ML17354B255

Calvert Cliffs Nuclear Power Plant, Units 1 and 2 - Redacted Updated Final Safety Analysis Report Revision 49

ML17354B255
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Site:	Calvert Cliffs
Issue date:	09/25/2017
From:	Calvert Cliffs
To:	Marshall M L Plant Licensing Branch 1
Marshall M L, NRR/DORL/LPLI, 415-2871
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{{#Wiki_filter:CALVERT CLIFFS UFSAR 1.1-1 Rev. 47

1.0 INTRODUCTION

AND SUMMARY

1.1 INTRODUCTION

Construction of Calvert Cliffs Units 1 and 2 was authorized by the Atomic Energy Commission (AEC) by issuance of Construction Provisional Permits CPPR-63 and CPPR-64 in Docket Numbers 50-317 and 50-318 on July 7, 1969. Unit 1 went into commercial operation in May 1975, and Unit 2 in April 1977.

On July 11, 1967, the AEC published in the Federal Register the Proposed General Design Criteria for Nuclear Power Plants. Prior to the issuance of the construction permit, Calvert Cliffs submitted the Preliminary Safety Analysis Report (PSAR) in which was reflected this plant's design intent based on these criteria. Design and construction proceeded accordingly. The Final Safety Analysis Report (FSAR) was submitted in support of the application for a license to operate the plant. Revision 0 of the Updated Final Safety Analysis Report was submitted in July 1982, and has been periodically revised since then. Subsequent to the initial startup of both units, 10 CFR Part 50, Appendix A containing 64 general design criteria was issued. These criteria reflected the original 70 criteria with revisions and regrouping. Design changes and modifications for Calvert Cliffs are evaluated for consistency with the proposed criteria except where specific Appendix A criteria have been required by the Nuclear Regulatory Commission (NRC). The Nuclear Steam Supply System (NSSS) for both units is identical, utilizing pressurized water reactors supplied by Combustion Engineering, Inc. (CE). The NSSS includes a control element assembly (CEA)-type reactor core with two steam generators (SGs), two reactor coolant loops and four reactor coolant pumps (RCPs). The geometry of the core is essentially identical to that used for the Main Yankee Atomic Power Station (Docket Number 50-309). The reactor coolant loops are very similar to those in the Palisades Plant (Docket number 50-255). The SGs are Babcock & Wilcox, Canada replacement steam generators. The replacement reactor vessel closure heads (RVCHs) were supplied by Babcock & Wilcox, Canada. An initial license was requested to operate each of the facilities at a core thermal output of 2,560 megawatts (MWt). An increase in power to 2700 MWt was authorized by license amendments (References 1 and 2). Rated thermal power was once again increased to 2737 MWth as part of a measurement uncertainty recapture modification which was approved by Reference 3. Site parameters and major systems and components, including the engineered safety features (ESFs) and containment structures, have been evaluated for operation at the higher power level. The postulated incidents considered in Chapter 14 are also evaluated at the higher power level.

Over the time the plant has been operated, numerous modifications have been made. In some cases, these changes were submitted to, and approved by, the NRC through the license amendment process. In other cases, changes were implemented under the provisions of 10 CFR 50.59 with notification to the Commission after the fact.

Each revision to the Updated Final Safety Analysis Report is intended to reflect, within the limitations of the report format, the configuration and operation of the plant at the end of the refueling outage preceding the revision date, as required by 10 CFR 50.71. On the basis of the information presented in this FSAR and referenced material, Calvert Cliffs Nuclear Power Plant (CCNPP) concludes that CCNPP Units 1 and 2 were designed and constructed and are operated without undue risk to the health and safety of the public.

CALVERT CLIFFS UFSAR 1.1-2 Rev. 47 1.

1.1 REFERENCES

1. Letter from D. K. Davis (NRC) to A. E. Lundvall, Jr. (BGE), dated September 9, 1977, Amendment No. 24 to Facility Operating License No. DPR-53 for Unit No. 1 2. Letter from D. K. Davis (NRC) to A. E. Lundvall, Jr. (BGE), dated October 19, 1977, Amendment No. 9 to Facility Operating License No. DPR-69 for Unit No. 2 3. Letter from D. V. Pickett (NRC) to J. A. Spina (CCNPP), dated July 22, 2009, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 - Amendment Re: Measurement Uncertainty Recapture Power Uprate (TAC Nos. MD9554 and MD9555) (Amendment Nos. 291/267)

CALVERT CLIFFS UFSAR 1.3-1 Rev. 47 1.3 COMPARISON WITH OTHER PLANTS Table 1-1 presents a summary of the original design characteristics of the CCNPP. The table includes similar data for Maine Yankee Unit 1, Turkey Point Units 3 and 4 and Palisades Unit 1. Bechtel Power Corporation (Bechtel) and CE are identified as contractors in Section 1.8. The Palisades plant is included in the table because its coolant system is similar to that of Calvert Cliffs, and because both Bechtel and CE were Palisades contractors. Maine Yankee is selected because its core is similar to that of Calvert Cliffs and it is a plant of vintage similar to Calvert Cliffs with which CE is associated. In particular, the reactor regulating system, the reactor coolant pressure regulating system, and the pressurizer level regulating system are essentially identical in design and function to the Maine Yankee systems except in the adaptation to the lesser number of similar inputs for a two-loop plant. Turkey Point is included because it is another comparable plant with which Bechtel is associated.

CALVERT CLIFFS UFSAR 1.3-2 Rev. 47 TABLE 1-1 COMPARISON OF ORIGINAL PLANT CHARACTERISTICS(a) CALVERT CLIFFS UNITS 1 & 2 MAINE YANKEE(b) TURKEY POINT UNITS 3 & 4(b) PALISADES UNIT 1 HYDRAULIC AND THERMAL DESIGN PARAMETERS Total Core Heat Output, MWt 2560 2440 2200 2200 Total Core Heat Output, Btu/hr 8740x106 8328x106 7479x106 7509x106 Heat Generated in Fuel, % 97.5 97.5 97.4 97.5 Maximum Overpower, % 12 12 12 12 System Pressure, Nominal psia 2250 2250 2250 2100 System Pressure, Minimum Steady State, psia 2200 2200 2220 2050 Hot Channel Factors, Overall Heat Flux, Fq 3.00 2.89 3.23 3.80 Enthalphy Rise, FWH 1.65 1.62 1.77 2.51 DNB Ratio at Nominal Conditions 2.18 2.45 1.81 2.00 Coolant Flow Total Flow Rate, lb/hr 122x106 122x106 101.5x106 125x106 Effective Flow Rate for Heat Transfer, lb/hr 117.5x106 117.5x106 97.0x106 121.25x106 Effective Flow Rate for Heat Transfer, ft2 53.5 53.5 41.8 58.7 Average Velocity Along Fuel Rods, ft/sec 13.6 13.9 14.3 12.7 Average Mass Velocity, lb/hr-ft2 2.20x106 2.9x106 2.32x106 2.07x106 Coolant Temperatures, °F Nominal Inlet 543.5 538.9 546.2 545 Maximum Inlet due to Instrumentation Error and Deadband, °F 548 546 550.2 548 Average Rise in Vessel, °F 52 51.1 55.9 46 Coolant Temperatures, °F Average Rise in Core, °F 54 53.1 58.3 47 Average in Core, °F 570.4 565.4 575.4 568.5 Average in Vessel 569.5 564.4 574.2 568 Nominal Outlet of Hot Channel 643 636 642 642.8 Average Film Coefficient, Btu/hr-ft2-°F 5240 5300 5400 4860 Average Film Temperature Difference, °F 33.5 33 31.8 30 CALVERT CLIFFS UFSAR 1.3-3 Rev. 47 TABLE 1-1 COMPARISON OF ORIGINAL PLANT CHARACTERISTICS(a) CALVERT CLIFFS UNITS 1 & 2 MAINE YANKEE(b) TURKEY POINT UNITS 3 & 4(b) PALISADES UNIT 1 Heat Transfer at 100% Power Active Heat Transfer Surface Area, ft2 48,416 47,700 42,460 51,400 Average Heat Flux, Btu/hr-ft2 176,000 170,200 171,600 142,400 Maximum Heat Flux, Btu/hr-ft2 527,900 502,300 554,200 541,200 Average Thermal Output, kW/ft 5.94 5.74 5.5 4.63 Maximum Thermal Output, kW/ft 17.5 16.7 17.9 17.6(c) Maximum Clad Surface Temperature at Nominal Pressure, °F 657 657 657 648 Fuel Center Temperature, °F Maximum at 100% Power 3780 3640 4030 4040 Maximum at Over Power 4070 3940 4300 4350 CORE MECHANICAL DESIGN PARAMETERS Thermal Output, kW/ft at Maximum Over Power 19.6 18.7 20.0 19.7(c) Fuel Assemblies Design CEA CEA RCC Cruciform Rod Pitch, in. 0.58 0.580 0.563 0.550 Cross-section Dimensions, in. 7.98x7.98 7.98x7.98 0.563 8.1135x8.1135 Fuel Weight (as UO2), lbs 207,269 203,934 176,200 210,524 Total Weight, lbs 282,570 279,235 226,200 295,800 Number of Grids per Assembly 8 8 7 8 Fuel Rods Number 36,896 36,352 32,028 43,168 Outside Diameter, in. 0.440 0.440 0.422 0.4135 Diametral Gap, in. 0.0085 0.0085 0.0065 0.0065 Clad Thickness, in. 0.026 0.026 0.0243 0.022 Clad Material Zircaloy Zircaloy Zircaloy Zircaloy Fuel Pellets Material UO2 Sintered UO2 Sintered UO2 Sintered UO2 Sintered Diameter, in. 0.3795 0.3795 0.367 0.359 Length, in. 0.650 0.650 0.600 0.600 CALVERT CLIFFS UFSAR 1.3-4 Rev. 47 TABLE 1-1 COMPARISON OF ORIGINAL PLANT CHARACTERISTICS(a) CALVERT CLIFFS UNITS 1 & 2 MAINE YANKEE(b) TURKEY POINT UNITS 3 & 4(b) PALISADES UNIT 1 Control Assemblies Neutron Absorber B4C/SS/Cd-In-Ag B4C/SS/Cd-In-Ag Cd-In-Ag (5-15-80%) Cd-In-Ag (5-15-80%) Cladding Material Inconel Inconel 304 SSO Cold Worked Stainless Clad Thickness, in. 0.040 0.040 0.019 0.016 Control Assemblies Number of Assemblies, full/part length 77/8 77/8 53 4 1/4 Cruciform Rods Number of Rods per Assembly 5 5 20 117 Tubes per Rod Core Structure Core Barrel ID/OD, in. 148/149.75 148/149.75 133.875/137.875 149.75/152.5 Thermal Shield ID/OD, in. None 156/162 142.625/148.0 None NUCLEAR DESIGN DATA Structural Characteristics Core Diameter, inches (Equivalent) 136.0 136.0 119.5 136.71 Core Height, inches (Active Fuel) 136.7 136.7 144 132 Reflector Thickness & Composition Top - Water plus steel, in. 10 10 10 10 Bottom - Water plus steel, in. 10 10 10 10 Side - Water plus steel, in. 15 15 15 15 H2O/U, Unit Cell (Cold) 3.44 3.44 4.18 3.50 Number of Fuel Assemblies 217 217 157 204 UO2 Rods per Assembly, unshimmed/shimmed 204 212/208 Batch A 176 176 Batch B 164 160 Batch C (176/164/164) (176/164/160) Structural Characteristics Performance Characteristics 3 Batch Mixed 3 Batch Mixed 3 Regions 3 Batch Mixed Loading Technique Central Zone Central Zone Non-Uniform Central Zone Fuel Discharge Burnup, MWD/MTU Average First Cycle 13,775 13,795 13,000 10,180 First Core Average 22,550 30,000 24,500 17,600 CALVERT CLIFFS UFSAR 1.3-5 Rev. 47 TABLE 1-1 COMPARISON OF ORIGINAL PLANT CHARACTERISTICS(a) CALVERT CLIFFS UNITS 1 & 2 MAINE YANKEE(b) TURKEY POINT UNITS 3 & 4(b) PALISADES UNIT 1 Feed Enrichments wt% Region 1 2.05 2.01 1.85 1.65 Region 2 2.45 2.40 2.55 2.08/2.54 Region 3 2.99 2.95 3.10 2.54/3.20 Control Characteristics Effective Multiplication (beginning of life) Cold, No Power, Clean 1.194 1.170 1.180 1.212 Hot, No Power, Clean 1.152 1.129 1.38 1.175 Hot, Full Power, Xe Equilibrium 1.094 1.075 1.077 1.111 Control Assemblies Material B4C/SS-Cd-In-Ag B4C/SS-Cd-In-Ag Cd-In-Ag (5-15-80%) Cd-In-Ag (5-15-80%) Number of Control Assemblies 85 85 53 45 Cruciform Number of Absorber Rods per CEA (or RCC) Assembly 5 5 20 117 Tubes Welded to Form 13.5 in. Span Total Rod Worth (Hot), % 9.6 9.9 7 8.6 Boron Concentrations To shut reactor down with no rods inserted, clean, Cold/Hot, ppm 1120/1095 945/935 1250/1210 1180/1210 To control at power with no rods inserted, clean/equilibrium xenon, ppm 960/725 820/590 1000/670 1070/830 Kinetic Characteristics, Ranger Over Life Moderator Temperature Coefficient k/k/F -0.20x10-4 to -1.96x10-4 -0.04x10-4 to 2.20x10-4 +0.3x10-4 to -3.5x10-4 -0.08x10-4 to -2.25x10-4 Moderator Pressure Coefficient, /psi Hot, Operating Beginning-of-Life End-of-Cycle +0.3x10-6 to 2.6x10-6 +0.65x10-6 to +2.39x10-6 -0.3x10-6 to +3.4x10-6 +0.10x10-6 to +1.7x10-6 CALVERT CLIFFS UFSAR 1.3-6 Rev. 47 TABLE 1-1 COMPARISON OF ORIGINAL PLANT CHARACTERISTICS(a) CALVERT CLIFFS UNITS 1 & 2 MAINE YANKEE(b) TURKEY POINT UNITS 3 & 4(b) PALISADES UNIT 1 Moderator Pressure Coefficient, Void, /% Void Hot, Operating Beginning-of-Life -0.1x10-3 -0.41x10-3 to +0.5x10-3 to 0.06x10-3 to End-of-Cycle -1.3x10-3 1.43x10-3 -2.5x10-3 -1.0x10-3 Doppler Coefficient(d) -1.46x10-5 -1.45x10-5 -1.0x10-5 -1.56x10-5 k/k/F -1.06x10-5 -1.07x10-5 1.6x10-5 -1.46x10-5 REACTOR COOLANT SYSTEM - CODE REQUIREMENTS Reactor Vessel(f) ASME III Class A ASME III Class A ASME III Class A ASME III Class ASteam Generator Tube Side ASME III Class A ASME III Class A ASME III Class A ASME III Class AShell Side ASME III Class A ASME III Class A ASME III Class C ASME III Class APressurizer ASME III Class A ASME III Class A ASME III Class A ASME III Class CPressurizer Relief (or Quench) Tank ASME III Class C ASME III Class C ASME III Class C ASME III Class CPressurizer Safety Valves ASME III ASME III ASME III ASME III Reactor Coolant Piping ANSI B 31.7 ANSI B 31.1 ANSI B 31.1 ANSI B 31.1 PRINCIPAL DESIGN PARAMETERS OF THE REACTOR VESSEL Operating Pressure, psig 2235 2235 2235 2085 Reactor Inlet Temperature, °F 544.5 540 546.2 545 Reactor Outlet Temperature, °F 599.4 592.8 602.1 591.1 Number of Loops 2 3 3 2 Design Pressure, psig 2485 2485 2485 2485 Design Temperature, °F 650 650 650 650 Hydrostatic Test Pressure (cold), psig 3110 3110 3107 3110 Total Coolant Volume - cu.ft. 11,101 11,026 9,088 10,809 CALVERT CLIFFS UFSAR 1.3-7 Rev. 47 TABLE 1-1 COMPARISON OF ORIGINAL PLANT CHARACTERISTICS(a) CALVERT CLIFFS UNITS 1 & 2 MAINE YANKEE(b) TURKEY POINT UNITS 3 & 4(b) PALISADES UNIT 1 PRINCIPAL DESIGN PARAMETERS OF THE STEAM GENERATORS Material SA-533, Grade B, Class 1, low alloy steel, internally clad with Type 304 austerities SS equivalent SA-533, Grade E Class 1 steel, forgings-A-508-64 Class 2, cladding-weld deposited 304 SS equivalent SA-302, Grade B, low alloy steel internally clad with type 304 austerities SS equivalent SA-302, Grade B low alloy steel internally clad with Type 304 austerities SS equivalent Design Pressure, psig 2485 2485 2485 2485 Design Temperature, °F 650 650 650 650 Operating Pressure, psig 2235 2235 2235 2085 Inside Diameter of Shell, in. 172 172 155.5 172 Outside Diameter Across Nozzles, in. 253 266-5/8 236 254 Overall Height of Vessel and Enclosure Head to Top of CRDM Nozzle, ft.-in. 41-11-3/4 42-1-3/8 41-6 40-1-13/16 Minimum Clad Thickness, in. 1/8 1/8 5/32 3/16 Number of Units 2 3 3 2 Type Vertical U-Tube with integral moisture separator Vertical U-Tube with integral moisture separator Vertical U-Tube with integral moisture separator Vertical U-Tube with integral moisture separator Tube Material Inconel Inconel Inconel Inconel Shell Material SA-533, Gr. B, Class 1 and SA-516 Gr 70 SA-533, Gr. B, Class 1 and SA-516 Gr 70 Carbon Steel Carbon Steel PRINCIPAL DESIGN PARAMETERS OF THE REACTOR COOLANT PUMPS Tube Side Design Pressure, psig 2485 2485 2485 2485 Tube Side Design Temperature, °F 650 650 650 650 Tube Side Design Flow, lb/hr 61x106 40.67x106 33.93x106 62.5x106 Shell Side Design Pressure, psig 985 985 1085 985 Shell Side Design Temperature, °F 550 550 556 550 Operating Pressure, Tube Side, Nominal, psig 2235 2235 2235 2085 Operating Pressure, Shell Side, Maximum, psig 885 885 1020 885 Maximum Moisture at Outlet at Full Load, % 0.2 0.2 1/4 0.2 CALVERT CLIFFS UFSAR 1.3-8 Rev. 47 TABLE 1-1 COMPARISON OF ORIGINAL PLANT CHARACTERISTICS(a) CALVERT CLIFFS UNITS 1 & 2 MAINE YANKEE(b) TURKEY POINT UNITS 3 & 4(b) PALISADES UNIT 1 Hydrostatic Test Pressure, Tube Side (cold), psig 3110 3110 3107 3110 Steam Pressure, psia, at full power 850 815 745 770 Steam Temperature, °F, at full power 525.2 520.3 510 513.8 Number of Units 4 3 3 4 Type Vertical, single stage centrifugal with bottom suction and horizontal discharge Vertical, single stage centrifugal with bottom suction and horizontal discharge Vertical, single stage radial flow with bottom suction and horizontal discharge Vertical, single stage radial flow with bottom suction and horizontal discharge Design Pressure, psig 2485 2485 2485 2485 Design Temperature, °F 650 650 650 650 Operating Pressure, nominal psig 2235 2235 2235 2085 Suction Temperature, °F 543.4 538.9 546.5 545 Design Capacity, gpm 81,200 108,000 89,500 83,000 Design Head, ft. 300 290 260 260 Hydrostatic Test Pressure, (cold), psig 3110 3110 3107 3110 Motor Type A-C Induction Single Speed A-C Induction Single Speed A-C Induction Single Speed A-C Induction Single Speed Motor Rating, hp 7200 (cold) 9000 6000 6250 (cold) PRINCIPAL DESIGN PARAMETERS OF THE REACTOR COOLANT PIPING Material SA516-gr 70 with SS clad SA516-gr 70 with SS clad Austenitic SS SA516-gr 70 clad with SS Hot Leg - ID, in. 42 33.5 29 42 Cold Leg - ID, in. 30 33.5 27 1/2 30 Between Pump & Steam Generator - ID, in. 30 33.5 31 30 CALVERT CLIFFS UFSAR 1.3-9 Rev. 47 TABLE 1-1 COMPARISON OF ORIGINAL PLANT CHARACTERISTICS(a) CALVERT CLIFFS UNITS 1 & 2 MAINE YANKEE(b) TURKEY POINT UNITS 3 & 4(b) PALISADES UNIT 1 CONTAINMENT SYSTEM PARAMETERS UNIT 1 Type Steel-lined, pre-stressed post tensioned concrete cylinder, curved dome roof Steel-lined, reinforced concrete flat bottom and hemispherical dome Steel-lined pre-stressed post tensioned concrete cylinder, hemispherical domed roof Steel-lined pre-stressed post tensioned concrete cylinder, hemis-cylinder curved dome roof Design Parameters Inside Diameter, ft. 130 135 116 116 Height, ft. 181-2/3 169-1/2 169 190 Free Volume, ft3 2,000,000 1,855,000 1,550,000 1,600,000 Reference Incident Pressure, psig 50 55 59 55 Concrete Thickness, ft. Vertical Wall 3-3/4 4-1/2 3-3/4 3 Dome 3-1/4 2-1/2 3-1/4 2-1/2 Containment Leakage Prevention & Mitigation Systems Leak-tight pene-tration & continuous steel liner. Automatic isolation where required. The exhaust from penetration rooms to vent Leak-tight pene-tration & continuous steel liner. Automatic isolation where required. Leak-tight pene-tration & continuous steel liner. Automatic isolation where required. Leak-tight pene-tration & continuous steel liner. Automatic isolation where required. Gaseous Effluent Purge Discharge through vent Discharge through stack Through particulate filter and monitors part of main exhaust system. Discharge through stack CALVERT CLIFFS UFSAR 1.3-10 Rev. 47 TABLE 1-1 COMPARISON OF ORIGINAL PLANT CHARACTERISTICS(a) CALVERT CLIFFS UNITS 1 & 2 MAINE YANKEE(b) TURKEY POINT UNITS 3 & 4(b) PALISADES UNIT 1 ENGINEERED SAFETY FEATURES Safety Injection System No. of High Head Pumps 3 3 (Charging) 4 (Shared) 3 No. of Low Head Pumps 2 2 2 2 Containment Fan Coolers No. of Units 4 6 3 3 Air Flow Capacity, each at emergency condition, cfm 55,000 Not Applicable 25,000 25,000 Containment Spray No. of Pumps 2 3 2 2 Emergency Power Diesel Generator Units 3 total for both units(e) 2 2 total for both units 4 Safety Injection Tanks, Number 4 3 3 4 _______________________ (a) The current design characteristics for Calvert Cliffs Units 1 and 2 may differ from those shown in this table. (b) The values listed for these plants were taken from public documentation. (c) Based on total heat output of the core rather than heat generated in the fuel alone. (d) Values shown are for hot, zero power/beginning-of-life, full power conditions. (e) The current design characteristics show four total emergency power diesel generators for both units. (f) See Table 4-9 for the design code of the replacement RVCH. CALVERT CLIFFS UFSAR 1.4-1 Rev. 47 1.4 PRINCIPAL ARCHITECTURAL AND ENGINEERING CRITERIA FOR DESIGN The principal architectural and engineering criteria for design of the plant are summarized below. 1.4.1 PLANT DESIGN Principal structures and equipment which may serve either to prevent incidents or to mitigate their consequences are designed, fabricated and erected in accordance with applicable codes to withstand the most severe earthquakes, flooding conditions, windstorms, ice conditions, temperature, and other deleterious natural phenomena which could be reasonably assumed to occur at the site during plant lifetime. Units 1 and 2 are sufficiently independent so that the safety of one unit will not be impaired in the unlikely event of an incident in the other unit. Principal structures and equipment are sized for the maximum expected NSSS and turbine outputs. Redundancy is provided in reactor and safety systems so that no single failure of any active component of the systems can prevent the action necessary to avoid an unsafe condition. The plant is designed to facilitate inspection and testing of systems and components whose reliability is important to the protection of the public and plant personnel. Provisions are made to protect against the hazards of such events as fires or explosions.

Systems and components which are significant from the standpoint of nuclear safety are designed, fabricated, and erected to quality standards commensurate with the safety function to be performed. 1.4.2 REACTOR The following apply to the reactor of either unit: a. The reactor is of the pressurized water type, designed to produce steam to drive a turbine generator. The reactor was initially licensed and operated at a core thermal output of 2,560 MWt; the license was later amended and the reactor now operates at 2,737 MWt. b. The reactor is fueled with slightly enriched uranium dioxide contained in Zircaloy, ZIRLO, or M5 tubes. c. Minimum departure from nucleate boiling ratio (DNBR) during normal operation and anticipated transients will not be below that value which could lead to fuel rod failure. The maximum fuel center line temperature evaluated at the design overpower condition will be below that value which could lead to fuel rod failure. The melting point of the UO2 will not be reached during normal operation and anticipated transients. d. Fuel rod clad is designed to maintain cladding integrity throughout fuel life. Fission gas release within the rods and other factors affecting design life are considered for the maximum expected exposures. e. The reactor and control systems are designed so that any xenon transients will be adequately damped. f. The reactor is designed to accommodate the anticipated transients safely and without fuel damage. g. The RCS is designed and constructed to maintain its integrity throughout expected plant life. Appropriate means of test and inspection are provided. CALVERT CLIFFS UFSAR 1.4-2 Rev. 47 h. Power excursions which could result from any credible reactivity addition incident will not cause damage to the pressure vessel either by deformation or rupture, or impair operation of the ESF. i. Control element assemblies are capable of holding the core subcritical at hot zero power conditions with adequate margin following a trip, even with the most reactive rod stuck in the fully withdrawn position. j. The CVCS is capable of adding boric acid to the reactor coolant at a rate sufficient to maintain an adequate shutdown margin during maximum design rate RCS cooldown following a reactor trip. The system is independent of the CEA system. k. The combined response of the fuel temperature coefficient, the moderator temperature coefficient (MTC), the moderator void coefficient, and the moderator pressure coefficient to an increase in reactor thermal power is a decrease in reactivity. In addition, the reactor power transient remains bounded and damped in response to any expected changes in any operating variable. l. Automatic and redundant reactor trips are provided to prevent anticipated plant transients from producing fuel or clad damage. 1.4.3 REACTOR COOLANT AND AUXILIARY SYSTEMS Heat removal systems are provided which can safely accommodate core heat output. Each of these heat removal systems is designed to provide reliable operation under all normal and expected transient circumstances. 1.4.4 CONTAINMENT STRUCTURE The Containment Structure, including the associated access openings and penetrations, is designed to contain the pressures and temperatures resulting from a LOCA in which the following occur: a. The total energy contained in the RCS water is assumed to be released into the containment through a double-ended break of one of the reactor coolant pipes adjacent to the reactor vessel outlet nozzle. b. External electric power is lost simultaneously. c. Heat is transferred from the reactor to the containment by water supplied from the Safety Injection System. d. Either the containment air recirculation subsystem or the containment spray subsystem functions. e. The containment ESF do not operate until 30 seconds following the incident. Means are provided for pressure and leak rate testing of the entire containment system including provisions for leak rate testing of individual piping and electrical penetrations that rely on gasketed seals, sealing compounds, or expansion bellows. 1.4.5 ENGINEERED SAFETY FEATURES The design for either unit incorporates redundant ESF systems. These, in conjunction with the containment systems, ensure that the release of fission products, following any credible LOCA, will not exceed the guidelines set forth in 10 CFR 50.67. The ESF systems include: (a) independent systems, each with redundant features, to remove heat from the Containment Structure in order to reduce containment pressure; (b) a Safety Injection System to limit fuel and cladding damage to an amount which would not interfere with adequate emergency core cooling (ECC) and to limit metal-water CALVERT CLIFFS UFSAR 1.4-3 Rev. 47 reactions to negligible amounts; and (c) a system to remove radioactive iodine for the post-incident containment atmosphere. The ESF are designed for all break sizes in the RCS piping up to and including the double-ended rupture of the largest reactor coolant pipe. 1.4.6 PROTECTION, CONTROL, AND INSTRUMENTATION SYSTEMS Interlocks and automatic protective systems are provided along with administrative controls to insure safe operation of the plant. An RPS is provided which initiates reactor trip if the reactor approaches an unsafe condition.

Sufficient redundancy is installed to permit periodic testing of the RPS so that failure or removal from service of any one protective system component or portion of the system will not preclude reactor trip or other safety action when required. 1.4.7 ELECTRICAL SYSTEMS Normal, standby, and emergency sources of auxiliary power are provided to assure both the safe and orderly shutdown of the plant and the ability to maintain a safe shutdown condition under all credible circumstances. 1.4.8 WASTE PROCESSING AND RADIATION PROTECTION The waste treatment systems are designed so that the discharge of radioactivity to the environment is in accordance with the requirements of 10 CFR Part 20. The plant is provided with a centralized Control Room having adequate shielding to permit occupancy during all credible accident conditions.

The radiation shielding in the plant and the radiation control procedures ensure that operating personnel do not receive radiation exposures in excess of the applicable limits of 10 CFR Part 20 during normal operation and maintenance. 1.4.9 FUEL HANDLING AND STORAGE Fuel handling and storage facilities are provided for the safe handling, storage and shipment of fuel and will preclude accidental criticality.

1.4.10 NIL DUCTILITY TRANSITION TEMPERATURE Components of the RCS are designed and will be operated so that no deleterious pressure or thermal stress will be imposed on the structural materials. Consideration is given to the ductile characteristics of the materials at low temperature.

1.4.11 FIELD RUNNING OF 2" AND SMALLER DIAMETER PIPE All 2" and smaller piping with the exception of portions of the charging, letdown, and some few branch connections tied into the Safety Injection System (these systems are classified in ANSI B 31.7 Class I and are routed by the engineering office) were field-run during plant construction and initial modification work. All piping for field-run essential systems (Section 1.8.1, II.E.4.2), including all ESF, were routed by experienced piping designers. This piping was routed and support points selected in accordance with the field installation manual. The spacing of supports and type of support used are such that the combination of stresses due to thermal, dead load, and seismic does not exceed the allowable stresses. Prior to installation, all field piping isometric drawings were routed to the engineering office for CALVERT CLIFFS UFSAR 1.4-4 Rev. 47 comments. After installation and before start-up, surveillance by the field quality assurance personnel and by the respective engineering specialist group was performed to ensure that all piping is installed per the design drawings. Prior to start-up, each system was checked off showing that all hangers and supports are located as designed.

During start-up, each system was observed under conditions which simulate operating conditions. This included the starting and stopping of pumps and opening and closing of valves.

CALVERT CLIFFS UFSAR 1.5-1 Rev. 47 1.5 RESEARCH AND DEVELOPMENT REQUIREMENTS 1.

5.1 INTRODUCTION

The design of the Calvert Cliffs Nuclear Plant is based upon concepts which have been successfully applied in the design of pressurized water reactor power plants. However, certain programs of theoretical analysis or experimentation (constituting "research and development" as defined in the Atomic Energy Act, as amended, and in AEC or NRC regulations) have been undertaken to aid in plant design and to verify the performance characteristics of plant components and systems. This section describes the results and status of those analytical and test programs which were conducted or in progress at the time of operating license application including experimental production and testing of models, devices, equipment and materials. No attempt is made to include all pertinent research and development programs conducted since operating license application.

In carrying out these programs, information which is derived from research and development activities of the AEC or NRC and other organization in the nuclear industry has been taken into account. 1.5.2 CEA TESTING 1.5.2.1 Critical Experiments An experimental program was completed to confirm techniques for calculating CEA worth and local nuclear peaking associated with the fuel assembly design. The work was performed in the CRX facility of the Westinghouse Reactor Evaluation Center at Waltz Mill, Pennsylvania, between June and August 1967. The basic core configuration was a 30x30 square array of Zr-4 clad UO2 fuel rods with an enrichment of about 2.7 wt% U-235; fuel rods were removed to create internal water holes or channels to accommodate absorber elements. The experiments demonstrated that the current CE methods of analysis (Section 3.0) accurately predict the CEA worth and local peaking in the small critical assemblies. This lends support to the use of these methods in the design of the Calvert Cliffs core. The significant conclusions which were drawn on the basis of the test program include the following: a. The standard CE design methods are capable of calculating clean, room temperature, critical lattices (lattices which contain no CEAs, water slots, or other heterogeneities) to an accuracy of within 0.03% reactivity on the average; b. The worths of various arrays of cylindrical absorbers containing boron carbide were predicted within 2% of the worth on the average, with errors for individual cases ranging from +6% to -2.2% of the worth. The arrays had worths ranging from 6 to 8% reactivity; c. In assemblies containing water holes, the calculated and measured power peaking agree within 2%. Occasional differences between calculated and measured power of up to 4% are seen, but only in fuel rods of low power, usually near the reflector of these highly buckled, small cores. 1.5.2.2 Mechanical Testing A series of tests were completed on single and dual CEAs to satisfy the following objectives: a. To determine the mechanical and functional feasibility of the CEA concept; CALVERT CLIFFS UFSAR 1.5-2 Rev. 47 b. To experimentally determine the relationship between CEA drop time and CEA drop weight, annular clearance between CEA fingers and guide tubes, and coolant flowrate within the guide tube; c. To experimentally determine the relationship between flowrate and pressure drop within the guide tube as a function of CEA axial position and of finger-to-guide tube clearance; d. To determine the effects on drop time of adding a flow restriction or of plugging the lower end of a guide tube (as might occur under accident conditions); e. To determine the effects of misalignment within the CEA guide tube system on drop time. Both types of CEAs were tested by operation for over 1000 hours at reactor operating temperature, pressure, and water chemistry with flows in excess of those anticipated in the reactor. This test and other tests demonstrated that the five-finger CEA concept is mechanically and functionally feasible and that the CEA meets criteria established for drop time under the most adverse condition. Both types of CEAs were examined after conclusion of these tests and no significant wear was observed. At no point were there any wear marks in excess of .001" in depth. The testing also verified that the analytical model used for predicting drop time gives uniformly conservative results. The effects on drop time of all possible combinations of frictional restraining forces in the CEDM, angular and radial misalignment of the CEDM, misalignments between the CEA and the guide tubes of as much as 0.4", misalignments of the CEA was experimentally investigated and defined. The conditions tested simulated all the effects of tolerance buildup, dynamic loadings, and thermal effects. The tests demonstrated that misalignments and distortions in excess of those expected from tolerance buildup or any other anticipated cause would still result in acceptable drop times. The results of the cold CEA testing were analytically extended into the hot operating range. Further testing of a design verification nature was performed scheduled on a complete CEDM-CEA system in the hot test facility at operating plant temperature, pressure, and flow. 1.5.3 CONTROL ELEMENT DRIVE MECHANISM TESTING The development of the magnetic jack CEDM was carried out over a period of approximately three years. As such, it was a closely coupled program consisting of design, testing and fabrication. The early design effort resulted in a prototype which was tested extensively and was modified during the testing process to improve operation. The results of this early testing led to the design and fabrication of a second improved prototype mechanism. This mechanism was again tested extensively in a manner similar to the first, and from this series of tests coupled with design modifications that Maine Yankee prototype mechanism resulted. The specifications for the Calvert Cliffs CEDM was similar to that used for the construction of the Maine Yankee mechanisms. The following design requirements were imposed on the CEDMs with regard to operation under severe service conditions: a. The pressure housing of the CEDM, for the replacement RVCH, is designed for service as a Class 1 appurtenance and code stamped for CALVERT CLIFFS UFSAR 1.5-3 Rev. 47 service at 2,500 psia and 650°F. Normal CEDM operating conditions are 2,250 psia and 608°F. b. The CEDMs are designed to withstand the combined mechanical loads associated with normal and abnormal operating pressures, temperatures and transients, plus the maximum earthquake, with no loss of function. c. When DBE reactor coolant pipe rupture loads are added to the loads given in (b) above, the CEDM need not function normally during the application of the DBE loads, but must prevent ejection of its CEA from the core and be capable of normal operation following the DBE. d. The CEDM is capable of performing its normal function after an inactive period of one month with the CEDM in the hold mode and the plant at normal pressure and temperature. e. Under the ambient conditions inside the containment building following a postulated main pipe rupture (DBE), the CEDM is capable of driving the CEA to fully inserted position from the fully withdrawn position, and transmitting all position indication signals for 15 minutes after rupture occurs. f. The CEDM is capable of withstanding complete loss of cooling service for a four-hour period with the plant at normal operating temperature and pressure. CEDM operation under these conditions is restricted to the "HOLD" and "SCRAM" modes. Upon restoration of cooling service the CEDM is capable of normal operation. The CEDM must be capable of passing the following tests: a. Single CEA - Accelerated life tests of at least 30,000' of travel and 200 full-height gravity drops at simulated reactor operating conditions and ambient external conditions. b. Dual CEA - Accelerated life test of at least 15,000' of travel and 200 full-height gravity drop tests at simulated reactor operating conditions and ambient external conditions. After installation and prior to operation, each CEDM was tested in the field to ascertain that the system, as constructed, meets all of the design requirements, as discussed in Section 13.1. 1.5.4 FUEL ASSEMBLY DESIGN 1.5.4.1 Prototype Tests Full size Zircaloy fuel assemblies filled with depleted UO2 have undergone high temperature (600°F), high pressure (2,250 psig) full flow tests in typical reactor chemistry coolant. These tests were performed in connection with a full-size CEA and a CEDM.

Full-size prototype fuel assemblies loaded with depleted UO2 were subjected to mechanical testing to evaluate their reaction to applied loads. Axial and lateral loading of assemblies supported in air between simulated upper and lower support plates, as well as free end twisting and lateral motion typical of refueling operation, were performed. Previous testing of a similar nature on CE fuel assemblies had shown them to be stable and mechanically sound for all expected reactor operating, casualty and refueling conditions. These tests were repeated for the Calvert Cliffs type fuel assemblies to verify that no significant mechanical characteristics changed. CALVERT CLIFFS UFSAR 1.5-4 Rev. 47 In 1966, a series of single-phase tests on coolant turbulent mixing was run on a "prototype" fuel assembly which was geometrically similar to the Palisades assembly. The model enabled determination of flow resistances and vertical subchannel flow rates using pressure instrumentation and the average level of eddy flow using dye-injection and sampling equipment. The tests yielded the value of inverse Peclet number characteristic of eddy flow (0.00366). The value was shown during the course of the tests to be insensitive to coolant temperature and to vertical coolant mass velocity. The design value of the inverse Peclet Number was established at 0.0035 on the basis of the experimental results. As part of a CE-sponsored research and development program, a series of single phase dye injection mixing tests were conducted in 1968. The tests were performed on a model of a portion of a CEA-type fuel assembly which was sufficiently instrumented to enable measurement (via a data reduction computer program) of the individual lateral flows across the boundaries of twelve subchannels of the model. Although these tests were not intended for that purpose, some of the test results could be used to determine the average level of turbulent mixing in the reference design assembly. The inverse Peclet number calculated from the average of 56 individual turbulent mixing flows (two for each subchannel boundary) obtained from the applicable data was 0.0034. With respect to general turbulent mixing, therefore, the more recent study on the CEA assembly verifies the constancy of the inverse Peclet number for moderately different fuel assembly geometries and confirms the design value of that characteristic. 1.5.4.2 Assembly Flow Distribution Tests Velocity and static pressure measurements were made in an oversize model of a CEA fuel assembly in order to determine the flow distributions present in that geometry. The effect of the distributions on thermal behavior and margin were evaluated, where necessary, with the use of CE's CORAL code, which is an extensively revised version of the COBRA thermal and hydraulic code. Subjects investigated include the following: a. Assembly inlet flow distribution, as affected by the core support plate and lower end fitting flow hole geometry. The flow distribution was measured and indicated that the desirable uniform condition is achieved within 10% of core height. The effect of the initial non-uniform condition on thermal behavior was analyzed; b. Assembly inlet flow distribution as affected by a blocked core support plate flow hole. The flow distribution was measured and indicated that flow has recovered to at least 50% of the uniform nominal value at an elevation corresponding to 10% of core height. The effect of the non-uniform flow pattern on thermal behavior was analyzed; c. Flow distribution within the assembly, as affected by complete blockage of one to nine subchannels. The flow distributions were measured and indicated very little upstream effect of such blockage, followed by recovery to normal subchannel flow conditions within 10 to 15% of core height, depending upon the number of subchannels blocked; d. Flow distribution below the upper end fitting as affected by the upper end fitting and fuel bundle alignment plate flow hole geometry and by the presence of the CEA shroud. Measurement of the flow pattern in the absence of the shroud showed no appreciable upstream effect of the flow holes in the active core region. CALVERT CLIFFS UFSAR 1.5-5 Rev. 47 1.5.4.3 DNB Testing on the Mark V CEA Fuel Assembly, 1969-1970 In 1968, CE initiated a series of tests at Columbia University on the departure from nucleate boiling (DNB) phenomenon. One purpose of the tests was to obtain experimental DNB data for verifying the combined accuracy of the thermal and hydraulic COSMO design code and the empirical W-3 DNB correlation in predicting the DNB condition for the CEA fuel assembly.

