

e REFUELING OPERATIONS COOLANT CIRCULATION LIMITING CONDITION FOR OPERATION 3.9.8 At least one shutdown cooling loop shall be in operation. APPLICABILITY: MODE 6. ACTION: a. With less than one shutdown cooling loop in operation, except as provided in b. below, suspend all operations involving an increase in the rea-tor decay heat load or a reduction in boron concentra-tion of the Reactor Coolant System. Close all containment penetra-tions providing direct access from the containment atmosphere to the outside atmosphere within 4 hours. The shutdown cooling pumps may be de-energized during the time intervals required for local leak rate testing of containment penetration number 41 pursuant to the requirements of Specification 4.6.1.2.d and/or to permit maintenance on valves located in the common shutdown cooling suction line, provided 1) no operations are permitted which could cause dilution of the reactor coolant e.ystem boron concentration, 2) all CORE ALTERATIONS are suspended, J) all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere a're maintained closed, and 4) the water level above the top of the irradiated fuel is greater than 23 feet. b. The shutdown cooling loop may be removed from operation for up to I hour per 8 hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs. c. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.9.8 A shutdown cooliig loop shall be determined to be in operation and circulating reactor coolant at a flow rate of > 3000 gpm* at least once per 24 hours.

> 1500 gpm when the Reactor Coolant System is drained to a level below the midplane of the hot leg.

CALVERT CLIFFS - UNIT 2 3/4 9-8 Amendment No. 7V, 31

3/4.'10 SPECIAL TEST EXCEPTIONS SHUT 00WN MARGIN-LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of CEA worth and shutdown margin provided reactivity equivalent to at least the highest estimated CEA worth is available for trip insertion from OPERABLE CEA(s). APPLICABILITY: MODE 2. ACTION: With any full length CEA not fully inserted and with less than a. the above reactivity equivalent available for trip insertion, immsdiately initiate and continue boration at > 40 gpm of 2300 l ppm boric acid solution or its equivalent untiT the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored. b. With all full length CEAs inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at > 40 gpm of 2300 ppm boric l acid solution or its equivalent until the SHUTDOWN MARGIN re-quired by Specification 3.1.1.1 is restored. SURVEILLANCE REQUIREMENTS 4 4.10.1.1 The position of each full length CEA required either partially or fully withdrawn shall be determined at least once per 2 hours. 4.10.1.2 Each CEA not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position with-in 24 hours prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1. CALV5RT CLIFFS - UNIT 2 3/4 10-1 Amendment No. 78, 31

SPECIAL TEST EXCEPTIONS GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 - The group height, insertion and power distribution limits of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.2, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.2.2, and 3.2.3 may be suspended during the performance of PHYSICS TESTS provided: The THERMAL POWER is restricted to the test power plateau a.- - which shall not exceed 85% of RATED THERMAL POWER, and b. The limits of Specification 3.2.1 are maintained and determined as specified in Specification 4.10.2.2 below. APPLICABILITY: MODES 1 and 2. ACTION: With any of the limits of Specification 3.2.1 being exceeded while the requirements of Specifications 3.1.1.4, 3.1. 3.1, 3.1. 3. 2, 3.1.3. 5, 3.1.3.6, 3.1.3.7, 3.2.2 and 3.2.3 are suspended, either: Reduce THERMAL POWER sufficiently to satisfy the requirements a. of Specification 3.2.1, or b. Be in HOT STANDBY wthin 6 hours. SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS in which the requirements of Specifications 3.1.1.4, 3.1.3.1, 3.1.~3.2, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.2.2 or 3.2.3 are suspended and shall, be verified to be within the test power plateau. a 4.10.2.2 The linear heat rate shall be determined to be within the limits of Specification 3.2.1 by monitoring it continuously with the Incore Detector Monitoring System pursuant to the requirements of i Specifications 4.2.1.3 and 3.3.3.2 during PHYSICS TESTS above 5% of RATED THERMAL POWER in which the requirements of Spscifications 3.1.1.4, 3.1.3.1, 3.1.3.2, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.2.2 or 3.2.3 are suspended. i CALVERT CLIFFS - UNIT-2 3/4 10-2 +

3/4.1 ~ REACTIVITY CONTROL SYSTEMS ^ BASES i! 3/4.1.1 BORATI0f1 CONTROL 1 . 3/4.1.1.1 and - 3/4.1.1. 2 SHUTDOWN-MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently j subcritical to, preclude inadvertent criticality in the shutdown condition. ~ .l SHUTDOWN MARGIN requirements vary throughout core life as a function of f fuel-depletion, RCS boron concentration and RCS T The minimum available i .SHUTOOWN MARGIN for no load operating conditions $E9beginning of life is 4.1%

ltk/k and at end of life is 4.3% Ak/k. - The SHUTDOWN MARGIN is based on the 2

.; safety analyses performed for a steam line rupture event initiated at no load l conditions. The most restrictive steam line rupture event occurs at EOC

, conditions.