The tests were conducted on a nine-foot long exact scale portion of a Mark Volt CEA fuel assembly, consisting of one guide tube and 21 electrically-heated "fuel" rods arranged five-by-five. There were three distinct test sections, one with a 7' heated length and a uniform lateral power distribution, and one with a 4' heated length and a non-uniform lateral power distribution. The axial power distribution was uniform for all test sections. Test conditions comprised a coolant inlet temperature range of 450°F to 650°F, a mass velocity range of 1x106 to 3x106 lb/hr-ft2 and a system pressure range of 1,500 to 2,200 psia. Approximately 90 data points were obtained from all three test sections. The COSMO/W-3 combination was used for predicting the corresponding Critical Heat Flux values for the experimental conditions. The measure of the accuracy of prediction was defined as the average value of the ratio of experimental to predicted Critical Heat Flux. The value was 0.983 with a sample standard deviation of 0.58; these compare satisfactorily with corresponding values in the literature. The result implies that COSMO and W-3 are acceptable by present standards for describing DNB in the CEA geometry. The remainder of the DNB analytical and experimental program was devoted in part to further aspects of predicting DNB for the reference design assembly. This program was comprised of: a. Refinement of the W-3 correlation or development of a new correlation to reduce the statistical error attendant on the prediction; b. Investigation of the case of the small systematic deviation; c. Investigation of DNB behavior over a wider range of system pressure and flow conditions. 1.5.4.4 Dynamic Loop and Vibration Testing Considerable testing was performed to evaluate the effects of assembly and fuel rod vibration or fretting. Dynamic loop testing under simulated reactor operating conditions and mechanically-induced autoclave vibration tests were carried out. Over 18,000 hours of test time were accumulated on subsize assemblies and over 14,000 hours on full-size test assemblies in dynamic test loops. Test conditions duplicated reactor temperature, water chemistry, pressure, and flow velocity. Intentional cross flow and forced bundle vibrations were used to accentuate any vibration between the fuel rods and the spacer grids. In addition, the spacer grid spring tabs were individually set to simulate relaxed spring conditions. In addition to the dynamic loop tests, forced vibration tests were also performed. Fuel rods supported by spacer grids were vibrated at various frequencies and amplitudes. The tests were conducted in a static autoclave at operating pressure and temperature. Test variables included, in addition to the vibration frequency and amplitude, spring preset of the spacer grids, and time under test. The spring tabs were varied from design interference fits to gaps. These tests did not CALVERT CLIFFS UFSAR 1.5-6 Rev. 47 reproduce reactor flow conditions but were designed to develop trends in the degree of fretting as a function of the test variables. Even under unreasonably severe conditions (i.e., high frequencies, large amplitudes, and gaps between the grid spring and fuel tube) no serious fretting was observed in these tests. 1.5.5 MODERATOR TEMPERATURE COEFFICIENT Analytical studies were completed as part of the detailed plant design to define the least negative MTC for the Calvert Cliffs reactor. The factors which affect the MTC are discussed in Section 3.0 of this report.

Analyses of MTC for the Connecticut Yankee reactor compared with measurements made during the course of the start-up experiments are shown Table 1-2. It will be observed from the data that the measured coefficient is at most 0.16x10-4 /°F more positive than the calculated value. This good agreement lends confidence in the ability of the methods used to predict MTCs.

1.5.6 FUEL ROD CLADDING A substantial amount of information was generated in the course of CE's continuing test program on Zr-4 cladding.

Creep collapse tests on unsupported Zr-4 specimens with t/OD ratios of between 0.050 and 0.071 were performed at 650°F and 750°F. All tests were performed at an external pressure of 2400 psia. Results of tests performed at 750°F show that specimens with a t/OD of 0.059 (the reference design of Calvert Cliffs) collapse between 100 and 1000 hours at this temperature. Tests conducted at 650°F show collapse will occur between 5000 and approximately 30,000 hours. Zircaloy-4 specimens supported with plenum springs and mandrels with machined defects to simulate chipping and separated pellets were creep collapse tested at 750°F. Zircaloy-4 specimens with plenum springs were creep collapse tested at 650°F. Tests at 650°F after 19,000 hours showed no indications of the cladding deforming into the spaces between the springs. Specimens tested at 750°F accumulated in excess of 10,000 hours and the cladding showed some deformation into the intentional defects. Grooves 3/16" wide representing separated pellets caused a maximum of 2.0 mils deformation of the clad into the groove. Grooves 1/16" wide showed no measurable deformation after 10,000 hours at 750°F. Long-term corrosion tests were performed on Zr-4 fuel cladding under simulated reactor coolant conditions. Results of weight gain and hydrogen pickup were evaluated with respect to results obtained using demineralized water. The tests were conducted in coolant which had approximately 1100 ppm boron and less than 10 ppm NH4OH added for pH control. Results after 12,000 hours of test at 650°F showed no effect of the coolant additives on corrosion rates or hydrogen pickup over similar tests conducted in demineralized water. Specimens tested included as-received Zr-4 tubing and 750°F steam autoclaved samples. Similar tests performed using LiOH as a pH control additive accumulated in excess of 8000 hours. Under the test conditions, the corrosion characteristics exhibited by the material were equivalent to those observed in demineralized water. Dynamic corrosion tests were also conducted at 600°F to determine the effect of contamination on non-autoclaved material corrosion rates. Samples intentionally contaminated with dilute acid and machine oils were tested. Results after 4,300 hours of CALVERT CLIFFS UFSAR 1.5-7 Rev. 47 tests showed no deleterious effect of these conditions on the corrosion behavior of the Zr-4. Samples tested included as-received and 750°F steam autoclaved Zr-4 tubing. In addition to the corrosion studies mentioned above, CE participated in two studies being conducted by American Society for Testing and Materials (ASTM).

The first study, "Corrosion Testing Zr-4 and its Effect on Hydrogen Absorption," was performed by the ASTM G Committee (Corrosion of Metals), Subcommittee No. VIII (Corrosion of Zirconium in Water Systems).

The second study, "Task Force on Hydride Orientation," was performed by the ASTM B10 Committee (Reactive and Refractory Metals and Alloys), Subcommittee No. II (Zirconium and Hafnium). Areas of concern to the committee included methods of hydriding and effect of fuel tubing fabrication on platelet orientation.

Calvert Cliffs began phasing in the Westinghouse ZIRLO cladding starting with Unit 1 Cycle 16 (Batch 1V). The AREVA M5 cladding is first used beginning with Unit 2 Cycle 19 (Batch 2Z) and Unit 1 Cycle 21 (Batch AB). 1.5.7 REACTOR VESSEL FLOW TESTS Tests were conducted with one-fifth scale models of CE reactors to determine hydraulic performance. The first tests were performed for the Palisades plant which has a RCS similar to that of Calvert Cliffs. The test investigated flow distribution, pressure drop and the tracing of flow paths within the vessel for all four pumps operating and various part-loop configurations. Air was used as the test medium.

Similar one-fifth scale model tests were performed for Maine Yankee, which has a core similar to that of Calvert Cliffs. These tests were conducted in a cold water loop. All components for the model were geometrically similar to those in the reactor except for the core where 217 cylindrical core tubes were substituted for the fuel bundles. The core tubes contained orifices to provide the proper axial flow resistance.

Combustion Engineering, Inc. also conducted tests on a one-fourth scale model of the Fort Calhoun reactor using air as the test medium. Flow characteristics for Calvert Cliffs were determined by taking into consideration similarities between Calvert Cliffs and other CE reactors, in conjunction with the experimental data from the flow model programs. 1.5.8 INCORE INSTRUMENTATION TESTS Tests on incore thermocouples and flux detectors were performed to insure that the instrumentation will perform as expected at the temperature to be encountered and that it does not excessively vibrate and cause excessive wear or fretting. Cold flow testing was completed; no adverse vibrations or wear effects were encountered. Hot flow testing was also completed; after 2,000 hours at 590°F and 2,100 psig in a test loop, no breach of mechanical integrity was observed. Mechanical tests of the insertion and removal equipment and instrumentation were performed to determine the necessary forms and procedures. The top entry incore instrumentation design provides a means of eliminating the need for handling instrument assemblies separately, thus minimizing down-time and personnel exposure. A full scale CALVERT CLIFFS UFSAR 1.5-8 Rev. 47 mock-up was built to accommodate three incore instrumentation thimble assemblies. Major components and subassemblies of the mock-up included: a. An incore instrumentation test assembly, including the upper guide structure support plate, three thimble guide sleeves, fuel alignment plate, three fuel bundle guide tubes, and the core support plate. b. A thimble assembly consisting of the instrument plate, three incore instrumentation thimbles and the lifting sling. c. An upper guide tube, with the guide tube attached to the thimble extension and the detector cable partially inserted in the guide tube. Insertion and withdrawal tests were performed to determine the frictional forces of a multi-tube instrument thimble assembly during insertion and withdrawal from a set of fuel bundles. This test simulated the operation that will be performed during the refueling of the reactor. To determine whether jamming of the thimbles would occur during this operation, bending loads were applied to the thimble assembly by tilting the instrument plate 0.5° increments up to a total of 5° from horizontal. Guide tubes were filled with water. The assembly was raised and lowered approximately five times for each tilt setting. Results showed no discernible difference in the friction forces for the various tilt settings, however, the friction forces varied during withdrawal and insertion, reaching a maximum value of 8 lbs. The tests demonstrated that the repeated insertion and withdrawal of incore instrumentation thimble assemblies into the fuel bundle guides can be accomplished with reasonable insertion forces.

Life cycle tests were performed to determine if the frictional forces increase as a result of 40 insertions and withdrawals. An automatic timer was installed in the crane electrical circuitry to automatically cycle the thimble assembly between the fully inserted and withdrawn position. The instrument plate was set for 5° tilt and the assembly was cycled 60 times. The insertion and withdrawal forces were measured during the first and last five cycles. No discernible difference was noticed. An off-center lift test was performed to determine if the thimble assembly can be withdrawn from the core region while lifting the assembly from an extreme off-center position. For a lifting point 11" off-center, insertion was accomplished without incident. The flexibility of the thimble is such that jamming of the assembly due to off-center lifting does not occur.

Cable insertion tests were performed to determine forces required to completely insert and withdraw a detector cable from the incore instrumentation thimble assembly. The guide tube routing included several 5" radius bends. The detector cable was passed through the wet guide tubing and into the thimble. For 540° of 5" radius bends, an insertion force of 15 lbs and a withdrawal force of 37 lbs was required. This force is reasonable for hand insertion.

1.5.9 MATERIALS IRRADIATION SURVEILLANCE Surveillance specimens of the reactor vessel shell section material are installed on the inside wall of the vessel to monitor the Nil Ductility Transition Temperature (NDTT) of the material during reactor operating lifetime. Details of the program are given in Section 4.1.5. CALVERT CLIFFS UFSAR 1.5-9 Rev. 47 TABLE 1-2 ANALYSIS OF MODERATOR TEMPERATURE COEFFICIENTS IN THE CONNECTICUT YANKEE REACTOR AT START-OF-LIFE REACTOR TEMPERATURE (°F) DISSOLVED BORON CONCENTRATION (ppm) ROD WORTH INSERTED (% ) MODERATOR TEMPERATURE COEFFICIENT (10-4/°F) CALCULATED MEASURED 260 2040 - 0.46 0.57 560 2305 - 0.84 1.00 551 2045 1.8 0.37 0.47 561 1730 4.5 -0.23 -0.25 551 1610 5.6 -0.30 -0.30 CALVERT CLIFFS UFSAR 1.6-1 Rev. 47 1.6 ACRS SPECIAL INTEREST ITEMS 1.6.1 GENERAL This section describes the status of programs conducted for the investigation of items which were identified by the Advisory Committee on Reactor Safeguards (ACRS) as being of special interest and pertaining to all large water-cooled power reactors.

In carrying out these programs, information derived from research and development activities of the AEC or NRC and other organizations in the nuclear power industry was considered. 1.6.2 QUALITY ASSURANCE The Baltimore Gas & Electric Company has traditionally retained full responsibility and maintained close control over all aspects of the design and construction of its power plants. This background of experience has been used by BGE in establishing a comprehensive quality assurance program to assure that the Calvert Cliffs Units 1 and 2 are designed, fabricated, and constructed in accordance with the requirements of applicable specifications and codes. The BGE program starts with the initial plant design and is continued through all phases of equipment procurement, fabrication, erection, construction, and plant operation. The program provides for review of specifications to assure that quality control requirements are included and for surveillance and audits of the manufacturing and construction efforts to assure that the specified requirements are met.

A summary description of the Calvert Cliffs quality assurance program is contained in the Quality Assurance Topical Report. This program fully meets the guidelines established by 10 CFR Part 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants." 1.6.3 FAILED FUEL DETECTION Early detection of gross failure of fuel elements is important in limiting the consequences of fuel element failure. Early detection permits early application of protective action. Combustion Engineering, Inc. has evaluated the following instruments for possible application as a failed fuel monitor: a. Delayed-neutron monitor; b. Ion exchange iodine monitor;

c. Cerenkov-detector monitor;

d. Gaseous-fission-product detector; e. Differential gamma monitor; f. Gross gamma plus specific isotope monitor. Based on this instrument evaluation and a study of the expected fission and corrosion product activities in the reactor coolant, it has been concluded that the gross gamma plus specific isotope monitor provides a simple and reliable means for early detection of fuel failures.

The design bases of the detection system include the following: a. Trends in fission product activity in the RCS are used as an indication of fuel element cladding failures. The minimum detectable activity is specified as 10-4 µc/cc I135; CALVERT CLIFFS UFSAR 1.6-2 Rev. 47 b. There is a time delay of less than five minutes before the activity, emitted from a fuel-element-cladding failure, is indicated by the instrumentation. This time delay is a function of the location of the monitor; c. The information obtained from this system is not used for automatic protective or control functions or for detecting the specific fuel assembly (or assemblies) which has failed; d. The high activity alarm is supplemented with radiochemical analysis of the reactor coolant for fission products to provide positive identification for a fuel element failure. The location and operation of the detector, designated as the process radiation monitor, is described in Section 9.1.3. 1.6.4 REACTOR VESSEL THERMAL SHOCK Large quantities of ECC water are available to flood the core region in the event of a major LOCA. The Calvert Cliffs design uses a section of each of the RCS cold legs to conduct the water from the safety injection nozzles to the reactor vessel. This water then flows into the downcomer annulus and into the lower plenum of the reactor vessel before flooding the core itself. Analytical investigations were performed to provide assurance that the resultant cooling of the irradiated inner surface of the thick-walled reactor vessel will not induce or propagate cracks sufficient to cause the reactor vessel to fail.

A detailed analysis of the reactor response to thermal shock was performed. A report describing the results of the investigation, "Thermal Shock Analysis on Reactor Vessels Due to Emergency Core Cooling System Operation," A-68-9-1, March 15, 1968, was prepared and submitted to the AEC as part of Amendment 9 to the Maine Yankee license application (AEC Docket No. 50-309). The conclusion as stated in the report was that the CE reactor vessels are capable of sustaining the thermal shock imposed by ECCS operation without gross failure. Further work was performed to refine the surface heat transfer coefficient and the brittle fracture model used in the evaluation. Combustion Engineering, Inc. first made an extensive review of temperature-quench data obtained during the heat treatment of several heavy section steel plates. With this background, additional quench tests were planned and conducted to develop experimental heat transfer coefficients to be used in the analysis of this problem. These tests were performed on a plate approximately 2'x2'x1/2" thick and instrumented with eleven thermocouples. The plate was heated to 550°F and quickly lowered into an agitated (turbulent) water bath at 80°F (nearly duplicating the temperature conditions present in the reactor). The temperature of all thermocouples were recorded throughout the cooldown of the plate. Subsequently, the transient temperature data were compared to a heat transfer computer model of the plate to obtain an effective heat transfer coefficient. A detailed report covering this work entitled, "Experimental Determination of Limiting Heat Transfer Coefficients During Quenching of Thick Steel Plates in Water," A-68-10-2, December 13, 1968, was submitted to the AEC and made part of the public record. The report concludes that an effective heat transfer coefficient of 300 Btu/hr-ft2 F provides a realistic upper limit for thick steel plates quenched in highly agitated room temperature water. Subsequent development of a more realistic brittle-fracture model involved the development of a finite-element analysis computer program. The finite element method was used to compute the stress near the tip of hypothetical axial and circumferential cracks in the vessel. The stress intensity factor for thermal, pressure and residual stress loadings was computed as a function of crack depth. A detailed report entitled, "Finite Element Analysis of Structural Integrity of a Reactor Pressure Vessel during Emergency CALVERT CLIFFS UFSAR 1.6-3 Rev. 47 Core Cooling," A-70-19-2, January 1970, has been submitted to the AEC and is part of the public record. Accuracy of the finite element solution was confirmed by comparisons between a boundary collocation solution and the intensity factor as a function of crack depth for an edge cracked tensile specimen. The results confirm the conservatism of the approach used by CE and verify that cracks in the vessel will not grow during the thermal shock transient associated with ECC operation. 1.6.5 BLOWDOWN FORCES ON CORE AND REACTOR COOLANT SYSTEM COMPONENTS In the event of a large break, the RCS would depressurize rapidly, developing local pressure differences and forces in excess of normal operating loads. The WATERHAMMER computer program was utilized to define the hydraulic transient during the initial subcooled portion of the blowdown; the MODFLASH-2 computer program was used to calculate the pressure variations during the saturated portion of the blowdown. The loadings on the system components were then calculated from the pressure forces so obtained.

The reactor vessel internals were evaluated on the basis of these transient loadings. All critical components were designed to withstand these loads so that the core will be kept in place and that there will be no significant interference with the subsequent cooling of the core. The analysis reported in Section 14.15 shows that the Calvert Cliffs reactor meets these design requirements; in addition, the analysis shows that all of the CEAs except those adjacent to the outlet nozzle nearest the break can be inserted into the core following the accident. In order to further refine the capability in this area, CE is assessing the possible use of third generation computer programs like FLASH-3 and RELAP-3B. The progress of AEC-or NRC-sponsored experimental programs like the LOFT program is being monitored, and the information derived is being used to confirm the adequacy of the analytical techniques currently in use and under development. 1.6.6 EFFECT OF FUEL ROD FAILURE ON ECCS PERFORMANCE Experimental results have indicated that the core conditions after a LOCA (high internal gas pressure and increasing clad temperature which decreases clad tensile strength) can induce deformation of the fuel cladding in the time interval between blowdown and refill. The deformation takes the form of localized swelling of a fuel rod continuing until the clad perforates or until the temperature transient is reversed. Analytical and experimental programs were undertaken by CE to provide assurance that this deformation will not significantly affect the ability of the ECCS to prevent fuel melting.

The analytical work program established the conditions within the core during the transient for various break sizes. The parameters which determine the extent of fuel deformation are the internal gas pressure and the clad temperature transient (the temperature, the rate of change and the duration). The experimental program was designed to establish the correlation between the variable (internal gas pressure and the temperature transient) and the extent of clad deformation and perforation. The testing indicated the degree of clad deformation which may take place before perforation, as a function of clad heating rate and internal pressure. The clad swelling was observed to be a localized phenomenon, and the clad perforation was observed as a longitudinal split less than 1/2" long and 3/16" wide. CALVERT CLIFFS UFSAR 1.7-1 Rev. 47 1.7 IDENTIFICATION OF CONTRACTORS Calvert Cliffs Nuclear Power Plant, Inc. retains full responsibility for the engineering and design of facilities, purchase of equipment, construction, and operation of the CCNPP. The procedure followed during construction was similar to that which has been used by BGE for most of its generating facilities now in service or under construction.

Baltimore Gas & Electric Company carried out its responsibilities either by performing the work with its own staff or by in-depth involvement in work delegated to its major contractors. Such responsibilities were divided internally within the Company as follows: The Electric Engineering Department had the overall responsibility for the design of the plant. Procurement was the responsibility of the Purchasing and Stores Department. The Electric Construction Department was responsible for all site construction activities. The Electric Production Department was responsible for preoperational testing and initial testing as well as operation and maintenance of the plant. Shop inspections and witness testing were the responsibility of the Electric Production Department, assisted by members of the Electric Test Department, under the direction of the Quality Assurance Engineer. All other departments of BGE were available as needed to assist in the design and construction of the plant. Baltimore Gas & Electric Company engaged CE to design, manufacture, and deliver to the site two complete NSSSs and to design and fabricate the initial core loads of fuel and two reload batches for each reactor. Combustion Engineering, Inc. also furnished technical and professional supervision for erection, initial fuel loading, testing, and initial start-up of the two NSSSs. Replacement steam generators and replacement RVCHs were provided by Babcock & Wilcox, Canada for Units 1 and 2. Bechtel Associates, an affiliate of the Bechtel Power Corporation, was engaged as the Architect-Engineer for this project and as such performed engineering and design work for the balance of the plant equipment, systems, and structures not included under CE's scope of supply. Bechtel Associates prepared specifications, subject to BGE's approval, for all material, equipment, and systems which were purchased. Bechtel Power Corporation also provided qualified inspectors for shop inspections. Baltimore Gas & Electric Company contracted with Bechtel Power Corporation to perform the on-site construction of the entire plant.

The firm of Dames and Moore was retained as a consultant in the fields relating to site acceptability; namely, population and land use, meteorology, geology, seismology, hydrology, local shoreline protection, and hurricane effects.

NUS Corporation was retained as a general nuclear and radiological consultant and assisted BGE in the area of environmental radiological monitoring. MPR Associates was retained to assist BGE in all aspects of the Quality Assurance Program. Their exact function in the Program is detailed in Appendix 1A.

A number of consultants participated in studies in connection with the use of Chesapeake Bay water for condenser cooling purposes. Sheppard T. Powell and Associates performed studies of the physical and chemical characteristics of the bay. Under the direction of Dr. Ruth Patrick, the Academy of Natural Sciences of Philadelphia performed extensive studies in the field of marine ecology. The Alden Research Laboratories of the Worchester Polytechnic Institute performed studies on a hydraulic model of a 34-mile portion of the Chesapeake Bay in the area of the plant site. In addition, Dr. John C. Geyer of the Johns Hopkins University served as a general water consultant. All of these studies were coordinated by BGE.

APPENDIX 1A QUALITY ASSURANCE PROGRAM FOR DESIGN AND CONSTRUCTION LIST OF EFFECTIVE PAGES PAGE REVISION PAGE REVISION CALVERT CLIFFS UFSAR LEP-1A-1 Rev. 47 LEP-1A-1 47 1A-1 47

CALVERT CLIFFS UFSAR 1A-1 Rev. 47 APPENDIX 1A 1A QUALITY ASSURANCE PROGRAM FOR DESIGN AND CONSTRUCTION Appendix 1A, Quality Assurance Program for Design and Construction, is historical information about the construction of the plant. Appendix 1A has been removed from the Safety Analysis Report and has been sent to Plant History. An image of the contents of the appendix can be accessed in the NORMs Records system under Document ID (DOC ID) "Appendix-1A." APPENDIX 1B QUALITY ASSURANCE PROGRAM FOR THE OPERATIONS PHASE LIST OF EFFECTIVE PAGES PAGE REVISION PAGE REVISION CALVERT CLIFFS UFSAR LEP-1B-1 Rev. 47 LEP-1B-1 47 1B-i 47 1B-ii 47 1B-1 47 1B-2 47 1B-3 47 1B-4 47 1B-5 47 1B-6 47 1B-7 47 APPENDIX 1B QUALITY ASSURANCE PROGRAM FOR THE OPERATIONS PHASE TABLE OF CONTENTS PAGE CALVERT CLIFFS UFSAR 1B-i Rev. 47 1B INTRODUCTION AND SUMMARY 1B-1 1B.1 REGULATORY GUIDE 1.8 1B-1 1B.2 REGULATORY GUIDE 1.16 1B-2 1B.3 REGULATORY GUIDE 1.26, REVISION 3 1B-2 1B.4 REGULATORY GUIDE 1.37 1B-3 1B.5 REGULATORY GUIDE 1.38 1B-3 1B.6 REGULATORY GUIDE 1.54 1B-6 1B.7 REGULATORY GUIDE 1.68 1B-6 1B.8 REGULATORY GUIDE 1.94 1B-7 APPENDIX 1B QUALITY ASSURANCE PROGRAM FOR THE OPERATIONS PHASE LIST OF ACRONYMS CALVERT CLIFFS UFSAR 1B-ii Rev. 47 ANSI American National Standards Institute CCNPP Calvert Cliffs Nuclear Power Plant SRO Senior Reactor Operator QA Quality Assurance CALVERT CLIFFS UFSAR 1B-1 Rev. 47 APPENDIX 1B 1B QUALITY ASSURANCE PROGRAM FOR THE OPERATIONS PHASE The Quality Assurance Program for the Calvert Cliffs Nuclear Power Plant is the latest version of the Quality Assurance (QA) Topical Report.

In addition the specific commitments in the Quality Assurance Topical Report, Calvert Cliffs has the following commitments and site-specific exceptions to the following Regulatory Guides and American National Standards Institute (ANSI) Standards. 1B.1 REGULATORY GUIDE 1.8 Personnel Selection and Training (September 1975). This endorses ANSI N18.1 (March 8, 1971). Calvert Cliffs Takes two exceptions to ANSI N18.1, as follows: Item 1 Requirement Paragraph 4.2.2 states that at the time of initial core loading or appointment to the active position, the Operations Manager shall hold a Senior Reactor Operator's (SRO) License. Paragraph 3.2.1 states that positions at the functional level of Manager are those to which are assigned broad responsibilities for direction of major aspects of a nuclear power plant. This functional level generally includes the plant manager (plant superintendent, or other title), his line assistants, if any, and the principal members of the operating organization reporting directly to the plant manager and having overall responsibility for operation of the plant or for its maintenance or technical service activities.

Response Calvert Cliffs has two positions in its organization, Manager-Operations and General Supervisor-Shift Operations. Neither of these positions needs to individually meet all of the requirements of both Paragraphs 3.2.1 and 4.2.2. The Manager-Operations will satisfy paragraph 3.2.1 and most of 4.2.2 except that he will not maintain an SRO license. Instead, the Manager-Operations will hold or have held an SRO license. The General Supervisor-Shift Operations will hold and maintain an SRO license. The General Supervisor-Shift Operations satisfies Paragraph 4.2.2, but he does not satisfy 3.2.1 because he does not report directly the plant manager. Reason The Manager-Operations will hold or have held an SRO license, as opposed to having a license at the time of appointment to the position. He will have an excellent understanding of plant operations. The General Supervisor-Shift Operations will not only hold an SRO license at the time of appoint to the position, but he will maintain the license. The General Supervisor-Shift Operations directly supervises the operating shift organization, whereas the Manager-Operations is also responsible for procedure development, modifications acceptance, and operations/ maintenance coordination. The Manager-Operations' level of supervision does not require current in-depth and plant specific knowledge that results from maintaining an SRO license. CALVERT CLIFFS UFSAR 1B-2 Rev. 47 Item 2 Requirement Paragraph 3.2.2 states that supervisors are persons principally responsible for directing the actions of operators, technicians, or repairmen. Those positions usually designated as intermediate and first line supervisors are included in this category. Paragraph 4.3.2 states that supervisors not requiring Atomic Energy Commission licenses shall have a high school diploma or equivalent and a minimum of four years of experience in the craft or discipline he supervises. Response Calvert Cliffs has three supervisory positions in its organization - Supervisors, and in some cases Assistant General Supervisors and General Supervisors - that are organizationally equivalent (when supervising technicians/repairmen) to the positions described in paragraph 3.2.2 of ANSI N18.1 (March 8, 1971). All these individuals need not possess the four years of craft/discipline experience required by paragraph 4.3.2. Instead, at least the first line supervisor shall possess four years experience in the craft/discipline he supervises while other supervisors in the organization may be selected to fill supervisory positions based on possessing a minimum of an Associate's Degree, with four years or related technical experience, and demonstrated supervisory ability. Additionally, all first line and intermediate supervisors shall have at least a high school diploma or equivalent.

Reason To provide a balanced and broad base of supervisory ability within the site organizations made up of technicians/repairmen, it is desirable to include as supervisors both individuals with extensive craft/discipline experience accrued through field work and individuals with related education and experience who have demonstrated the ability to effectively supervise. 1B.2 REGULATORY GUIDE 1.16 Reporting of Operating Information - Appendix A Technical Specifications (Revision 4, August 1975). Calvert Cliffs is committed to this standard, as modified by the Calvert Cliffs Technical Specifications. 1B.3 REGULATORY GUIDE 1.26, REVISION 3 Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants.

Calvert Cliffs takes the following alternative: Calvert Cliffs' Quality Assurance program is applied to structures, systems, components, and activities that have been designated safety-related because they prevent accidents or mitigate the consequences of postulated accidents that could cause undue risk to the health or safety of the public. The QA program is also applicable to designated non-safety-related structures, systems, components, activities, and services as required by regulations. Designated non-safety-related program requirements are based on a graded approach to Quality Assurance required to meet applicable regulatory designated requirements and guidance. The level of QA program controls placed on designated non-safety-related items are defined in QA program documents and/or implementing procedures. Controls have been established for specifying on a Quality List all safety-related structures, systems, components, and activities that are subject to the requirements of the QA program. CALVERT CLIFFS UFSAR 1B-3 Rev. 47 1B.4 REGULATORY GUIDE 1.37 Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants (March 16, 1973). This endorses ANSI N45.2.1 (February 26, 1973). Calvert Cliffs takes one exception to ANSI N45.2.1, as follows: Requirement Subsection 3.2 outlines requirements for demineralized water. Response Calvert Cliffs specifications for demineralized water are different from the specifications outlined in the standard. Reason Calvert Cliffs specifications for demineralized water are consistent with guidelines provided by the Nuclear Steam Supply System supplier. Calvert Cliffs specifications are generally more restrictive than those specified by ANSI N45.2.1. 1B.5 REGULATORY GUIDE 1.38 Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Power Plants (Revision 2, May 1977).

This endorses ANSI N45.2.2 (December 20, 1972). Calvert Cliffs takes seven exceptions to ANSI N45.2.2, as follows: Item 1 Requirement Subsection 2.4 could be interpreted to mean that on-site and off-site personnel who perform any inspection, examination, or testing activities related to the packing, shipping, receiving, storage, and handling of items for nuclear power plants shall be qualified in accordance with ANSI N45.2.6.

Response Calvert Cliffs requires that only persons who are responsible for approving items for acceptance shall be qualified in accordance with Regulatory Guide 1.58 (which endorses ANSI N45.2.6) and that personnel who verify that storage areas meet requirements will be qualified to either Regulatory Guide 1.58 (which endorses ANSI N45.2.6) or ANSI N45.2.23. Reason Our receipt inspection procedures require persons who approve items for acceptance to be qualified in accordance with Regulatory Guide 1.58 (which endorse ANSI N45.2.6). Quality and Performance Assessment assessors verify that storage areas meet requirements. All other inspection, examination, and testing activities are subject to review by persons qualified to Regulatory Guide 1.58 (which endorses ANSI N45.2.6). Item 2 Requirement The second sentence of Subsection 2.4 requires that: CALVERT CLIFFS UFSAR 1B-4 Rev. 47 Off-site inspection, examination, or testing shall be audited and monitored by personnel who are qualified in accordance with ANSI N45.2.6. Response Calvert Cliffs uses personnel qualified in accordance with ANSI N45.2.23 to perform auditing and monitoring functions. Reason The qualification requirements for auditors cannot always be met by persons qualified to Regulatory Guide 1.58 (which endorses ANSI N45.2.6). Item 3 Requirement Subsection 2.7 requires that activities covered by the Standard shall be divided into four levels, though recognizing that within the scope of each level there may be a range of controls depending on the importance of the item to safety and reliability.

Response 1. The level of protective measures defined by Subsection 2.7 are applied to Basic Component Purchases. 2. Engineering or Supply Chain personnel will determine the level of protective measures to be applied to Commercial Grade purchases. Reason Calvert Cliffs' position is as follows: 1. For Commercial Grade items, it is not always possible to assign a level of classification in accordance with ANSI N45.2.2, as many items are purchased after they have been packaged by the manufacturer and shipped to his local agent, the wholesaler. 2. Experience has shown that the level of protection assigned to Commercial Grade items by vendors is adequate. Item 4 Requirement Subsection 3.0 specifies detailed requirements for packing items for each level defined in Subsection 2.7. Response Calvert Cliffs has replaced Section 3.0 with the following: 1. Packaging for Shipment to Calvert Cliffs Nuclear Power Plant (CCNPP) Engineering or Supply Chain personnel shall ensure that procurement documents for Basic Component and Commercial Grade item purchases either indicate that the normal methods of packaging and shipment used by industry in general are acceptable for the items being procured or specify the level of protection assigned to the item and the requirement that the vendor conform to applicable requirements for items in that classification defined in Regulatory Guide 1.38, Revision 2 - March 1977.

2. The normal methods of packaging used by the industry in general are acceptable for items being procured as Commercial Grade.

CALVERT CLIFFS UFSAR 1B-5 Rev. 47 3. Packaging for Storage by CCNPP In general, the packaging used by the vendor to ship items for all types of purchases to CCNPP need not be retained after the item is received by CCNPP, provided that the item is stored in an area that meets the requirements for a storage area for the level of protection assigned to the item. Special or unique items, however, may require special protective measures. For such unusual items, the Department that initiated the purchase, together with Engineering or Supply Chain personnel shall identify if any of the requirements of Section 6.4.2 of ANSI N45.2.2-1972 apply. Reason 1. This substitution will ensure that the item will receive adequate protection during shipment and storage, thus eliminating unnecessary restrictions and enabling CCNPP to use commercial sources to the utmost. 2. Experience shows that industrial practices for packaging Commercial Grade items are adequate for most applications. Item 5 Requirement Section 4.0 defines shipping requirements related to the protection levels assigned to items. Response Calvert Cliffs has replaced Section 4.0 with the following: 1. Shipping to CCNPP Calvert Cliffs will invoke the requirements for shipping specified in Section 4.0 of ANSI N45.2.2-1972 on Basic Component purchases only when Engineering or Supply Chain personnel have specified in procurement documents that the item shall be packaged in conformance with ANSI N45.2.2, Section 3.8.

Calvert Cliffs will not invoke the requirements of ANSI N45.2.2-1972, Section 4.0, on Commercial Grade item purchases. 2. Shipping from CCNPP Items shipped from CCNPP need not conform to any of the requirements of ANSI N45.2.2, but the organization that packs and handles the item shall provide roughly the same level of protection that the item was given during shipment to CCNPP. Reason If engineering personnel have determined that the vendor's methods of packaging are acceptable, they have already determined that the suppliers's methods of shipping are adequate. As items are shipped from CCNPP only for repair, the detailed requirements specified in Section 4.0 of ANSI N45.2.2 are not necessary. Item 6 Requirement Subsection 6.4 gives detailed requirements for care of items in storage, according to the protection levels assigned to the items. Response CALVERT CLIFFS UFSAR 1B-6 Rev. 47 Calvert Cliffs does not require items to be stored in the packing used for shipment if the storage level in the area provides the same protection as the level of packing assigned to the items. Caps, covers, etc., will be required only if specified by Engineering or Supply Chain personnel during the procurement process. If an item is taken from one storage area to another, however, the persons who move it are responsible for ensuring, as applicable, that additional packing is supplied to give adequate protection during transportation.

Reason The degree of protection given an item during storage should be tailored to the importance of the item to safety and the probability of deterioration during storage; to base storage requirements purely on the categories in Subsection 2.7 of ANSI N45.2.2-1972 is impractical. Calvert Cliffs requires Engineering and Supply Chain personnel to specify requirements more closely related to the actual function of items and to storage conditions.

Item 7 Requirement Subsection 7.3.3 requires compliance with a series of ANSI documents.

Response Calvert Cliffs controls for the use of hoisting equipment are compatible with the Standards listed in Subsection 7.3.3 of ANSI N45.2.2, although at the discretion of the Plant General Manager, they need not be compatible with documents referred to in these documents.

Reason Lower-level documents referred to in the documents listed in Subparagraph 7.3.3 will not necessarily affect the ability of CCNPP personnel to properly handle safety-related items and could lead to confusion. 1B.6 REGULATORY GUIDE 1.54 Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants (June 1973). This endorses ANSI N101.4 (November 28, 1972). Calvert Cliffs takes an exception to ANSI N101.4, as follows: Requirement Section 1.2 specifies applicability requirements for the Standard.

Response Calvert Cliffs requires that only activities performed inside containment structures and related to protective coatings applied to ferritic steels, aluminum, stainless steel, zinc-coated (galvanized) steel, concrete, or masonry surfaces shall conform to applicable Section of ANSI N101.4.

Reason Deterioration of protective coatings applied to surfaces outside containment structures would have no detrimental effects on the safe operation of the plant. 1B.7 REGULATORY GUIDE 1.68 Preoperational and Initial Startup Test Programs for Water-Cooled Power Reactors (November 1973). CALVERT CLIFFS UFSAR 1B-7 Rev. 47 Calvert Cliffs is committed to this standard. The preoperational and initial startup test program was completed prior to each Unit startup and is described in Section 13.1. 1B.8 REGULATORY GUIDE 1.94 Quality Assurance Requirements for Installation, Inspection, and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants (Revision 1, April 1976). In lieu of this Regulatory Guide, Calvert Cliffs conforms to ANSI N45.2.5, Draft 3, Revision 1 (November 1973).

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CHAPTER 2 SITE AND ENVIRONMENT TABLE OF CONTENTS PAGE2.0 SITE AND ENVIRONMENT 2.1 GENERAL DESCRIPTION 2.2 POPULATION AND LAND USE 2.3 METEOROLOGY CHAPTER 2 SITE AND ENVIRONMENT TABLE OF CONTENTS PAGE 2.4 GEOLOGY 2.5 HYDROLOGY 2.6 SEISMOLOGY CHAPTER 2 SITE AND ENVIRONMENT TABLE OF CONTENTS PAGE2.7 SUBSURFACE AND FOUNDATIONS 2.8 CHESAPEAKE BAY STUDIES CHAPTER 2 SITE AND ENVIRONMENT TABLE OF CONTENTS PAGE2.9 ENVIRONMENTAL RADIATION MONITORING 2.10 OTHER DESIGN CONSIDERATIONS CHAPTER 2 SITE AND ENVIRONMENT LIST OF TABLES TITLEPAGE CHAPTER 2 SITE AND ENVIRONMENT LIST OF TABLES TITLEPAGE CHAPTER 2 SITE AND ENVIRONMENT LIST OF FIGURES FIGURE CHAPTER 2 SITE AND ENVIRONMENT LIST OF FIGURES FIGURE CHAPTER 2 SITE AND ENVIRONMENT LIST OF FIGURES FIGURE CHAPTER 2 SITE AND ENVIRONMENT LIST OF FIGURES FIGURE CHAPTER 2 SITE AND ENVIRONMENT LIST OF ACRONYMS CALVERT CLIFFS UFSAR 2.1-1 Rev. 47 2.0 SITE AND ENVIRONMENT 2.1. GENERAL DESCRIPTION This section presents data on the site and environs for the Calvert Cliffs Nuclear Power Plant (CCNPP). These data were used to establish a basis for the selection of design standards for the plant and to determine the adequacy of concepts for controlling routine and accidental release of radioactive effluents to the environment. A series of studies (population and land use, meteorology, geology, hydrology, and seismology) has been conducted.

The site is located in Calvert County, MD, approximately 10-1/2 miles southeast of Prince Frederick, MD, and on the west bank of the Chesapeake Bay. Cooling water for the plant is drawn from and returned to the Chesapeake Bay.

The exclusion area around the plant has a minimum radius of 1150 meters. The distance to the nearest permanent residence is approximately one mile. A summer camp presently included in the plant site was abandoned in December 1971. The closest major metropolitan area is Washington, DC, approximately 45 miles to the northwest.

The structures are founded on Miocene sediments of the Coastal Plain Physiographic province. The foundation materials presented no special problems in design or construction. Foundations have been designed in accordance with the considerations discussed in the report on subsurface conditions and foundations, Section 2.7.

CALVERT CLIFFS UFSAR 2.2-1 Rev. 47 2.2 POPULATION AND LAND USE 2.2.1 GENERAL This section of the report presents the results of a population and land-use study. References 1 through 14 used for this study include United States Census data for Maryland, Virginia, Delaware, and Washington, DC; planning reports for various areas of Maryland; maps and aerial photographs of the site and surrounding area; and discussions with various individuals (References 15 through 18). A list of references is presented in Section 2.2.7.

2.2.2 LOCATION The site is located in Calvert County, MD, on the west bank of the Chesapeake Bay, approximately 10-1/2 miles southeast of Prince Frederick, MD. It originally covered an area of approximately 1135 acres and was owned by Baltimore Gas and Electric Company (BGE). The site boundary is posted and a fence has been erected around the immediate plant area. This site originally included Camp Conoy, a summer camp previously operated by the Baltimore YMCA. The camp was operated through December 1971, at which time it was abandoned. The camp was used by BGE for various recreational purposes. Camp Bay Breeze, also a summer camp, is two miles to the southeast and has a seasonal population of approximately 140. Nearby communities include: Calvert Beach and Long Beach, approximately 3 miles to the northwest; Cove Point, approximately 4-1/2 miles to the southeast; Chesapeake Ranch Estates, approximately 6 miles to the south-southwest; and the Patuxent Naval Air Test Center (NATC), approximately 10 miles to the south. Cultural features in the region and area are shown on Figures 2.2-1 and 2.2-2, Regional Map and Site Vicinity Map, respectively. The low population zone as defined in 10 CFR 100.2(b) is shown on Figure 2.2-13.

The metropolitan centers closest to the site are: Washington, DC, approximately 45 miles to the northwest; Baltimore, MD, approximately 60 miles to the north; Richmond, VA, approximately 80 miles to the southwest; and Norfolk, VA, approximately 110 miles to the south. 2.2.3 PRESENT POPULATION The estimated 1970 population density is shown on Figure 2.2-3, Regional Map, Showing Present and Future Population Density 0-50 Miles, and on Figure 2.2-4, Site Vicinity Map, Showing Present and Future Population Density 0-10 Miles. Estimated 1970 population distribution is shown on Figure 2.2-5, Regional Map, Showing Present and Future Population Distribution 0-50 Miles, and on Figure 2.2-6, Site Vicinity Map, Showing Present and Future Population Distribution 0-10 Miles. The 1970 population estimates are predicated upon 1970 census data and extrapolation of past population trends for cities, towns, election districts, and minor civil divisions. The population within each of these various political subdivisions is assumed to be uniformly distributed. All estimates include both seasonal and permanent population. (Seasonal population estimates are based on housing data from the United States Census of Housing and on residence classification data.) Population estimates within a five-mile radius of the site are based on a count for houses shown on the 1959 Calvert County General Highway Map, assuming four people per house. The estimates were extrapolated to 1970, based on the growth history of the area. House counts have been confirmed by recent aerial photographs for the immediate vicinity of the site. CALVERT CLIFFS UFSAR 2.2-2 Rev. 47 The population estimates indicate that the site and surrounding area are sparsely populated with the exception of localized areas along the coast of Chesapeake Bay which attract many summer residents. The summer seasonal residents account for approximately 20% of the total population within 10 miles of the site. The region surrounding the site is predominantly rural in character. Table 2-1 lists the communities within 30 miles of the site with 1970 populations greater than 1,000.

The total 1970 population, including seasonal residents, within 10 miles of the site was estimated to be 16,827. The population is distributed throughout the area and includes many small communities with population less than 1,000. The County Seat, Prince Frederick, is located 10-1/2 miles northwest of the site and has a total population of about 605. The population of the larger communities located within 10 miles of the site is presented in Table 2-2. A new community, known as Chesapeake Ranch Estates, is located 6 miles south-southeast of the site. The present (1970) permanent population of this development is approximately 180. In the summer, the population is approximately 1,000 during the week, and reaches a maximum of approximately 2,000 on weekends. All of the above populations are included in the data presented on Figures 2.2-3 through 2.2-6.

The character of the area begins to change from rural to suburban as the major population center of metropolitan Washington, DC, is approached. As indicated in Table 2-3, Accumulative Population Summary - 1970, the rate of change is greatest within 10 to 20 miles of Washington, DC, more than 30 miles from the site. The sector containing the maximum 1970 population is bounded by the west and northwest radial lines as shown on Figure 2.2-5. The estimated population of this sector, out to 50 miles from the site, is approximately 1,160,000. The data show that 96.6% of the present population in this sector is located more than 30 miles from the site. The population estimates also indicate that 97% of the 2010 population in this sector will be located more than 30 miles from the site. 2.2.4 FUTURE POPULATION Estimates of population density and distribution for the year 2010 are presented on Figures 2.2-3 through 2.2-6. The population estimates are based on an extrapolation of the past growth history of the region and on future population estimates made by the Maryland State Planning Department.

The published future population estimates extend through 1985. These were extrapolated to 2010 for purposes of this study. In areas where the present population was less than 50, estimates of the 2010 population are based on an anticipated population density of 250 persons per square mile. This method of computation was necessary only for certain areas within 10 miles of the site. Estimates of the future development of Chesapeake Ranch Estates indicate a maximum future population of 28,000. This estimate is included in the data presented on Figures 2.2-3 through 2.2-6. With continued growth of the Washington, DC metropolitan area, moderate population gains can be expected in the outlying regions, including portions of northern Calvert County and eastern Charles County which are within 15 to 25 miles of the site. Considerable population gains are expected in the area near Washington and Baltimore as part of the growth of the Boston to Washington "megalopolis." Table 2-4 presents the accumulative 2010 population within various distances from the site.

CALVERT CLIFFS UFSAR 2.2-3 Rev. 47 2.2.5 LAND USE In 1959, 85,400 acres (61%) of the land in Calvert County was devoted to farms, 51,200 acres (36.5%) to forests, and 3,500 acres (2.5%) to other uses. Of the 3,500 acres not used for farms or occupied by forests, 79% was residential; 5% was commercial; 3% was industrial; and 13% was devoted to public and semi-public use.

Dairy farming is of minor importance in Calvert County. In 1959, there were five dairy farms in the county and in 1964 the number decreased to one. There is, however, no dairy farm within a 5 mile radius of the plant. As stated previously, approximately 61% of the land in Calvert County was devoted to farms in 1959. In 1964, it declined to approximately 53%. The amount of harvested cropland declined slightly over this period, from 16,800 acres to 16,100 acres. The majority of the harvested cropland was used for growing tobacco, corn, and hay, as shown in Table 2-5, Agricultural Land Use - 1959 and 1964. Within 25 miles of the site, over 90% of the land area is located in Calvert, Charles, Dorchester, and St. Mary's Counties. In 1959, approximately 512,700 acres (49%) of the land in these counties was devoted to farms. In 1964, it declined to approximately 45%. The amount of harvested cropland declined over this period from 180,000 acres to 170,500 acres. The major crops grown in the four-county area are shown in Table 2-6, Agricultural Land Use - 1959 and 1964.

With continued population growth, it is anticipated that the percentage of land devoted to farms will continue to decline and will be accompanied by increased residential and commercial use. However, the overall character of the area is expected to remain essentially rural.

The waters adjacent to the site are used for commercial fishing, primarily for shellfish, such as clams, oysters, and crabs. Calvert County accounted for approximately 2% of the State's total fish catch in 1963.

CALVERT CLIFFS UFSAR 2.2-4 Rev. 47 2.2.6 SUMMARY The site is located in an undeveloped, sparsely populated area. The present population within 30 miles of the site is small. However, moderate increases, on the order of 1.5%/year, are estimated over the next 40 years. At present, more than 90% of the CALVERT CLIFFS UFSAR 2.2-5 Rev. 47 population within 50 miles of the site is located at distances greater than 30 miles. This trend is expected to continue for the expected life of the CCNPP. At the present time, the major portion of the land in the area surrounding the site is devoted to agricultural and forest uses. Although the amount of land devoted to farming is declining, agriculture should continue to be a primary land use during the life of the proposed nuclear plant. Small increases in the amount of land devoted to residential and commercial use will occur with increased population growth. The waters of the Chesapeake Bay are now and should remain a source of sea food, primarily clams, oysters, and crabs.

From a population and land-use standpoint, the site is suitable for the location of a nuclear power plant. 2.

2.7 REFERENCES

1. Gladstone, Robert & Associates, Washington, DC, April 1965 - The Economy and Population of Southern Maryland, prepared for the Maryland State Planning Department 2. Maryland State Planning Department, April 1967 - Proposed Comprehensive Plan, Calvert County 3. Maryland State Planning Department Newsletter, July 1967, Volume XX, No. 4 4. Maryland State Roads Commission, General Highway Map, Calvert County, Maryland, 1959 5. Regional Planning Council, Baltimore, Maryland, October 1964 - Regional Data Book, Publication 2 6. Regional Planning Council, Baltimore, Maryland, September 1965 - Regional Plan Alternatives, Publication 5 7. United States Bureau of the Census, 1964 Census of Agriculture, Statistics for the State and Counties, Maryland, Volume 1, Part 23, United States Government Printing Office, Washington, DC 8. United States Bureau of the Census, 1950 Census of Housing, Volume 1 General Characteristics, Parts 2, 3, and 6, United States Government Printing Office, Washington, DC 9. Unites States Bureau of the Census, 1970 Census of Housing, Volume 1 States and Small Areas, Parts 3, 4, and 8, United States Government Printing Office, Washington, DC 10. United States Bureau of the Census, 1970 Census of Population, Volume 1, Characteristics of the Population, Parts 9,10, 22, and 46, United States Government Printing Office, Washington, DC 11. United States Bureau of the Census, U.S. Census of Population: 1970, Number of Inhabitants -- Delaware, 1971 12. United States Bureau of the Census, U.S. Census of Population: 1970, Number of Inhabitants -- Maryland, 1971 13. United States Bureau of the Census, U.S. Census of Population: 1970, Number of Inhabitants -- Virginia, 1971 14. United States Department of Commerce, Bureau of the Census, Election Districts, 1960 Minor Civil Division Maps for the States of Delaware, Maryland and Virginia CALVERT CLIFFS UFSAR 2.2-6 Rev. 47 15. Chesapeake Ranch Club; Lusby, MD; Mr. Jetmore 16. Girl Scouts Council of the Nations Capital; Arlington, VA; Mr. Slover

17. Patuxent NATC; Lexington Park, MD; Housing Personnel 18. Young Men's Christian Association; Baltimore, MD; Mr. Moss 19. Final Environmental Impact Statement, Increased Flight and Related Operations in the Patuxent River Complex, Patuxent River, Maryland: Department of the Navy, Naval Air Warfare Center Aircraft Division, December 1998 20. Letter from Ms. D. M. Skay (NRC) to Mr. C. H. Cruse (CCNPP), dated August 29, 2001, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 - Issuance of Amendment Re: Aircraft Hazards Analysis (TAC Nos. MA7229 and MA7230)

CALVERT CLIFFS UFSAR 2.2-7 Rev. 47 TABLE 2-1 COMMUNITIES WITHIN 30 MILES OF THE SITE WITH POPULATION OF 1,000 OR GREATER POPULATION(a) COMMUNITY 1940 1950 1960 1970 DISTANCE AND DIRECTION FROM SITE (miles) Patuxent NATC (c) (c) l,900(b) 2,100(b) 10-S Lexington Park (c) (c) 7,039 9,136 12-S Leonardtown 668 1,017 1,281 1,406 14-SW Cambridge 10,102 10,351 12,239 11,600 21-ENE St. Michaels 1,309 1,470 1,484 1,456 26-NNE Waldorf (c) (c) 1,048 7,368 27-WNW Easton 4,428 4,836 6,337 6,809 30-NE La Plata 488 780 1,214 1,561 30-WNW (a) Based on United States Census Bureau Statistics. (b) Estimated. (c) Population less than 1,000. Exact number not available.

CALVERT CLIFFS UFSAR 2.2-8 Rev. 47 TABLE 2-2 POPULATION OF COMMUNITIES NEAR THE SITE COMMUNITY ESTIMATED 1970 POPULATION DISTANCE AND DIRECTION FROM SITE Calvert Beach and Long Beach 935 3 NW Cove Point 385 4-1/2 SE Kenwood 275 5-1/2 NW Scientists Cliffs 660 7 NNW Solomons 350 8 S Dares Beach 825 10 NNW Prince Frederick 605 10-1/2 NW CALVERT CLIFFS UFSAR 2.2-9 Rev. 47 TABLE 2-3 ACCUMULATIVE POPULATION SUMMARY - 1970 RADIAL DISTANCE FROM SITE ACCUMULATIVE POPULATION (miles) 5 3,425 10 16,827 20 83,495 30 188,755 40 518,825 50 2,305,635 CALVERT CLIFFS UFSAR 2.2-10 Rev. 47 TABLE 2-4 ACCUMULATIVE POPULATION SUMMARY - 2010 RADIAL DISTANCE FROM SITE ACCUMULATIVE POPULATION (miles) 5 11,253 10 59,750 20 187,470 30 379,830 40 1,040,750 50 4,757,810

CALVERT CLIFFS UFSAR 2.2-11 Rev. 47 TABLE 2-5 AGRICULTURAL LAND USE - 1959 AND 1964 (Calvert County) PERCENT OF HARVESTED CROPLAND CROP 1959 1964 Tobacco 42 44 Corn 27 27 Hay 16 15 Small Grains 8 8 Other 7 6 CALVERT CLIFFS UFSAR 2.2-12 Rev. 47 TABLE 2-6 AGRICULTURAL LAND USE - 1959 AND 1964 (Calvert, Charles, Dorchester, and St. Mary's Counties) PERCENT OF HARVESTED CROPLAND CROP 1959 1964 Corn 28 30 Soybeans 25 22 Small Grains 17 17 Tobacco 12 12 Hay 10 8 Other 8 10

CALVERT CLIFFS UFSAR 2.3-1 Rev. 47 2.3 METEOROLOGY 2.

3.1 INTRODUCTION

This section summarizes the meteorological studies that have been conducted since the start of the engineering and design of the CCNPP. The meteorological studies performed include work in the following main categories, listed in chronological order: a. Preliminary Data Collection

b. Initial Site Weather Data Program

c. Special Vertical Wind Standard Deviation Tests

d. Land-Sea Wind Speed Investigation e. Extended Onsite Penetration Wind Study f. Calculation of Incident and Routine Long-Term Relative Concentrations 2.3.2 PRELIMINARY DATA COLLECTION Proximal long-term weather station data were used from the Patuxent NATC - PAX (now Patuxent Naval Air Station - NHK) for periods of record from 1955-1960, and 1949-1964. In addition, meteorological data from Washington National Airport (DCA); Byrd Field, Richmond, VA, (RIC); and Annapolis, MD, (ANA) were used to evaluate the frequency of various weather parameters and certain meteorological extremes, respectively. See Regional Map, Figure 2.2-1. Also, statistical data for severe weather parameters were obtained from numerous official records issued by the Environmental Science Services Administration (ESSA), Department of Commerce, Asheville, NC.

The following weather information from the above sources was evaluated and related to the Calvert Cliffs Nuclear Plant Site: Tornadoes, Freezing Precipitation, Tropical Storms, Hurricanes, and Diffusion Conditions. 2.3.2.1 Tornadoes Five tornadoes were observed during the period 1953-1962 in the general vicinity of a single latitude-longitude square near the proposed plant site. The mean annual frequency was 0.5 tornadoes per year and the probability of a tornado striking a single point within a single latitude-longitude square near Calvert Cliffs, using a method originally derived by H.C.S. Thom of ESSA, was calculated to be 3.75x10-4. The recurrence frequency was calculated to be once about every 2,700 years.

2.3.2.2 Thunderstorms Thunderstorm day statistics indicate that about 40 thunderstorms per year can be expected in the area. Fifteen years of records at Patuxent showed 814 observations of thunderstorm activities. From these data one can calculate the average duration of a thunderstorm to be 1.356 hours for a point. A study of 10 years of records for transmission subtransmission feeders was conducted. This study showed that transmission and subtransmission feeder losses were 4 minutes and 423 minutes, respectively, due to storms in a 10-year period. The subtransmission feeders covered an area of approximately 180 square miles.

2.3.2.3 Freezing Precipitation The Patuxent NATC records (1949-1964) list 910 hours of snow and 265 hours of frozen or freezing precipitation, other than snow, for a total of 1175 hours (or 70,500 minutes) in 15 years. Interpolating for a 10-year span yields 47,000 minutes. The outages due to snow and/or freezing precipitation were 182 minutes and 122 minutes in 10 years, for transmission and subtransmission feeders, respectively. It is interesting to note that 9 of 12 outages occurred during CALVERT CLIFFS UFSAR 2.3-2 Rev. 47 a single snowstorm in March 1958. Certain design changes were made as a result of this storm and it is unlikely that outages of this magnitude would again occur. 2.3.2.4 Tropical Storms and Hurricanes Approximately one hurricane per year poses a threat to the area, and about one hurricane every 10 years produces a significant effect. Northeasters, or extratropical storms, also can influence the area in terms of flooding of low-lying land. The detrimental effects of northeasters are considerably less than those postulated for hurricanes in the site area. 2.3.2.5 Preliminary Diffusion Climatology The frequency of various Pasquill classes of diffusion was initially assessed through the use of the familiar Pasquill-Turner method. The proximal Patuxent NATC data were used for a five-year period of record, which yielded the following results: Pasquill Condition Annual Percent Occurrence A and B 2.6 C 10.4 D (day) 35.0 D (night) 28.2 E 11.8 F 8.0 G 4.0 Since it is possible to take advantage of offshore waterways in considering a site boundary, it was considered reasonable to limit discussion interest and calculations to onshore winds at the Calvert Cliffs site. The onshore wind directions, by sector, are as follows: a. North

b. North-Northeast c. Northeast d. East-Northeast

e. East

f. East-Southeast The frequency of all winds from these directions was documented (over a five-year period) to be: a. Patuxent NATC 23.0% b. Washington National Airport 24.0% c. Byrd Field, Richmond, VA 30.0% d. Annapolis, MD 24.0% Based on these total wind frequency samples, it was calculated that the frequency of inversion winds associated with onshore flow was as follows: Pasquill Pasquill Station "E" "F and G" a. Patuxent NATC 2.04% 1.35% b. Annapolis 1.97% 3.40%

CALVERT CLIFFS UFSAR 2.3-3 Rev. 47 In order to confirm the initial conclusions drawn above and to get a first approximation of typical Pasquill "F" conditions at the site, two additional station records were examined in detail. Five years of records from RIC and DCA were examined and a computer program was written to produce the frequency of winds equal to or less than X knots for the 0100 EST hour of the day when the cloud cover was equal to or less than .4 coverage. Values for wind speeds from 10 knots to calm were documented. The results indicated that the average Pasquill wind speed for RIC was 1.88 m/sec, while that for DCA was found to be 2.02 m/sec. The frequency of these conditions with onshore wind directions was found to be 2% at RIC and 3% at DCA. Since neither of these stations had as good exposure as would be anticipated for the Calvert Cliffs site due to the unrestricted fetch over the Chesapeake Bay, it was deemed conservative to select a wind speed of 1.5 m/sec as a typical onshore Pasquill "F" site condition. In general, the site's low-level winds under a temperature inversion drain toward the Chesapeake Bay. It would not be possible for a ground-released effluent to get to the minimum site boundary under these conditions, and highly improbable that the ground release could get to the other inland boundaries due to terrain slope and other effects. Wind persistence maxima for all wind sectors based on five-year record summaries at Annapolis, MD were as follows: 1 Sector 3 Sectors 5 Sectors 48 hrs 140 hrs 220 hrs Washington National, Byrd Field, and Patuxent showed less persistent winds MAXIMUM WIND PERSISTENCE FOR ONSHORE WINDS IN A SINGLE SECTOR (5 YEARS OF RECORD) Station Pasquill Condition Maximum Persistence DCA 1-3 knots winds 6 hrs RIC 1-3 knots winds 6 hrs PAX "E" and "F" 12 hrs ANA "E" and "F" 12 hrs PAX All Speeds 27 hrs ANA All Speeds 37 hrs Onsite low-level diffusion measurements were made at two primarily coastal locations during the periods from September 14, 1967, through November 9, 1968 at site N1W; and November 9, 1967 and through November 9, 1968, at site S1W. See Figure 2.3-1, Figure 2.3-2, and Table 2-10 for station locations and the description of meteorological instrumentation, respectively. In addition, temperature gradient approximations were made using two inland ground-level thermograph stations at the site at locations approximately 120' above mean sea level (MSL) and 40' MSL. These data also extended from November 9, 1967 to November 9, 1968. The two coastal sites were selected initially because they: a. offered good to excellent exposure to onshore winds; and

b. offered the only initial long-term exposure to winds unmodified by terrain and extensive tree cover.

CALVERT CLIFFS UFSAR 2.3-4 Rev. 47 The results of these onsite data comparison evaluations indicated the following: a. Frequency of inversions derived from 1. onsite data 31% 2. long-term data 24% b. Air drainage was toward the Bay under inversion conditions. c. Average wind speed during inversion conditions was 2.6 MPS. d. Standard deviation of horizontal wind direction () during worst, single-season wind sector inversion conditions averaged 6.6°. e. When wind speed decreased, increased, in general. f. For on/or along-shore winds, the average value of was 0.209 rad meters/sec. g. /Q values at the 0.5% level of all conditions was 1.17x10-4 sec/m3 for the 1150 meter minimum site boundary. 2.3.3 SPECIAL VERTICAL WIND STANDARD DEVIATION TESTS 2.3.3.1 General Two sets of special diffusion tests were conducted at the Calvert Cliffs site. In both cases, both horizontal and vertical standard deviations of the wind conditions were measured.

2.3.3.2 Test Set 1 (October 17 to November 1, 1968) In order to simulate actual reactor site location data, a standard anemometer was set up on a 40' bluff at Camp Conoy - just south of the reactor site. The anemometer permitted recording of wind direction and wind horizontal direction and its deviation. In addition, an e meter was installed to evaluate the vertical standard deviations during this period. The wind sensors were about 10' above the cliff area and about 40' inland. Results of e Test Set 1 were as follows: Onshore Inversion Wind e Offshore Inversion Wind e "Neutral" Winds e Cases 16 122 157 e 13° 8° 14.3° Lowest e 1° 1° 1° Cases <5° 1 35 9 2.3.3.3 Test Set 2 (February 11 through 20, 1969) A second set of readings was taken during this period at Station 2 (about 2000' from the coastline). The companion statistics for Test Set 2 were follows: Onshore Inversion Wind e Offshore Inversion Wind e "Neutral" Winds e Cases 36 28 104 e 10° 13° 6° Lowest e 6° 2° 4° Cases <5° 0 5 1 These readings were also taken about 10 to 12' above the ground, but with an unobstructed trajectory from an onshore viewpoint.

CALVERT CLIFFS UFSAR 2.3-5 Rev. 47 2.3.3.4 Conclusions The sigma e values measured during these two test series both indicated that a. Onshore inversion winds tend to produce near-neutral (Pasquill "D") e values. b. Offshore inversion winds tend to produce lower standard deviations than onshore cases near the coast, but somewhat larger inland. c. Only one case in the total showed e values as low as Pasquill "F". 2.3.4 LAND-SEA WIND SPEED INVESTIGATION There was some concern expressed that the wind speed for onshore flow at the Station 4 (S1W) site was not representative for inland locations because the anemometer was in an area that is subject to a "Venturi" effect when the wind direction is onshore. In order to explore this possibility, this study compared the wind speed and diffusion values at the Station 4 (S1W) site to those on a raft anchored about one mile offshore. For approximately one month of data, the diffusion parameter () (STUB) was compared at each site where simultaneous onshore flow occurred at the sites. The average wind speeds at the two sites were also compared. Table 2-11 gives the results of the 256 simultaneous onshore winds and compares them to the classical Pasquill inversion classification values. The data indicated the following: a. Only 1 observation of 256 at S1W gave value equivalent to Pasquill "F". b. Wind speeds were generally lower at S1W than at the raft, but wind deviations were larger. c. The only possible Venturi effect noted at S1W was when the wind was onshore and the speeds were 3 mph or less. 2.3.5 THE PENETRATION ONSHORE WIND STUDY The primary purpose of this extended meteorological investigation at the Calvert Cliffs site was to further refine the atmospheric dispersion parameters obtained from the initial site weather data program for use in the calculation of the relative concentration, /Q, at the site boundary nearest the reactor. Of secondary importance was to examine any anomalous flow features detected at the site and discuss its relevance to site diffusion characteristics.

Three inland meteorological stations were set up along with Station 4 (S1W). All four stations became active January 10, 1969 at the Calvert Cliffs site. In addition, temperature gradient systems were installed at Stations 2 and 4. See Figure 2.3-1 and Table 2-12 for station locations and instrumentation. A computer program was developed to analyze the wind flow across the site using the simultaneous wind observations from the four stations as input. The wind speed at each of the four stations was conservatively read to the lowest whole mph. Standard techniques for evaluation of short-term releases (Pasquill "F", wind direction invariant, = 1 MPS), were compared with measured parameters to determine, within conservative limits, the proper values applicable to this specific location. The low percent probability level of values was considered over the area collectively. The procedure was to select only those hours when the wind at Station 1 (K) was blowing onshore and also where at least two of the stations a value of 0.200 existed. Results for the one year extended study showed inversion conditions for 35% of the total observations, neutral conditions for 47%, and lapse for 17%, with 1% of the observations CALVERT CLIFFS UFSAR 2.3-6 Rev. 47 missing. The winds showed a definite tendency to drain offshore during inversions; for the onshore winds, nearly 18% were in the neutral category, 9% in the unstable, and less than 4% in the stable category. The cumulative frequency distribution by wind speed category of on/or along-shore inversion winds for the four stations is given in Table 2-13 in terms of the total observations. 2.3.6 CALCULATION OF INCIDENT AND ROUTINE LONG-TERM RELATIVE CONCENTRATIONS Two types of relative concentration calculations are of interest at the Calvert Cliffs site. The first are the 0-2 hour, 2-24 hour, and 1-30 day values which are used to determine the resulting radiation exposure from all of the postulated incidents. The second type is that pertinent to routine gaseous releases at the site. 2.3.6.1 Calculation of the Zero to Two-Hour Relative Concentration For the first two hours following a postulated "maximum hypothetical accident," the relative concentration is calculated by the Gifford wake model for a ground release: where: = relative concentration, seconds/m3 = average wind speed, meters/sec yz = standard deviations of the distributed material in the lateral and vertical directions, in meters c = wake factor (dimensionless) A = cross-sectional area of structure from which material is presumed to be released, square meters From the data in Table 2-13 it was determined that 5% of the time the on/or along-shore winds at Station 1 had speeds of 3.2 MPS or less; the comparable speeds at the 5% level for Stations 2, 3, and 4 are 1.1 MPS, 1.7 MPS, and 2.1 MPS, respectively. The average of the four stations at the 5% level is 2.0 MPS. This shows that relatively strong flow is available for on/or along-shore inversion wind directions even at the 5% frequency level at the site. The 0-2 hour relative concentration was evaluated at various frequency levels of the statistic STUB, the product of sigma theta and u-bar, using a very conservative technique. The technique was to select the average of the two lowest of the four simultaneous values of STUB observed for on/or along-shore winds, and to array these averages in the order of frequency of occurrence. Assuming that the wind speed was one meter per second, the corresponding values of were tabulated, and the corresponding values of for a distance of 1150 meters (the distance to the nearest site boundary) were selected. A wake factor of cA = 0.5x1640 M2 = 820 M2, and a z value of 24 meters were used. The relative concentrations are shown in Table 2-14 for the 1% through 10% frequency levels. The value of z = 24 meters was selected as being compatible with the Pasquill "E" category for the 1% STUB level, using a wind speed of one meter per second. The previously referred to measurements of e showed that a selection of Pasquill "E" for the vertical fluctuations was highly conservative. CALVERT CLIFFS UFSAR 2.3-7 Rev. 47 A value for the 0-2 hour /Q of 1.3x10-4 sec/m3 was selected for the radiation exposure calculations in Chapter 14 resulting from the containment wall release pathway. Meteorological conditions resulting in this value or higher for the 0-2 hour relative concentration will occur less than 5% of the time. For releases from the plant vent stack, main steam gooseneck, and refueling water tank vent, a 0-2 hour /Q of 1.44x10-4 sec/m3 was calculated based on a zero cross-sectional area. 2.3.6.2 Calculation of the 2-24 Hour and 1-30 Day Average Relative Concentrations Average relative concentrations for periods of 10 hours, 12 hours, and 29 days were calculated utilizing the onsite data acquired at Calvert Cliffs. No credit was taken for the wake factor of the plant structure and a minimum site boundary of 1150 meters was assumed in all 16 sectors.

The meteorological station with the lowest STUB value, Station 4 (S1W), was selected for this study. No Pasquill class with more diffusion than Pasquill "C" (slightly unstable) was considered and a ground-release accident model was assumed. As was done with the 0-2 hour /Q, the 2-24 hour and the 1-30 day values were also selected at the 5% frequency level. The resulting values were as follows: Time Period 5% Probability Level /Q at 1150 meters (sec/m3) 2-24 hrs 9.10x10-6 1-30 days 2.70x10-6 The 5% values are shown as a function of distance on Figure 2.3-3 for all of the incident-related time periods.

2.3.6.3 Calculation of Routine Long-Term Concentrations The average annual relative concentrations, /Q which are applicable to routine venting or other routine operational gaseous effluent releases, have been determined for the final annual data record in accordance with the following equations: Where R = = relative concentration (sec meter-3); at a distance D (meters) from the effluent source; in direction sector i p = Pasquill class (A through G) f(k) = percent frequency wind blows toward sector i, within speed interval k, during Pasquill class condition p µ(k) = Wind speed value representative of speed class interval, k, MPS z(p,D) = vertical dispersion coefficient, meters, for Pasquill class p, at distance D B = spread of wind sector, radians = /8, for 22-1/2° sectors. CALVERT CLIFFS UFSAR 2.3-8 Rev. 47 These equations and resultant calculations are appropriate for evaluating ground releases over longer time intervals. They do not include a wake factor term. The Isopleths of the average annual concentration, shown in Figure 2.3-4 were calculated using the wind data and T Pasquill class data of the final annual record. The maximum average on-shore relative concentration is 2.2x10-6 seconds meter-3 in the southeast sector at a distance of 1300 meters, which occurs as a result of the northwest winds and associated stability conditions. The site boundary in this direction is 2100 meters (Figure 1-1).

2.3.6.4 Average Annual Concentration at the Milk Samples Location Milk samples were obtained from a location 4.2 miles southwest of the reactor site, during the period December 23, 1971 through June 5, 1976. Since this time, no samples have been available in the area. The model used in the above section has been applied to this location. The average annual /Q is 7.0x10-8 and occurs with a northeast wind. 2.3.6.5 Continuing Studies Additional studies were made to further refine the diffusion parameters. Included in these studies was an analysis of the data obtained at Station 2 (IS) between November 12, 1971 and November 11, 1972. This analysis showed that the diffusion characteristics specified in Section 2.3.6 are conservative. Comparisons were made of T data from the 12' to 48' system installed on the pole at Station 2 (IS) and the T data from the "Sky Needle" 30' to 98' system also located at Station 2. The 12' to 48' system was continued in operation until September 1974.

A comparison study was made during the summer of 1974 to determine the correlation between meteorological data obtained from the "Sky Needle" 30' to 98' system at Station 2 and data obtained from the microwave tower system. The results of this study have been evaluated, and the remaining meteorological systems at Stations 2 and 4 were discontinued. The "Sky Needle" system was taken out-of-service August 14, 1975. 2.3.7 METEOROLOGICAL MEASUREMENT SYSTEMS In accordance with the requirements of NUREG-0654 and Generic Letter 82-33 (Supplement 1 to NUREG-0737), a meteorological tower (Figure 1-1) was installed to provide the essential parameters used in support of dose assessment calculations for emergency preparedness. The meteorological tower and instrumentation design meets the intent of Safety Guide 23, February 1972, and Regulatory Guide 1.97, Revision 3, for primary meteorological measurements systems. The instrumentation on the meteorological tower is described in Table 2-12. Signals from the wind and temperature sensors are transmitted to the plant Control Room where T, Ws, Wd, , and rain water level can be continuously monitored by the operator. The meteorological tower, located at the end of Road B-1, has been operational since April 1982. Subsequently, the Technical Specifications were amended to designate the new meteorological system as the "primary" meteorological system as addressed in Regulatory Guide 1.23, Revision 1, and the old microwave tower became a backup system. The meteorological instrumentation on the old microwave tower was taken out-of-service in the fall of 1993. The current primary and backup meteorological CALVERT CLIFFS UFSAR 2.3-9 Rev. 47 measurement systems are described in the Emergency Response Plan and its implementing procedures. 2.3.8 INVESTIGATION OF RELATIVE CONCENTRATION FREQUENCIES USING THERMAL STABILITY PARAMETERS During the investigation of the meteorological conditions at Calvert Cliffs, the almost universal acceptance of sigma theta to define diffusion qualities was questioned. This was in part due to the uncertainties of the sigma theta measurements in defining vertical plume growth. Also with winds at 2 to 3 mph or less, the measurement of sigma theta becomes difficult. Yet, in evaluations of the accident hazards, the periods of low wind speeds are the most critical. For these reasons the need for a Calvert Cliffs diffusion climate evaluation which does not depend upon sigma theta measurements was assessed. 2.3.8.1 The Requirement For Additional Meteorological Evaluation at Calvert Cliffs The Calvert Cliffs site analyses in Section 2.3.6 use sigma theta measurements to define horizontal plume growth only. The uncertainties of the relationships between these measurements and vertical plume growth do not, therefore, cloud the validity of these analyses. Further, the 5% worst weather conditions of most concern for the accident evaluations are those with on-shore winds at low speeds. With on-shore directions conservatively defined to include nine 22-1/2° sectors, NW through SE clockwise, at Station 2 (IS) at 12' above grade, on-shore winds at 3 mph or less occur 12% of the time. This 12% frequency includes the unstable and neutral as well as the stable (winds have subsequently been measured at 33' above grade. At this elevation, on-shore winds of 3 mph or less occur less than 5% of the time.) It is unlikely, therefore, that the analysis based on sigma theta measurements are significantly biased by difficulties of measuring sigma theta at low wind speeds. Nevertheless, to remove the residual uncertainties in 1969 Baltimore Gas and Electric Company began to measure and record vertical temperature gradients near the ground for use in classifying site stability characteristics into inversion, neutral, and unstable conditions.

2.3.8.2 The Weather Data for the Independent Evaluation The vertical temperature gradient (T) was measured continuously between 12 and 50' above grade at Station 2 from 1969 through September 14, 1974. Concurrently, an MRI 2040 wind instrument was installed at 33' above grade (as opposed to the prior wind instrumentation at 12' above grade) to measure wind speed and direction and values. Hourly averages of wind speed and direction, and one-an-hour 20 minute averages of were recorded. There were two sources of data available with the MRI 2040 wind instrument, sigma meter readings and wind range measurements. The data were compared and wind range measurements, divided by six to obtain , in accordance with the standard procedures, gave uniformly-lower values at the smaller readings. The range measured values were, therefore, used in this analysis because they provide more conservative estimates of the site diffusion quality. CALVERT CLIFFS UFSAR 2.3-10 Rev. 47 Subsequent to the initiation of this program, it became an accepted practice to classify stability conditions into the standard Pasquill classes by the use of T values in accordance with the following table of values. PASQUILL CLASS T° C/100 meters A -1.9 B -1.9 to -1.7 C -1.7 to -1.5 D -1.5 to -0.5 E -0.5 to +1.5 F +1.5 to +4.0 G +4.0 To take advantage of this accepted practice, the validity, for Pasquill classification purposes, of the 12 to 50' T data has been investigated by a comparison with concurrently observed 12 to 97' data, as shown in Figure 2.3-5 and Table 2-15. For the shallower layer, 2 to 3% more of the observations fell in the critical Pasquill E, F and G classes, and 2 to 3% less in classes B, C, and D. Because the atmospheric layer upward from 30' above grade was becoming the standard layer for determination of thermal stability Pasquill classes, the validity of the 12 to 50' layer data was further investigated by a comparison with concurrently observed 30 to 97' Ts at Station 2, as shown in Figure 2.3-6. On this figure, the dashed line is the line showing equal lapse rates for both layers. It is apparent that the assignment of Pasquill classes using the 12 to 50' layer Ts is very conservative in comparison with the use of the standard layer based at 30' above grade. Data observed at Station 2, from November 1969 through October 1970, were selected for the Primary Year of Record. There were gaps in this record caused by equipment malfunctions. To complete the record and eliminate a potential seasonal bias, 1971 data were added to it, thereby creating the Final Annual Record. These added data are limited to dates and hours of the day which coincide with the data gaps in the Primary Year of Record. 2.3.8.3 The Zero to Two-Hour Relative Concentration Determined by T and Parameters Relative concentrations for each hour of the final annual record have been calculated using the equations in Section 2.3.6.1 but with the uncertainties associated with measurements eliminated. y and z values were fixed by the Pasquill classes, as before. However, two sets of Pasquill Classes were defined; one set based upon the T measurements using the table in Section 2.3.8.2 and the other set using the measurements as recommended in Meteorology and Atomic Energy. All vertical dilution factors (z), plus those horizontal dilution factors (y) associated with wind speeds at 3 mph or less, were determined by the T Pasquill classes. The horizontal dilution factors (y) associated with winds greater than 3 mph were determined by the Pasquill Classes. On-shore and along shore wind directions were conservatively selected to include the nine sectors NW through SE, clockwise. The hourly relative concentration values occurring with these wind direction were ranked and placed in a cumulative frequency distribution in accordance with the accepted practice at coastal sites for evaluation of accident conditions as shown in Figure 2.3-7. The relative concentration which is exceeded only 5% of the time during the year is 1.3x10-4 CALVERT CLIFFS UFSAR 2.3-11 Rev. 47 seconds per cubic meter. In view of the very conservative nature of the 12 to 50' T data, as evidenced in Figure 2.3-6, and of the conservatism of the data as evidenced by a comparison with the concurrently observed sigma meter readings, this 5 percentile relative concentration value is conservative indeed. It is concluded that, considering both the T and data observed at the Calvert Cliffs site, a relative concentration of 1.3x10-4 seconds per cubic meter is a very conservative value, and is appropriate for the 0-2 hour accident evaluations. This relative concentration is equivalent to a meteorological condition which may be defined as Pasquill E and a wind at 1.4 MPS. Details of the concurrent values of wind speed and direction and T and Pasquill Classes for the Final Annual Record are presented in Figure 2.3-10, Sheets 1 through 14. The same data are presented in Figure 2.3-10, Sheets 13 through 28 except that the 1971 data observed after the Primary Year of Record have been omitted. 2.3.8.4 A Critique of the Data Record for the Independent Evaluation A calendar of data availability is presented in Table 2-30. It can be seen from this table that the Primary Year of Record, November 1969 through October 1970, provides the most complete 12-consecutive-month data record during the November 1969 through October 1971 period. To fill in the data gaps which might be the cause of a seasonal bias in the Primary Year of Record, 1971 data coincident with the dates and times-of-day of the data gaps have been added in this Final Annual Record used in the analysis. A calendar of data in this Final Annual Record is presented in Figure 2.3-8, which shows the dates for which no data is available from November 1969 through October 1971. The sequence of overall data availability is presented in Figures 2.3-8 and 2.3-9.