For the steam line rupture event at beginning of cycle conditions', lla minimum SHUTDOWN MARGIN of less than 4.1% Lk/k is required to control the 6 l reactivity transient, and end of cycle conditions require 4.3% Ak/k. Accordingly, !,!theSHUTDOWNMARGINrequirementisbaseduponthislimitingconditjonandis 200 F, the 1: consistent with FSAR safety analysis assumptions. With T ! reactivity transients resulting from any postulated accid 8 U are minimal and a

I'. '.k/k shutdown margin provides adequate protection.

With the pressurizer 'e.e'. 'ess than 90 inches, the sources of non-borated water are restricted to increase the time to criticality during a boroi dilution event. 3 /4.1.1. 3 BORON DILUTION A minimum flow rate of at least 3000 GPM provides adequate mixing, i: crevents stratification and ensures that reactivity changes will be

gradual during boron concentration reductions in the Reactor Coolant l' System.

A flow rate of at least 3000 GPM will circulate an equivalent

Reactor Coolant System volume of 9,601 cubic feet in approximately

,24 minutes. The reactivity change rate associated with boron conc.an-

ration reductions will therefore be within the capability of operator irecognition and control.

) I ,{3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC) i l~ The limitations on MTC are provided to ensure that the assumptions

csed in the accident and transient analyses remain valid through each: fuel cycle.

The surveillance requirements for measurement of the MTC l: curing each fuel cycle are adequate to confirm the MTC value since this ~ j'coefficientchangesslowlydueprincipallytothereductioninRCSboron { concentration associated with fuel burnup. The confirmation that the 4 i measured MTC value is within its limit provides bssurances that the-l Igcoefficient will be m2intained within acceptable values throughout each i uel cycle. f CALVERT CLIFFS - UNIT 2 B 3/4 1-1 Amendment No. 78, 31 4 1 t 1 -.--.c,, ,7 r

REACTIVITY CONTROL SYSTEMS BASES 3 /4.1.1. 5 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made crigical with the Reactor Coolant System average temperature 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 pressure vessel is above its minimum RTNDT temperature. 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 failu:res during the repair period. The boration capability of either system is sufficient to provide a SHUTOOWN MARGIN from all gperating conditions of 3.0% ak/k after xenon l 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 cequirements, the RWT is required to have a minimum contained volume of 400,000 gallons during MODES 1, 2, 3 and 4. The maximum boron 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 i 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 2 B 3/4 1-2 Amendment No. $, $,31 ~w w

REACTIVITY CONTROL SYSTEMS BASES The boron capability required below 200 F is based upon providing a 3% ak/k SHUTDOWN MARGIN af ter xenon decay and cooldown from 200*F to 140*F. This condition requires either 737 gallons of 7.25% boric acid solution from the boric acid tanks or 9,844 gallons of 2300 ppm borated water from the refueling water tank. The OPERABILITY of one boron injection. system during REFUELING ensures that this system is available for reactivity control while in MODE 6. 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power . distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of a CEA ejection accident are limited to acceptable levels. The ACTION statements which permit ' limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original criteria are met. The ACTION statements applicable to a stuck or untrippable CEA and to a large misalignment (> 15. inches) of two or more CEAs, require a ~ prompt shutdown of the reactor since either of these conditions may be indicative of a possible loss of mechanical functional capability of the CEAs and in tne event of a stuck or untrippable CEA, the loss of SHUT-DOWN MARGIN. For small misalignments (1 15 inches) of the CEAs, there is 1) a small degradation in the peaking factors relative to those assumed in generating LCOs and LSSS setpoints for DNBR and linear heat rate, 2) a small 'effect on the time dependent long term power distributions rela-tive to those used in generating LCOs and LSSS setpoints for DNBR and linear heat rate, 3) a small effect on the available SHUTDOWN MARGIN, and 4) a small effect on the ejected CEA worth used in the safety analysis. Therefore, the ACTION statement associated with the small misalignment of a CEA permits a one hour time interval during which attempts may be made to restore the CEA to within its alignment require-ments prior to initiating a reduction in THERMAL POWER. Tne one hour time limit is sufficient to (1) identify causes of a misaligned CEA, (2) take appropriate corrective action to realign the CEAs and (3) minimize i-the effects of xenon redistribution. Overpower margin is provided to protect the ;nre in the event of a large misalignment (> 15 inches) of a.CEA. However, this nisalignment would cause distortion of the core power distribution. The reactor CALVERT CLIFFS - UNIT 2' B 3/4 1-3 Amendment No. 6, 31 l l I m-