The Final Annual Record data is quite complete with less than 10% missing observations; 8% occurring in consecutive-day lots, and 2% in periods of less than a day duration. The consecutive-day lots range from four to six days duration, occurring in January, April, August, and November. Because of the distribution of this missing data, it is very unlikely that it has contributed a seasonal or diurnal bias to the data record. The data added to the Primary Year of Record to fill in its gaps, thereby producing the Final Annual Record, were added to ensure that a potential seasonal bias in the data record was eliminated. Data were added only to replace data lost because of equipment malfunctions, and they were only added to the extent that 1971 data, coincident with the dates and times-of-the-day of the equipment malfunction, were available. Although added data constitute 20% of the Final Annual Record, they could not create a bias in the record. 2.3.9 RECENT DATA COLLECTION The following sections summarize the meteorological studies that were conducted to obtain information for use in the design of the Diesel Generator Building for Diesel Generator 1A. 2.3.9.1 Strong Winds As illustrated in Reference 2, the average velocity for CCNPP's "fastest mile" of wind with a mean return period of 100 years is 100 mph. Reference 2 used CALVERT CLIFFS UFSAR 2.3-12 Rev. 47 records of the fastest mile as published by the United States Weather Bureau from data obtained at airport stations. 2.3.9.2 Snow Storms Monthly snowfall depth data from the weather stations at Baltimore, Maryland (1958 to 1989) and the Patuxent River Naval Air Station, Lexington Park, Maryland (1976 to 1992) were used to estimate the 100-year ground snow pack level at the CCNPP site. Frequency analyses were performed on the monthly snowfall records for the months of December, January, February, March, and the combined snowfall total for the months of January and February. The snowfall total for the combined months of January and February, 59", was chosen to represent the 100-year snow pack on the ground. 2.3.10 REFERENCES 1. Meteorology and Atomic Energy 1968, USAEC Division of Technical Information 2. S. C. Hullister, The Engineering Interpretation of Weather Bureau Records for Wind Loading on Structures, Cornell University, Ithaca, NY CALVERT CLIFFS UFSAR 2.3-13 Rev. 47 TABLE 2-10 FIRST YEAR ONSITE METEOROLOGICAL STATIONS AND INSTRUMENTATION CALVERT CLIFFS NUCLEAR POWER PLANT DESIGNATION LOCATION ELEVATION PERIOD INSTRUMENTATION Station 1(a) "N1W" North N11,916 E10,403 100' MSL

+10' Mast 09/14/67-11/11/68 Packard Bell (Beckman-Whitley, Inc.) Model K-100 with Quick-D Vane Wind System 09/14/67- 11/15/67 Cassella Thermograph 12/14/67- 11/11/68 Standard US Weather Bureau Rain and Snow Gauge Station 4 "S1W" Conoy South N8,400 E1,060,000 90' MSL

+50' Mast 11/09/67 Use Discontinued, Date Unknown Packard Bell Electronics Corporation (Beckman-Whitley, Inc.) Model K-101 with Quick-D Vane Wind System Station UT1 Upper N10,000 E8,162 120' MSL 

+4' Shelter 11/15/67 to 12/31/68 Cassella Thermograph installed in standard US Weather Bureau Cotton-Region type shelter Station LT1 Lower N8,642 E9,590 40' MSL

+4' Shelter 11/15/67 to 12/31/68 Same as Station UT1 Test Site Camp Conoy N7,600 E1,055,000 N7,625 E1,000,000 40' MSL +12' Masts 60' MSL 10/17/68 11/01/68 Meteorology Research, Inc. (MRI) Mechanical Weather Station Model 1072 with rain gauge; (2) MRI vector vane Sigma Meter Model 1053 and (3) MRI Mechanical Weather Station Model 1071 (a) Temporary Location CALVERT CLIFFS UFSAR 2.3-14 Rev. 47 TABLE 2-11 RESULTS OF 256 SIMULTANEOUS ONSHORE WINDS IN RAFT STUDY AS COMPARED TO CLASSICAL PASQUILL INVERSION CLASS VALUES SITE PASQUILL CLASS AVG. (rad m/sec) Ave. (m/sec) AVERAGE (degrees) Raft 0.434 4.23 (9.5 mph) 7.2 Station 4 (S1W) 0.492 3.41 (7.6 mph) 8.3 Classical "F" 0.044 1.00 (2.2 mph) 2.5 Classical "E" 0.175 2.00 (4.5 mph) 5.0 CALVERT CLIFFS UFSAR 2.3-15 Rev. 47 TABLE 2-12 ONSITE METEOROLOGICAL SYSTEMS AND INSTRUMENTATION CALVERT CLIFFS NUCLEAR POWER PLANT DESIGNATION LOCATION ELEVATION PERIOD INSTRUMENTATION Microwave Tower N9,770 E8,809 75' MSL +40' & 125' & 220' 8/8/73 - Fall 1993 125' & 200' MRI 2040 Wind Diffusion System 8/8/73 - Fall 1993 40', 125' & 200' Weathermeasure Corporation Aspirated Radiation Shields with Rosemount Sensors (Temperature Gradient System) 8/23/73 - Fall 1993 125' Weathermeasure Corporation Dewpoint System Meteorological Towers a. Primary Tower N10,560 E7,710 110' MSL +33' & 197' 1982 - Current

1982 - Current 1982 - Fall 1995 197' & 33' Wind Sensors

197' & 33' Temperature Sensors 33' Dewpoint Sensor b. Backup Tower N10,422 E7,709 110' MSL +33' 1982 - Current

2005 - Current 2005 - Current 0' Rain Gauge

33' Wind Sensor 33' Temperature Sensor Station 1 "K" Knoll N10,895 E10,435 48' MSL +12' Mast 1/3/69 to 11/4/70 Meteorology Research, Inc. (MRI) Mechanical Weather Station, Model 1072 Wind System with Precipitation Gauge Station 2 "IS" Inner South N9,530 E8,720 48' MSL +12' Mast 1/9/69 to 1/12/70 MRI Mechanical Weather Station, Model 1071 2/11/69 to 2/20/69 MRI Vector Vane Sigma Meter Model 1053 48' MSL +12' & 49.5' 5/15/69 to 9/4/74 Temperature Gradient System, Packard Bell Corp. (Beckman-Whitley) Model 327 Aspirated Radiation Shields 48' MSL +12' Mast 6/1/69 to 8/7/69 MRI 2040 Wind Diffusion System 48' MSL +33' 8/7/69 to 5/15/71 MRI 2040 Wind Diffusion System CALVERT CLIFFS UFSAR 2.3-16 Rev. 47 TABLE 2-12 ONSITE METEOROLOGICAL SYSTEMS AND INSTRUMENTATION CALVERT CLIFFS NUCLEAR POWER PLANT DESIGNATION LOCATION ELEVATION PERIOD INSTRUMENTATION 48' MSL +33' & 97' 5/15/71 - 8/14/75 9/29/71 - 8/14/75 7/17/71 - 8/14/75 MRI 2040 Wind Diffusion System Temperature Gradient System. Weathermeasure Corp. Aspirated Radiation Shields with Rosemount Sensors Beckman-Whitley Model WS-101 Quick Vane Wind System 3/13/72 - 1/11/74 Gill Anemometer Bivane Station 3 "BW" Boundary West N12,375 E6,735 115' MSL +10' Mast 1/9/69 to 1/11/70 MRI Mechanical Weather Station Model 1071 Station 4 "S1W" Conoy South N8,500 E10,550 90' MSL +50' Mast +12' & 49' Mast 1/10/69 to 5/13/75 This station was shut down for reworking during the summer of 1971. It was reactiviated 9/1/71. Beckman-Whitley Model WS101 Wind System

Packard Bell (Beckman-Whitley) Model 327 Aspirated Radiation Shields CALVERT CLIFFS UFSAR 2.3-17 Rev. 47 TABLE 2-13 CUMULATIVE FREQUENCY DISTRIBUTION, PERCENT OF TOTAL OBSERVATIONS, FOR ON/ALONG-SHORE INVERSION WINDS STATION SPEED CLASS 1 2 3 4 meters/sec (K) (IS) (BW) (S1W) 0.01 - 0.50 0.30% 2.52% 1.40% 0.55% 0.51 - 1.00 0.82 4.92 2.92 1.54 1.01 - 2.00 2.46 7.66 6.11 4.91 2.01 - 3.00 4.28 9.66 7.94 8.53 3.01 - 4.00 5.78 10.67 8.55 11.01 4.01 - 5.00 7.44 11.23 8.68 12.59 5.01 - 6.00 8.01 11.36 8.71 13.02 6.01 - 8.00 8.57 11.37 8.71 14.14 8.01 -10.00 8.87 11.37 8.71 14.17 10.01 8.98 11.37 8.71 14.17 8291a 8386a 8399a 6743a _______________________ a Number of valid observations in each record. CALVERT CLIFFS UFSAR 2.3-18 Rev. 47 TABLE 2-14 LOW-FREQUENCY /Q VALUES FOR ON-ALONG SHORE INVERSION WINDS AT CALVERT CLIFFS NUCLEAR STATION % LEVEL OF OCCURRENCE STUB (radian-M/sec) y (M)(1150M) z (M)(1150M) /Q (0-2 hrs) (sec/m3) 1 .097 59 24 1.89x10-4 2 .130 66 24 1.72x10-4 3 .158 74 24 1.56x10-4 4 .185 83 24 1.41x10-4 5 .208 92 24 1.29x10-4 6 .228 103 24 1.17x10-4 7 .243 110 24 1.09x10-4 8 .258 116 24 1.045x10-4 9 .273 124 24 9.85x10-5 10 .287 133 24 9.20x10-5 CALVERT CLIFFS UFSAR 2.3-19 Rev. 47 TABLE 2-15 THE FREQUENCY OF CONCURRENTLY OBSERVED T VALUES FROM THE 50-12 FT AND 97-12 FT LEVELS ABOVE GRADE 97-12 ft 50-12 ft T T 0.5 -0.4 -0.3 -0.2-0.10+0.10.20.3 0.40.50.60.7+0.80.90.9 924 52 41 614668203920 10793115-0.8 12 8 5 533010 002110-0.7 11 4 6 430101 010101 -0.6 19 6 9 813011 213026 -0.5 16 12 9 9216200 000222 -0.4 14 16 11 8321642 010213-0.3 6 12 11 13534832 000011-0.2 4 4 3 9212621 001102 -0.1 4 2 5 949231 1000020 4 3 4 139342 000004+0.1 3 1 5 114162 0111030.2 0 0 0 403222 010105 0.3 0 0 0 100213 110016 0.4 0 0 0 003021 211002 0.5 1 1 0 000131 111006 0.6 1 0 1 313122 02111100.7 0 0 1 002110 0000060.8 1 0 0 000001 005025 0.9 0 0 0 001211 310115 1.0 0 0 0 100000 130117 1.1 0 0 1 001010 001209 1.2 0 0 0 000011 0001251.3 0 0 0 000010 2000281.4 0 0 0 000000 111005 1.5 0 0 0 100000 010006 1.6 0 0 0 000010 022107 1.7 0 0 0 000000 0100181.8 0 0 0 000000 0001061.9 0 0 1 215313 074127215 Observations were made between May 14, 1971 and September 29, 1971. Tables 2-16 through 2-29 were deleted, see Figure 2.3-10. CALVERT CLIFFS UFSAR 2.3-20 Rev. 47 TABLE 2-30 CALENDAR OF METEOROLOGICAL DATA AT STATION 2 FROM NOVEMBER 1969 THROUGH OCTOBER OF 1971 Number of Days with 12 or more hours of valid data. PARAMETER NOV 69 DEC 69 JAN 70 FEB 70 MAR 70 APR 70 MAY 70 JUN 70 JUL 70 AUG 70 SEP 70 OCT 70 T 23 30 24 28 31 14 20 30 31 28 30 28 Wind Dir 27 31 31 28 31 30 31 26 31 27 8 0 Sigma Theta 27 31 31 28 31 30 31 26 31 27 8 0 Wind Speed 30 31 31 27 31 30 31 26 31 27 15 4 All 23 30 24 27 31 14 20 26 31 24 8 0 Running 12-Month Totals T -- -- -- -- -- -- -- -- -- -- -- 317 Wind Dir -- -- -- -- -- -- -- -- -- -- -- 301 Sigma Theta -- -- -- -- -- -- -- -- -- -- -- 301 Wind Speed -- -- -- -- -- -- -- -- -- -- -- 314 All -- -- -- -- -- -- -- -- -- -- -- 258 PARAMETER NOV 70 DEC 70 JAN 71 FEB 71 MAR 71 APR 71 MAY(a) 71 JUN 71 JUL 71 AUG 71 SEP 71 OCT 71 T 30 27 30 28 30 30 31 20 30 24 30 29 Wind Dir 14 14 26 28 20 19 28 20 26 25 30 29 Sigma Theta 14 14 26 28 20 19 28 20 26 25 30 29 Wind Speed 21 31 22 26 20 19 28 28 29 24 27 29 All 8 12 16 26 19 19 28 17 25 16 27 29 Running 12-Month Totals T 324 321 327 327 326 343 354 344 343 339 339 340 Wind Dir 288 271 266 266 255 244 241 235 230 228 250 279 Sigma Theta 288 271 266 266 255 244 241 235 230 228 250 279 Wind Speed 305 305 296 295 284 273 270 272 270 267 279 304 All 243 225 217 216 204 209 217 208 202 194 213 242 _______________________ (a) Much of the wind data from May through September 1971 was observed at 100' above grade. CALVERT CLIFFS UFSAR 2.4-1 Rev. 47 2.4 GEOLOGY 2.

4.1 INTRODUCTION

AND SUMMARY This section of the report presents the results of the geologic phase of the environmental study. This phase of the study included a geologic investigation of the site and surrounding area, a review of pertinent geologic literature (References 1 through 23), and interviews with personnel from government agencies and private organizations (References 24 through 33). Subsurface geologic conditions within the site were investigated in detail by exploratory borings. The site is underlain by approximately 2,500' of southeasterly dipping sedimentary strata of Cretaceous and Tertiary age. Underlying these sediments are crystalline and metamorphic rocks of Precambrian and Early Paleozoic age.

Sediments of the Chesapeake Group of Miocene age underlie the proposed plant area to a depth of about 200'. The material in this group consists of essentially horizontally-stratified sandy and clayey silt with occasional interbeds of sand and shells. It is relatively impervious and dense and provides adequate foundation support for the nuclear power plant. The Miocene sediments are underlain by dense, relatively pervious glauconitic sand and silt of Eocene age. No known or suspected faults are present in the sedimentary strata underlying the site. The closest known faults are located in the Piedmont Province in Western Maryland, approximately 50 miles from the site.

The site is considered satisfactory, from a geologic standpoint, for construction and operation of a nuclear power plant. 2.4.2 REGIONAL GEOLOGY 2.4.2.1 Physiography The site lies within the Coastal Plain Physiographic Province about 50 miles east of the Fall Zone. The Fall Zone separates the low-lying gently rolling terrain of the Coastal Plain from the higher relief of the Piedmont Physiographic Province. The provinces are shown on Figure 2.4-1, Regional Physiographic Map.

The Coastal Plain in Maryland is a low plain rising from sea level to about Elevation +250' at the Fall Zone. Relief in the region ranges generally from about 20 to 100'. The regional slope of the Coastal Plain is to the east at approximately 1.5 ft/mile. The topography of the region is characterized by a series of broad, step-like terraces. The terraces are successively less dissected by stream erosion from west to east. The region is well drained by a large number of small streams. 2.4.2.2 Stratigraphy The general geologic characteristics of the region are shown on Figure 2.4-2, Regional Geologic Map. The Piedmont Province consists of a complex of igneous and metamorphic rocks of Precambrian and Early Paleozoic age with areas of sedimentary and igneous rocks of Triassic age. Beneath the Coastal Plain Province these rocks are concealed by younger strata of Cretaceous and Tertiary age. The buried surface of the basement igneous and metamorphic rocks slopes to the southeast at about 50 ft/mile. In the vicinity of the site, the surface of the basement complex is located approximately 2,500' below sea level. The Cretaceous and Tertiary strata consist of sedimentary deposits of silt, clay, sand, and gravel which exhibit considerable lateral and vertical variations in lithology and CALVERT CLIFFS UFSAR 2.4-2 Rev. 47 texture. The strata form a wedge-shaped mass which thickens to the southeast and pinches out to the northwest toward the Fall Zone. A generalized geologic cross-section of the Coastal Plain is presented on Figure 2.4-3, Regional Geologic Section. A detailed description of the stratigraphy at the site is presented on Figure 2.4-4, Geologic Columnar Section - Site Area.

2.4.2.3 Structure The thick sedimentary strata of the Coastal Plain in the vicinity of the site have remained essentially undeformed since they were deposited up to 135 million years ago. They are believed to have been affected only by slow regional crustal downwarping during their deposition. No known faults have been identified within the Cretaceous and Tertiary sedimentary deposits in the site area. Some local, very shallow folds have been recognized in the Coastal Plain sediments about 40 miles south of the site. These structures are possibly related to depositional conditions rather than to post-depositional tectonic activity. The strata exposed for many miles along the Chesapeake Bay shoreline show no visible signs of faulting or deformation. There is no known fault or geologic evidence of faulting in the deep crystalline rocks in the area. The absence of deformation in the overlying sediments indicates that no major faults are present in the area. Significant tectonic features of the region are shown on Figure 2.4-5, Regional Tectonic Map.

The closest known faults to the site are more than 50 miles to the west in the Precambrian and Early Paleozoic rocks of the Piedmont Physiographic Province. The rocks in the Piedmont are highly folded, and many zones of major faulting have been identified. Most earthquake activity in the region can be related to them. Some of these faults, the closest of which are located about 60 miles southwest of the site, theoretically could be projected beneath the Coastal Plan strata toward the general location of the site. However, such faults are local rather than regionally continuous and appear to be associated with individual fault troughs containing Triassic sediments. Concealed local faults of this type in the basement rock may be responsible for part of the minor earthquake activity in the Coastal Plain of Maryland. 2.4.2.4 Geologic History The recognizable geologic history of the region begins with the deposition of Paleozoic sediments on a Precambrian granitic and metamorphic basement complex. Thick sequences of sedimentary rocks, which accumulated during the Cambrian and Ordovician Periods of geosynclinal deposition were subsequently uplifted, folded, faulted, and metamorphosed during the late Paleozoic Period of mountain building. This activity was followed by another period of uplift along the axis of the Appalachian Mountain chain at the end of the Triassic Period.

Slow regional downwarping of the Coastal Plain started during Early Cretaceous time and continued intermittently through Tertiary time. South and east of the Fall Zone the Piedmont was depressed below sea level providing a base on which the sediments were deposited. Several periods of submergence and emergence resulted in alternate deposition and erosion of continental and marine deposits throughout Cretaceous and Tertiary times. Near the end of the Tertiary Period (Pliocene time) the area is believed to have been above sea level. This resulted in erosion of the sediments deposited CALVERT CLIFFS UFSAR 2.4-3 Rev. 47 previously during Early Pliocene and Late Miocene time, so that Miocene sediments are presently exposed in the site area. During Early Pleistocene time, the ocean advanced westward to the Fall Zone, completely covering the Coastal Plain. Fluctuating sea levels, occurring during Pleistocene time, resulted in alternating periods of erosion and deposition along what are now the major terraces and scarps of the region. A veneer of Pleistocene soils covers most of Coastal Plain. At present, the land is again being submerged by a very slow rise of the sea level. 2.4.3 SITE GEOLOGY 2.4.3.1 General The site is located on the west shore of the Chesapeake Bay in an area characterized by densely wooded, low, flat to gently rolling terrain of low to moderate relief. Ground surface elevations at the site range from sea level to about +130', with an average Elevation of approximately +100'. Nearly vertical cliffs, over 100' high in places, are located along the shore of the Chesapeake Bay. The plant is located in an area near the east edge of the site where the preexisting ground Elevation was about +65'. The final grade Elevation is about +45'. 2.4.3.2 Surficial Deposits The upland areas of the site (areas above Elevation +70') are underlain by sediments of Pleistocene age. These sediments consist primarily of silt and sand, and as encountered at the boring locations, range up to about 50' in thickness. The portion of the site below Elevation +70', which includes the plant area, is underlain by relatively impervious sediments of the Chesapeake Group of Miocene age. The contact between the Pleistocene and Miocene sediments is relatively even and slopes very gently toward the southeast. The surficial geology of the site is shown on Figure 2.4-6, Site Geologic Map. 2.4.3.3 Subsurface Deposits The details of the subsurface geology were investigated primarily by means of ten exploratory borings at the locations shown on Figure 2.4-7, Plot Plan.

The borings ranged in depth from 146' to 332' and were drilled with truck-mounted rotary drilling equipment. Data were obtained from the borings through continuous observation of drill cuttings and examination of undisturbed samples collected by Dames & Moore geologists and engineers.

The soil samples were obtained at intervals in each boring ranging between 3-1/2 and 15', utilizing the Dames & Moore soil sampler illustrated on Figure 2.4-8, Soil Sampler Type U. A few samples were obtained using a standard split-spoon sampler. The number of hammer blows required to drive the sampler a distance of 1' into undisturbed material is recorded in the column entitled "blow count" on the left side of each boring log. The energy used to advance the samplers was greater than that in a standard penetration test, resulting in generally lower blow counts. All samples were examined and logged in the field and then shipped to Dames & Moore's New York office for further examination and appropriate laboratory testing. Detailed descriptions of the materials encountered in the borings are shown on Figures 2.4-9A through 2.4-9J, Logs of Borings. The type of sampler used and data relative to the energy used to advance the sampler are presented on the logs CALVERT CLIFFS UFSAR 2.4-4 Rev. 47 of borings. The depth of ground water after completion of drilling operations and the date on which the borings were completed also are presented on the logs. The site is underlain by a relatively simple sequence of strata, which is shown on Figures 2.4-10A, 2.4-10B, and 2.4-10C, Geologic Sections A-A, B-B, and C-C, respectively. Details of the strata exposed along the shore of the Chesapeake Bay are illustrated on Figure 2.4-11, Schematic Cliff Section, Plant Area.

The Chesapeake Group is approximately 270' in thickness and occurs between Elevation +70' and Elevation -200'. It is composed primarily of gray and green, fine sandy and clayey silt which is relatively impervious. Occasional interbeds of sand and small shells are present, particularly in the upper portion of the group. The upper 15 to 30' of the Chesapeake Group, where exposed in the plant area, have been highly oxidized by weathering. The Chesapeake Group in the region has been divided from top to bottom into the St. Mary's formation, the Choptank formation, and the Calvert formation. For purposes of this study, these formations are essentially identical. Eocene deposits consisting of about 350' of dense, relatively pervious, green, interbedded glauconitic sands and silts with some clays are present below Elevation -200'. The uppermost Eocene deposit, the Piney Point formation, is approximately 40' thick and is composed primarily of glauconitic sand. Because this formation is continuous and distinctive, it provided an excellent horizon for correlation stratigraphy at the site. The contact between the Chesapeake Group and the Piney Point formation occurs at about Elevation -200' and is essentially horizontal throughout the site.

The deeper sediments underlying the Piney Point formation (below an Elevation of about -240') were not investigated, but they have been identified in nearby water wells. The names and descriptions of these formations are shown on Figure 2.4-4. No evidence of faulting was observed at the site in surface outcrops, in the borings, or in the results of the geophysical surveys. A good correlation of subsurface stratigraphy was obtained between the borings. The strata exposed along several miles of the western Chesapeake Bay shoreline in the vicinity of the site show no visible deformation. A view of the slightly dipping strata is shown on Figure 2.4-12, Cliff Face Photograph - Plant Site Vicinity.

A poorly developed crack pattern, believed to be related to desiccation, is exposed in outcrops of the Chesapeake Group strata. These cracks are noticeable in places along the cliffs facing the Chesapeake Bay in material where the effects of weathering are pronounced. 2.4.4 SHORE EROSION The cliffs bordering the Chesapeake Bay along the east side of the site have receded due to shoreline erosion at a maximum rate of about 2' per year. This rate of erosion was calculated from records of measurements made along the shore from 1848 to 1945. The measurements were updated to 1967 by means of recent topographic maps and aerial photographs.

The data indicate that the shoreline at the site receded a maximum of 200' between 1848 and 1945. The average rate of recession along various sections of the eroded coast, including part of the site, has ranged up to 2.1' per year. Changes along the shoreline are shown on Figure 2.4-13, Shoreline Changes. CALVERT CLIFFS UFSAR 2.4-5 Rev. 47 A field check on the rate of shore erosion was provided by an unidentified monument located near the southeast corner of the site. The inscription on this monument reads: "The bank was 55' from this line1 in August 1936." On September 8, 1967, the bank was only 36' from the monument. Therefore, 19' of bank recession has occurred at this location since 1936, representing an average of 0.6' per year.

Shoreline recession along the site is due mainly to wave erosion, particularly storm waves, undercutting the cliff. This results in sloughing of the overlying material. Generally, only the surficial 1'-2' of the cliff face slough at any one time. In the proximity of localized jointing, up to 5' of the cliff face may slough at one time.

Records show that 645 acres of land along a 31.3 mile section of the shoreline in Calvert County were lost between 1848 and 1945 due to erosion. However, during the same interval of time, 115 acres were gained by redeposition. Approximately 3700 lineal ft of shore protection has been placed in front of the plant area, as shown on Figure 1-3A. The shore protection consists of onsite material placed in front of the cliffs and faced with filter cloth and layered riprap. 2.

4.5 REFERENCES

1. J.L. Anderson, 1948, Cretaceous and Tertiary Subsurface-Geology, Maryland Geological Survey Bulletin 2 2. V.K. Bennion, D.F. Dougherty, and R.M. Overback, 1957, Water Resources of Calvert County, Maryland Geological Survey Bulletin 8 3. D.J. Cederstrom, 1945, Structural Geology of Southeastern Virginia, A.A.P.G. Bulletin, Volume 29 4. W.B. Clark, and E.B. Mathews, 1906, The Physical Features of Maryland, Maryland Geological Survey Bulletin, Volume VI, Part 1 5. W.B. Clark, and others, 1904, The Miocene Deposits of Maryland, Maryland Geological Survey 6. C.W. Cooke, 1930, Pleistocent Seashore, Washington Academy of Science Journal, Volume 20 7. N.H. Darton, 1951, Structural Relations of Cretaceous and Tertiary Formations in Part of Maryland and Virginia, GSA Bulletin, Volume 62 8. A.J. Eardley, 1962, Structural Geology of North America, Harper and Row, New York 9. M. Ewing, et al, 1937, Geophysical Investigations in the Emerged and Submerged Atlantic Coastal Plain, GSA Bulletin, Volume 48 10. M. Ewing, J.L. Worzel, and C.L. Pederis, 1948, Propagation of Sound in the Ocean, GSA Memoir 27 11. V.T. Hack, 1957, Submerged River System of Chesapeake Bay, GSA Bulletin, Volume 68 12. W. Harrison, et al, Possible Late Pleistocene Uplift Chesapeake Bay Entrance, Journal Geology, Volume 73 13. P.B. King, 1950, Tectonic Framework of the Southeastern United States, A.A.P.G. Bulletin, Volume 34 1 Line refers to an engraved line on top of the monument.

CALVERT CLIFFS UFSAR 2.4-6 Rev. 47 14. G.E. Murray, 1961, Geology of the Atlantic and Gulf Coastal Province of North America, Harper and Row, New York 15. C.B. Officer, and M. Ewing, 1954, Geophysical Investigations of the Emerged and Submerged Atlantic Coastal Plain, Part 7; GSA Bulletin, Volume 65 16. E.G. Otton, 1955, Ground Water Resources of the Southern Maryland Coastal Plain, Maryland geologic Survey, Bulletin 15 17. H.G. Richards, 1945, Deep Oil Test at Salisbury, Wicomico County, Maryland, A.A.P.G. Bulletin, Volume 29 18. H.G. Richards, and S. Judson, 1965, The Atlantic Coastal Plain and the Appalachian Highlands in the Quaternary of the United States, Princeton University Press, Princeton 19. J.D. Ryan, 1953, The Sediments of Chesapeake Bay, Maryland Geological Survey Bulletin 12 20. G.B. Shattuck, 1906, The Pliocene and Pleistocene Deposits of Maryland, Maryland Geological Survey 21. J.T. Singewald, Jr., and T.H. Slaughter, 1949, Shore Erosion in Tidewater Maryland, Maryland Geological Survey Bulletin 6 22. W.P. Spangler, and N. Peterson, 1959, Geology of the Atlantic Coastal Plain in Delaware, New Jersey, Maryland and Virginia, A.A.P.G. Bulletin Volume 34 23. H.E. Vokes, 1956, Geography and Geology of Maryland, Maryland Geological Survey, Bulletin 19 24. Georgia Institute of Technology, Atlanta, GA, Dr. H. W. Straley 25. Maryland Department of Geology, Mines, and Water Resources, Baltimore, MD, H. Hansen, T. Slaughter 26. Virginia Polytechnic Institute, Blacksburg, VA, Dr. C. E. Sears, Dr. B. N. Cooper 27. American University, Washington, DC, Professor J. Grace 28. George Washington University, Washington, DC, Dr. Geza Teleki

29. Johns Hopkins University, Baltimore, MD, Dr. F. J. Pettijohn, Dr. Taylor (Biologist, Chesapeake Bay Institute) 30. University of Delaware, Newark, DE, Dr. J. J. Groot 31. University of Massachusetts, Amherst, MA, Dr. R. Bromery 32. College of William and Mary, Williamsburg, VA, Dr. K. F. Bick

33. U.S. Geological Survey, Water Resources Division, Baltimore, MD, J. Weigle (Project Geologist for Calvert County)

CALVERT CLIFFS UFSAR 2.5-1 Rev. 47 2.5 HYDROLOGY 2.

5.1 INTRODUCTION

AND SUMMARY This section of the report presents the results of the surface and ground water hydrology phase of the environmental study performed by Dames & Moore. Research for this phase included a review of available pertinent hydrologic literature (References 1 through 5) and interviews with representatives of government agencies and other individuals possessing knowledge of the local area (References 6 through 10).

A study of the hydrologic features of the site and the surrounding area was conducted. This study included an inventory of water wells, an inspection of surface drainage features, a study of aquifers, measurement of ground water levels in exploratory borings, and an analysis of the depth, direction, and rate of ground water movement.

The site is well drained and not susceptible to flooding. Surface runoff is moderately high and accounts for about 35% of the total annual precipitation. Average annual precipitation in the region ranges from about 40.6" at the Patuxent NATC to about 44" at Prince Frederick. A drainage divide extends across the site in a general north-south direction. The area east of the divide (20% of the site) drains into the Chesapeake Bay, whereas the area to the west drains into local tributaries and eventually into the Patuxent River. The plant is located east of the divide where surface drainage is toward the Chesapeake Bay. The plant area is underlain by over 200' of relatively impermeable deposits which effectively confines the underlying artesian aquifers and minimizes their possible contamination by the downward percolation of an accidentally discharged contaminated liquid. The vertical component of ground water movement through the Chesapeake Group is upward. This precludes the possibility of contamination of the aquifers due to downward percolation of a contaminant.

Most of the potable water used in the region is obtained from the artesian aquifers underlying the Chesapeake Group. The aquifers are composed of glauconitic sand and silt of the Piney Point, Nanjemoy, and Aquia formations. The piezometric surfaces of these water-bearing formations slope to the southeast at about 2' per mile. Based upon this hydraulic gradient and coefficients of permeability for these formations, the estimated average rate of natural ground water movement is less than 1" per day.

A limited quantity of potable water is obtained from shallow wells completed in the surficial Pleistocene deposits which overlie the Chesapeake Group throughout most of the area surrounding the site. The areas in which these materials are utilized as a source of water are up-gradient from the plant and cannot be affected by the accidental release of contaminated liquids at or below the ground surface in the plant area. The possibility of adversely affecting the available ground water resources or existing wells in the area, by the construction and operation of a nuclear facility, is remote. The hydrologic characteristics of the site are favorable for the construction and operation of a nuclear power plant. 2.5.2 SURFACE HYDROLOGY 2.5.2.1 General Calvert County is a peninsula bounded on the east by the Chesapeake Bay and on the west by the Patuxent River. The area is characterized by gently rolling terrain with a dendritic drainage pattern. A drainage divide extends longitudinally across the county. The county is well drained by a relatively large number of streams, although most are less than seven miles long. Many streams have moderately CALVERT CLIFFS UFSAR 2.5-2 Rev. 47 steep valley walls, while others form estuaries to the Patuxent River. Swampy areas and tidal flats are common along the coastal areas. Stream flow in Calvert County is measured at two gauging stations maintained by the U.S. Geological Survey. Their locations are shown on Figure 2.5-1, Map of Area Showing Surface Hydrology. The gauge on St. Leonard Creek is a continuous recording station, while the Hellen Creek station is a partial recording site. Average monthly discharges at the continuous recording station are presented in Table 2-31, Average Monthly Discharge at Gauging Station on St. Leonard Creek (1958-1964). The average runoff measured at the St. Leonard gauge from 1958 to 1964 was 15-1/2"/year. The average annual precipitation for the same period as measured at Prince Frederick, about 10-1/2 miles north of the site, was about 44". Thus, runoff accounts for about 35% of the total precipitation. Evapotranspiration also accounts for a large portion of the annual precipitation. Relatively little precipitation percolates into the surficial materials to recharge the phreatic surface.

The reason for the high runoff and low infiltration probably can be attributed to the impermeable nature of the Miocene subsoils which retard downward percolation of water. The surficial soils are Pleistocene or Recent deposits which are relatively pervious. Rainfall absorbed by them is soon discharged as stream runoff or lost through evapotranspiration. Many lowland areas of the county are not mantled by Pleistocene deposits and the relatively impermeable Miocene deposits are exposed. Precipitation falling on these areas is discharged almost immediately as surface runoff. 2.5.2.2 Site Conditions The topography at the site is gently rolling with steeper slopes along stream courses. Local relief ranges up to about 130'. The site is well drained by short, intermittent streams. A drainage divide, which is generally parallel to the coastline, extends across the site as shown on Figure 2.5-1. The area to the east of the divide comprises about 20% of the site and includes the plant area. This area drains into the Chesapeake Bay. The western area is drained by tributaries of Johns Creek and Woodland Branch, which flow into St. Leonard Creek and subsequently into the Patuxent River. Grading performed during construction has not substantially altered the present drainage system. The average Elevation of the site is about 100' above mean sea level. The site occupies the head-water area of several small drainage basins, and is not subject to flooding. It is possible that high intensity rain storms may cause water to back up in some valleys due to local constrictions in the stream beds, but this would be a temporary situation. The plant area has an Elevation of about +45' and has a storm drain system to handle runoff. High water levels in the bay due to storm conditions are discussed in Section 2.8.3.

Site grading in the vicinity of the Diesel Generator Buildings provides a system of swales that direct overland flow of the probable maximum precipitation runoff without producing drainage or flooding problems for the buildings. The system of swales direct runoff to the Chesapeake Bay without any dependence on the site's storm drain system. The results of the runoff and backwater analyses indicate that during the probable maximum precipitation storm the swale system at the Diesel Generator Building site will convey the surface runoff with a maximum water level of 44.8' above sea level near the Diesel Generator Buildings. This water level is below the floor grade of the Diesel Generator Buildings, which is 45.5' above sea CALVERT CLIFFS UFSAR 2.5-3 Rev. 47 level, and thus precludes the potential for flooding of the Diesel Generator Buildings during the probable maximum precipitation. 2.5.3 GROUND WATER HYDROLOGY 2.5.3.1 Regional Conditions Ground water occurs in the surficial soils and is tapped by many shallow dug and driven wells. Ground water in deeper aquifers occurs under artesian conditions. These aquifers, the Piney Point, Nanjemoy, and Aquia formations, are separated from the surficial deposits by an aquiclude averaging about 270' in thickness. Recharge to these aquifers occurs in their outcrop areas about 15 to 30 miles west of the site. The geologic position of these aquifers relative to other formations in Calvert County is presented in Table 2-32, Geologic Units in Calvert County. The hydrologic characteristics of the Piney Point, Nanjemoy, and Aquia formations are discussed in greater detail in the following subsections. 2.5.3.1.1 Piney Point Formation The Piney Point formation consists of glauconitic sand interspersed with shell beds and a little clay. Well cuttings and particle-size analyses indicate that the aquifer is composed mainly of medium to fine sand. The formation occurs as a wedge-shaped geologic unit and is known only in Southern Maryland. It is about 30' thick in the vicinity of the site and increases in thickness to the southeast.

The Piney Point formation is widely utilized as a source of ground water in Calvert County and adjoining St. Mary's County. It is estimated that more than 500 domestic wells in these two counties tap the Piney Point and the underlying Nanjemoy formations. These two aquifers are hydrologically connected. The yields of domestic wells generally range from about 3 to 20 GPM. At the Patuxent NATC, located about 10 miles south of the site, large-capacity wells tap the Piney Point aquifer and the upper part of the Nanjemoy formation. These wells each yield as much as 190 GPM. The specific capacities of 25 selected wells in these formations range from 0.1 to 3.3, and average 1.2 GPM/ft of drawdown.

2.5.3.1.2 Nanjemoy Formation The lower part of the Nanjemoy formation consists of an impermeable red clay known as the Marlboro Clay. The remainder of the formation consists chiefly of greensand, but contains some clayey greensand. A limited number of particle-size analyses indicate that the sand is predominantly medium- to fine-grained. The Nanjemoy formation is an important aquifer in Calvert County where it is tapped by several hundred wells. Most wells are completed in the permeable water bearing sands occurring in the uppermost 80' of the formation and yield less than 10 GPM. The specific capacities of 11 wells in Calvert County tapping this aquifer range from 0.2 to 2.4, and average 0.8 GPM/ft of drawdown. On the basis of water level recovery measurements made during a pumping test, a coefficient of transmissibility of approximately 2,000 GPD/ft has been computed. The field coefficient of permeability was 66 GPD/ft2. The results of a similar test in Prince George's County indicate coefficients of transmissibility ranging from 260 to 840 GPD/ft. CALVERT CLIFFS UFSAR 2.5-4 Rev. 47 2.5.3.1.3 Aquia Formation The Aquia formation is characterized by an abundance of glauconitic sand with some quartz sand and clay. The thickness of permeable sandy beds in the formation ranges up to slightly more than 40' in parts of Calvert County. Particle-size analyses of nine samples at the Aquia formation show that the sand is medium- to fine-grained.

The most productive wells tapping this formation are at the Patuxent NATC. Yields of individual wells range from 125 to 350 GPM. The specific capacities of eight of these wells range from 0.8 to 4.2 and average 2.5 GPM/ft of drawdown. The results of six pumping tests indicate field coefficients of permeability ranging from 130 to 1,340 GPD/ft2. Coefficients of transmissibility determined from the tests range from 5,500 to 33,000 GPD/ft2. Plant wells tap this formation.

2.5.3.1.4 Water Levels The artesian head of the three principal aquifers in Calvert County is generally above sea level. The effect of tidal fluctuations on water levels is noticeable in two observation wells, completed in the Nanjemoy and Piney Point formations, at Solomons Island about 7 miles south of the site. Recorder charts from these wells, which are 248 and 493' deep, respectively, show semi-diurnal fluctuations of about 1/2'. The approximate configurations of the piezometric surfaces of the Aquia and Nanjemoy formations are shown on Figure 2.5-2, Piezometric Surfaces in Calvert County. The regional hydraulic gradient in the vicinity of the proposed plant site is to the southeast. However, local minor variations occur. The cone of depression at the southern end of Calvert County is the result of ground-water extraction at the Patuxent NATC. Records of observation wells maintained by the U.S. Geological Survey indicate that water levels in the area have remained essentially unchanged since 1963. If the future rate and amount of ground-water extraction in the area is not significantly changed, it is likely that the cone of depression will remain constant and the existing hydraulic gradient in the vicinity of the site will be maintained. 2.5.3.2 Water Use 2.5.3.2.1 General Nearly all potable water used in Calvert County is from subsurface sources. Since little industry is located in this area, the major use of water is for domestic and agricultural purposes.

2.5.3.2.2 Public Water Supplies In 1967, there were 12 towns in Calvert County with public water supplies. The output from these systems is relatively small, but increases substantially in the summer to accommodate the seasonal population increase. Data concerning the public water supplies are presented in Table 2-33, Public Supply Wells in Calvert County. The locations of these supplies are shown on Figure 2.5-3, Public Water Supplies in Calvert County. CALVERT CLIFFS UFSAR 2.5-5 Rev. 47 2.5.3.2.3 Private Wells Most domestic water supplies in Calvert County are obtained from private wells greater than 300' in depth. In some instances, other wells are less than 50' deep and are of limited capacity. The locations of the deep wells, in the vicinity of the site, are shown on Figure 2.5-4, Map of Area, Showing Known Water Wells. Information pertaining to these wells is presented in Table 2-34, Known Water Wells, Vicinity of Site.

Shallow dug or driven wells are not tabulated or shown on Figure 2.5-4. The shallow wells will not be affected by changes in the ground water regimen at the site since they are at a higher elevation than the proposed plant grade.

Wells numbered 2 & 10 are located on BG&E property and were previously owned by YMCA. One is in use supplying water to our recreational pool facility for authorized personnel including BGE employees and the Red Cross. The same aquifer supplies water to this well and three others located close to the reactors on the plant site. Output from the three onsite wells referred to in Section 2.5.3.3 is pumped to two storage tanks from the well water treatment building. There is no comprehensive source of public information on dug wells in Calvert County. The Calvert County Health Department records and retains dug-well records for only five years. Permits are not required by the Maryland Department of Water Resources nor by the Maryland Department of Health. Several dug wells are listed in Reference 2 and Section 2.5.4.2; however, accurate public information is not available for most of them. It is believed that dug wells in the area are upgradient from the plant site or are across drainage area boundaries. 2.5.3.3 Site Conditions The depth of ground water at the site was measured in piezometers installed in seven of the Dames & Moore exploratory borings. The piezometers consisted of small-diameter steel pipe equipped with a well point, or perforated PVC pipe. They were installed in borings DM-1, DM-2, DM-3, DM-5, DM-7, DM-8, and DM-9 immediately after completion of the drilling operations. The locations of the borings are shown on Figure 2.4-7, Plot Plan. The water level recorded in each piezometer is shown on the Log of Borings, Figures 2.4-9A through 2.4-9J in Section 2.4, Geology.

An in-situ soil percolation test was performed at the site in Miocene soils typical of those underlying the proposed plant. The test was conducted in a one-foot square hole in accordance with the procedure used for the Corps of Engineers Soil Absorption Test (Reference 5). Results of this test indicate a permeability of less than 1 GPD/ft2. The location of the test is shown on Figure 2.4-7. Representative samples extracted from the exploratory borings were subjected to a laboratory testing program in order to evaluate the permeability characteristics of the natural soils and the physical properties of the material for correlation purposes. The laboratory program included the following tests: a. moisture and density determinations;

b. particle-size analyses; c. permeability tests; and d. cation exchange and X-ray diffraction analyses.

CALVERT CLIFFS UFSAR 2.5-6 Rev. 47 The moisture and density determinations were performed on undisturbed samples for correlation purposes. The results of these determinations are shown on the Log of Borings in Section 2.4, Geology. Selected soil samples were tested in order to measure their grain-size distribution. The results of these analyses were used in evaluating soil permeability and for classification purposes. The results are presented on Figures 2.5-5A and 2.5-5B, Particle Size Analyses. Two permeability tests were performed on materials typical of those underlying the plant. The results of these tests, which were performed in accordance with the American Society for Testing and Materials (ASTM) procedures, are presented in Table 2-35, Laboratory Permeability Tests. The clay mineral content and the total cation exchange capacity of seven selected soil samples was analyzed. The results of these tests are presented in Table 2-36, Cation Exchange and X-ray Diffraction Analyses. Data obtained from the geologic exploratory borings indicate that a large portion of the site is mantled by relatively permeable Pleistocene soils. These soils have been eroded from a portion of the site exposing the Chesapeake Group which includes the St. Mary's, Choptank, and Calvert formations. The Chesapeake Group consists of about 270' of impervious sandy and clayey silts of Miocene age. Underlying this material are the Piney Point, Nanjemoy, and Aquia formations of Eocene age. The Pleistocene deposits consist mainly of silts and sands which have fairly good infiltration characteristics. A few domestic wells in the area obtain water from this material, but are up-gradient from the plant and cannot be affected by a change in the ground water regimen at the site. Grain-size analyses of the surficial soil samples indicate a maximum permeability coefficient of about 400 GPD/ft2. The elevation of the phreatic surface changes with the surface topography and can be expected to fluctuate slightly as a result of climatic changes. The water table occurs generally within 30' of the ground surface. East of the topographic divide, the direction of ground water movement is toward the Chesapeake Bay. The direction of ground water flow west of the divide is toward the existing stream valleys. Piezometers installed in the borings show that the ground water gradient at the site is generally less than 1%. The average rate of ground water flow probably does not exceed a few feet per day. The underlying impervious sandy and clayey silts of the Chesapeake group extend to about 200' below mean sea level. A percolation test conducted near the plant indicates a permeability of less than 1 GPD/ft2. Particle-size analyses indicate that the permeability of the Chesapeake Group averages about 3 GPD/ft2. The rate of ground water movement is extremely low (much less than 1" per day). The formation is an aquiclude which effectively confines the underlying artesian aquifers. Regional studies by the U.S. Geological Survey have shown that the head in the artesian aquifers is above sea level. The result is vertical upward leakage through the Chesapeake Group. The rate of leakage is extremely low because of the low permeability of the Miocene sediments. At the site, the combined thickness of the aquifers within the Piney Point and Nanjemoy formations is about 80'. They occur at Elevations ranging between 200' and 300' below mean sea level and are separated from the deeper Aquia formation by a layer of clay (Lower Nanjemoy) about 150' thick. The general CALVERT CLIFFS UFSAR 2.5-7 Rev. 47 direction of ground water movement in the Aquia formation is toward the southeast with a piezometric gradient of about 2' per mile. Grain-size analyses of samples of the Piney Point formation collected at the site indicate a permeability of about 150 GPD/ft2. This value is probably typical of both the Piney Point and Nanjemoy aquifers. It is estimated that the permeability coefficient of the Aquia formation may be on the order of 1,000 GPD/ft2. The computed rate of flow of ground water through these aquifers ranges from about .07 to .004' per day. The possibility of accidental contamination of the Eocene aquifers beneath the site is remote because; (a) the aquifers are covered by over 200' of relatively impervious soils and, (b) the vertical component of ground water movement is upward.

Cation exchange and X-ray diffraction analyses were performed on seven samples ranging in depth from 5 to 115' in order to evaluate the cation retention characteristics in the vicinity of the site. The cation exchange capacity of these soils is relatively high and would effectively absorb radioactive cations.

Three wells have been developed for plant use. They extend to a depth of approximately 640' and are each capable of producing 300 GPM from the Aquia formation. Casings for the three plant wells are continuous and sealed with grout to the top of the screens. The effect of pumping these wells will be to create a cone of depression in the Aquia formation such that the direction of groundwater flow will be toward the site rather than away from it. This will further minimize the possibility of lateral migration of any possible release of contaminated liquids beyond the site boundary. For other aquifers, the upward component of groundwater movement and the overlying aquiclude prevent contamination due to downward percolation. 2.

5.4 REFERENCES

1. J.L. Back, 1966, Hydrochemical Facies and Ground Water Flow Patterns in the Northern Part of the Atlantic Coastal Plain, USG Professional Paper 498-4 2. V.K. Bennion, D.F. Dougherty, and R.M. Overback, 1951, Water Resources of Calvert County, Maryland, Geological Survey, Bulletin 8, pp 100 3. E.G. Otton, 1955, Ground Water Resources of the Southern Maryland Coastal Plain, Maryland Geological Survey, Bulletin 15 4. H.E. Vokes, 1956, Geography and Geology of Maryland, Maryland Geological Survey, Bulletin 19 5. U.S. Corps of Engineers, 1959, Sewage Treatment Plants, Engr. Manual 345-243 6. U.S. Geological Survey Water Resources Division; Towson, MD; Mr. James Weigle, Mr. Wayne E. Webb 7. Maryland State Department of Health; Baltimore, MD; Mr. Charles Gross 8. Maryland Department of Geology Mines and Water Resources; Baltimore, MD; Mr. H. Hansen 9. Maryland Division of Water Resources; Annapolis, MD; Dr. Bruce Martin

10. Numerous local residents; Vicinity of Site CALVERT CLIFFS UFSAR 2.5-8 Rev. 47 TABLE 2-31 AVERAGE MONTHLY DISCHARGE AT GAUGING STATION ON ST. LEONARD CREEK (1958-64)(a) January 1.74" May 1.35" September 0.48" February 1.92" June 0.88" October 0.56" March 2.46" July 1.00" November 1.23" April 2.04" August 0.83" December 0.99"

_______________________ (a) Expressed in inches of runoff. CALVERT CLIFFS UFSAR 2.5-9 Rev. 47 TABLE 2-32 GEOLOGIC UNITS IN CALVERT COUNTY GEOLOGIC UNIT APPROXIMATE RANGE IN THICKNESS (ft) PHYSICAL CHARACTERISTICS WATER BEARING PROPERTIES Pleistocene surficial deposits 0-150 Silt and sand with some clay gravel Yields small quantities of water to relatively shallow dug or driven wells. Chesapeake Group: St. Mary's, Choptank, and Calvert Formations 30-325 Sandy and clayey silt with inter-bedded sand and fossiliferous layers An aquiclude. Yields small supplies of water to a few dug wells. Piney Point Formation 0-60 Glauconitic sand Yields up to 200 GPM are reported from drilled wells. An important aquifer in Calvert County. Nanjemoy Formation 40-240 Glauconitic sand with clayey layers. Basal part is red or gray clay Yields of individual wells reported up to 60 GPM. An important aquifer in Calvert County. Aquia Formation 30-200 Green to brown glauconitic sand Yields up to 300 GPM reported from wells. An important aquifer in Southern Maryland. Brightseat Formation 0-40 Gray to dark gray micaceous silty and sandy clay Not known to be an aquifer in Southern Maryland. Monmouth and Matawan Formations 20-135 Sandy clay and sand, dark gray to black, with some glauconite Not a major aquifer in Southern Maryland, but yields up to 50 GPM have been reported. Magothy Formation 0-40 Light-gray to white sand and gravel with interbedded clay layers A few wells reportedly yield up to 1,000 GPM but average yields are considerably less. This aquifer is not used in Calvert County because of its depth. Raritan Formation 100 Interbedded sand and clay with iron-stone nodules Yields up to a few hundred GPM reported. Not utilized in Calvert County due to depth. Patapsco Formation 100-650 Interbedded sand, clay, and sandy clay Large-diameter wells yield up to 1,000 GPM. Not used in Calvert County because of depth. Arundel Clay Formation 25-200 Red, brown, and gray clay Not generally a water-bearing formation. CALVERT CLIFFS UFSAR 2.5-10 Rev. 47 TABLE 2-32 GEOLOGIC UNITS IN CALVERT COUNTY GEOLOGIC UNIT APPROXIMATE RANGE IN THICKNESS (ft) PHYSICAL CHARACTERISTICS WATER BEARING PROPERTIES Patuxent Formation 100-450+ Chiefly gray and yellow sand with interbedded clay Yields of several hundred GPM reported. Not used in Calvert County due to great depth. Precambrian Unknown Gneiss, granite, gabbro, meta-gabbro, quartz diorite, and granitized schist Yields moderate supplies of ground water, generally not more than 50 GPM. Not used in Calvert County because of its great depth. CALVERT CLIFFS UFSAR 2.5-11 Rev. 47 TABLE 2-33 PUBLIC SUPPLY WELLS IN CALVERT COUNTY (1967) TOWN POPULATION SERVED(a) NUMBER OF CONNECTIONS AVERAGE OUTPUT (mgd) WELL TOTAL DEPTH (ft) DIA (in) Calvert Beach 60 15 .006 1 475 5 Chesapeake Beach 500 155 .05 1(b) 2 3 4(b) 5 400 400 400 400 400 8 8 6 2 2 Chesapeake Ranch Estates 150 1 2 3 400 750 750 4 4 4 Dares Beach 600 175 .05 1(b) 2 217 210 1 1/2 2 1/2 Hunting Hills 30 8 .003 1 4 Kenwood Beach 250 62 .025 1 365 1 1/2 Long Beach 720 180 .07 1 2 3 4 525 500 475 500 3 4 4 4 Prince Frederick 125 35 .02 1 552 8 St. Leonard 60 14 .006 1 550 5 Scientists Cliffs 120 185 .025 1 240 6 Western Shores 1 325 3 White Sands(c) 160 (approx.) Comm. 40 2 .012 .015 1 2 402 315 6 2 1/2 _______________________ (a) Does not include seasonal increases, with the exception of Chesapeake Ranch. (b) Not in use. (c) Appropriate permits data per Maryland Department Natural Resources ground water. NOTES: 1) Information based on 1963 data from United States Public Health Service and Maryland Department of Health. 2) Total population of a community is not necessarily served by the public water supply. CALVERT CLIFFS UFSAR 2.5-12 Rev. 47 TABLE 2-34 KNOWN WATER WELLS, VICINITY OF SITE OWNER WELL NUMBER WELL DEPTH (ft) WELL DIAMETER (in) AMOUNT OF CASING (ft) YIELD (GPM) K. C. Gerard 1 365 4 280 22 BGE 2 585 6 560 60 Yacht Club 3 315 2 1/2 189 25 H. Krellen 4 285 2 180 15 E. Zinn 5 384 4 237 25 B. Foot 6 399 2 1/2 252 20 H. C. Wilder 7 399 2 1/2 273 20 H. J. Mishou 8 404 2 1/2 277

• D. Wood 9 315 2 273 5 BGE 10 540 6 520 10 R. C. Hall 11 274 2 1/2
• 50 G. D. Wait 12 252 2
• 3 D. Adams 13 300 2 1/2 *
• William Rekar 14 461 3 *
• E. Bowen 15 525 3 *
• E. Daniels 16 450 * *
• W. Jenkin 17 435
• 140
• Knotty Pine Bar & Grill 18 * * *
• Bay Breeze Camp 19 340 * *
• Mrs. Faron 20 * * *
• Mrs. Moran 21 * * * *
• 22 365 2 1/2 *
• Mr. McQueen 23 300 * *
• Mr. Street 24 375 * *
• P. Andjiano 25 360 * *
• Mr. Bancroft 26 * * *
• Mr. Wohlgenmuth 27 * * *
• Mrs. Mansfield 28 * * * *
• 29 * * * *
• 30 * * * *
• 31 * * *
• _______________________
• Information not available.

NOTE: Well locations are shown on Figure 2.5-4.

CALVERT CLIFFS UFSAR 2.5-13 Rev. 47 TABLE 2-35 LABORATORY PERMEABILITY TESTS BORING DEPTH (ft) SOIL TYPE DRY DENSITY (lbs/ft3) COEFFICIENT OF PERMEABILITY (ft/day) DM-6 45 Sandy silt with shells 93 3.18 DM-8 75 Sandy silt 82 0.86 CALVERT CLIFFS UFSAR 2.5-14 Rev. 47 TABLE 2-36 CATION EXCHANGE AND X-RAY DIFFRACTION ANALYSES BORING DEPTH SOIL TYPE GRADATION IN % FINER(a) % OF TOTAL CLAY MINERALS(b) TOTAL CATION EXCHANGE CAPACITY(c) (ft) Montmorillonite and Mixed Clay .074 .048 .005 .002Illite Minerals Chlorite Test 1 Test 2 (in millimeters) DM-1 45 Gray Silty Clay 98 44 23 17 30 50 20 23.8 24.5 DM-6 43 1/2 Green Silty Sand 28 25 21 15 40 60 - 10.0 11.0 DM-6 115 Green Clayey Silt 98 91 44 31 30 50 20 24.3 27.3 DM-8 30 Green Silty Sand 28 25 22 15 35 65 - 13.8 12.8 DM-8 45 Green Clayey Silt 90 80 45 35 20 60 20 25.0 30.6 DM-9 5 Reddish-Brown Sandy Clay 80 65 38 34 - - 100 10.0 10.3 DM-10 47 Gray Sandy Silt 59 22 16 11 20 80 - 14.4 10.0 _______________________ (a) Soil samples soaked for 24 hours in 0.4% hours in 0.4% sodium hexametaphosphate before hydrometer analysis. (b) X-ray diffraction analyses of minus 2 micron material. (c) Because of the CaCO3 in some of the samples, the ammonium acetate method was used. Total Cation Exchange was determined on the minus 40 micron material. NOTE: Swell observed in samples DM-1 (45'), DM-6 (115'), and DM-8 (45').

CALVERT CLIFFS UFSAR 2.6-1 Rev. 47 2.6 SEISMOLOGY 2.

6.1 INTRODUCTION

AND SUMMARY This section of the report presents the results of the engineering seismology phase of the environmental study. This phase of the study included literature research to compile a record of the seismicity of the area (References 1, 2, 6, and 8), evaluation of the geologic structure and tectonic history of the region, a program of dynamic laboratory soil testing, and analyses to evaluate the response of the foundation materials to earthquake-type loading. Field geophysical studies were performed to evaluate the in-situ dynamic properties of the foundation materials. The site is located in a region which has experienced infrequent and minor earthquake activity. Most of the reported earthquakes are related to known faulting more than 50 miles west of the site in the Piedmont Physiographic Province. No known faults occur in the vicinity of the site. The closest earthquake (Intensity VII) which caused any structural damage occurred about 80 miles southwest of the site. Several minor shocks (no greater than Intensity V) have been reported within 50 miles of the site. Because of very limited data, it is not possible to determine whether or not these were of tectonic origin. The foundations of major plant structures are established in dense Miocene soil which will not undergo reduction in strength or increased settlement under safe shutdown earthquake (SSE) conditions.

Significant earthquake ground motion is not expected at the site during the economic life of the nuclear facility. On a conservative basis, the power plant was designed to respond elastically, with no loss of function, to horizontal ground accelerations as high as 8% of gravity and vertical ground accelerations as high as 5-1/3% of gravity.

For safe shutdown of the reactor, a maximum horizontal ground acceleration of 15% of gravity was used in the design. A maximum vertical acceleration of 10% of gravity was used in the design. These accelerations are based on an assumed Intensity VII earthquake originating near the site.

The results of the regional study of seismicity and tectonics show that the aforementioned ground accelerations are conservative. Therefore, the nuclear power plant designed to these parameters meets all safety requirements. It is not expected that the plant will be subjected to a significant tsunami effect. The maximum expected tsunami would not result in more than minor wave action at the site and, thus, was not significant in the design. Geological and geophysical investigations performed since the late 1970s have led to speculation that many large Cretaceous and Cenozoic fault zones may exist in the Atlantic Coastal Plan. The closest of these are the Stafford and Brandywine fault systems, located about 45 and 29 miles west of the site, respectively.

An update of the seismic activity within 200 miles of the site focuses on two areas. The first is the occurrence of any seismic events since the last investigation that would be of greater significance than those previously considered. The second area is the evaluation of the earthquake catalog data that would lead to significant changes to the original SSE assessment. A significant earthquake catalog was developed by investigators and consultants of the Electric Power Research Institute in 1988 and the National Center for Earthquake Engineering Research in 1992. An independent evaluation of regional seismicity, performed by Bechtel Power Corporation in 1992, found that while more seismic events have been cataloged than were previously considered in the Updated CALVERT CLIFFS UFSAR 2.6-2 Rev. 47 Final Safety Analysis Report, none of the additional events were larger or were in new or significantly different areas than those used to develop the SSE design basis. The basic premise of SSE specifications at Calvert Cliffs, an intensity MMI VII earthquake in the Atlantic Coast Plain province in the vicinity of the site, remains unchanged. Reference 13 discusses the results of additional analyses of vibratory ground motion which were performed to determine the design ground acceleration level for the Diesel Generator Buildings. This supplemental information applies to the power block as well as to the Diesel Generator Buildings. 2.6.2 TECTONICS The site is located in the Coastal Plain Physiographic Province. This province is bounded on the east by the Atlantic Ocean, and on the west by the Fall Zone and the Piedmont Physiographic Province. The Coastal Plain consists of easterly dipping Cretaceous and Tertiary sediments which are about 2,500' thick at the site. Crystalline basement rock outcrops near the Fall Zone about 50 miles west of the site. A graphical representation of the subsurface conditions at the site is shown on Figure 2.6-1, Columnar Section, Showing Geophysical Data. On the basis of regional data, the Cretaceous and Tertiary sediments are undeformed. The absence of folding and faulting in the sedimentary strata indicates that displacements along unknown faults which may be present in the basement have been negligible.

No known faults occur within the basement rock or sedimentary deposits in the vicinity of the site. The closest known fault systems are found in the rocks of the Piedmont, more than 50 miles west and northwest of the site. The Piedmont Province consists of igneous and metamorphic rocks of Precambrian and early Paleozoic Age, with areas of sedimentary and igneous rocks of Triassic Age. Major tectonic activity that has occurred in the Mid-Atlantic Region can be related to known faults in the Piedmont Province.

2.6.3 SEISMICITY The site is situated in a region which has experienced only infrequent minor earthquake activity. No shock within 50 miles of the site has been large enough to cause significant structural damage. Since the region has been populated for over 300 years, it is probable that all earthquakes of moderate intensity, say VI or greater, would have been reported during this period. It is very likely that all earthquakes of Intensity V or greater which occurred within the last 200 years have been reported.

The first report of earthquake occurrence in the general area of the site dates back to the late 18th Century. Since then, only 14 earthquakes with epicentral intensities of V or greater on the Modified Mercalli(a) Scale have been reported within about 100 miles of the site. None of these shocks was greater than Intensity VII. Few were of high enough intensity to cause structural damage and only one of these shocks can be considered more than a minor disturbance. This was an Intensity VII shock near Wilmington, DE in 1871 about 100 miles from the site. A list of earthquakes of Intensity V or greater with epicenters located within a distance of about 100 miles of the site is presented in Table 2-37, Significant Earthquakes within 100 miles of the site. Several smaller earthquakes, which are significant because of their proximity to the site, are also included in Table 2-37. The locations of these and other earthquakes in the region surrounding the site are shown on Figure 2.6-2, Epicentral Location Map. Several small shocks are shown on the Epicentral Location Map, but not indicated in Table 2-37. Little information is available regarding these shocks. The indicated epicentral locations are those suggested by G.P. Woollard (Reference 12). CALVERT CLIFFS UFSAR 2.6-3 Rev. 47 Most of the reported earthquakes in the region have occurred in the Piedmont Physiographic Province west of the Fall Zone. The closest approach of the Fall Zone to the site is about 50 miles. These shocks were generally related to known faults in the Piedmont rocks. There have been several large shocks with epicenters in the Coastal Plain, some of which were damaging. The largest of these is the Charleston, South Carolina, earthquake of 1886, which has an epicentral intensity of about IX. Geological and seismological research in the meizoseismal vicinity of Charleston appear to support the view that the Charleston earthquake occurred in association with a specific seismogenic structure located near Charleston, and there is no need to consider a site intensity of MMI X for seismic design at Calvert Cliffs.

While the Giles County, Virginia Seismic Zone is just over 200 miles from the Calvert Cliffs site, it is relevant to discuss this seismically active zone in that it was the location of a MMI VII intensity earthquake on May 31, 1897. However, the seismic activity around Giles County appears to be very distinct in character as compared to seismicity in the nearer Central Virginia Seismic Zone. There appears to be no reason to assume the 1897 Giles County earthquake is applicable to the Calvert Cliffs site seismicity. The largest earthquake in the Coastal Plain close enough to the site to be of significance in the current study occurred in 1927 near the northern New Jersey coast, about 180 miles northeast of the Calvert Cliffs site. The epicentral intensity of this earthquake was VII. Three shocks were felt over an area of about 3,000 square miles from Sandy Hook to Tom's River. Highest intensities were felt from Asbury Park to Long Branch where several chimneys fell, plaster cracked, and articles were thrown from shelves. This shock has not been related to any known geologic feature.

An earthquake which occurred near Wilmington, DE, in 1871 is the largest reported earthquake within 100 miles of the proposed plant site. It is not possible to accurately locate the epicenter of this shock with the limited data available, but it is probable that the shock occurred along the Fall Zone about 100 miles northeast of the site. The epicentral intensity of this shock is rated at VII. At Wilmington, chimneys toppled and windows broke. Damage was also reported at Newport, New Castle, and Oxford, DE. The earthquake was felt over a relatively small area of northern Delaware, southeastern Pennsylvania, and southwestern New Jersey. Only one earthquake of Intensity V or greater has been reported within 50 miles of the proposed plant site. This shock, which had a rated epicentral intensity of V, caused no structural damage. Its epicenter was located near Seaford, DE, about 45 miles northeast of the site. Several small shocks have been reported in the Coastal Plain in the region surrounding the site. Four such shocks were reported east and south of the site. Several other small shocks were reported in the vicinity of Annapolis, MD, northwest of the site. These reported earthquakes were considered in this investigation; however, available data regarding these shocks are limited, and it is impossible to estimate their maximum intensities or to precisely locate their epicenters. None of these earthquakes caused structural damage and they are of interest only in that they may indicate the possible presence of unidentified faulting in the basement rock of the Coastal Plain. However, it is possible that some of these reports may refer to relatively distant earthquakes which were felt in eastern Maryland. Furthermore, it is also possible that these shocks resulted from causes other than tectonic activity. The probable epicenters of these small shocks are shown on the Epicentral Location Map, Figure 2.6-2.

CALVERT CLIFFS UFSAR 2.6-4 Rev. 47 An independent evaluation of regional seismicity performed in 1992, to support design of the Diesel Generator Building, identified more earthquakes than were previously identified. However, these additional earthquakes were neither larger nor in different areas than those used to develop the original SSE design basis. 2.6.4 GEOPHYSICAL INVESTIGATIONS 2.6.4.1 General Geophysical studies were performed at the Calvert Cliffs site in order to evaluate the dynamic properties of the foundation materials. The dynamic soil properties are used in evaluating the response of the foundation materials to earthquake loading.

A seismic refraction survey and an uphole velocity survey were performed in order to measure the velocity of compressional wave propagation at the site. The shear wave velocity was estimated from the field measurements of surface waves (predominantly Rayleigh waves). Micromotions were measured in order to indicate the pattern of vibration at the site due to background "noise." These measurements assist in estimating the natural period of vibration at the site. Laboratory shockscope tests were performed for correlation with the field measurements. Shear and compressional wave velocity measurements were derived for the upper strata during the geophysical investigations. Compressional wave velocities for the deeper strata near the site were measured during a geophysical survey performed by Ewing and Worzel in 1943. Shear wave velocities were computed from these data using an estimated Poisson's ratio. Geophysical data for the entire stratigraphic section and presented on Figure 2.6-1.

2.6.4.2 Refraction Seismic Survey Seismic refraction surveys were performed along two lines, 2,000' and 2,100' in length, as shown on Figure 2.4-7, Plot Plan. The purpose of these surveys was to evaluate the compressional wave velocities of the sediments underlying the site. The work was conducted with an Electro-Technical Labs M4E seismograph. Dynamite charges, from 10 to 50 lbs at each end of the seismic lines, were used as a source of energy. The charges were buried at depths of 25 to 60'. Geophones were located at intervals ranging from 20 to 100' along each line. Data from these field measurements indicate that the velocity of compressional wave propagation in the upper surficial Pleistocene silts and sands is about 2,200 fps, and in the Miocene sandy and clayey silt about 5,900 fps. Extensive geophysical surveys in the Coastal Plain were made in 1943 by Ewing and Worzel. These surveys included measurements to the crystalline basement rock at a point several miles south of the site. The results of these measurements and the measurements made at the site by Dames & Moore are summarized in Table 2-38 Geophysical Data. 2.6.4.3 Uphole Seismic Velocity Survey An uphole seismic survey was performed in Boring DM-4 in the proposed plant area. The purpose of this survey was to correlate the compressional wave velocities of the materials in the plant area with the compressional wave velocities measured by the refraction survey approximately 1,000' to the northwest. CALVERT CLIFFS UFSAR 2.6-5 Rev. 47 The uphole seismic survey was performed using the Electro-Technical Labs ER-75-12 seismograph. The seismic energy was provided by either blasting caps or charges of one to three ounces of dynamite detonated at 10' intervals in the boring to a maximum depth of 148'. Geophones were placed at 5 to 60' intervals at distances up to 220' from the boring. The results of this survey are presented on Figure 2.6-3, Uphole Seismic Survey.

2.6.4.4 Shear Wave Measurements The velocity of shear wave propagation was evaluated at the site. The compressional wave and shear wave velocities were used to compute Poisson's ratio and the dynamic properties of the soil. The shear waves were estimated from surface waves (predominantly Rayleigh waves) measured with a Sprengnether velocity meter. These measurements indicate that the velocity of shear wave propagation at the site is about 1,600 fps in the Miocene sediments. 2.6.4.5 Micromotion Measurements Micromotion measurements were made at three locations at the site using the Dames & Moore Microtremor Equipment. The micromotion measuring stations are shown on Figure 2.4-7, Plot Plan. The equipment used is a highly sensitive recording device capable of magnifying ground motions up to 150,000 times and is accurate over a frequency range of 1 to 30 CPS.

The microtremor records indicate a predominant period of background vibration on the order of one-half to three-quarters of a second. The low intensity levels are consistent with what would be expected in a reasonably dense material. The predominant ground period and intensity of ambient motion at the site will present no special problems in design of the facility.

2.6.4.6 Laboratory Shockscope Tests Several representative samples of the soil underlying the site were tested in the Shockscope. The Shockscope is an instrument developed by Dames & Moore to measure the velocity of propagation of compressional waves in soil. The velocity of compressional wave propagation observed in the laboratory was correlated with the field measurements and used as an aid in evaluating the dynamic elastic properties. In the Shockscope test, the samples were subjected to physical impulses under a range of confining pressures while the time necessary for the shock wave to travel the length of the sample was measured using an oscilloscope. The velocity of compressional wave propagation was then computed. The results of these tests are presented in Table 2-39, Shockscope Test Data. 2.6.4.7 Diesel Generator Building Siting Surveys Additional analyses of vibratory ground motion were performed to support design of the Diesel Generator Buildings. A significant earthquake catalog was developed by investigators and consultants of Electric Power Research Institute and the National Center for Earthquake Engineering Research. A subsequent independent evaluation of this regional seismicity was performed by Bechtel Power Corporation in 1992. CALVERT CLIFFS UFSAR 2.6-6 Rev. 47 2.6.5 ASEISMIC DESIGN 2.6.5.1 Foundations The foundations of the major plant structures are established in the Miocene sandy and clayey silts of the Chesapeake Group. These soils are apparently preconsolidated as a result of deposition and subsequent erosion of younger sediments, as well as desiccation and increase in effective pressure caused by lowering of the water table. Some appurtenant structures at the site are founded in the surficial Pleistocene silt and sand which overlies the Miocene sediments. 2.6.5.2 Operating Basis Earthquake (No Loss of Function) On the basis of the seismic history of the area, it does not appear likely that the site will experience significant earthquake ground motion during the economic life of the proposed facility. The nuclear power plant was conservatively designed to respond elastically, with no loss of function, to horizontal earthquake ground accelerations of 8% of gravity, and vertical earthquake ground accelerations of 5-1/3% of gravity. It is not believed that this level of ground motion will be exceeded at the site during an earthquake similar to any historical event. This ground acceleration is considerably greater than what might be expected due to an Intensity V shock (Magnitude 4 on the Richter Scalea) close to the site, or to an Intensity VII (Magnitude 5) shock at an epicentral distance of about 15 to 20 miles.

2.6.5.3 Safe Shutdown Earthquake (Safe Reactor Shutdown) For a safe shutdown of the reactor, the facility was designed using a horizontal ground acceleration of 15% of gravity at foundation level, and a corresponding vertical ground acceleration of 10% of gravity. It is not believed that this level of earthquake ground motion would be exceeded during the maximum potential earthquake. The magnitude of vertical ground motion was estimated on the assumption that vertical particle motions due to compressional waves have magnitudes of about one-half to two-thirds of the horizontal particle motions due to shear waves. The SSE for the site is considered to be a shock similar to one of the following: a. A shock equivalent to the Intensity VII, 1871 Wilmington earthquake as close to the site as its related geologic structure. This earthquake was probably a Magnitude 5. It is likely that this earthquake was related to faulting in the Piedmont west of the Fall Zone. However, since it is impossible to precisely locate the epicenter of this shock from the limited available data, and since the earthquake was felt in portions of the Coastal Plain, it was considered that the epicenter of this shock may have been located somewhat east of the Fall Zone. b. A shock equivalent to the Intensity VII northern New Jersey earthquake of 1927 occurred close to the site. This shock occurred in the Coastal Plain and has not been related to known geologic structure. Therefore, the conservative assumption was made that it could occur along a hypothetical geologic structure in the basement rock near the site. This earthquake was probably about a Magnitude 5 to 5-1/2. c. A shock equivalent to the Intensity IX Charleston earthquake of 1886 recurring at or near the original epicenter. An Intensity IX (Magnitude 7) a Earthquake magnitudes in this section refer to the magnitude scale developed by Dr. C. F. Richter. The magnitude scale is a means of indicating the size of an earthquake based on instrumental records. The magnitude scale is further described in Section 2.6.8. CALVERT CLIFFS UFSAR 2.6-7 Rev. 47 earthquake centered near Charleston, about 480 miles southwest of the site, would not be of significance at the site. Based on the foregoing statements, the very conservative assumption has been made that the SSE would be a shock as large as Intensity VII (Magnitude 5 to 5-1/2) originating in the basement rock close to the site.

Although a later, independent evaluation of region seismicity performed in 1992 (in support of the Diesel Generator Project) identified more recent earthquakes than those presented here, these were not larger nor were they in different areas than those used to develop the original SSE basis.

2.6.5.4 Response Spectra Ground motion response spectra are presented on Figures 2.6-4 and 2.6-5, Response Spectra-Operating Basis Earthquake (OBE) and Response Spectra-SSE, respectively. These spectra conform to the average spectra developed by Dr. G. W. Housner for the frequency range higher than about 0.33 CPS. These average spectra were originally presented in TID-7024 (Reference 5). The spectra presented herein are based on a later revision by Dr. Housner, presented for the H. B. Robinson Nuclear Power Plant of Carolina Power and Light Company (Reference 9). The spectra for frequencies lower than about 0.33 CPS were prepared utilizing data suggested by Dr. N. M. Newmark (Reference 7).

The spectra have been normalized to a horizontal ground acceleration of 8% of gravity for the OBE and 15% of gravity for the SSE. The response spectra indicate the estimated response of a structure subjected to earthquake ground motion. The spectra are presented over a range of frequencies corresponding to the natural frequencies of the various structural elements and represent the maximum amplitude of motion in the various elements of the structure for typical degrees of damping. The digitalized El Centro Earthquake (1940, East-West), normalized to a ground acceleration of 0.08g horizontally and 0.053g vertically, acting simultaneously, was used in the analysis of Category I equipment. See paragraph 5.1.3.2 for description of dynamic response spectra. 2.6.6 TSUNAMIS The occurrence of tsunamis is infrequent in the Atlantic Ocean. Other than the tidal fluctuation recorded on the New Jersey shore during the Grand Banks earthquake of 1929, there has been no record of tsunamis on the northeastern United States coast. The earthquake of November 18, 1929, on the Grand Banks about 170 miles south of Newfoundland, resulted in a tsunami which struck the south end of Newfoundland, about 750 miles northeast of the Massachusetts Coast. This tsunami occurred at a time of abnormally high tide and resulted in some loss of life and destruction of property. The effect of this tsunami was recorded on tide gauges along the east coast of the United States as far south as Charleston, SC. A tidal fluctuation of approximately nine-tenths of one foot was noted at Atlantic City, NJ and Ocean City, MD (Reference 11). The Lisbon earthquake of November 1, 1755, produced great waves which contributed heavily to destruction on the coast of Portugal. These waves were noticeable in the West Indies. It has been reported that the Cape Ann, MA, earthquake of November 18, 1755, caused a tsunami in Saint Martin's Harbor in the West Indies. However, there is no record of tsunami occurrence along the east coast of the United States at this time, and it appears that the Saint Martin's Harbor report actually refers to the tsunami caused by the CALVERT CLIFFS UFSAR 2.6-8 Rev. 47 Lisbon earthquake, which occurred less than three weeks before the Cape Ann shock. Some tsunami activity has occasionally followed earthquakes in the Caribbean, but none of these was reported in the United States (References 4 and 10). It is not believed that the site will be subjected to a significant tsunami effect. The maximum expected tsunami would result in only minor wave action, and the maximum expected storm wave effect, discussed in Section 2.8.3, Hurricane Tidal Effects, was a more critical factor in the design.

2.6.7 MODIFIED MERCALLI INTENSITY SCALE The Modified Mercalli Intensity scale of 1931 is described in Table 2-40. The intensity scale is a means of indicating the relative size of an earthquake in terms of its perceptible effect. The intensities presented in this report indicate the damage caused by an earthquake at its epicenter. 2.6.8 RICHTER MAGNITUDE SCALE Magnitude Scale is a means of indicating the size of an earthquake based on instrumental records. Dr. C. F. Richter developed a magnitude scale which is based on the maximum recorded amplitude of a standard seismograph located at a distance of 100 kms from the source of a shallow earthquake. The magnitude is defined by the relationship M=log A - log Ao. In this equation, A is the recorded trace amplitude for a given earthquake at a given distance written by the standard instrument, and Ao is the trace amplitude for a particular earthquake selected as a standard. The zero of the scale is arbitrarily fixed to fit the smallest recorded earthquakes. The largest known earthquake magnitude is on the order of 8-3/4. This magnitude is the result of observations and not an arbitrary scaling. The upper magnitude limit is not known, but is estimated to be about 9.

Empirical relationships between earthquake magnitude and energy release have been developed by several investigators (Reference 10). There is no exact relationship between earthquake magnitude and energy for large earthquakes, and these empirical relationships should be considered no more than approximations. 2.

6.9 REFERENCES

1. C.E. Dutton, 1889, The Charleston Earthquake of August 31, 1886, Ninth Annual Report of the U.S. Geological Survey, Washington, DC 2. Earthquake History of the United States - Part I, 1965, United States Department of Commerce, Coast and Geodetic Survey, Washington, DC 3. M. Ewing and L. Worzel, 1948, Explosion Sounds in Shallow Water, Geological Society of America, Memoir 27 4. B. Gutenberg and C.F. Richter, 1954, Seismicity of the Earth and Associated Phenomena, Princeton University Press, Princeton 5. G.W. Housner, 1963, Response of Structures to Earthquake Ground Motion, Nuclear Reactors and Earthquakes (TID-7024), United States Atomic Energy Commission, Division of Technical Information 6. G.R. MacCarthy, 1964, A Descriptive List of Virginia Earthquakes Through 1960, Journal of the Elisha Mitchell Scientific Society, Volume 80, No. 2 7. N.M. Newmark, 1969, Design Criteria for Nuclear Reactors Subjected to Earthquake Hazards, Proceedings of the International Atomic Energy Agency Panel on Aseismic Design and Testing of Nuclear Facilities, Tokyo, Japan, 1967 CALVERT CLIFFS UFSAR 2.6-9 Rev. 47 8. Preliminary Determination of Epicenters - (Card Series 1966 through 1967) United States Department of Commerce, Coast and Geodetic Survey, Washington, DC 9. Preliminary Safety Analysis Report, H.B. Robinson Nuclear Power Plant, Carolina Power and Light Company 10. C.F. Richter, 1958, Elementary Seismology, W.H. Freeman and Company, San Francisco, CA 11. United States Earthquakes (Serial Publications, 1928 through 1965) United States Department of Commerce, Coast and Geodetic Survey, Washington, DC 12. G.P. Woollard, 1958, Areas of Tectonic Activity in the United States As Indicated By Earthquake Epicenters, Transactions of the American Geophysical Union, Volume 39 13. Letter from R.E. Denton (BGE) to Document Control Desk (NRC), dated December 18, 1992, Emergency Diesel Generator Project-Civil Engineering Design Report CALVERT CLIFFS UFSAR 2.6-10 Rev. 47 TABLE 2-37 SIGNIFICANT EARTHQUAKES WITHIN 100 MILES OF THE SITE YEAR DATE TIME INTENSITYLOCATION N. LAT. W. LONG.AREA FELT DISTANCE FROM SITE (sq. mi.) (miles) 1733 June 14 -- (a) Vicinity of Annapolis, MD -- -- -- -- 1758 April 24 -- (a) Vicinity of Annapolis, MD -- -- -- -- 1774 Feb. 21 14:00 VI Richmond, VA 37 1/2 77 1/2 -- 80 1833 Aug. 27 06:00 VI Central Virginia 37 3/4 78 52,000 90 1871 Oct. 9 09:40 VII Wilmington, DE 39 3/4 75 1/2 -- 100 1875 Dec. 22 23:45 VI Near Richmond, VA 37 1/2 77 1/2 50,000 80 1876 June 19 -- (a) Vicinity of Annapolis, MD -- -- -- -- 1879 Mar. 25 19:30 IV-V Northern Delaware 39 3/4 75 1/2 600 100 1883 Mar. 11 18:57 IV-V Harford County, MD 39 1/2 76 1/2 Local 80 Mar. 12 00:00 IV-V Harford County, MD 39 1/2 76 1/2 Local 80 1885 Jan. 2 21:16 V Frederick County, 39 1/2 77 1/2 3,500 80 1897 Dec. 18 18:45 V Ashland, VA 37 3/4 77 1/2 7,500 75 1906 May 8 12:41 V Seaford, DE 38 3/4 75 3/4 400 45 1908 Aug. 23 04:30 V Powhatan, VA 37 1/2 78 450 95 1919 Sept. 5 21:46 VI Front Royal, VA 38 3/4 78 1/4 -- 95 1930 Jan. 18 -- IV(a) Pines of the Sermon, MD -- -- -- -- 1930 Nov. 1 01:34 I-III(a) Anne Arundel County, MD 39.0 76.5 Local -- 1949 May 8 06:01 IV-V Richmond, VA 37 1/2 77 1/2 1,800 80 1966 May 31 06:19 IV-V Central Virginia 37.6 78.0 -- 100 _______________________ (a) Several small shocks in Maryland are included in this table. Little information is available regarding these reports, and the indicated epicenters are uncertain. See text of report for discussion.

CALVERT CLIFFS UFSAR 2.6-11 Rev. 47 TABLE 2-38 GEOPHYSICAL DATA SURFICIAL SEDIMENTS (PLEISTOCENE) COMPRESSIONAL UNCONSOLIDATED SEDIMENTS (TERTIARY) COMPRESSIONAL INTERMEDIATE SEDIMENTS (CRETACEOUS)(a) COMPRESSIONAL BASEMENT ROCK COMPRESSIONAL STATION WAVE VELOCITY (fps) THICKNESS (ft) WAVE VELOCITY (fps) THICKNESS (ft) WAVE VELOCITY (fps) THICKNESS (ft) WAVE VELOCITY (fps) DEPTH (ft) Solomons Shoal(b) -- -- 5900 3080 -- -- 15,170 3130 Solomons Deed(b) -- -- 6080 1070 6980 1900 18,100 3080 Site(c) 2200 40 5500 -- -- -- -- -- Site(c) -- -- 5900 -- -- -- -- -- _______________________ (a) These measurements refer to a "masked" arrival and the results are questionable. (b) Adapted from Ewing and Worzel (Reference 3). (c) Measurements by Dames & Moore.

CALVERT CLIFFS UFSAR 2.6-12 Rev. 47 TABLE 2-39 SHOCKSCOPE TEST DATA BORING DEPTH (ft) CONFINING PRESSURE (lbs/ft2) COMPRESSIONAL WAVE VELOCITY (fps) DM-2 5 0 2000 4000 6000 1,000 1,200 1,400 1,700 DM-9 15 0 2000 4000 6000 1,200 1,300 1,500 1,700 DM-1 30 0 2000 4000 6000 1,400 1,500 1,800 2,100 DM-10 68 0 2000 4000 6000 2,600 2,600 3,200 3,200 DM-10 111 0 2000 4000 6000 2,600 2,600 3,000 3,000 DM-10 156 0 2000 4000 6000 1,800 1,800 1,900 1,900 DM-10 211 0 2000 4000 6000 1,600 1,700 1,700 1,700 DM-10 256 0 2000 4000 6000 2,100 2,100 2,200 2,200 DM-10 271 0 2000 4000 6000 2,000 2,200 2,300 2,600

CALVERT CLIFFS UFSAR 2.6-13 Rev. 47 TABLE 2-40 MODIFIED MERCALLI INTENSITY (DAMAGE) SCALE OF 1931 (Abridged) I. Not felt except by a very few under especially favorable circumstances. (I Rossi-Forel Scale) II. Felt only by a few persons at rest, especially on upper floors of buildings. Delicately suspended objects may swing. (I to II Rossi-Forel Scale) III. Felt quite noticeably indoors, especially on upper floors of buildings, but many people do not recognize it as an earthquake. Standing motorcars may rock slightly. Vibration like passing of truck. Duration estimated. (III Rossi-Forel Scale) IV. During the day felt indoors by many, outdoors by few. At night some awakened. Dishes, windows, doors disturbed; walls make creaking sound. Sensation like heavy truck striking building. Standing motorcars rocked noticeably. (IV to V Rossi-Forel Scale) V. Felt by nearly everyone, many awakened. Some dishes, windows, etc., broken; a few instances of cracked plaster; unstable objects overturned. Disturbances of trees, poles, and other tall objects sometimes noticed. Pendulum clocks may stop. (V to VI Rossi-Forel Scale) VI. Felt by all, many frightened and run outdoors. Some heavy furniture moved; a few instances of fallen plaster or damaged chimneys. Damage slight. (VI to VII Rossi-Forel Scale) VII. Everybody runs outdoors. Damage negligible in buildings of good design and construction; slight to moderate in well-built ordinary structures; considerable in poorly built or badly designed structures; some chimneys broken. Noticed by persons driving motorcars. (VIII Rossi-Forel Scale) VIII. Damage slight in specially designed structures; considerable in ordinary substantial buildings with partial collapse; great in poorly built structures. Panel walls thrown out of frame structures. Fall of chimneys, factory stacks, columns, monuments, walls. Heavy furniture overturned. Sand and mud ejected in small amounts. Changes in well water. Persons driving motorcars disturbed. (VIII+ to IX Rossi-Forel Scale) IX. Damage considerable in specially designed structures; well-designed frame structures thrown out of plumb; great in substantial buildings, with partial collapse. Buildings shifted off foundations. Ground cracked conspicuously. Underground pipes broken. (IX+ Rossi-Forel Scale) X. Some well-built wooden structures destroyed; most masonry and frame structures destroyed with foundations; ground badly cracked. Rails bent. Landslides considerable from river banks and steep slopes. Shifted sand and mud. Water splashed (slopped) over banks. (X Rossi-Forel Scale) XI. Few, if any, (masonry) structures remain standing. Bridges destroyed. Broad fissures in ground. Underground pipelines completely out of service. Earth slumps and land slips in soft ground. Rails bent greatly. XII. Damage total. Waves seen on ground surface. Lines of sight and level distorted. Objects thrown upward into the air. CALVERT CLIFFS UFSAR 2.7-1 Rev. 47 2.7 SUBSURFACE AND FOUNDATIONS 2.

7.1 INTRODUCTION

This summarizes the results, analyses, and evaluation of the subsurface and foundation investigations. The field exploration and the laboratory testing were done under the supervision and direction of Bechtel Associates. These studies included site and area reconnaissance, field supervision of the boring operations, a review of pertinent literature, and the foundation analysis and evaluation. The initial graphic boring logs and laboratory test data cited were presented in the Preliminary Safety Analysis Report (PSAR). Subsequent data and subsurface profiles are presented here. The Civil Engineering Design Report for the Emergency Diesel Generator Project describes: (1) exploration, (2) the properties of subsurface materials, (3) groundwater conditions, (4) response of soil and rock to dynamic loading, (5) liquefaction potential, and (6) static and dynamic stability at the site of the Diesel Generator Buildings. 2.7.2 EXPLORATION 2.7.2.1 Field Reconnaissance The geologic field work for this site was started concurrently with a drilling program. The site reconnaissance was a continuation of the field work done in the early part of 1967. Local soil outcrops were examined on the Chesapeake Bay bluffs and the soils, types, orientation, and variations noted.

In addition to the site reconnaissance, the area was examined as the drilling progressed. Also, geological information was gained from the Maryland Geological Survey and various publications (References 1, 2, 3, 4, and 6). The geologic information gained from the site reconnaissance and literature review was used in conjunction with the borings to prepare the site geology portion of this section. Also, Section 2.4, Geology, gives a comprehensive presentation of the site and regional geology. 2.7.2.2 Boring and Sampling Investigations In June and July 1967, a preliminary foundation exploration for the proposed nuclear power plant was conducted at the site by BGE. The field program included five borings, B1 through B5, and numerous split-spoon soil samples.