REACTIVITIY CONTROL SYSTEMS BASES 4 protective system would not detect the degradation in radial peaking factors and since variations in other system parameters (e.g., pressure and coolant temperature) may not be sufficient to cause trips, it is probable that the reactor could be operating with process variables less conservative than those assumed in generating LC0 and LSSS setpoints. Therefore, the ACTION statement associated.with the large misalignment of a CEA requires a prompt and significant reduction in THERMAL POWER prior to attempting realignment of the misaligned CEA. The ACTION statements applicable to misaligned or inoperable CEAs include requirements to align the OPERABLE CEAs in a given group with the inoperable CEA. Conformance with these alignment requirements bring the core, within a short period of time, to a configuration consistent with that assumed in generating LC0 and LSSS setpoints. However, extended operation with CEAs significantly inserted in the core may lead to perturbations in 1) local burnup, 2) peaking factors and 3) available shutdown margin which are more adverse than the conditions assumed to exist in the safety analyses and LC0 and LSSS setpoints determination. Therefore, time limits have been imposed on operation with inoperable CEAs to preclude such adverse conditions from developing. Operability of the CEA position indica'. ors is required to determine CEA positions and thereby ensure compliance with the CEA alignment and insertion limits and ensures proper operation of the rod block circuit. The CEA " Full In" and " Full Out" limits provide an additional independent means for determining the CEA positions when the CEAs are at either their fully inserted or fully withdrawn positions. Therefore, the ACTION statements applicable to inoperable CEA position indicators permit continued operations when the positions of CEAs with inoperable position indicators can be verified by the " Full In" or " Full Out" limits. CEA positions and OPERABILITY of the CEA position indicators are required to be verified on a nominal basis of once per 12 hours with more frequent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LC0's are satisfied. The maximum CEA drop time restriction is consistent with the assumed Measur'ement with T CEA drop time used in the accident analyses. l 515'F and_ with all reactor coolant pumps operating ensures that f8@ ~ t I I l CALyERT CLIFFS - UNIT 2 B 3/4 1-4 ~

a 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 exceec' 2200*F. Either of the two core power distribution monitoring systems, the ~ Excore Detector Monitoring System and the Incore Detector Monitoring System, provide adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its limits. The Excore Detector Monitoring System performs this function by continuously 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 Ficure 4.2-1, 3) the AZIMUTHAL POWER TILT restrictions of Specification 3.E.3 are satisfied, and 4) the TOTAL 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 engineering uncertainty factor of 1.03, 4) an allowance [ of 1.01 for axial fuel densification and thermal expansion, and 5) a THERMAL POWER measurement uncertainty factor of 1.02. l 3/4.2.2, 3/4.2.3 and 3/4.2.4-TOTAL PLANAR AND INTEGRATED RADIAL PEAKING T T FACTORS - F AND F AND AZIMUTHAL POWER TILT - Tq xy T The limitations on F and T are provided to ensure that the assumptions used in tre aElysis 90r establishing the Linear Heat Rate l and Local Power Density High LCOs and LSSS setpoints remain valid l during operation at yhe various allowable CEA group insertion limits. l The limitations on F and T are provided to ensure that the assumptions l usedintheanalysisestablishingtheDNBMarginLCO,andThermal r l l CALVERT CLIFFS - UNIT 2 8 3/4 2-1 Amendment No. 9, JE, 24 l l