The field exploration for the PSAR and plant design began with the initial reconnaissance of the site by personnel from BGE and Bechtel Associates. Subsurface exploration started on August 17, 1967, and was completed on September 22, 1967. This exploration included drilling 22 soil test borings in the vicinity of the plant site and one geologic boring on a bluff approximately 2,000' north of the site for the Maryland Academy of Science. During November and December 1968, a supplementary subsurface investigation was undertaken. A series of 18 borings were drilled in the plant area. The primary purpose of the supplementary program was to provide additional information for the plant foundation analyses. In order to obtain additional information necessary for the design of the Intake Structure and switchyard foundations, a final subsurface investigation was undertaken during spring 1969. Five test borings were drilled in switchyard area and borings WB-1 through WB-35 were drilled from a barge offshore of the Intake Structure. The location plan of the borings is shown on Figures 2.7-1 to 2.7-3. CALVERT CLIFFS UFSAR 2.7-2 Rev. 47 During all field investigations, a geologist or soil engineer from Bechtel Associates continuously supervised and inspected drilling operations and modifications in the boring programs as deemed necessary. Split-spoon samples and undisturbed Shelby tube samples were obtained at desired sampling intervals. All samples were initially visually inspected and classified in the field. Samples were then either forwarded to Bechtel Associates' Washington Area Engineering Office for further examination, or to a soils laboratory for testing. The boring logs subsequent to the PSAR are shown on Figures 2.7-4 to 2.7-26. 2.7.2.3 Geotechnical Investigations to Support Diesel Generator Building Siting Two field investigations were conducted, one in 1980/1981 and another in 1992, which assessed the geotechnical conditions at the proposed site of the Diesel Generator Buildings. In 1980/1981, a field investigation was performed to assess the North Parking Area's suitability as a location for a generic Category I structure. The investigation consisted of sample borings, cone penetration soundings, and observation wells. In 1992, a second field investigation was performed which consisted of sample borings, dilatometer soundings, a crosshole seismic survey, and field soil resistivity tests. In both investigations, standard penetration test samples were obtained in accordance with ASTM D 1586. Thin-walled tube samples were obtained in general accordance with ASTM D 1587. The crosshole seismic survey was performed in accordance with the requirements of ASTM D 4428/4428M-84 by using the "preferred method." Field soil resistivity testing was conducted using the "Wenner four-electrode method" in accordance with ASTM G 57. Additional details about the geotechnical investigations can be found in the Civil Engineering Design Report for the Diesel Generator Project (Reference 29). 2.7.3 SITE CONDITIONS 2.7.3.1 Area Geology A geology summary for the purpose of understanding the foundation evaluation is presented herein. The main geologic presentation is in Section 2.4.

The CCNPP site lies in the Atlantic Coastal Plain physiographic region of Maryland. It is in an area of sedimentary deposits formed by the ancient rivers which carried large quantities of solids from the northern and western uplands of the Piedmont and Appalachian physiographic provinces into the once larger Atlantic Ocean. These deposits were formed in both a freshwater (fluvial) and a saltwater (marine) environment. The upper deposits in this coastal plain area are the Recent and Pleistocene deposits of tan and brown silts, sands and clays with some inclusions of seashell fossils. Below the Pleistocene deposits lie the older Miocene sediments. The Miocene deposits represent the soils of significant interest for the foundations of this power plant. The foundation properties of both the Pleistocene and Miocene are more than adequate for the plant loads. The crystalline basement rocks are approximately 2500' below the present ground surface. 2.7.3.2 Soil Conditions For foundation engineering purposes, the soils at the site can be divided into an upper zone and a lower zone. These soils are a mixture of marine and fluxial deposits. Each zone has both continuous and discontinuous strata with isolated pockets or lenses of slightly different materials. The soils are non-uniform CALVERT CLIFFS UFSAR 2.7-3 Rev. 47 sedimentary deposits of silty sands, sandy silts, clayey sands and sandy clays with layers of shell fossils. The upper zone is generally yellowish tan, brownish tan and light brown in color. This color is caused by oxidation of the mineral constituents. The soil is firm to dense in consistency. The predominant soil types are sandy silts and silty sands. The average thickness is 18' and varies in elevation dependent upon the topography. This zone is primarily the Pleistocene deposit. The lower zone, the Miocene deposit, is greenish gray in color with several occurrences of medium to light gray soil at the top of the zone. The soil below +5' MSL is very dense to extremely dense with a few lenses, which are isolated but dense in consistency. The major soil types are sandy silts, silty sands, and slightly clayey sands. The lower part of this zone is classified as fine sands and silts. The upper zone of soil can support light loads, on the order of 2000 to 3000 psf, with a small amount of anticipated consolidation, while the lower zone can support heavy loads on the order of 15,000 to 20,000 psf with slight consolidation. The original groundwater surface was between +15' and +20' MSL in the plant area; however, a permanent pipe drain system, subsurface drain system, surrounding the plant will maintain the ground water below Elevation +16'. Additional information concerning groundwater appears in Section 2.5. Subsurface profiles are shown on Figures 2.7-27 and 2.7-28. 2.7.4 LABORATORY TESTING The laboratory testing program provided the soils' physical characteristics for foundation design. The testing was conducted in accordance with currently accepted procedures (References 7 through 12). The testing program was divided into three parts, to determine the soil parameters under static, dynamic, and remolded (fill) conditions. The laboratory program included: grain size and specific gravity tests to determine particle size and distribution; Atterberg limit tests to determine soil plasticity characteristics; consolidation tests, to determine the soil settlement characteristics; unconfined compression and static triaxial shear tests to aid in the evaluation of foundation bearing capacity and slope stability analysis; dynamic triaxial shear tests to determine the dynamic properties used in the evaluation of liquefaction potential of foundation materials; compaction tests; and numerous moisture-density, void ratio and relative density determinations.

2.7.5 STRUCTURAL DATA The foundations for the Turbine Building, Auxiliary Building, Containments, Turbine-Generators, and Circulating Water System are mat foundations on the Miocene soils. Individual bearing capacities were required because such a value depends on load, elevation of foundation, settlement tolerance, foundation size, and proximity to other loads. The final design bearing loads are as follows: Structure Contact Pressure Containment Structure Mat 8000 psf Auxiliary Building Mat 8000 psf Turbine Pedestal Mat 5000 psf Turbine Building Column Footings 5000 psf Intake & Discharge Structure Mat 2500 psf CALVERT CLIFFS UFSAR 2.7-4 Rev. 47 In all the above cases, the allowable soil bearing capacity exceeds the contact pressure. Two groups of structures, the circulating water structures (i.e., intake and discharge structures), and the switchyard structures have been studied since the PSAR and are discussed below. 2.7.5.1 Circulating Water Structures The intake structure is located between the Turbine Building and shoreline, and is approximately 90 x 385' in size. The total effective load due to the structure is approximately 42,000 tons. As a result, net soil pressures due to the structure will be approximately 2500 psf. The size and total loads were increased due to design changes necessitated by the structure being changed from a partial to a total Category I Structure. A 300' segment of anchored sheet piling extends from the intake structure to the inside intake channel at the shoreline. The invert inlet elevation of the intake structure is Elevation -26' MSL and the elevation at the junction of the anchored sheet piling and the intake channel is -51' MSL. The approximate slope of the excavation from the intake inlet to the channel junction is approximately 10 horizontal to 1 vertical. The channel extends 4500' offshore from the shoreline with 5 horizontal to 1 vertical side slopes excavated in the dense, silty slightly, cemented sand. The discharge facility is located north of the plant. It consists of four conduits extending from the Turbine Building to a point 850' offshore. The top of the conduit is Elevation -6' MSL and the invert is at Elevation -19.5' MSL at the point of discharge. The conduit will be constructed and buried by cut and cover methods. The plans of the intake and discharge scheme are shown on Figure 2.7-29. 2.7.5.2 Switchyard Structures The switchyard area is located approximately 500' west of the plant area. The north half of the yard is in a cut area, the maximum cut being approximately 30'. The south half of the yard is a fill area with a maximum thickness of about 20'. Analyses show an allowable soil bearing capacity of 2000 psf in the fill area and 3000 psf in the cut excavated area. Drilled piers were used where higher loads and/or uplift conditions required deep foundations. 2.7.6 FOUNDATION EVALUATION The soils at this site are suited for the construction of the plant. The upper zone, Pleistocene soils, will support light loads of 2000 to 3000 psf without adverse settlements. The lower zone, the Miocene, is exceptionally dense and will support heavy foundation loads on the order of 15,000 to 20,000 psf. 2.7.6.1 Site Excavation and Earthwork The general site grading and excavation was done with conventional earth moving equipment. Soil compaction requirements were prepared and based on the proposed utilization of a filled area. In general, the bearing capacity of the fill was not the controlling engineering parameter; but rather, it was found that settlement controls. A maximum settlement of 1" was the limit set in the computations to determine the contact pressure for 10'x10' and smaller footings to be placed on the compacted fills.

CALVERT CLIFFS UFSAR 2.7-5 Rev. 47 The following criteria were formulated for the various loading conditions based on the soil test data for the proposed fill materials. Fill Compaction (minimum) Areas Where Criterion Is Used 85% Standard Proctor (ASTM D698; AASHO T-99) Shore protection landfill except within 100' of bulkhead or anchor sheet piling and within 25' of culverts. 90% Standard Proctor General, nonstructure supporting, fill areas and plant parking lot lower than 5' below finished grade. 95% Standard Proctor Shore protection landfill within 100' of bulkhead or anchor sheet piling and within 25' of culverts and switchyard fill. 97% Standard Proctor Structural backfill areas supporting facilities with footings 10'x10' or smaller with contact pressure of 4000 psf or less. 95% Modified Proctor (ASTM D1557: AASHO T-180) Roadway embankments and subgrades. 95% Modified Protector (ASTM D155; AASHO T-180) Structural fill for the Diesel Generator Buildings (consisting of well graded, sound, dense and durable crushed stone). 100% Modified Proctor Structural backfill areas supporting facilities with footings 10'x10' or smaller with contact pressures of 5000 psf or less. The above criteria covered the majority of the backfill conditions. Unique conditions of footing size and load were evaluated on an individual basis. The minor plant excavation and embankment slopes were constructed according to the following tabulation: Slope Height Temporary Slope Permanent Slope 0-30' 1:1 1 1/2:1 30-50' 1 1/4:1 2:1 These slopes have a factor of safety in excess of 1.5. In addition to the minor slopes, five other slopes adjacent to the plant were evaluated. Attached, at the end of this section, are the five cross-sections of the slopes around the plant. The locations of these slopes are indicated on Figure 2.7-30, "Slope Cross-Sections at Plant." These cross-sections show the range of topographic conditions that exist. The backfill on the north, west, and south sides of the plant is to Elevation 45', and on the east side to approximately Elevation 45' but sloping to the Chesapeake Bay. Stability analyses (Reference 24) were made for the design slopes shown on the cross-sections. The slopes shown on Section DD and EE are the maximum in height within the immediate plant area. These slopes have a safety factor in excess of 1.5. The other slopes around plant area are flatter or of less height; therefore, by inspection, safety factors greater than 1.5 can be assigned to these slopes. All of the slopes are acceptable for permanent slopes (Reference 22). Two dynamic slope stability analysis methods were used to evaluate the safety of the slopes for conditions resulting from the SSE.

CALVERT CLIFFS UFSAR 2.7-6 Rev. 47 In the conventional method of dynamic slope stability analysis (Reference 24), the severity of the earthquake is expressed by relating the ground acceleration to the acceleration of gravity as a percentage. The horizontal severity is 8% g and the vertical severity is taken as two-thirds of this, or 5.3% for the OBE. For the SSE these values are 15% g for the horizontal severity and 10% g for the vertical severity. During the earthquake, all parts of the mass of soil are assumed to be acted on by a steady vertical and horizontal force, equal to unit weight times acceleration, in addition to all other forces to which the slope is subjected. These forces act in the direction of instability. With these two forces determined, the analysis was completed similarly to the conventional static analysis. The other analysis method used was the procedure proposed by N.M. Newmark (Reference 23). This is fully explained in the cited reference; therefore, it is not discussed here. The factor of safety for the dynamic conditions is approximately equal for both methods of analysis. These design slopes have a factor of safety of 1.3 or more, which is acceptable (References 13 and 22). Also, an extensive slope stability analysis of the intake channel and structure was undertaken with the assistance of a computer program. The factor of safety was computed using the Swedish slip circle method of analysis. Both static and the SSE dynamic conditions were analyzed. The intake channel slopes away from the intake structure at a gradual slope of approximately 10 horizontal to 1 vertical. The minimum factors of safety for the intake structure were computed to be 2.7 and 1.6 for the static and dynamic conditions, respectively. The minimum factors of safety for the intake channel side slopes, which are 5 horizontal to 1 vertical, were computed to be 6.5 for static conditions and 2.0 for dynamic conditions. The minimum factors of safety for the critical circle perpendicular to the anchored sheet pile section of the intake were computed to be 3.0 for the static conditions and 1.6 for the dynamic conditions. During investigations done to support siting the Diesel Generator Building, the stability of the western slope of the site was evaluated to determine a factor of safety under both static and dynamic conditions. For the static condition, an analysis was made of both the total and effective stress cases. Actual conditions will fall somewhere in between the two. Since dynamic conditions are developed under undrained conditions, only the total stress case was evaluated for the dynamic conditions. The Simplified Bishop method of computing the factor of safety was used to perform the slope analysis. The results of the evaluation are summarized as follows: Condition Factor of Safety Static, Total Stress 1.72 Static, Effective Stress 1.74 Dynamic, Total Stress 1.22 Results of an earlier analysis performed on the western slope yielded similar results. The results of both analyses demonstrate that an adequate factor of safety exists against mass failure of the western slope of the site under static and dynamic conditions.

Since the crib wall adjacent to the Diesel Generator Buildings was not seismically designed, it is possible that some localized failure may occur during a seismic event. The postulated worst case is a complete failure of the crib wall, which CALVERT CLIFFS UFSAR 2.7-7 Rev. 47 would result in fill material sloughing against the wall of the Diesel Generator Building. The resulting wedge of fill material would reach a maximum height of 7.5' above grade. The static lateral loadings which result from this failure were found to be enveloped by the loading imposed by the design basis tornado. The west wall of the Safety-Related Diesel Generator Building is, therefore, designed for the fill loading under dynamic conditions. 2.7.6.2 Plant Foundations The foundation elevations, original ground surface and amount of stress unloading by excavation are shown below: Structure Average Ground Elevation Foundation Elevation Average Excavation Unloading North Containment Structure +75' MSL -1' MSL 8400 psf South Containment Structure 60 -1 6600 Auxiliary Building West End 70 -14 8300 East End 70 -19 8850 North Turbine Building 60 -11 7300 South Turbine Building 40 -11 4900 Intake Structure 80 -30 10800 Discharge Structure 20 -27 4050 Generally, the weight of soil removed by site grading and pit excavation for the structures is greater than the loads imposed by plant construction. This verified the results of the analyses made using the triaxial shear data, i.e., that bearing capacity is no problem. The ultimate bearing capacity of the foundation strata is in excess of 80,000 psf. The allowable bearing capacity is in excess of 15,000 psf. In addition to bearing capacity, settlement of the proposed structures was also considered. The settlement of the foundations can be divided into two categories: (1) elastic settlement; and (2) time-dependent or hydrodynamic settlement.

Elastic expansion of the confined soil occurred as a result of excavation unloading. This resulted in a slight upward movement. During construction, the soil moved downward as load was applied. This elastic movement is small and was complete when construction was completed. It had no effect on the structures or function of the plant. The excavation unloading and structural loading caused a small change in void ratio. This change allowed a very small amount of hydrodynamic settlement to occur. The time-dependent or hydrodynamic settlement will be very small or negligible because the structural load is either less than the overburden removed, or only slightly greater than the removed overburden weight. Considering the types of soils present, contact pressures of 1500 to 2000 psf greater than the overburden removed would not result in large consolidation settlements. The magnitude of maximum possible post-construction settlement is 1/2". The excavation for the power plant structures was below the water table. Conventional dewatering was done to maintain a dry and stable condition during construction.

CALVERT CLIFFS UFSAR 2.7-8 Rev. 47 2.7.6.3 Liquefaction Potential If a loose, saturated sandy soil (a soil with less than 10 to 15% silt and clay fines and less than 200 to 300 psf cohesion) is subjected to ground vibrations, as during an earthquake, it tends to compact and decrease in volume. If the soil cannot drain during the rapid load fluctuations imposed by an earthquake, there is a buildup in pore pressure until it is equal to the overburden pressure. The effective stress then becomes zero, the soil loses its strength, and develops a "quick" or liquefied condition. If this condition is of a general areal extent and the pressure not otherwise relieved, it can cause a flow or bearing capacity failure (References 14, 16, 17, 18, 19, 20, and 21). For the evaluation of liquefaction potential at this site, data were used from the dynamic triaxial testing, standard penetration resistances from the borings, in-place density determinations and geologic origin of the sedimentary soils at the site. All of these data showed that the soil at the site was not of a liquefaction potential. The dynamic tests showed exceptional strength under constant cyclic stress.

Other characteristics also support that there is no liquefaction potential at this site. The amount of material passing the No. 200 sieve in the gradation analysis (the silt and clay fines) and the amount of cohesion observed in the static triaxial shear tests supports the conclusion that there is no potential to liquefy. This is concluded because the soil is not truly cohesionless or reasonably suited to be susceptible to the liquefaction phenomenon. The last significant indicator that a liquefaction potential does not exist is the geologic origin of the site soils. The areas where liquefaction is believed to have happened are areas of flacial outwash, recent alluvium, or loose artificial fills (Reference 15). The Calvert Cliffs site soils are preconsolidated deposits several million years old.

During the siting of the Diesel Generator Buildings, the liquefaction potential of various strata of the North Parking area were evaluated using the standard penetration test blow counts obtained during subsurface explorations. Factors of safety against liquefaction were computed for all standard test blow counts performed for the loose-to-medium dense granular sand strata (considered to be the most susceptible to liquefaction). These computed factors of safety ranged from 1.3 to 2.4 with a median value of 1.8. Previous analyses performed in 1981 indicated a minimum factor of safety of 1.37 against liquefaction based upon standard penetration test blows, and between 1.6 and 2.0 based upon laboratory cyclic triaxial tests data. The results indicated that the site of the Diesel Generator Buildings has an adequate factor of safety against liquefaction under design earthquake conditions. 2.7.6.4 Lateral Earth Pressure The lateral earth pressures for the walls of this plant were evaluated for both the static and dynamic conditions.

The rigidity of the walls and the fact that backfilling was done after the walls were framed at the top do not allow sufficient movement for developing the active earth pressure case.

Therefore, the at-rest condition was developed. For convenience of design, the earth pressure has been converted to an equivalent fluid pressure method. This utilizes the Ranking approach which is a conservative estimate of lateral earth pressures. The earth pressure has been determined based on the characteristics CALVERT CLIFFS UFSAR 2.7-9 Rev. 47 of the material stockpiled for use as backfill. This backfill was sand, silty sand, and gravely silty sand. The equivalent fluid unit weight above the water table is 47 lbs/ft3. Below the water table, the equivalent fluid unit weight is 85 lbs/ft3. The pressure distribution will be hydrostatic. The dynamic earth pressures were considered for this plant. The analysis was based on work by N. M. Newmark, Y. Ishii, et al., K. Terzaghi, and the Corps of Engineers (References 25, 26, 27, and 28, respectively). These references provided at-rest coefficients of earth pressure which are dependent on the magnitude of the earth shock acceleration. Based on this information, an increase of the equivalent fluid unit weight should be 10% and 17% for the OBE and SSE, respectively. 2.

7.7 REFERENCES

1. C. Schuchert, Historical Geology of the Antillean-Caribbean Region, John Wiley & Sons, New York, 1935 2. G.E. Murray, Geology of the Atlantic and Gulf Coastal Province of North America, Harper and Brothers, New York, 1961 3. E.G. Otton, Ground-Water Resources of the Southern Maryland Coastal Plain, State of Maryland, Dept. of Geology, Mines, and Water Resources, Bulletin 15, 1955 4. The Physical Features of Charles County, State of Maryland, Dept. of Geology, Mines and Water Resources, 1948 5. Part II, Bituminous Materials for Highway Construction, Waterproofing, and Roofing; Soils; Skid Resistance, 1966 Book of ASTM Standards with Related Materials, American Society for Testing and Materials, Philadelphia, 1966 6. Dames & Moore, "Report of Geology," Calvert Cliffs Nuclear Power Plant, BGE, 1967 7. Procedures for Testing Soils, American Society for Testing and Materials, Fourth Edition, Philadelphia, 1964 8. T.N.W. Akroyd, Laboratory Testing in Soil Engineering, Geotechnical Monograph No. 1, Soil Mechanics, Ltd., London, 1964 9. A.W. Bishop and D.J. Henkel, The Measurement of Soil Properties in the Triaxial Test, Edward Arnold (Publishers Ltd., Second Edition, London, 1964) 10. T.W. Lambe, Soil Testing for Engineers, John Wiley & Sons, Inc., New York, 1951 11. K.A. Healy, Triaxial Tests Upon Saturated Fine Silty Sand, Massachusetts Institute of Technology, Department of Civil Engineering, Boston, 1962 12. H.B. Seed and R.L. McNeill, "Soil Deformations Under Repeated Stress Applications," Conference Papers, Conference on Soils for Engineering Purposes, American Society for Testing Materials Special Technical Publication STP No. 232, Mexico, D. F., Mexico and Philadelphia, 1957 13. W. Ellis and V.B. Hartman, "Dynamic Soil Strength and Slope Stability," Journal of Soil Mechanics and Foundations, ASCE, Paper 5321, New York, 1967 14. K.L. Lee and H.B. Seed, "Cyclic Stress Conditions Causing Liquefaction of Sand," Journal of Soil Mechanics and Foundations, ASCE, Paper 5058, New York, 1967 CALVERT CLIFFS UFSAR 2.7-10 Rev. 47 15. H.B. Seed and I.M. Idriss, "Analysis of Soil Liquefaction: Niigata Earthquake," Journal of Soil Mechanics and Foundations, ASCE, Paper 5233, New York, 1967 16. H.B. Seed and K.L. Lee, "Liquefaction of Saturated Sands During Cyclic Loading," Journal of Soil Mechanics and Foundations, ASCE, Paper 4972, New York, 1966 17. K.L. Lee and H.B. Seed, "Dynamic Strength of Anisotropically Consolidated Sand," Journal of Solid Mechanics and Foundations, ASCE, New York, 1967 18. F.S. Brown, "Foundation Investigations for the Franklin Falls Dam," Journal of the Boston Society of Civil Engineers, Volume 28, No. 2, Boston, 1941 19. Nuclear Reactors and Earthquakes, U. S. Atomic Energy Commission, Division of Technical Information, Publication TID-7024, Washington, DC, 1963 20. Symposium of Earthquake Engineering, Proceedings, Vancouver, BC, 1965 21. "Project Report on Soil Mechanics and Foundation Engineering, Bolsa Island Nuclear Desalting Facility," prepared by the Bechtel Corporation for Los Angeles Metropolitan Water District, et al, 1967 22. G.F. Sowers, Earth and Rockfill Dam Engineering, Asia Publishing House, New York, NY, 1962 23. N.M. Newmark, "Effects of Earthquakes on Dams and Embankments, "Fifth Rankine Lecture, Institution of Civil Engineers, Geotechnique, London, 1965 24. J.L. Shrard, et al. Earth and Earth-Rock Dams, John Wiley and Sons, Inc., New York, NY, 1966 25. N.M. Newmark, "Principle and Practices for Design of Hardened structures," Technical Documentary Report Number AFSWC-TDR-62-138, Kirtland Air Force Base, New Mexico, December 1962 26. Y. Ishii, et al., "Lateral Earthpressure in an Earthquake," Proceedings of the Second World Conference of Earthquake Engineering, Science Council of Japan, Tokyo, and Kyoto, 1960 27. K. Terzaghi, Theoretical Soil Mechanics, John Wiley & Sons, Inc., New York, 1943 28. "Retaining Walls," EM1110-2-2502, Corps of Engineers, U.S. Army, Washington, DC, 1961 29. Letter from R. E. Denton (BGE) to Document Control Desk (NRC), dated December 18, 1992, Civil Engineering Design Report for the Emergency Diesel Generator Project

CALVERT CLIFFS UFSAR 2.8-1 Rev. 47 2.8 CHESAPEAKE BAY STUDIES The Calvert Cliffs site is located on the western shore of the Chesapeake Bay approximately at the mid-point of its 195 mile length. The Chesapeake Bay is the largest tidal estuary on the Atlantic Coast, roughly comparable in size to Lake Ontario. Its width ranges from 3 miles to 35 miles. Major tributaries include the Susquehanna, Patapsco, Choptank, Patuxent, Potomac, Rappahannock, York, and James Rivers.

The site is located on the western shore of the Bay, approximately 10 miles north of the Patuxent River, and 25 miles south of the Choptank River which enters from the eastern side of the Bay. At the site, the Bay is approximately 6 miles wide. Water depths up to 110' are found in the main channel of the Bay at the site, with depths of 6, 12, and 18' found at distances of 750, 1200, and 1500' offshore, respectively. 2.8.1 USES OF THE BAY The Chesapeake Bay in the general vicinity of the site is utilized for fisheries' resources, navigation, and recreation. Shellfish (oysters and clams), crabs, and finfish are taken commercially from the area. A natural oyster bar of 680 acres had extended in front of the site. However, in 1969, BGE relocated approximately 500 acres of the bar to a location in the Patuxent River. Navigational interests, both ocean-going and local Bay vessels, utilize the Chesapeake Bay extensively. Major shipping channels are situated some 3500' or more from the Calvert Cliffs site.

Recreational use of the Bay in the vicinity of the site is primarily boating and sport fishing. Water-contact recreation at the site location is negligible. 2.8.2 COMPREHENSIVE STUDY PROGRAM The need to control and minimize adverse effects on the Chesapeake Bay from the construction and operation of the Calvert Cliffs Plant was recognized at the start of the project planning. In keeping with this objective and to assure conformance with the water quality standards of the State of Maryland, shortly after the site was purchased a team of research consultants was assembled to develop the information needed for guiding the design, construction and operation of the plant. The study program and its findings have been discussed frequently with appropriate State and Federal agencies during the course of the work. The principal consultants were Sheppard T. Powell Associates, Academy of Natural Sciences of Philadelphia, Alden Research Laboratories of Worchester Polytechnic Institute, Dr. John C. Geyer of Johns Hopkins University and NUS Corporation of Rockville, MD. During studies concerning water quality in 1979-1981 Ecological Analysts, Inc. and J. E. Edinger Associates, Inc. were added to the team of consultants. The effect of the condenser cooling water discharge on temperatures in the Chesapeake Bay was studied by the Alden Research Laboratories through operation of hydraulic models which simulated flows in a 34 mile stretch of the Bay and the areas within a few thousand feet from the plant site. The objective of the studies was to determine the optimum arrangement of the cooling water system to achieve rapid dispersion of effluents and minimize water temperature variations in the area of plant influence. Results of tests on the final design arrangement were reported in December 1969.

In order to provide a baseline for assessment of the effects of plant operation on the aquatic environment, the Academy of Natural Sciences of Philadelphia engaged in a seven-year program of compiling a comprehensive inventory of the population, species, CALVERT CLIFFS UFSAR 2.8-2 Rev. 47 and condition of the aquatic life in the area, and detailed information on the physical, chemical, and bacteriological characteristics of the Chesapeake Bay near Calvert Cliffs. These studies have been completed for five years of operating conditions. In accordance with the requirements of the Calvert Cliffs PSAR and the conditions of the Atomic Energy Commission (AEC) construction permits, BGE initiated in 1969 work on the design and development of the environmental radiological monitoring program for Calvert Cliffs. Concurrent with the design, development and operation of the monitoring program, BGE in contract with NUS initiated several studies to assess the potential radiological impact of expected radioeffluents from Calvert Cliffs. A review of the results of these studies was made in conjunction with other known environmental data. The purpose of this review was to ascertain the significance of the various exposure pathways and to identify the "potential critical pathways" in the area of the facility. The results of this review and the mandatory compliance requirements, based on the AEC limitations on 1971 (proposed 10 CFR Part 50, Appendix I), determined the scope of the monitoring program described in Section 2.9.

In June 1974, BGE received discharge permits from both the Water Resources Administration of the State of Maryland and from the Environmental Protection Agency authorizing the discharge of heated condenser cooling water into the Chesapeake Bay. In June 1976, BGE received a National Pollution Discharge Elimination System (NPDES) permit which consolidated the previous two permits. The NPDES permit is periodically renewed on a schedule established by the Maryland Department of the Environment. On July 31, 1974, the AEC issued a facility operating license authorizing the power operation of CCNPP in accordance with a set of Environmental Technical Specifications. The Non-Radiological section of the Environmental Technical Specification (Appendix B) was amended by the Nuclear Regulatory Commission (NRC) by deletion of four sections dealing with environmental monitoring on March 25, 1981. Some of the aquatic programs had been completed and the remainder are the responsibility of the NPDES program of the State of Maryland.

The above-mentioned documents require that the plant is to be operated in such a manner as to ensure compliance with the State of Maryland Water Quality Regulations, and also require that monitoring programs are conducted to determine the effects of the operation of the plant on the physical, chemical and biological characteristics of the Chesapeake Bay. In order to implement these requirements, the Academy of Natural Sciences of Philadelphia and Radiation Management Corporation of Philadelphia are continuing the programs initiated before the operation of the plant. Baltimore Gas and Electric Company has performed programs to study specialized areas, such as the impingement of organisms on the traveling screens and the passage of organisms through the condenser cooling water system. The results of these studies indicate the plant meets State mixing zone criteria, the entrainment does not impact a spawning or nursing area of consequence, and the impingement cannot be cost effectively mitigated any further. The PPSP has concurred with these findings.

2.8.3 HURRICANE TIDAL EFFECTS 2.8.3.1 Historic Storms and Tides Historic accounts of early hurricanes affecting the Chesapeake Bay area date back to the 17th Century. Early chronologies of tidal flooding record extreme events which occurred in August 1667, October 1749, September 1769 and July 1788. In his report (Reference 1) on hurricane tides and tidal flooding in the Chesapeake Bay area, the District Engineer, Baltimore Corps of Engineers District, notes that U.S. Weather Bureau records show at least 80 tropical hurricanes or their CALVERT CLIFFS UFSAR 2.8-3 Rev. 47 remnants have affected the bay area in the 75 year period since 1889. By far the most destructive hurricane in recent years to affect the Chesapeake Bay region was the hurricane of August 23, 1933. Other notable storms were hurricanes "Hazel" in October 1954, "Connie" and "Diane" in August 1955 (only 5 days apart), and "Donna" in September 1960. The "Great Atlantic Hurricane" of September 1944, which passed some 50 miles offshore of Chesapeake Bay, was also a storm of major size and intensity. As noted above, the relative frequency of hurricane occurrence for this area is slightly more than one hurricane per year. Numerous studies have been made of the more significant hurricanes to have affected the Middle Atlantic and New England States (References 2, 3, 4, 8, 9, and 11) in which their paths, intensities, forward speeds, resulting tides and other associated phenomenon have been well documented. In general, record hurricanes passing over or near Chesapeake Bay have had central pressures of from 27.8 to 28.5"; peak wind speeds over the ocean approaching 100 mph, and maximum winds over the bay area of up to 75 mph. Following recurvature in the middle latitudes the forward speed of these storms has ranged from 10 to 36 knots. Northeast storms also affect the Chesapeake Bay area, however, because of the general orientation of the bay the magnitude of tides reached in the Bay is not as great as those generated by record hurricanes. The northeast storm of March 6-8, 1962, which resulted in 4.9' mean low water (MLW) tide in the lower Potomac River, was about the worst experienced along the Atlantic Coast.

2.8.3.2 Tides and Storm Surges Normal Tides - Normal tides in the bay area are the semidiurnal type having two highs and lows roughly every 23-1/2 hours, with a higher high and lower low as a daily occurrence. Information contained in Reference 5 shows normal and spring tide ranges at various locations in Chesapeake Bay as follows: Mean Tidal Range (ft) Spring Tidal Range (ft) Kiptopeke Beach (Ocean) 2.7 3.2 Hampton Roads 2.5 3.0 Cape Charles Harbor (Bay) 2.4 2.8 Point Lookout 1.2 1.4 Cove Point 1.2 1.4 Taylors Island 1.3 1.5 Oxford 1.4 1.6 The mean and spring tide ranges to be expected at the site are 1.2 and 1.4', respectively. The time occurrence of daily high and low tides within the Chesapeake Bay, related to time of occurrence at Hampton Roads and Baltimore, is also given in Reference 5. From the data given in Table 2 of that reference, it was ascertained that the travel time of high and low tide occurrence from the Bay entrance to the site area is approximately 5 hours.

Storm Surges and Extreme High Tides - Storm surges and extreme high tides have been recorded at numerous locations in Chesapeake Bay and in the various rivers entering the Bay. Plate 2 of Reference 1 shows peak tide elevations above MLW of 8.2' at Solomons Island near the mouth of Patuxent River and 7.4' at Point Lookout at the mouth of the Potomac River. Tide levels of 4.1' and 5.1', respectively, were recorded at these locations in the October 1954 hurricane. In the August 23, 1933 hurricane, a peak tide of about 8.5' (MLW) occurred at Norfolk, VA. A generalized time-frequency curve was developed for the CALVERT CLIFFS UFSAR 2.8-4 Rev. 47 Chesapeake Bay area in Reference 1 (Plate 3) utilizing observed hurricane surge elevations. A reproduction of that relation is shown on Figure 2.8-1, Generalized Time-Frequency Curve. The tide elevations noted above include the cumulative effects of tidal surge, pressure effect, local wind effect, wave effect (in open Bay areas) and the astronomical tidal component. The contribution of the latter can add as much as 3' to the total recorded hurricane tide height (Hampton Roads) if the peak ocean surge entering the Bay coincides with a peak spring tide condition. The time of translation of tides up the Bay, noted above, is also an important consideration in determining the coincidence of both ocean and Bay peak surge heights with normal and spring high and low tides. Analysis of observed hurricane tide hydrographs in References 2 and 3 shows this effect to be quite pronounced. For example, in the August 1949 hurricane, which passed to the west of Chesapeake Bay, below-normal tides were experienced at Hampton Roads and Norfolk. Similarly, in the June 1945 hurricane, which passed east of the bay, a peak tide of 3' was recorded at Hampton Roads, whereas a low tide of over a foot below normal was recorded at Baltimore.

2.8.3.3 Tide and Storm Surge Analysis A comprehensive investigation of hurricane surge problems for the Chesapeake Bay area was made using parameters from Memorandum HUR 7-97, "Interim Report - Meteorological Characteristics of the Probable Maximum Hurricane, Atlantic and Gulf Coasts of the United States" (Reference 13).

2.8.3.4 Probable Maximum Hurricane Parameters describing the maximum probable hurricane were selected from Reference 13 at the approximate latitude of the Chesapeake Bay entrance (36.1°). Definition of each of those parameters for a maximum probable hurricane is given below. Central Pressure (Po) - Minimum central pressures in hurricanes passing over or near the Chesapeake Bay area have been as low as the 27.88" of mercury for the September 1944 hurricane. Except for hurricanes occurring within the last several decades, sufficient information on central pressures to establish a reliable pressure-frequency relationship for the area is not available. The standard project hurricane derived in Reference 6 as hurricane "B" had a pressure anomaly of 2.2" of mercury which would mean a central pressure of about 27.75" of mercury. A probable maximum hurricane (PMH) was derived in Reference 10 for the New York Bay area. The pressure and wind patterns constructed for that storm are shown on Figure 43 of that report. The minimum pressure of that hurricane, when opposite the entrance to Chesapeake Bay, is 27.02" of mercury. A central pressure of 26.94" was selected. Asymptotic Pressure (Pn) - A value of 30.92" was taken from the envelope curve shown on Figure 6 of Reference 13 to represent the peripherical pressure of the PMH. Radius of Maximum Winds (R) - A value of 30 statute miles was used for this parameter, being considered representative of severe storm occurrences in the general area. Its use results in a storm of reasonable size for transposition purposes. That value lies between moderate and large radius values recommended in Table 1 of Reference 13. Forward Speed (T) - The value selected for forward speed is 23 mph, a moderate speed of translation. The forward speed of the storm affects not only the peak 30'- overwater wind speed, but also the height of peak ocean tide at or near shore and CALVERT CLIFFS UFSAR 2.8-5 Rev. 47 the shape of the resulting hydrograph. In the case of Chesapeake Bay, the forward speed is especially important in that it is related to the development of surge elevation within the Bay, the speed of the free wave up the Bay, and the resulting surge height at the plant site. A very slow moving storm would permit the Bay surge to crest at the site before the maximum effect of crosswinds could reinforce and increase that height. A fast moving storm would result in the converse.

Maximum Winds at Radius R would be 124.7 mph; adding half the forward speed results in a peak isovel wind speed of 136.2 mph.

Path - The path selected for the PMH. is shown in Figure 2.8-2. It would approach the coast from the east, curving northward on passing inland west of Chesapeake Bay. Parametric relationships describing the wind speed profile, pressure profile, pressure effect profile and basic wind data used in constructing the isovel pattern for the PMH were derived using a computer program developed and employed by personnel of the Jacksonville District Corps of Engineers, and run on a GE 415 Computer. The output of the program can be seen on Tables 2-41 and 2-42. Methods used to derive the PMH conform to those given in Reference 13. Graphical representation of the overwater wind profile, the pressure and pressure effect profiles can be seen on Figure 2.8-3. An isovel pattern was constructed for transposition purposes using data given in Table 2-42. That pattern is shown on Figure 2.8-4. 2.8.3.5 Tidal Surge Computations General - Procedures used in the tidal surge analysis for the open ocean across the Continental Shelf are those described in the U.S. Army Coastal Engineering Research Center publication, "Shore Protection - Planning and Design," Technical Report No. 4 (Reference 14). Formula (1-65) shown on page 140 of that report was used for the computations. Peak tide at the Chesapeake Bay entrance will occur at time To. Basic offshore depths, fetch, wind speed data, and Cos c values, together with general map features from wind speed data, and Cos c values, together with general map features from C&GS Chart No. 1222 can be seen on Figure 2.8-5. Peak winds in the zone of maximum winds were oriented over the shallow Bay entrance channel area to obtain the maximum surge height. Based on the selected forward speed, the surge hydrograph at the coast would have about a 12- to 14-hour rise from slightly above normal tides to the peak surge at To. This is based on a comparison of storm features of the PMH with that of the August 1933 hurricane which affected the area. Data for that storm and its resulting tide can be found in Table 1 and Figure 2 of Reference 6. Procedures - An offshore bottom profile along the fetch noted on Figure 2.8-5 is plotted on Figure 2.8-6. Average depths offshore along a fetch of about 88 statute miles range from 22' to 800'. The normal tide relations used are as follows: a. At Hampton Roads: Mean Tide Range = 2.5' MLW Normal High Tide (NHT) = 1.2' + MLW

b. At Cove Point (near Plant Site): Mean Tide Range = 1.2' MLW NHT = 0.6' MLW CALVERT CLIFFS UFSAR 2.8-6 Rev. 47 The basic assumptions made for the surge computations are as follows: a. PMH surge at bay entrance coincident with NHT at entrance. b. Beginning elevation of surge computations E1 (Mile 105) = NHT + PE E1 = 1.2' + 1.5' = 2.7' MLW c. Five-mile fetch lengths used with miles shown on Tables 2-43 and 2-44 as mid-point. d. Average depth values, , taken from offshore depth profile, Figure 2.8-6, at mid-points of 5-mile reaches. e. Values of µ (wind speed) from isovel pattern at 5-mile fetch increments. f. Surge computation procedures from Reference 14 (Formula 1-65, page 140). g. Pressure effect values taken from PMH parameters, Figure 2.8-3. h. PMH approach speed = 23 mph. Speed of free wave in Chesapeake Bay for 40 to 50' average bay depth = 24 to 27 mph. i. Assume PMH forward speed overland increases slightly from 23 mph to speed of free wave in bay. Storm accompanies surge up the bay to plant site. j. Distance from bay entrance to Calvert Cliffs plant site = 110 miles (statute). At 24-27 mph, the surge would travel from bay entrance to plant site in 4+

to 5 hours. k. Normal high tide will occur at plant site 4+ to 5 hours after NHT in channel entrance. Coincident peak surge at NHT will thus occur at both locations, with 4+ to 5-hour time difference. l. Overland reduction in PMH intensity - used factors given in Table 2a, Reference 13; at T+5 hours reduction factor = 80% µ max at site area = 136x.80 = 110 mph. It was assumed that NHT at shore would occur coincident with the peak hurricane surge. Basic data along the fetch for time To are given in Table 2-43; those data were employed in the computational procedures and formula as shown on Table 2-44.

Results - The peak tidal surge elevation that would occur at the Bay entrance was computed to be 18.67' MLW (17.32' MSL). It should be noted that wave effect was not considered to be applicable. Water depths in the channel entrance to the Bay would be on the order of 40' (22' depth + 18' surge). That depth would sustain a 30-32' wave which would move into the bay area to break farther inland.

Chesapeake Bay Surge Analysis - A comprehensive analysis of the interrelationships involved in tidal surge movement up Chesapeake Bay was presented in Reference 6. In that report, it was shown that a reduction in ocean surge occurs in its passage into and up the Bay due to the comparative dimensions and hydraulic characteristics of the entrance channel and the various sections of the Bay between Hampton Roads and Baltimore. The results of that analysis were utilized for prediction purposes. The relationship shown on Figure 15 of that report relates the maximum surge on the open coast to that to be CALVERT CLIFFS UFSAR 2.8-7 Rev. 47 expected in the southern portion of the Bay. The relation extends within the range of the computed PMH surge elevation. Using the mean prediction curve and the value of 18.67' MLW, the value of 13.2' +/- 0.8' would be obtained for the surge Elevation of 14.0' MLW. Movement of the surge up the Bay to the plant site area will occur at approximately the speed of the free wave in the Bay (about 24 to 27 mph depending on depth changes) and at a speed coincident with the speed of the hurricane. The presence of large rivers with added storage volume was found in Reference 6 to result in a further minor reduction in surge height in its passage up the Bay. Table IV of that report indicates a factor of 0.96 times the surge elevation in the lower Bay will give the value of the surge elevation to be expected in the vicinity of the plant site. Using that factor gives a surge Elevation of 13.44' MLW (14.0x0.96). That elevation represents the height of the surge in the Bay as it moves northward past the plant site. To that value must be added the additional effect of hurricane winds blowing from east to west across the Bay, and the effect of coincident occurrence of NHT at the site, plus any wave effect. Surge Elevation at Plant Site - Movement of the PMH inland and overland will result in a reduction in intensity and wind speed. Table 2 of Reference 12 lists the reduction factors to be applied with respect to travel time overland. At T +5 hours, a factor of 80% is considered applicable. Wind directions slightly ahead of the zone of maximum winds will be oriented generally east-to-west over the Bay in the vicinity of the plant site at the time the peak surge reaches that area. An evaluation of wind speed and direction was made for that condition. Wind speeds of 115 to 120 mph (117 mph average) were found to be applicable for the wind direction and fetch conditions shown on Figure 2.8-7. An effective crosswind of 94 mph (117x0.80) was, therefore, used to compute the additional height of Bay setup. An average bottom profile shown on Figure 2.8-8 was constructed using data from Figure 2.8-7. The total fetch length is approximately 10 Statute miles. Preliminary estimates indicated the node line along the fetch would be at about Mile 4.0 east of the western Bay shore. A summary of setup computations is given in Table 2-45. The computed setup elevation in the vicinity of the plant site was determined to be 15.21' MLW. A value of 1' was added to that elevation for estimated wave effect, giving a total peak surge elevation at the site of 16.21' MLW (15.6' MSL).