POWER DISTRIBUTION LIMITS i BASES the analysis establishing the DNB Margin LCO, and Thermal Margin / Low Pressure LSSS setpoints remain valid duringT peration at the various o allowable CEA group insertion limits. If F F' or T exceed their basic limitations, operation may continue uNe,r [he ad8itional 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 Heat Rate, Thermal Margin / Low Pressure and Local Power Density - High LCOs and LSSS.setpoints r=ain 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 that must be used in the equation F = F*Y (1 + T ) and F}he value of Tr(1+T)Ssthemeasuredtilt. 9 =F r q T T The surveillance requirements for verifying that F F and T aye r T 4 F within their limits provide assurance that the actual vEu,es yf F i r and T do not exceed the assumed values. Verifying F and F afNr each9uelloadingpriortoexceeding75%ofRATEDTHEblPOWERprovides additional assurance that the core was properly loaded. 3/4.2.4 FUEL RESIDENCE TIME This specification deleted. 3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal 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 1.195 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 l operation. The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the l flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification,of flow rate on a 12 hour basis. i l CALVERT CLIFFS - UNIT 2 B 3/4 2-2 Amendment No. 9, 75, 78,31

'3/4.9 REFUELING OPERATIONS BASES l 3/4.9.1 BORON CONCENTRATION The limitations on minimum boron concentration (2300 ppm) ensure that: l

1) the reactor will remain subcritical during CORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity control in 'he water volumes having direct access to the reactor vessel.

The limitation on Keff of no greater than 0.95 which includes a conservative allowance for uncertainties, is sufficient to prevent reactor criticality during refueling operations. 3/4.9.2 INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactiv.ity condition of the core. 3/4.9.3 DECAY TIME r The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assumptions used in the accident analyses. 3/4.9.4 CONTAINMENT PENETRATIONS The requirements on containment penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressur-ization potential while in the REFUELING MODE. 3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the j facility status or core reactivity condition du, ring CORE ALTERATIONS. CALVERT CLIFFS - UNIT 2 8 3/4 9-1 Amendment No. 31 i l

REFUELING OPERATIONS BASES 3/4.9.6 REFUELING MACHINE OPERABILITY The'0PERABILITY requirements for the refueling machine ensure that: 1

1) the refueling' machine will be used for movement of CEAs and fuel assemblies, 2) the refueling machine.has sufficient load capacity to lift a CEA or fuel assembly, and 3) the core internals and pressure vessel are protected from excessive lifting-force in the event they are inadvertently engaged during lifting operations.

3/4.9.7-CRANE TRAVEL - SPENT FUEL STORAGE BUILDING The restriction on movement of loads in excess of the nominal weight of a fuel assembly and CEA over other fuel assemblies in the storage pool ensures that in the event this load is dropped (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. Tnis assumption is consistent with the activity release assumed in the accident analyses. 3/4.9.8 COOLANT CIRCULATION The requirement that at least one shutdown cooling loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel 3 below 140*F as required during the REFUELING MODE, and (2) sufficient 4 coolant circulation is maintained through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification. i 3/4.9.9 CONTAINMENT PURGE VALVE ISOLATION SYSTEM l The OPERABILITY of this system ensures that the containment purge j valves will be automatically isolated upon detection of high radiation levels within the containment. The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment. CALVERT CLIFFS - UNIT 2- .B 3/4 9-2

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i l DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 _ The reactor containment building is designed and shall be main-tained for a maximum internal pressure of 50 psig and a temperature.of 276 F. 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 217 fuel assemblies with each fuel assembly containing a maximum of 176 fuel rods clad with Zircaloy-4. Each fuel rod shall have a nominal active fuel length of 136.7 inches and contain a maximum total weight of 3000 grams uranium. The initial core loading shall have a maximum enrichment of 2.99 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 3.7 weight percent U-235. 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 The reactor core shall contain 77 full length and no part' length l cor. trol element assemblies. 5.4 REACTOR COOLANT SYSTEM j DESIGN PRESSURE AND TEMPERATURE l i 5.4.1 The reactor coolant system is designed and shall be maintained: I { a. In accordance with the code requirements specified in Section 4.2 of the FSAR with allowance for normal degradation pursuant of the applicable Surveillance Requirements, b. For a pressure of 2500 psia, and For a temperature of 650 F, except for the' pressurizer which c. l is 700 F. CALVERT CLIFFS - UNIT 2 5-4 Amendment No. 78,31 .}}

ML19341C319

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