Wave Analysis - The significant wave height that can be expected to occur in the vicinity of the plant site during the PMH peak surge will be a function of wind speed, water depth, and length of available fetch (Reference 7). Evaluation of average water depth with fetch length in the Bay offshore indicates a 50'-depth for about 7 miles; a 40'-depth for about 9 miles. Using a wind speed of 94 mph and Figures 1-40 and 1-42 of Reference 14, results in a significant wave height of 11.4'; with a corresponding wave period of 9 seconds (Figure 1-43). The wave will break in approximately 14-1/2' of water. The height of the wave above still-water level would be 6.8' (11.4x0.6). Added to the peak surge Elevation of 16.2', the elevation of the top of that wave, unbroken, would be 23.0' MLW. Wave Run-up - The maximum wave run-up elevation at the intake structure was previously calculated to be 28.1' MLW or 27.5' MSL. A series of scale model tests was performed at the University of Florida with an adverse slope on the top of the pump room wall. Six tests were performed to represent still-water levels of 16.2', 17.2' and 18.2', each with and without the baffle wall. Results (Figure 2.8-9), indicate that there will be no overtopping of the intake structure. The calculated run-up elevations on the slopes north and south of the intake structure are well below the plant grade of Elevation 45'.

CALVERT CLIFFS UFSAR 2.8-8 Rev. 47 2.8.3.6 Wave Run-up at Intake Structure Introduction - The saltwater cooling pumps, which are essential for safe shutdown of the CCNPP, are housed in the intake structure. Since these are Class I components, it was decided to design the enclosure as a Category I structure for seismic, tornado, and hurricane conditions. The hurricane surge calculations in Section 2.8.3.5 are based on implementation of References 13 and 14. Using a number of conservative assumptions and the results of a series of model wave run-up tests at the University of Florida, it is concluded that the structural integrity of the intake structure will be maintained under PMH conditions. Thus, the saltwater cooling pumps can continue to operate under PMH conditions. Configuration of the Intake Structure - The intake structure has an open deck at Elevation 10.0' MSL on the Bay side. The deck is about 50' wide and has openings for the trash rakes and racks, stop logs, and traveling screens. Behind this open deck is an enclosure housing the circulating water pumps and saltwater cooling pumps. The roof of the pump room is at Elevation 28.5' MSL and has watertight hatches to provide access to the pumps for maintenance. An intake structure air supply unit is mounted on each saltwater pump hatch, and an air exhaust vent is mounted on each circulating water pump hatch. To minimize entry of moisture into the pump room, each air supply unit and air exhaust vent housing is provided with louvers designed for high moisture separation efficiency. Similarly, a separate ventilation system draws outside air through the service building (west wall) to minimize water entry. The personnel door located at the north end of the intake structure is of a watertight design. Model Tests - The configuration of the intake structure is not one of the classic profiles which has been tested in the past. That is, it is not a curved or vertical wall, a stepped wall, or a riprap-covered wall like those shown in Figures 3-6 through 3-10 in Reference 14. However, T. E. Haeussner prepared a report predicting a maximum run-up Elevation of 28.1' MSL based on the curves of Figure 3-6, with an appropriate reduction factor for the case of waves breaking on the front edge of the intake structure.

There were questions on this approach and it was decided to perform two-dimensional model tests at the University of Florida to determine the run-up elevation. These tests were run with a prototype wave period of 9.0 seconds and wave height of 11.4', the significant wave height. Both of these values appear in Section 2.8.3.5. Three still-water Elevations were tested: 16.2', 17.2', and 18.2' MSL. The lowest of these, 16.2' MSL, was 0.6' higher than the value calculated in Section 2.8.3.5, due to confusion between MSL and MLW. The higher still-water elevations were to account for some differences of opinion as to the appropriately-conservative value.

During the tests, considerable overtopping resulted in the early runs, so the adverse slope (upper 5' of pump room sloping out at 20% angle) was added. This adverse slope appeared to eliminate the overtopping. Even with the still-water at 18.2' MSL, an analysis of the test runs indicates that the run-up carries only to Elevation 26.5' MSL, a run-up of 8.3'. Figure 2.8-9 shows the output of the final test runs. A color film was available for viewing. Table 2-46 is a summary of the model test results compared with run-up computed from Figure 3-6 in Reference 14. The highest model/computed ratio is 0.50. Predicted Wave Run-up - There have been questions as to the conservatism of basing wave run-up on the run-up from the significant wave. Therefore, the results CALVERT CLIFFS UFSAR 2.8-9 Rev. 47 of the University of Florida tests were used as a predictive tool along with the wave height corresponding to H1, the average of the highest 1% of waves. This is the wave height recommended for design of rigid structures in Reference 14. In Reference 15, C.L. Bretschneider found the expression where: Hmax = maximum wave height Hs = significant wave height g = acceleration of gravity, 32.2 ft/sec2 d = depth, in ft U = wind velocity, in ft/sec (138 for Calvert Cliffs) Solving this expression gives Hmax = 1.27 Hs for 40' depth, and Hmax = 1.30 Hs for 50' depth. Fitting these values to a normal curve approximation yields H1 - 1.23 Hs for 50' 1.23 (11.4') = 14.0. Thus, the design wave for the intake structure is 14.0' high.

As explained previously, the still-water levels tested at the University of Florida were 0.6' high. Using a still-water Elevation of 17.6' MSL, which was calculated by the AEC's consultant, using the 14.0' wave with a period of 9 seconds and using the curves of Reference 14, Figure 3-6 with a reduction factor of 0.50 (the highest model/computed ratio in Table 2-46), the calculated wave run-up is to Elevation 27.1' MSL, 1.4' below the pump house roof. As a comparison with the approach, the following check was made, again based on model test results:

Run-up with SWL of 17.2' MSL = 7.3' Run-up with SWL of 18.2' MSL = 8.3' Interpolating, run-up for 17.6' MSL = 7.7' Multiplying by H1/Hs (1.23) = 9.5' for 14.0' wave Add still-water level = 17.6' Run-up for 14.0' wave = 27.1' MSL Analysis was also made of the same conditions with periods of 5 and 12 seconds and resulted in somewhat less run-up. There was a question of model scale effects. The predicted run-up of 9.5' came to 1.4' below the pump house roof, so that a run-up of 10.9' would not overtop the structure. (10.9/9.5) = 1.15. Thus, the structure allows for model scale effects of 15%. This is a greater increase in run-up than predicted from use of the method of effective slope outlined in Reference 14. Structural Analysis of the Intake Structure - The intake structure has been analyzed for seismic and tornado loadings. In addition, although it is not expected that the pump house portion of the structure will see breaking or broken waves, the intake structure has been analyzed for the following hurricane loading conditions: a. Nonbreaking wave for water to the top of the roof, Elevation 28.5' MSL; still-water level of 17.6' MSL; and wave periods of 5, 9, and 12 seconds. b. Broken wave for a wave height of 14.0'; still-water level of 17.6' MSL; and wave periods of 5, 9, and 12 seconds. c. Breaking wave for the highest wave which could continue unbroken across the front edge of the structure. Since the still-water level is 17.6' MSL and CALVERT CLIFFS UFSAR 2.8-10 Rev. 47 the deck Elevation is 10.0' MSL, the controlling depth is 7.6'. Thus, the maximum wave height is 0.78x(7.6) or 5.9'. This condition also has been examined for periods of 5, 9, and 12 seconds. For each of the above loading conditions, the analysis shows that structural integrity will be maintained.

Additional Considerations - Curbs a minimum of 6" high are provided around the roof hatches. The roof and hatch covers are designed for a live loading of 250 psf. The louvered housings for intake structure air supply units and exhaust vents are designed for a live load of 100 psf.

The baffle wall in the intake channel is designed for conditions less than PMH (85 mph wind with a 9-second period, 10.5' high wave). However, the Florida tests indicate that there would be no overtopping of the intake structure whether or not the baffle wall is in place during the PMH. An analysis showed that even if sections of the wall came loose, they would not damage or block the intake structure. In addition, the saltwater cooling pumps are redundant (two out of three required for each unit) and each can take suction from either of two screen wells.

The concrete stop logs were stored in a recess to the south of the pump house. Thus, they were not considered to be a missile for the intake structure design. Scour at the front edge of the structure is not expected since there is a very low velocity past this point (about 1 ft/sec), and since the foundation soil is a dense silty sand or sandy silt.

There will be no resonant vibration of the structure due to the waves. The structure's natural frequency is about 3 CPS. Conclusions - The following conclusions are drawn from the studies performed: a. The predicted wave run-up is to Elevation 27.1' MSL, 1.4' below the pump room roof elevation. b. The structural integrity of the intake structure will be maintained during hurricane conditions. c. The saltwater cooling pumps can continue to operate under hurricane conditions. 2.8.3.7 Extreme Low Tide Considerations Normal Tides and Tidal Datum - Normal tides in Chesapeake Bay are semidiurnal, or of the mixed type with two highs and two lows occurring daily; and with a higher high and lower low tide level as a daily occurrence. Information available in Reference 5 for Taylors Island, located across the Bay from the site, gives a mean range of 1.3', a spring range of 1.5', and a mean tide level of 0.6'. The latter value indicates that sea level is 0.6' above low water datum. The extreme annual range is approximately 2.3', occurring in December. Low Tide Considerations - Various factors affect and, to a large extent, control the value of extreme low tide elevation at the site. Essentially, they are as follows: a. Hurricane wind direction, duration and intensity in storms passing offshore of Chesapeake Bay. Counterclockwise rotating winds from the northeast to CALVERT CLIFFS UFSAR 2.8-11 Rev. 47 northwest over the Bay will prevail for such a storm path and have the greatest effect in lowering Bay tide levels. b. The location of the plant site with respect to the total length of the Bay. c. Depth of water in the Bay, especially along available west-east tide fetches for generation of maximum set-down at shore. d. General orientation of the Bay, length of the Bay and degree of curvature of the longitudinal axis corresponding to wind streamline curvature of hurricane winds over the Bay. The general orientation of the Chesapeake Bay is north-south, its length from Baltimore to Norfolk is about 165 miles. The site is about two-thirds of that distance from Norfolk. Records of hurricane tides, shown on path and tide maps in References 2, 3, and 5, indicate that for hurricanes passing offshore of the Bay, maximum drawdown in the Bay occurs at the windward end, i.e., Baltimore, becoming less with distance southward toward the mouth of the Bay. For example, in Hurricane Donna of September 1960 (Figure 3, page 188 of Reference 5), the drawdown at Baltimore was to -2.6' MLW; to -1.5' at Annapolis; to -1.0' at Solomons; and to -0.8' at Portsmouth. Analyses of other hurricanes with similar wind conditions over the Bay, such as the October 1944, September 1947, and October 1954 hurricanes, show much the same variation in tidal drawdown within the Bay. Because of its location, Calvert Cliffs does not experience the maximum effects of Bay drawdown, as is indicted by observed data. The maximum drawdown elevation presumably occurred in Hurricane Donna, and is estimated to have been about -1.2' MLW (1.7' MSL). The extreme low tide elevation occurrence believed possible at Calvert Cliffs is predicated on an occurrence of the PMH on a path offshore of Chesapeake Bay, similar to that of Hurricane Donna. Correlating wind intensity over the Bay of the PMH with that of Donna and the drawdown elevations experienced in Donna, an estimated drawdown Elevation of -3.0' could be expected to occur at Baltimore with a value of -1.6' at the site. With a counterclockwise wind shift from north to west over the Bay, an additional setdown of about 0.5' could be expected to occur, giving an extreme low tide Elevation of -2.1' MLW (-2.7' MSL) at the site.

The predicted extreme low tide Elevation is -3.6' MSL (-3.0' MLW). However, the plant has been designed for -4.0' MSL and can continue to operate with an extreme low water Elevation of -6.0' MSL. The top of the saltwater pump intakes is at Elevation -9.5' MSL. 2.

8.4 REFERENCES

1. House Document No. 350, 88th Congress, 2nd Session, Tidewater portions of the Patuxent, Potomac, and Rappahannock Rivers Including Adjacent Chesapeake Bay Shoreline - Interim Hurricane Survey, August 1964 2. D.L. Harris, An Interim Hurricane Storm Surge Forecasting Guide, National Hurricane Research Project, Report No. 32, U.S. Weather Bureau, August 1959 3. D.L. Harris, Some Problems Involved in the Study of Storm Surges, National Hurricane Research Project, Report No. 4, U.S. Weather Bureau, December 1956 4. Inter-Agency Committee Report, The Resources of the New England - New York Region, Part Two, Chapter XXXIX, Special Subjects Regional, Hurricanes, Volume 4, December 1954 5. Tide Tables, 1967 East Coast, North and South American, U.S. Department of Commerce, ESSA, Coast and Geodetic Survey CALVERT CLIFFS UFSAR 2.8-12 Rev. 47 6. Miscellaneous Paper No. 3-59, Hurricane Surge Predication for Chesapeake Bay, BEB, September 1959 7. Technical Memorandum No. 83, Reid, R. D, Approximate Response of Water Level on a Sloping Shelf to a Wind Fetch Which Moves Toward Shore, BEB, June 1956 8. N.H.R.P. Report No. 50, part II, Proceedings of the Second Technical Conference on Hurricanes, U.S. Weather Bureau, March 1962 (pp 341-354) 9. N.H.R.P. Report No. 14, B. I. Miller, On the Maximum Intensity of Hurricanes, U.S. Weather Bureau, December 1957 10. Technical Memorandum No. 120, The Predication of Hurricane Storm-Tides in New York Bay, BEB, August 1960 11. N.H.R.P. Report No. 50, Part I, Proceedings of the Second Technical Conference on Hurricanes, U.S. Weather Bureau, March 1962 12. Miscellaneous Publication, Revisions in Wave Forecasting; Deep and Shallow Water, C.L. Bretschneider, BEB, 1957 13. Environmental Science Services Administration Memorandum HUR 7-97, "Interim Report - Meteorological Characteristics of the Probable Maximum Hurricane, Atlantic and Gulf Coasts of the United States" 14. U.S. Army Coastal Engineering Research Center, Shore Protection Planning and Design, Technical Report No. 4 15. Technical Memorandum No. 46, Field Investigations of Wave Energy Loss in Shallow Water Ocean Waves, BEB, September 1954

CALVERT CLIFFS UFSAR 2.8-13 Rev. 47 TABLE 2-41 BASIC INFORMATION FOR CONSTRUCTING PMH ISOVEL PATTERN DISTANCE FROM CENTER OVERWATER WIND PROFILE PRESSURE EFFECT PRESSURE 7.5 42.05 4.45 27.01 15.0 81.16 3.92 27.48 22.5 111.92 3.34 27.99 30.0 124.71 2.87 28.40 40.0 110.14 2.39 28.82 50.0 96.07 2.05 29.12 60.0 87.81 1.79 29.35 70.0 81.24 1.58 29.53 80.0 74.47 1.42 29.68 90.0 68.90 1.29 29.79 110.0 62.11 1.08 29.97 130.0 56.48 0.94 30.10 150.0 51.50 0.82 30.20 170.0 47.03 0.73 30.28 190.0 42.96 0.66 30.34 210.0 39.22 0.60 30.39 230.0 35.74 0.55 30.43 250.0 32.48 0.51 30.47 R T Po Pn 30.00 23.00 26.94 30.92 CALVERT CLIFFS UFSAR 2.8-14 Rev. 47 TABLE 2-42 ANGLES MEASURED FROM LINE OF FORWARD MOTION (USED FOR CONSTRUCTING ISOVEL PATTERN) DIST. 25 55 85115145175205 2352652953253557.5 42.0 47.8 52.053.552.047.842.0 36.332.130.532.136.315.0 81.2 86.9 91.192.791.186.981.2 75.471.269.771.275.422.5 111.9 117.7 121.9123.4121.9117.7111.9 106.2102.0100.4102.0106.230.0 124.7 130.5 134.7136.2134.7130.5124.7 119.0114.7113.2114.7119.040.0 110.1 115.9 120.1121.6120.1115.9110.1 104.4100.298.6100.2104.450.0 96.1 101.8 106.0107.6106.0101.896.1 90.386.184.686.190.360.0 87.8 93.6 97.899.397.893.687.8 82.177.876.377.882.1 70.0 81.2 87.0 91.292.791.287.081.2 75.571.369.771.375.580.0 74.5 80.2 84.486.084.480.274.5 68.764.563.064.568.790.0 68.9 74.7 78.980.478.974.768.9 63.258.957.458.963.2110.0 62.1 67.9 72.173.672.167.962.1 56.452.250.652.256.4 130.0 56.5 62.2 66.468.066.462.256.5 50.746.545.046.550.7150.0 51.5 57.2 61.563.061.557.251.5 45.741.540.041.545.7170.0 47.0 52.8 57.058.557.052.847.0 41.337.135.537.141.3190.0 43.0 48.7 52.954.552.948.743.0 37.233.031.533.037.2210.0 39.2 45.0 49.250.749.245.039.2 33.529.327.729.333.5230.0 35.7 41.5 45.747.245.741.535.7 30.025.824.225.830.0250.0 32.5 38.2 42.444.042.438.232.5 26.722.521.022.526.7 CALVERT CLIFFS UFSAR 2.8-15 Rev. 47 TABLE 2-43 BASIC DATA - PMH - FETCH FETCH MILE µ (mph) Cos c VALUE µ Cos c = µx (mph)2 DIST. FOR Pe (miles) Pe (ft) Pe (ft) DEPTH @X (ft MLW) X = 0 116 0.42 49 5,684 39 2.42 -.10 0 Hampton Roads 5 122 0.52 63 7,690 37 2.52 -.13 19 10 127 0.60 76 9,650 34 2.65 -.10 23 15 131 0.70 92 12,040 32 2.75 -.11 25 17.5 Bay Entrance 20 133 0.80 106 14,100 30 2.86 0 24 25 136 0.90 122 16,600 30 2.86 0 32 30 136 0.93 127 17,300 30 2.86 0 46 35 136 0.97 132 17,950 30 2.86 0 60 40 136 0.98 133 18,100 30 2.86 0.11 62 45 135 0.95 128 17,300 32 2.75 0.10 65 50 130 0.93 121 15,700 34 2.65 0.13 70 55 126 0.88 111 14,000 37 2.52 0.10 76 60 120 0.85 102 12,230 39 2.42 0.14 82 65 115 0.80 92 10,620 43 2.28 0.11 90 70 110 0.75 83 9,130 46 2.17 0.12 99 75 105 0.70 73 7,660 50 2.05 0.12 100 80 100 0.65 65 6,500 54 1.93 0.08 106 85 97 0.60 58 5,630 57 1.85 0.05 112 90 94 0.56 53 4,980 60 1.80 0.13 130 95 90 0.52 47 4,230 65 1.67 0.09 180 100 87 0.50 44 3,828 70 1.58 0.07 300 105 84 0.45 38 3,200 74 1.51 -800 -Deep Water CALVERT CLIFFS UFSAR 2.8-16 Rev. 47 TABLE 2-44 PMH SURGE COMPUTATION - OCEAN TO BAY ENTRANCE E1 = 2.7 (Mid Point) (ft) Pe (S1) (ft) + S1 (ft) T (ft) 2 (mph) Si (ft) S2 (ft) E2 (ft MLW) 105 800 0.07 800.+ 803 32,000 0.024 0.024 2.724' 100 300 0.07 300.+ 303 3,828 0.077 0.10 2.80' 95 180 0.09 180.+ 183 4,230 0.128 0.23 2.93' 90 130 0.13 130.+ 133 4,980 0.227 0.46 3.16' 85 112 0.05 112.+ 116 5,630 0.296 0.76 3.40' 80 106 0.08 106.+ 110 6,500 0.36 1.12 3.82' 75 100 0.12 100.+ 104 7,660 0.45 1.57 4.27' 70 99 0.12 99.+ 104 9,130 0.53 2.10 4.80' 65 90 0.11 90.+ 95 10,620 0.68 2.78 5.48' 60 82 0.14 83 89 12,230 0.83 3.61 6.31' 55 76 0.10 77 84 14,000 1.01 4.62 7.32' 50 70 0.13 71 79 15,700 1.21 5.83 8.53' 45 65 0.10 66 75+ 17,300 1.39 7.22 9.92' 40 62 0.11 62.+ 75 18,100 1.46 8.68 11.38' 35 60 0 60.+ 74 17,950 1.47+ 10.15 12.85' 30 46 0 46.+ 61 17,300 1.72 11.87 14.57' 25 32 0 32.+ 48 16,600 2.10 13.97 16.67' 20 24 0 24.+ 43 14,100 2.00 15.97 18.67' S0 (Max.) - Surge Elevation at Bay Entrance (Mile 17.5) - 18.67' MLW or 17.32' MSL. CALVERT CLIFFS UFSAR 2.8-17 Rev. 47 TABLE 2-45 BAY SETUP COMPUTATIONS AT PLANT SITE Used Parametric Relationship (Corps of Eng. - Jax. Dist) Based on Formula: (N factor not included) SETUP PORTION E1 = 13.44' + 0.6' = 14.04' Fetch F miles av mph Dav ft W/T Slope(1) ft/M1. Setup S ft s ft 2DS+ ft Setup S(2) ft ft E2 ft MLW 17 2.70 94 56.4 0.20 +0.54 0.54 56.7 0.51 0.51 14.55' 0.30 94 40.4 0.20 0.06 0.60 40.7 0.07 0.58 14.62' 0.65 94 28.4 0.38 0.25 0.85 28.8 0.24 0.82 14.86' 0.35 94 19.4 0.54 0.19 1.04 19.9 0.19 1.01 15.05' 0.15 94 8.4 1.40 0.21 1.25 9.0 0.16 1.17 15.21' 4.15 Surge in bay + crosswind effect = 15.21' Added wave effect = 1.00' (est.) Total Tide = 16.21' MLW (15.6' MSL) CALVERT CLIFFS UFSAR 2.8-18 Rev. 47 TABLE 2-46 WAVE RUN-UP AT INTAKE STRUCTURE Comparison of Model and Computer Wave Run-up STILL-WATER ELEVATION (MSL) COMPUTED RUN-UP (T.R. 4, Figure 3-6) MODEL RUN-UP MODEL/ COMPUTED RATIO 16.2' 14.8' 6.7' 0.45 17.2' 16.2' 7.3' 0.45 18.2' 16.7' 8.3' 0.50

CALVERT CLIFFS UFSAR 2.9-1 Rev. 47 2.9 ENVIRONMENTAL RADIATION MONITORING 2.9.1 GENERAL The objectives of the radiological environmental monitoring program at CCNPP are to: - Measure actual radiation exposure to the general population at the fence line and beyond. - Observe any sudden or unexpected rise in radiation levels in the vicinity of the plant. - Document for legal and regulatory purposes actual radiation exposure levels and radionuclide concentrations in air, Bay surface water, sediment, fish, invertebrates, and vegetation. - Provide monitoring services in emergency situations. In order to fulfill these objectives, the radiological environmental monitoring program must differentiate between naturally-occurring and artificially-introduced radioactivity in the environment, and between plant-related and unrelated radioactivity. The radiological monitoring program is carried out in two phases: preoperational and operational. 2.9.2 PREOPERATIONAL RADIATION MONITORING In accordance with the requirements of the Calvert Cliffs Safety Analysis Report and the conditions of the construction permits, BGE initiated work in 1969 on the design and development of the radiological environmental monitoring program for Calvert Cliffs. Concurrent with the design and development of the monitoring program, BGE, in contract with NUS, initiated several studies to assess the potential dose impact of expected radio-effluents from Calvert Cliffs. These studies addressed the following topics:

- Build-up of radionuclides in the aquatic environment; - Relative biological significance of radionuclides; 

- Estimate of potential dose to a maximum exposed individual via seafood ingestion; - Estimate of potential immersion dose from noble gases, and thyroid inhalation dose to a hypothetical individual at the site boundary; - Estimate of potential adult-thyroid and child-thyroid dose via the air-pasture-cow-milk pathway; and, - Potential dose to population within 50-mile radius. A review of the results of these studies was made in conjunction with other environmental data. The purpose of this review was to ascertain the significance of the various exposure pathways, and to identify the "potential critical pathways" in the area of the facility. The results of this review and the mandatory requirements based on the regulatory limitations on dose, radiation/radioactivity levels as published on June 7, 1971 (10 CFR Part 50, Appendix I) determined the design of the monitoring program. The preoperational phase provided both seasonal and annual information about the distribution of natural radioactivity in the region, defined the ambient gamma-radiation levels, and obtained baseline data for some of the more important radionuclides, both natural and man-made. 2.9.3 OPERATIONAL RADIATION MONITORING With the issuance of the operating license for Calvert Cliffs Unit 1 on August 1, 1974, BGE began the operational phase of the monitoring program.

Between 1974 and 1985, the program was carried out based on the environmental monitoring network designed in the preoperational phase. On February 22, 1985, the CALVERT CLIFFS UFSAR 2.9-2 Rev. 47 NRC issued the Technical Specifications associated with the environmental monitoring program to assure the compliance with the provisions of 10 CFR Part 20, 10 CFR Part 50, 40 CFR Part 190, and NUREG-0472. The new operational program started on March 1, 1985, as per Table 2-47 and Figure 2.9-1.

In its present form, the radiological environmental monitoring program requires sufficient sample locations, types of samples, and analytical sensitivities which, in conjunction with the preoperational and background data, permit verification of the effectiveness of station radio-effluent control. The program provides data on changes in use of unrestricted areas and meets quality assurance criteria. The results of the program provide an indication of a measurable change, if any, in radiation and radioactivity in the environment, and provide reasonable assurance that the releases are within the limits specified in the Offsite Dose Calculation Manual for plant operation. The program is periodically reviewed to determine any changes that may be warranted in its content.

CALVERT CLIFFS UFSAR 2.9-3 Rev. 47 TABLE 2-47 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE PATHWAY AND/OR SAMPLE NUMBER OF REPRESENTATIVE SAMPLES AND SAMPLE LOCATIONS SAMPLING AND COLLECTION FREQUENCY TYPE AND FREQUENCY OF ANALYSIS 1. DIRECT RADIATION 23 routine monitoring stations (DR1-DR23) either with two or more dosimeters, or with one instrument for measuring and recording dose rate continuously, placed as follows: At least quarterly Gamma dose at least quarterly an inner ring of stations, one in each meteorological sector in the general area of the SITE BOUNDARY (DR1-DR9)(1); an outer ring of stations, one in each meteorological sector in the 6- to 8-km range from the site (DR10-DR18); the remaining stations (DR19-DR23) to be placed in special interest areas such as population centers, nearby residences, schools, and in 1 area to serve as a control station 2. AIRBORNE Samples from 5 locations (A1-A5): Continuous sampler operation with sample collection weekly, or more frequently if required by dust loading Radioiodine Canister I-131 analysis weekly Radioiodine and Particulates 3 samples (A1-A3) from close to the 3 SITE BOUNDARY locations, in different sectors of the highest calculated annual average ground-level D/Q(1) Particulate Sampler: Gross beta radioactivity analysis following filter change; Gamma isotopic analysis of composite (by location) quarterly 1 sample (A4) from the vicinity of a community having the highest calculated annual average ground-level D/Q 1 sample (A5) from a control location, as for example 15-30 km distant and in the least prevalent wind direction CALVERT CLIFFS UFSAR 2.9-4 Rev. 47 TABLE 2-47 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE PATHWAY AND/OR SAMPLE NUMBER OF REPRESENTATIVE SAMPLES AND SAMPLE LOCATIONS SAMPLING AND COLLECTION FREQUENCY TYPE AND FREQUENCY OF ANALYSIS 3. WATERBORNE a. Surface 1 sample at intake area (Wa1) 1 sample at discharge area (Wa2) Composite sample over 1-month period Gamma isotopic analysis monthly Composite for tritium analysis quarterly b. Sediment from shoreline 1 sample from downstream area with existing or potential recreational value (Wb1) Semi-annually Gamma isotopic analysis semi-annually 4. INGESTION a. Fish and Invertebrates 3 samples of commercially and/or recreationally important species (2 fish species and 1 invertebrate species) in vicinity of plant discharge area (Ia1-Ia3) Sample in season, or semi-annually if they are not seasonal Gamma isotopic analysis on edible portions 3 samples of same species in areas not influenced by plant discharge (Ia4-Ia6, Patuxent River) b. Food Products Samples of 3 different kinds of broad leaf vegetation grown near the site boundary at 2 different locations of highest predicted annual average groundlevel D/Q (Ib1-Ib6)(1) Monthly during growing season Gamma isotopic and I-131 analysis 1 sample of each of the similar broad leaf vegetation grown 15-30 km distant in the least prevalent wind direction (Ib7-Ib9) Monthly during growing season Gamma isotopic and I-131 analysis

_______________________ (1) Exception to these locations is in the South Sector where DR7, A1, lb4, lb5 and lb6 are located approximately 0.7 km from the release point. This location is conservative with respect to the site boundary, which is located approximately 2.1 km from the release point. CALVERT CLIFFS UFSAR 2.10-1 Rev. 47 CALVERT CLIFFS UFSAR 2.10-2 Rev. 47 CALVERT CLIFFS UFSAR 2.10-3 Rev. 47 2.2-1 REGIONAL MAF REGIONAL MAP " ,.,,

• c" r
• r" r rr *u* .. *=' .. 1 ,.,,., ....,. ,..,, "t*dl6 ,_ r .. * .,., IC'f.fl:'M.tlt.. ""!"...._ .... ,.._ &;>-**: _ ...... ""' ---....... Rev.a SITE VICINITY MAP STATUTE MILES 0 I 2 **!!**!!** ::::::==::====== R E F E RE N C E: THIS MAP WAS PREPARED FROM A PORTION OF USG$ WASHINGTON, o.c.-MARYLAND AND VIRGINIA ICJ57 TOPOGRAPHIC W.Po DAM*88MOOR* FIGURE 2.2-2 i 1 w..--* ' /J I <<t S.t 1 J; ,l I *I " l i 1 l * ,.., l .? '*-j ****-** ! TE'
• t!.1'-. -*-*-* ! j POflJL.ATfON OENStTIES AR£ BA.SEO ON LA'f> -'REA* -1f , .i/ !? t .* 'Jj ,_, 1 ti ._, r*>-._,, l .. .... ______ ... .i.. KEY = ESTIMATED 1970 POPULATION PER SQUARE MILE 78 ESTIMATED 2010 POPULATION PER SQUARE MILE REGIONAL SHOWING PRESENT POPULATION DENSITY IO 1R C F C R C N C E: STATUTE lllLH 0 10 MAP ! THIS Ml\P WAS PREPARED FROM THE FOLLOWING UHS SEC1'10NAL. jA£ROf<A\ITICAL CHARts: WASMIHGTQN AND NORFOLK. ' !;$VISION 5 ---***--** FIGURE Z.2-3 0 0 N 0 0 POPULATION DENSITIES ARE BASED ON LAND AREA* KEY: 52 _ ESTIMATED 1970 POPULATION PER SQUARE MILE 79-ESTIMATED 2010 POPULATION PER SQUARE MILE SITE VICINITY MAP SHOWING PRESENT AND FUTURE PO*lJLA TI01' DEMJilT'I' 0-10 MILBS STATUTE MILES 0 I I ----------**********-----------II!: f' ER!: NC 1:: THIS MAP WAS l'REPAREO l"ROU A l'ORTION OP' USGS WASHINGTON, o.c."WIRYLAND AND VIRGINIA 1957 TDl'DGRAl'HI C MllP* .............. REVISION 5 FIGURE 2.2-4 KEY: = ESTIMATED 1970 POPULATION 261 ESTIMATED 2010 POPULATION REGIONAL MAP SHOWING PRESENT AlfD FUTUltB POP:Ul.ATION DISTRIBUTIOlf O *SO *ILBS 11 E F E A E N C e:: STATUTE MILH 0 10 IO THIS UAP WAS PREPARED ,ROM THE f'OLLDWING USGS SECTIONAL AERONAUTICAL CHAAts: WASlllNGTON AND NORFOLK* ................. Fl8UR£ 2.2-8

... . .... ... ... .. \ 0 0 N 0 0 KEY: 210 ESTIMATED 1970 POPULATION 58i) c ESTIMATED 2010 POPULATION SITE VICINITY MAP SBOWJNC PRBSBRT ARD FUTURE POPULATION DISTRIBUTION 0*10 MILBS STATUTE MILES 0 I I ------------**********----------* r, E R r "c E: THIS '-'AP WAS PREPARED FROM A PORTICIH Ill' UHS WASHINGTON* D*C*._YUIND MD VlllGllllA 1957 TDPDQRAPHIC -* a1V1SION 5 ----*-**** ,IGURE 2.2-* 1 -, ' ..... _ FIGURE 2..2-1 AIRPORTS IN THE VICINITY OF CALVERT CLIFFS NUCLEAR POWER PLANT HOBBS 71'i / HOUGH 45 Calvert Cliffs Nuclear Power Plant METEOROLOGICAL INSTRUMENTATION LOCATIONS z n -< . ...

• Figure 2.3-1 Revision 32 BALTIMORE GAS & ELECTRIC CO. Calvert Cliffs Nuclear Power Plant Figure 2.2-13 LOW POPULATION ZONE Rev.18

\ \ DAM*& 8 MOOll*

N ' 2500 METERS w E s AVCRAGE ANNUAL VENTING RELATIVE CONCENTRATION FIGURE 2.3 -4 (Rev. 3/3/72) I a MOO** ._ ____________________________________________________________________________ ! D,.. -Ci) c: ;o f""I N . w I VI a , I Ill Ul °' I g II !I I-' 0 -N N --...J I-' A COMPARISON or 12-50 FEET ABOVE VALUES WITH THOSE BETWEEN 12-97 FEET. so -+*----------------*---------* -0 0} I 0 f OBSERVATIONAL INTERVAL BOUNDARIES. ,_ z w u 40 +----' --1--------------+-I a:: w 11.. -18o6 CONCURRENT OBSERVATIONS >-u z w ::> 30 1-------f * -----------i I ....---12-50.FOOT LAYER 0 w a:: "'-20 w > 12-97 FOOT LAYER -,_ < _. ::> 2:: ::> IJ 10 1 ASQUILL CLASSES f-0 A __ _l ___ __ L_I I F I I G 1 0 1 -2 3 4 Oc/100 METERS ( -} I

• I
• I I ,, G5 c :::a fTl N (JJ I (J) -g' < w -w --..J N '-' 91' -30' llT *r H!.I cl D G 1-.8 1-.7 1-.6 1-.5 1-.4 1-.3 -.2 1-.1 0 +.1 +.2 +.3 f+.4 l+.5 l+.6 l+.7 l+.8 l+.91+1.ol+1.1!+1.2!+1.3!+1.lil+1.5l+1.6h1.7 !.5 .... -.4 Gl .. "<-I:.: ". -.2 i=:===r***1==-***1*****1:;===1=-**...:*, .. .... --.1
• 0 : : :: *1: :: ::1::: ::1:::: :1: *.*. ' ' :-,,.., + .. 1 ==***1gm,.
• 1::::: 1- u 0:: UJ a.. (/) c z 0 u UJ (/) .. (/) z 0 <( 0:: 1-z UJ u z 0 u L&J :: <( ...J Li.I 0:: FREQUENCIES OF ON-SHORE RELATIVE CONCENTRATIONS ASSUMING AN INVARIANT CENTERLINE WIND CUMMULATIVE FREQUENCY PERCENTAGE OF TOTAL TIME trzDETERMINED BY 12 -50 FT. 6 T PASQUILL CLASSES BY 12 -50 FT. PASQUILL CLASSES AT WIND <3 MPH AND BY 33 FT. LEVEL CLASSES AT WINDS >3 MPH (Rev. 3/3/72) FIGURE 2.3 -7 DAM*S & MOOR*

DATA CALENDAR FINAL ANNUAL RECORD 1969 DATE NOV DEC I I I I : : i ,__fl I 2 J l * ! f j Ir ! i 21 I 3 ' I ' ' I l --4 +-: 1_t-: 5 .... -:-; I : : 6 I ra--r--1 -:-Bi 7 21 12111 ! f; 8 21 ! 111 ; l 9 I ' :fl: I I 10 I i i i ; J IJ II I I I Z-1 ! ! 12 i i I I i 13 I ! ! I. a: 14 I l ! I ! -r I l=l=J I Ii: 2:-i 17 -I 2H-i I :. ,,. . i. ' 18 a: 19, I : 1H 20: i IH-21 1 i

• I : 1 1 22 1 i a ; : a 1 : i 22431 I i :.* ! i i ! : H : ! I _.__.__.__.._._I I : : : I : : H 21: i i ; ; M' 28 ! *. :.' !. ' == PRIMARY YEAR OF RECORD DATA
• 1971 DATA FIGURE 2 3-8 (Rev. 3/3/72) DAM*eBMOOR*

DATA CALENDAR NOVEMBER 1970 THROUGH OCTOBER 1971 1970 1971 DATE NOVbEC JAN FES1rviAR:APR1MAY1JUN JUL1AUG:SEPIOCT I I * -= I I 2 i = = = I = 3 : I = I I ! I = ! -l = = = ! = I 1--5 : = = 6 I : = = >-----I i 7 I ; = i = 8 ' ' -J I ; i w 9 I I I -! ! : I = -=--* ' 10 I j ;;;;j I P= '-II I .: i -! : 12 = : i ' 13 I ! i ! : I -I 14 I i I I i ! ' i = ' I i 15 ! i ! = i ! 16 I ! I = l i 17 I ! I : l =i ' I 18 I I ! ! = ! -I -' : =! I 19 ! I = i : 20 ! = ! I= I = I I 'I= ! --..._ 21 j ! l -i i I i I I 22 I i I i -, I = ! i I -23 I I I i i I 1 -' = _; 24 i i I ! = I = i I i -I 25 ! I i = : = --I = 26 I ' ::;;;;; I = I I I = I --I 27 I ! I = -i I = ! -29 I --= I 30 Clltiit CIC 31 ---33 FT LEVEL WIND, AND 50-12 FT LEVEL .6 T DATA ARE NECESSARY FOR X/Q ANALYSIS ALL NECESSARY DATA AVAILABLE _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

• 50-12 FT .6T DATA AVAILABLE 100 FT LEVEL WIND DATA ARE __ -AVAILABLE INSTEAD OF 33 FT LEVEL DATA FIGURE 2 3-9 (Rev. 3/3/72) DAM*&& MOOR*

WIND FREQUENCY DISTRIBUTION ( FR[QUCNCY IN NUMnER or OCCURRENCES) CALVERT CLIFFS PLANT SITE: STATION 2(IS) DATA PERIOD: NOVEMBER 1, 1969 THROUGH OCTOBER 31, 1970 -I (WITH 1971 DATA SUBSTITUTIONS) :::r: rn 8 "' (Rev. 3/3/72) -I ;o !'.'.;" .... PlllOOILL II 1'110t1 llRC/Dlt.Tll T ClllTtlllll*50*1t nttl PlllGUILI. 11 ,,_ llHIDD.Tll T ClltTIRllh50*11 FHU OJ c: PllSOUILI. II 1'1IOW "EC/UOttA THl:Tll ClllURl/11 llMIO! UUDI PllSOUILI. I IP-AJ:CISlllPIA THUii ClllTllll"' llN!OI USEDI I > ..... SECTOR UPP!ll CLASS IHTiRVAl..S or VUID SPEED (HP!t> 11£1111 SECTOR UPM:ll ct.llSS IHfl!llVl\LS or VIND SPEED IHP!O HtNI "' a .0 :z: I I 3 " 5 6 7 9 9 10 II > 11 10TAI.. 8Ptr.D I 9 3 " 9 6 1 8 9 10 II *II TOTI\.!. SPUD Si a 0 0 I 111n: 0 0 0 I I 0 0 r "Tl Jiii! 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0

• 3*75 r HE 0 0 0 0 0 0 0 0 0 0 0 0 0 I NI 0 0 0 0 0 0 0 0 0 0 0 0 0 I n DIE I 0 0 0 0 0 0 0 0 0 0 0 I ,90 !Ill 0 0 0 0 0 0 0 0 0 0 0 0 0 I r :z: I 0 0 0 0 0 0 0 0 0 0 0 0 0 I I! 0 0 0 0 I 0 0 0 0 0 0 0 I .ti.50 > c:::> 151 0 0 0 I 0 0 0 0 0 0 0 0 I 3,go IB! 0 0 0 0 0 0 0 0 0 0 0 0 0 I V> V> "' :::0-21 IE 0 0 0 0 0 0 0 0 0 0 0 0 0 I SE 0 0 0 0 I 0 0 0 0 0 0 0 I 5*00 -0 ISi 0 0 I 0 0 0 0 0 0 0 0 0 I 8*70 SS! D 0 I 0 0 0 0 0 0 0 0 0 I 1*30 > rn rn
• 0 I 0 I I a I 0 0 0 0 0 T **94 s 0 *o 0 I II 9 I 0 0 0 0 0 9 9*01 n c:::> JISll 0 0 I 3 I 9 I 0 0 0 0 0 II 4,90 ssv I I 0 I 0 0 0 0 0 0 0 0 3 l*IT a . ti> ...., av 0 0 I 6 a I 0
• 0 0 0 0 0 10 3*86 &II 0 I 0 0 0 I 0 0 0 0 0 0 I 3**5 :z: -* (1) c:::> a 0 N vsv 0 0 " e " " 0 0 0 0 0 0 14 **O* vsv 0 0 0 e I 0 0 0 0 0 0 I " 9,51 3 ;;; :::s
• v 0 I II " 9 6 0 I 0 0 0 I Bl **71 II 0 0 0 0 3 I I 0 0 t I e ID 7*9" rn VJ lnlll 0 0 I 0 0 0 I 0 0 0 0 0 I **75 ll!IV 0 0 0 I 0 I 0 I 0 0 0 0 3 5*30 a n ...... I HV 0 0 0 I I 0 0 0 0 0 0 0 R 4*05 1111 0 0 I 0 0 0 0 0 0 I 0 0 t s.eo :z: ;::! 00 ...... NNV 0 0 0 0 0 0 0 0 0 0 0 0 0 I HNV 0 0 0 0 0 0 I 0 0 I 0 0
• e.os = a 0 " 0 0 0 0 0 0 0 0 0 0 0 0 0 I If 0 0 0 0 0 0 0 0 0 0 0 0 0 I -I ti> TOTAL I 3 10 IS ** lft " I 0 0 D I 70 4,37 TOTAi. I 2 8 6 9 8 3 I 0 " I 3 40 ,,,, :::r: > :::;" rn :z: (1) c:::> (1) "Tl ...... <l>q ...... r -0 ,,. > :z: "' z .0 PllSGUILL A crnoH A!C/D!:l.TA T CRIURll\*90-lft n:tr> PllSQUILI. A ITROll A!CID!:l.Tll T CRIT!RIA*50*19 f'!l!:U* c: c > ..... PllSQUILL C IP1!0H AlCISIOHA TH&TA CRltEAllll RANG! USED> PASQUILI. D HROff A!CISIOHA THETA CRITERIA! HAllOE US!DI r r I r-UCfOR UPPF.R CLASS l!ltr.RVAl.5 01 lllHD SPEED IHP111 HEAN SECTOR UPPF.R CLASS INT!AVALS O' VlllD SrtED (Hl'l!I HEii/i ;o rn n I g 3 4 ' 6 7 9 9 10 11 >I I TOTAL SPUD I a 3 4 9 6 1 ft 9 10 II > 11 TOTH. SPEr.o n r 0 > Hll!: 0 0 0 0 3 8 I 0 0 0 0 0 & &*M 11n 0 I 0 3 1 5 1 ' ;o V> I
• I 0 3t 6ol0 0 V> Ht 0 0 I Q I 0 0 0 D 0 0 0
• 3, 6B Hr. 0 0 I 1 10 3 6 R RI Q 0 0 31 , . ., rn lrff!: 0 0 0 I 8 R 0 0 0 0 0 0 ' mr. 0 0 I 0 ,10 8 3 0 I 0 I 0 IS , .. o V> It 0 0 0 I 0 R 0 0 0 u 0 0 3 **03 ! 0 0 0 3 ft * ' R 3 0 0 0 u s.e3 a ESE 0 0 I 0 0 0 I 0 0 0 . 0 0 3 3,93 !St 0 I I I e
• a 3 I I 0 0 5oeR c: SE 0 0 0 I 0 0 0 0 0 I 0 0 I e.so sr. 0 0 I 3 3 e 0 I I I 0 0 IA ,. Q5 ;o ...... ssr. 0 0 0 I a 0 0 I a 0 0 I 1 7* 19 SSE 0 I 3 2 8 I I I ' 0 0 0 P.1 s.at :z: ( s 0 0 0 0 e 0 I 0 0 0 0 0 3 s 0 3 2 I I 3 2 " 0 0 I 0 IA t\1:°!1) G> SSll 0 e 0 I 0 0 I 0 tl 0 0 0 4 3.00 SSV 0 a I 0 I I a I \) 0 0 0 9 4,51 SV 0 0 0 I 0 0 0 0 0 0 D 0 I 3,70 I 3 0 I I 8 0 I 0 0 0 0 9 3*5R llSV 0 0 g 0 I 0 0 0 0 0 0 I " ,. 30 ll!V 0 I 3 0 0 0 0 0 0
• 0 8 10 7.qe a v 0 I I g e 0 I 0 I I I e 10 6071 v 0 I Q 3 R 0 I I 0 3 I 3 IT ,,. VHll 0 0 a g 3 e 9 0 e I 0 1 15 6* no 11!111 0 0 0 3 3 3 3 0 I 0 R -19 7*50 3: HV 0 0 0 0 I 0 R 4 I 0 n I II 7, 95 HV 0 0 I I 4 5 3 4 5 4 5 la 44 9*83 rn HHV 0 0 0 I e q u 0 I I 0 I 10 47 tltlll-0 0 0 I e ' 4 3 6 I 10 38 8*81 "' H 0 0 0 (J I 3 3 0 0 0 0 u 7 5, ?Q N 0 0 0 3 6 1 10 3 g e 3 7ol0 {;IQ 3: TOTfoL 0 3 13 ao 15 IR 7 3 1 97 9,95 0 TOTAL I 13 16 as 6G 4U .,, 3U :'?"I 03 14 3* 361 6061 0 ;o rn

"' w u :z: w c:: §§ u u 0 u... 0 * : > Q) -'=-::i::l a: cc ., ........ .... ...... .,,_, :; .. : :: E : = * -000000-00000000 cooooooo--oocooo -QOOOOOO<<OOOOCCOO oooooco-00-00000 -coooocooo-000000 -Figure 2.3-10, sheet 2 Revision 18 O **f--* . ---oi-= a 0 :z ::l:I tr. cc "' ..... ., Q : OOOQOCOOOOQOOOO--o-oooooooooeoo-m ooooooooooocooom c -0000000000-oooeo -oocooooooooooooo 0 -ooooooooooocoooo c THE DISTRIBUTION OF \/IND SPEED, DIRECTION, ANO o-0 PASQUILL CLASSES DURING M PASQUILL CLASS A CONDITION IN THE FINAL ANNUAL RECORD ,., .. 0 .. DA.HES & MOORE *

z: 8 1-:::i co a:'. 1-(.1") 0 Ll :z: w =o er w a:'. u... 0 :z: (.I") w Ll :z: w a:'. "" =o Ll Ll 0 u.. o a:'. w c::i ::>;: :::> :z: Ll :z: w =o er w a:'. ....._ N -...... VJ f5 u... en u... ::>;: w _, > Ll 0 :z: a:'. w > 0 ;i 8 u a:'. w c.. c:: et: Cl > Q) -------No--*o---o ., .... .. . " ! e 0000000000000000 c = 0000000000000000 0 0000000000000000 0 0000000000000000 Q oooocooooooooooo 0 0000000000000000 c 00000000000--000

• 00000000000-0000 coooooooooo-0000 cooooooooooocooo Q 00000000-00-0000 * . " OOQOOOOOOOOOOOOO 0 SS :5 ... .. H ; * .;.;.; .0 .., ! 0000-00"*-o*omoo e 0000000000000000 0 = 0000000000000000 0 0 0000000000000-00 -§ i-0000000000000000 0 ..... Ei .
• c
• e .. .... .. l: cl_ ... ss cc *c c .. .. oooocoooocoooooo 0 c 0000000000000000 0 "' .. 00000000000-0000 i 00000000-0000-00 .. .. -. 0000000000000000 0 : c dn 0000-00-0-000000 ii !* 00000000-00-0000 0000000000000000 0 :!:! SS n u! Figure 2.3-10, sheet 3 Revision 18 .... _, _, _, _, << E .. .... lllO'OSICl ...... 4I09l Sll_,,., .. 0 .. .. 0 .. 0000-00-oa-oooc-oo-r>--caooccooooa *-00-000000--oc-0--000-oooooaooo 0000000000000000 0.000000-00-00000 0000000000000000 .,.i.a:w>:..:11> "' .. ... "' "' 0 0 .. ;i V\4'\ -1AW'I ""'"""° Cl "'"' a*oo--00000-ano-! Hi "'"' .... *c _, _, _, _, e :: e e 0000000000000000 0 0000000000000000 0 -ooooocooooco-oo .. 0-000000000-0000 0000--000000-oow --0000000000-000 0000000000000-00 ooaaooocoocaoooo 0 0000000000000000 0 0000000000000000 0 THE DISTRIBUTION OF VIND SPEED. DIRECTION, AND 0-0 PASQUILL CLASSES DURING 6.T PASQUILL CLASS B CONDITION IN THE FINAL ANNUAL RECORD DAMES & MOORE WIND FREQUENCY DISTRIBUTiON (FREQUENCY IN NUMBER OF OCCURRENCES) CALVERT CLIFFS PLANT SITE: STATION 2(IS) DATA PERIOD: NOVEMBER 1, 1969 THROUGH OCTOBER 31, 1970 -I (WITH 1971 DATA SUBSTITUTIONS) ::i:: m c ..... en -I P0411JLL * ( "'°" lllQIDa.. TA T C"ITllll ... ll0* 11 rnn PAIQIJILL
• CPAOll AICIDn.tA 1 C"l1111IA. IO*ll Plttl (Rev. 3/3/72) PAIQllJLL II Al:ll/SI 0:111 TllUll Cl\I TDllAI llAHOll USIU>I PASQUILI. P C lllC/llOMA 'lltllTll ClllTllJIJAI llMIOll UUl>I co c: SEC TO II i.>PDI CLllSS UTtllV""-S or ¥1110 SPtl!D CHPlll lf!Alf SECTOR lll'PIOI a.AU !HTDIV.U.S or VllfD IPttD (lfPllJ lfl:Alt .,, -I > ..... I I 3 " 5 6 1 9 9 10 II *II TOTAi. SPEED I ll 3
• s ' T * ' 10 II *II TOTAL SPUD en 0 .0 z HNI 0 0 0 0 I 0 3 0 0 0 0 0 " 5,95 HllE 0 0 0 0 I 0 0 0 0 0 0 0 I 4,40 c: ..... 0 HI 0 0 0 I 0 I 0 0 0 0 0 0 I "*90 HE 0 0 0 I I 0 0 0 0 I 0 0 3 r .,, IMI 0 0 0 0 I 0 0 0 0 0 0 0 I **10 !Nit 0 0 0 0 0 0 0 0 0 0 0 0 0 I r c II 0 0 0 0 0 I 0 0 0 0 0 0 I s.10 IC' 0 0 I 0 0 0 I 0 0 0 0 0 I ..... n ..... .. I 0 0 0 0 I 0 I 0 0 0 0 0 8 s.*s [51! 0 0 0 0 0 0 0 0 0 0 q 0 0 I r z SI 0 0 0 0 0 0 0 I 0 0 0 0 I 7*70 Sit 0 0 0 0 0 0 0 0 0 0 0 0 0 I > c cn sn 0 0 0 0 0 0 0 0 I I 0 0 a 9o3D SS! 0 0 0 0 0 0 0 0 I 0 0 0 I lo30 cn en I 0 0 0 0 0 0 0 0 0 0 0 0 0 I 5 0 0 0 0 0 0 0 0 0 0 0 0 0 I .,, :;d :::1 UV 0 0 0 0 I 0 0 0 0 0 0 0 I Al*10 SSV 0 0 0 0 0 0 0 0 0 0 0 0 0 I co m m IV 0 0 0 0 0 0 0 0 0 0 I 0 I II *00 sv 0 0 0 0 0 0 0 0 0 0 0 0 0 I n 0 vsv 0 0 0 I 0 0 0 0 0 0 0 0 I 3.ao VSV 0 0 0 0 0 0 0 0 0 0 0 0 0 I 0 . ::z: en ....., v 0 0 0 0 0 0 0 0 0 0 0 0 0 I v 0 0 0 0 0 0 0 0 0 0 0 0 0 I 0 0 -* (1) YllV 0 0 0 0 0 0 0 I 0 0 0 0 I To30 lr!IV 0 0 0 0 0 0 0 0 0 0 0 0 0 I '.:::1 ;o 0 N NV 0 0 0 0 0 0 II I I 0 I 3 8 9,90 NV 0 0 l 0 0 0 0 0 0 0 0 0 I **90 ..... m ::l
• NH\/ 0 0 I 0 0 I 0 I I 9 I 3 II 9,95 HHV 0 0 0 0 0 0 0 0 I 0 0 I I '"*OO 0 n w II 0 0 0 a I 3 I I I 0 0 I II 60 65 H 0 0 0 I 0 I 3 0 0 0 I I II 11.01 ::z: ;j ........ I ..... 0 00 ,.._. TOTN.. ::z: 0 0 l
• 5 6 1 5 6 3 3 7 47 1.116 TOTN.. 0 0 ft e 3 I Al 0 I I
• I " 7. 31 -I > en ::i:: m ::z: ::r 0 (1) ,, (1) % ci ...... :>> Cb .j:>. r .,, > > ::z: en ::z: .0 c c l'lllQU!LL * ( fllOll AEC/t>SL TA T Clll UJll ll, SO-I I rnt> l'llSQUlLL B UftOlt A[CIDl'.l.TA t CAIT!AIA150*11 rtol > ..... r r l'AIGU!Lt. a c rnoit AtO/UOltA nn:TA C!ll Tl:ll!AI RA/IOI US!O) VINOS At 33 PUT llllOVI ORADt r ;o n sr:cto11 Ul'l'rll Cl.AU ltltr.RVAl.S 01" VlllD SPEED IHP!ll ltlWI UCTOI\ 'UM m IJl'M:ll Cl.ASS ltlTl:lll/111.S 0, VIND SPUD CltPlll n ;: 1 I 3 " a 6 1 II 9 10 II *I l TOT/II. SP!ID 1 g 3 " 5 6 1 II 9 10 II *II TOTAL SPEtt> 0 ;o en 0 en Hiil 0 0 0 0 0 0 0 0 0 0 0 0 0 I fltlr. 0 0 0 0 6 I 3 3 0, 0 0 0 13 Soll m t/) NI 0 0 0 0 l 0 0 0 0 0 0 0 I If[ 0 0 0 3 " d I 0 o' I II 0 13 5,9 .. llt!: 0 0 0 0 0 0 0 0 0 0 0 0 0 I DI! 0 0 0 I I 0 I 0 0 0 0 q 3 a.oo 0 II 0 0 0 0 0 0 0 0 0 0 0 0 0 I IC 0 0 I 0 0 ll " I 0 0 0 0 B 5.11 Ill 0 0 0 0 0 0 0 0 0 0 0 0 0 I [5[ 0 0 I 0 a I g 0 I. l 0 0 B 6001 ..... n: 0 0 0 0 0 0 0 0 0 0 0 0 0 I sr. 0 0 0 0 0 II I I I ,o 0 0 n 6090 ::z: Ill 0 0 0 0 0 0 0 0 0 0 0 0 0 I sst. 0 0 0 I 0 0 Q I 8 0 0 1 1. Ci) I I 0 0 0 0 0 0 0 0 0 0 0 0 0 I s 0 I I 0 0 0 0 0 0 I 0 0 3 4, '1 UV 0 0 0 I 0 0 0 0 0 0 0 0 I 3.90 ssv 0 ll 0 I g 0 0 0 0 0 0 0 ' )oAIO SV 0 0 0 0 0 0 0 0 0 0 0 0 0 I SV 0 0 l 0 0 0 0 0 0 0 I 0 e 6*95 0 VSV 0 0 0 0 0 0 0 0 0 0 0 0 0 I vsv 0 I 0 I 0 0 0 0 0 I 0 0 3 "*83 II 0 0 0 0 0 0 0 0 0 0 0 0 0 I v 0 g 0 I I
• I 0 0 0 0 I. 9 ,, 08 m V!IV 0 0 0 0 0 0 0 0 0 0 () 0 0 I VIII/ 0 () 0 g g a () I 0 0 0 7 Al*91 en 1111 0 0 0 0 0 0 0 0 () 0 0 0 0 I 0 0 I I I, R I " I I 6 ti 9,71 RO "'"' 0 0 0 0 0 0 0 0 0 0 0 I I 13.30 llllV 0 0 I 0 0, I 0 *
• I 8 9 80 10.91 II 0 0 0 0 0 I I I 0 0 0 0 3 H 0 0 0 3 R 7 7 3 R I 3 e 30 7ol0 3: 0 TOTN.. 0 0 0 I I I I I 0 0 0 I 6 4hft3 CAI.II 0 0 ;o m 101AI, 0 8 H u ae R-16 18 9 9 IT 157 7.01 c...l :z: UJ :::> w a:: U-C> :z: :;; </) UJ c...l :z: w a:: a:: :::> c...l c...l 0 C> a:: w co ::E: ::::> % c...l :z: w ::::> CT w a:: u.. f-a:: 0 ...... en .... w >Cl -' 0 -:i:: c...l °" w c... > OJ
• ftft** * ., .. <> .. .... ...... :;:; :: .. c s .. .. ! 0000000000000000 0 : 0000000000000000 0 0000000000000000 0 0000000000000000 0 oocooooooooooooo 0 000000000000-000 -000000000000-000 0-00000-000-0000 0-0000000---0000
• 000000000000-000 000000000000-000 0000000000000000 c !. ., .. oc .JJ ...... SS !! ii &.&. .. n 0000000000000000 0 0000000000000000 0 0000000000000000 0 000000000-000000 -0000000000000000 0 000000000-0-0000
• 0000000----*oooo 0000000000000000 0 0000000000000000 0 Figure 2.3-10, sheet 5 Revision 18 0: .. r,. o. : 0000000000000-00 ooOooo-ooooooaoo
• 00-00-000-00----OOOCIOOOOOOOO-ooo 0-,.,000000-000 --o'P-o-oooom-oo-o o---ooooooooooa-000-000000000000 H .. 000000-000000000 0000000-00000000 000000000000000-OOOQOOOOOOOQOOOO 0000000000000000 0000000000000000 000000000000-000 0000000000000-00 0--0000000-oOilOOO coaoooooooooooo-000--00000000000 coooooo--0000000 oooooooooooccooc 0 0 .. " THE DISTRIBUTION OF SPEED, DIRECTION, AND ()"8 PASQUILL CLASSES DURING L'l.T PASQUILL CLASS C CONDITION IN THE FINAL ANNUAL RECORD DAMES & MOORE

.... WIND FREQUENCY DISTRIBUTION (FREQUENCY IN NUMBER OF OCCURRENCES} CALVERT CLIFFS PLANT SlTE: STATION 2{IS) DATA PERIOD: NOVEMBER 1, 1969 THROUGH OCTOBER 31, 1970 -< (WITH 1971 DATA SUBSTITUTIONS) ::r: ,.,, c ..... (./) 01\0tf APllOIJ.TI\ T C"l1t1UA. run (Rev. 3/3/72) ;d PAIQU1Lt. C J>.UOlltU. 0 I '1'1011 AltCIDlll. TA f lllllfllll II> SO* II rtlU ;;; PAllGlllLL I 11ftOll tlllf.11 CRlfllll.ill llltlfOI l/J!ll) l'ASQ(JILL ' l'M>t AICCllla.tA tlllTA ClllTtlllAI IWIOI UJll>I c:: HCTO!I llPPtll Ct.AU llfft!lllA!.S or VIN!) SPUD (r!P!fl ttlNI !ECTOR Ul'PlA Cl.AU lllJl!lllM..9 OP 111111> Sl'HD flll'ff> !ft.All "O -< > 0 1 I 3 s

• T 8 9 10 II *II TOTA!. SPUD I * ' * '
• 1 II ' ID II *II TOTAL !Pl!D "' .0 z lfU 0 0 0 *
• I I I 0 I I 0 9 4, 30 lllft 0 0 0
• I 3 0 0 0 0 0 I T S*H c: .... 0 lllC 0 0 0 0 3 0 0 0 0 0 I 0 * &*OB 11g 0 0 0 0 0 0 0 0 0 I I 0
• 10.0 r "T1 1:1111: 0 I 0 I I 0 0 I 0 0 0 0 ' 4*RO "11 0 0 0 0 0 0 0 0 0 0 0 0 0 I r t 0 0 I 0 0 ' 0 0 0 0 0 0 a 3,55 I 0 0 0 0 0 0 0 0 0 0 0 .. 0 0 I ("') ..... lt.U 0 0 0 0 0 I II 0 0 0 0 0 3 d*GO tSI 0 0 () 0 0 I 0 0 0 0 0 ". 0. I 9*10 :z: 31 H 0 0 0 0 0 0 0 0 I I 0 0 3 t.10 u 0 I) 0 0 0 0 0 0 0 0 0 0 0 I c "' (,/') qg ISi 0 Q 0 Q 0 Q 0 I 0 I
• 0 ' t.u SS! 0 0 0 0 0 0 0 0 0 0 0 0 0 I (,/') "O
• I 0 0 0 0 0 I 0 I 0 0 0 3 s.u s 0 0 0 0 0 0 0 0 0 0 0 0 0 I C""l ,.,, Ul UV 0 0 0 0 0 0 I 0 0 0 0 0 I d*SO SSll 0 0 0 0 0 0 0 0 0 0 0 0 0 r l'T1 -* (l) sv Q 0 0 0 0 0 0 0 0 0 0 0 0 I S'I 0 0 0 I 0 0 0 0 0 0 0 0 I 3*10 n t:l 0 . 0 IV 11811 0 0 0 0 0 0 0 0 0 0 0 0 0 r 11511 0 0 0 0 0 0 0 0 0 0 0 0 0 I :;i: 0 :::l
• II 0 0 0 0 0 0 0 0 0 0 0 0 0 I II 0 0 0 0 0 0 0 0 0 0 0 0 0 I 0 w Vltll 0 0 0 I 0 0 0 0 0 0 0 0 I 4,00 11!111 0 0 I 0 0 0 0 0 0 0 0 0 I **tO .... ;.; -I ...... I 1111 0 0 0 I 0 I 0 I
• 4 0 1 U IO*RI 1111 0 0 0 0 0 0 I 0 0 0 0 0 I d*30 0 ,.,, 00 ...... lflfV I 0 0 I I I I 0 I 0 4 e 16 10*114 H>lll 0 0 0 0 I 0 0 0 0 I I t ' ..... n If 0 0 0 0 0 0 I I I I 0 I s 9*411 II 0 0 0 I ,, I 0 0
• 0 , 4 14 , .. , :z: -I .... 8 Ul TOTAi. II I I 1 1 s 7 s 8 9 II ,. 13 e.as 101AI. 0 0 I 4 II 9 I 0 t
• 5 7 31 s.41 :z: .:z: ;::l"' -< (l) ::r:: > m :z: 0 0\ -n <l>q r -o ,,,. r;; 2: :z: .0 c:: c:: ;p. p PA5-CIU1Lt. C CFllO!f llti:ll!>lt.11\ f CA1tt11iAo &0*11 'Ull MSQIJILL C I Pllll!f AIUJ'Dll. 111 t CRITtlll A> SO* II '1.) ,.... ,.... "'!IQU?LL a C '110>4 Ati:llll <!1111 CAltll\1111 IWIOt USP.DI IJ1"DS Af 33 '21!1 A!IOVI ORADI! ;c ,.,, C""l 111'1'1:11 IHUllVALS Of' I/IHI> SPUD lr!P!I> lft.llf St>::tOA ll!'Ptf\ a.l\SS 11111:1\VN..S 0,. llJHI> SPUO llfl>lll llL'llf " ,... SIP:CTOI? 0 > I I 3 4 S 6 1 S 9 10 II *II TOTA!. SPUO I e 3 .. II e 7 II 9 10 II *I I tor111. enrn "" V> C> V> 0 0 0 0 0 0 0 I 0 I
• 11.10 Hiit 0 0 0 .. 5 s I I
• n 8*'19 m Mlf I 0 0 I I ' V> Hit 0 a 0 0 0 Q 0 0 0 0 0 0 0 I Ht 0 0 0 e II I 0 0 0 I
• 0 II 8*04 Cl l'.>llt 0 a 0 a 0 0 0 0 0 0 0 0 0 I mt 0 I 0 6 I II 0 r *O 0 0 0 II 4dS I! 0 0 0 0 I 0 0 0 0 0 0 0 I **30 t 0 0 3 I ,, I II
• I 0 0 0 ,,, ... ,, ltU 0 0 0 0 0 0 0 0 0 0 0 0 0 I rsr 0 0 I 0 I 3 3 0 0 0 0 0 e 9,119 = st 0 0 0 0 0 I 0 0 0 0 0 0 I IE 0 0 0 0 0 .. I 0 I
• 0 0 " 7.J I m ' au 0 0 0 0 0 0 0 0 0 0 Q 0 0 I sn 0 I 0 I 0 3 0 I 0 * .. 0 .. ,, .,
• 0 0 0 0 0 0 0 0 0 0 0 0 0 I 5 I I 8 0 * .. I I ' 0 I 0 Id s.03 UV 0 0 0 0 0 0 0 0 0 0 0 0 0 I SSll 0 I 0 I
• 0 0 0 0 0 0 0 " 3,3y IV 0 0 0 0 0 0 0 0 0 0 0 0 0 I Sii 0 0 0 3
• I I Q I 0 0 *o s l*O* "'" 0 0 0 0 0 0 0 0 0 0 0 0 0 I vsv 0 I
• I
• 0 0 0 0 0 0 Q e 3,95 v 0 0 0 0 0 0 0 0 0 0 0 0 0 I v 0 0 I 9 ,A I I I lo 0 0 0 II ... o ,.,, llHV 0 0 0 0 0 0 0 0 0 0 0 0 0 I VllV 0 0 0 0 3 6 Q I 0 I 0 I ... f,H V> 0 0 0 0 0 0 0 0 0 0 0 0 0 I Hll 0 0 0 I 'I I .. I ' ' I 14 31 ,.o 12" Hlf V 0 0 0 0 0 0 a 0 0 0 0 0 0 I llHV I 0 0 :> 3 I I 0 II I 5 ' u 10.01 If 0 0 0 0 0 a 0 0 0 0 I I 4 I) tRO H 0 0 0 J 3 1 I I ' I 4 ,, t9 **01 :i:: 0 101111. 0 0 0 0 I 3 0 0 0 I I I ,, d.:,. CN.H 0 0 "" m
• 9 31 41 37 19 10 IS u 19 39 139 1.01 N .-.... ...... M 0 ;;; "' ...... 1: :;; M ..; .,., C' ..... OOCll'P *4'nf'>""'°.,f') r., ... > _, 0 QJ .. -0000000-00000-00 -: ; 000000000000-000 2; 0 000000000000--00 0 "' c c i-"' ... iE ::: .. !l r -* 0000000-0000-000 ::> -* .. ;: .. s "' -"' ... C> "'"' "' .. ,. .. .. "Z .. .... 0000-00-00000000 ... :a A.tO 0: m-n -o--_, n .. .. ., <r . .. "' 0 -; <
• c. :z ooooooo*ooooGooo o-! ... .. .,_ _,.. ".: 15 > """ .. ... . "' .., .. ... .... ... eB 0 .. it c: C'., :;; WU .. .. ;1 -c _ .. O:>-"' .... "' .... ... <>"' ., .. .. i5 .. i5 0 .. cc ? cc ..... ..... ... --0000*-wo--o-oo .... ... "' !::l : d:: ::l .. .... < < ...... cl .. 6., ., ...... s ii cc 0: "' M cc .. (/) .. "' .., .. .. c:= l; .. 0000000**-oaoo-o 00 o--o--ooow-00-0-U) w :: f "'"' l..Ll N cc ...... u 0 (/) :z: :z: :z: I-:z: 000000-000000000 ooopoooccooo-ooo l..Ll 0 u 3 0111 oc 0 c:= ;:: 0 I-c:= I-_, _, ,_, _, ::> < ::c: ::> _, .... _, _, ::::> u I-<.!> ::: 0: _, "' cc V> ::::> 0 0 u 0 I-.... .. .. .. ,. > .. E .. .. .. .. .. .. .. > c:= 0 c:= (./) << .... I? I-..... ::c: co ...... ..... V> 0 LL.I I-::> Q :::: (/) ""' en >-LL.I V> CD c:: co en I-u ::>: I-...... c:: :z: ::> :z: 0 " OJ 0 ... w :z < n "' ,, .. ... :::> _, ...... .:.; i.H'. ..'; c:::r ;::; c.. ...... .... .., l..Ll c:= CJ) ..... .. ..., c:= Cl) l..Ll ...... --oonn**-"'*10,...,-o .. ... LL. ,,_ LL. cc .. --_., --c:cs::-., Cl u LL. :i:: ::c: :z: w !:; 0 :z: w _, > .. 3 :::> u 0 CY :z: 0000000-00000000 l..Ll I--: ; ""' ""' LL. LL.I > 0 000000000000-000 _, 0 < u c:= UJ 0 0000000000000000 0 .. a... Q i-a l-< .. I-x 0 .. r "" :> _ .. 0000000000000000 :> _ .. 0 .... fl :: .. e ... , ..., .. ., .. :! .. .. :2 :;"' cooooooooo-0-000 l:i"'
• 0 ':"C 0 -; c ! .. "' s-s:-_ .. ... .. * :i .. .. .. c>-.. 0 ., ! o.., .. .. -c -c a: ... :> o; ... > ., .. .. uw fi .. :: .. j: ..i: % c< * .... .. * .. d::! ,:_ .. .. a .. .. ,, .. c d .. ....... _, ;; lHl ., .. cc s cc :! ... ,, .. ., ,, .. 00 !; .. ! 00 !; .. : a: a; o;o; .... ..... 0000000000-moooo ... , 00000-0--0000000 CIC ,_,_, ... _, ...... "' _,_, a: SS 0 Ml M W IW :. ' '> ! ;:; 0 .. .. "' .. _, :: .. !II '> :.. cc 0 ..... &. ... .. .. ... ... Figure 2.3-10, sheet 7 Revision 18 THE DISTRIBUTION OF 'JIND SPEED, DIRECTION, AND O"o PASQUILL CLASSES DURING M PASQUILL CLASS 0 CONDITION IN THE FINAL ANNUAL RECORD DAMES & MOORE

z: 0 ,_ :::> co c:: I-"' 0 u :z: w :::> = w c:: l.r.... 0 :z: :; "' w (_) :z: w c:: c:: ::::> (_) (_) 0 l.r.... 0 c:: w = ::::> :z: u = w ::::> = w a:: l.r.... l-a:'. w ::>Cl u a:'. w 0.. > QJ !::. : C*** c : ! 0 § f-;: .. .. ., .. ., """' .. "'

• c o-"'"' ... c .. :i;:: ..... .. i!' n c:a: ..... ..... _, _, _, _, SS 00 .. :: 2 0. 0. "' ""' _, .J .J.J 55 :: i n-o---oooooooon-00000-0-oooooooQ oooooomooooooooo oooooooooooooooQ o " ................ 000000--0000-00-0-00000-00000000 w..,1.:1.,>>>=-... .. " Figure 2.3-10, sheet 8 Revision 18 .. ... : . c E .. iC " .. .. <[C .. do Q ...... '"' a :I: c ., .. a:w .. .. _ .. ! ... .. ., "' d.., fi 0. .... ::> cMOO 0 ... o-"-o-m 0 ID ---cw-0-CI 0 4'--t-e> S) _,O ---aa-------ca ea--------* ------0 " Qn _, .. "'< s .. oQ tn:Z <<-0. .. a: 0 g a c c ., ., .> .J _, .J ;>S ........ :; --ooooooooooco----00000000000000 = -noocooooooooo---0000000000000--n ooocoooooo-coooc 0000000000000000 c ooococoooooooooo THE DISTRIBUTION OF WIND SPEED, DIRECTION, AND o-8 PASQUILL CLASSES DURING fi.T PASQUILL CLASS D CONDITION IN THE FINAL ANNUAL RECORD ., ... .. .. 0 ., "' ... 0 .. ., DAHES & MOORE

z: a ,__ => c:i c:: 1-V'l ;::; u :z: w => er w c:: u.. 0 :z: :; V> w u = w "" "" => w u 0 '-0 u :z: w => = w "" u.. c:: UJ N c:J 0 :z: l-o u 0 I-C::: ::c I-<.!J (.I) => 0 c:: :i:: UJ ,__ ::::; (/) V> :z: C) ..... => I-(I) c:i => (I) ...... ..... c:: CJ) (.I) w ...... LL. c:i LL. ::i;: w _, > u 0 :z: C:: UJ > b _, 0 c:::: u c:: w Cl.. c:::: C:::: Cl ::c ..... :; > QJ 0 !E 'IC .. : 2 0000000-00000-00 a = 00000000000-00-0 0 000000000000-000 § f-.. .., :> -* .... Ei *

• c -:i c .. .. -c 0: .. ., .. .. ?; .... " .. .. :; .. ooocoo**-000--00 -.0000-00--oncoooo m n 2 000000000000-000 00000000-0000000 o n : i-= n 0000000--co---oo °' 000000-mo-oooooo
• Figure 2.3-10, sheet 9 Revision 18 "--o.4'.,_.n,..mne1C',... Clll hO-.,Oe-*"'°"" on.o.., °" o. -s.n.e.nw .rn.;.: .:...; 2 .. .. "' ... ... .., -* ... -n.,-oo-oei:i .-i. n
• oco ... 2n--o-oooowoa-00000 2 00000--oooooc-o--f"')OOOCCll QO oooq __ -n "<:cunoo-o o o oo ca 0i1 o er -C"') 0 Cil OOI l?'OI 0 o-w in o-Cf". C") .... o--'° o--0 -aoo-ooat"J-00--000 Q N:r:: > ':!: 0 "' 0 "' .. ... "' .., .. .. .. ., ... ... .., " : THE DISTRIBUTION OF WIND SPEED, DIRECTION, AND o-8 PASQUILL CLASSES DURING b.T PASQUILL CLASS E CONDITION IN THE FINAL ANNUAL RECORD DAMES & MOORE WIND FREQUENCY DISTRIBUTION (FREQUENCY IN NUMOER OF OCCURRENCES) CALVERT CLIFFS PLANT SITE: STATION 2(IS) DATA PERIOD: NOVEMBER 1, 1969 THROUGH OCTOBER 31, 1970 ;i (WITH 1971 DATA SUBSTITUTIONS) f"T1 0 PllSQUILL I U110!4 AIC/DII.Tll t CRITIRIA.50*1Q run PASOUILL I ""°" ll!C/DII.tll t CRlttlllll.50*18 PllTI (Rev. 3/3/72) [:> PllSQUILL .I! 1'110H lltC/SIO:iA TKITll CRITVl!AI U5t01 PllSOUILL F HRO" AF.WSlo:fll TI!ETll CRITVllM IWIO! US!OI -I S!CtOR IJPPtll CLUI IHTtAllALJ or VIND SPIED 1!1Plll Ill.All sr.ctOR UPPER ct.ASS IH1ERVALI 01 VIND SPUO IHPHI . :1tM g I I 3 4 9 6 1 8 9 I 0 II >II TOTAL SPUD I g 3 " S 6 1 9 9 I 0 11 *II TOTAL SPUD 0 :z: NII! a 7 9 e 9 3 1 I 0 6 6 3 8 to 6* Niii: 0 0 0 I I ' I ' R 3 3 3 u 7,9 6 c::: 0 II! I 0 I 6 I 7 a e a 3 0 4 ll9 6*56 NE 0 0 I e 4 0 .. I " 0 0 0 16 5,94 r "Tl r>n: 0 0 5 1 3 I I 9 0 0 I 5 25 6ol9 l?lE 0 0 I 3 0 0 0 0 0 0 0 0 4 3,eg r . g 3 O 5 ft 3 3 3 I I 0 0 PD 97 4, 04 E I 0 I I 0 I 0 0 0 0 0 0 .t 3* 10 n 1tn: o a I T 3 6 3 " 1 n o o 119 39 nr. o o o o e o o o o o o o 1 4, ao r :z: n o a a s a 3 " 1 3 a 1 o es s. 6D SE o e o o 1 o o o o 1 o o 4 ... eo "" 0 ,..,... >--rj SSI 0 a " 9 I 0 13 II u 6 ' 4 ' u 6* ST Sst 0 I I 9 0 I I I I 0 0 0 9 5, 13 vi I"' ....,, S I .t 3 6 6 I 9 3 I 3 0 R 3' 5086 S O O I R 9 0 0 0 0 0 0 0 9 3,79 " fJQ HV I 3 I 3 I 0 0 O 0 0 0 O II R*86 SSU I O 0 0 0 0 0 0 0 0 I 0 R 9,45 rri Fil ;::: . $::= 1\1 I 0 I I I 0 0 0 0 0 0 0 6 a. 65 sv I 0 0 I 0 0 0 0 0 0 0 0 a ** 10 n 0 ti> VSV 0 I I I 0 0 0 0 0 0 0 0 .. e. 47 11511 0 0 0 0 I 0 0 0 0 0 0 0 I 4, 60 0
• o' '" II I 0 I R 4 0 I 0 I 0 0 0 10 4,15 II I I 0 0 0 0 0 0 0 0 O 0 8 loOO S o ::3 N VNV 0 0 I 9 1 3 3 8 I 0 0 3 ll9 6°03 111111 0 0 8 I 0 I 0 0 0 0 0 0 4 3*80 ,_, ..... w NII 0 I a I 9 11 30 05 18 II 16 98 160 9,35 NV 0 0 0 R 0 3 a 3 0 I 3 I 16 7*88 ;j -I NllV 0 Q 0 R a 10 1 6 IS ta 17 19 99 9.19 I I 0 R 3 6 6 e g 7 7 84 64 9,95 o n 00 ,_. N o 3 3 10 1 4 s 7 6 2 7 13 70 7, 69 N o I 3 6 9 s s e "' 3 9 20 67 9,41 :z: ;j 0 ,.... o u TOTAL II 28 38 86 65 71 85 81 61 *9 ft7 708 6099 T07AL 9 6 10 U 33 H BO U 16 15 R3 O Ret 7,91 :z: :z: ti> . ::;--I l> ('f) :z: ('f) 0 .... .,, -q,q 0 r " )> "" :z l/) :z: 0 §;; S PASQUILI. E IPRO!f Al:C/Dll.TA T CRITD\fA,BO*IR FUTI PASQUILL E (1ROH AEC,DII.Tll t 50*18 n.1 r F= PASOUILL a 1 FRO!f AEC/SIG!fll TllETll CAI TERIAI RAllOE USIDI lllHDS AT 33 TUI llOOVE 011110£ ;;:: n Sl:CTOR UPPVI Cl.ASS INTDIVAl.S or VIND SPUD IHPlll 'IF.llN StCtOR VPPEA Cl.ASS IHTEnVAl.S or I/IND SPUD IHPHI HEAN n ; I
• 3 ... ' 6 1 e 9 10 II >II TOTAL SPl!:ED I e 3 "' 5 6 1 8 9. 10 II *II TOTAi. SPUD !NE 0 0 0 I I I a e 4 9 I 3 17 9,31 1111! " 6 9 13 10 IR 16 90 Hi 16
• II 14 153 7o09 (;l NI: 0 0 0 I 0 I 0 I I 0 0 0 4 6049 HE 3 I 4 14 7' 11 10 12 11 IS 3 4 104 6*84 !lilt 0 0 I 0 0 0 0 0 0 0 0 0 I g,30 IN! I 0 13 14 18 I e 7 6 I e 1 19 5,55 O I! 0 0 0 0 0 0 0 0 0 O 0 0 0 I E 4 ft IO HO II II 11! 6 O
• O I O 91 4* !18 Ill 0 0 0 0 0 0 0 0 0 0 0 0 0 I ESE I 9 1 IS 17 15 11 e R I O O 67 4, 79 ,.... 1 5! 0 0 0 0 0 0 0 0 0 0 0 0 0 I SE 0 7 II 16 10 16 II 3 7 I 3 I O 85 4,95 il'i 0 0 0 0 I 0 0 0 0 0 0 0 I 5.oo SS! e II !18 65 51 35 34 38 IT 9 13 16 30 5,45 S 0 0 O 0 0 O O 0 O O O O 0 I S 9 30 70 70 79 43 26 13 3 6 3 1 361 ** 30 SSV 0 0 0 0 0 I 0 0 0 0 0 0 I 5,70 SSll 3 35 50 40 34 86 1 5 0 O I O 906 3, 61 sv o o 1 o o o o o o o o o I e. 10 sv 4 es 31 P.o et "' 1 e a o o o 110 3,94 0 11511 I 0 I 0 0 0 0 0 0 0 0 0 a ** so VSI/ ... 18 R3 I 6 17 5 . I 0 e 0 I 0. 87 3, eg :i:o v o o o o o o o o o o o o o r 11 s 19 3n es 01 9 "' 6 e. o 1 o 131 3, s1 VIII/ 0 0 0 0 0 0 O 0 o U 0 0 0 I WllU 9 6 16 40 33* H 11 II 9J 10 5 1 181 9,54 Vl NII 0 0 0 0 I 0 0 0 I 0 0 0 e 60 s 5 NII 0 3 14 17 50 !18 47 35 96 39 sa 314 7, 90 R" Hill 0 0 0 0 I g 3 R 4 I I 11 es 10.e3 NNll I 9 0 u 11' 88 ea 20 29 83 32 59 851 5,69 ti 0 0 a g I I 5 I J J 31 9, 55 " 0 7 10 a3 19 II 19 19 205 7o9R :i:: 8 TOIAL I 0 5 & 6 10 6 IJ 6 6 R3 85 9*01 C/\l.11 ;o rri TOTAL 47 196 36R 406 957 P06 17" IP.6 130 R04 5, 11

z: 0 => CD a::: I-"' 0 u :z: w ::::> O' w a::: u.. 0 :z: :.:; u :z: LL! ::::> O' w a::: u.. 0 ..... Cl'l ...... ...... <") U) :z: 0 => !:: V> CD ::::> V> ...... ..... a::: Cl'l .--4 u.. :L LL! -' > u 0 :z: a::: LLJ > 0 _, 0 <I: u a::: w n.. <I: ei: 0 e -0000000000000000 -0 Q oooooooooocooooo f-o ::.w ..:o .... ,. .... :.: ... .,_ .,., ... c .. -c a: .. uw :c .... cc .. :c i3 """ ... ... uu ..... cc lC :c oo "'a: ... ... ., ..>.,, _, _, = oooooooooooooooc 0 0 -oooooooonn-ooooc ... i g e -COOOOOOO'OOOooooo 0 -0 o Ooooooooooocooo:> o i-H ..... .. " A < _, ..... A.a ::> OOOOOCIOOOOOOOC;),:> -: ... < _, _, _, _, -00000-00-00000-0 :;:; :: (,) << ... ...... .. .. ., _, _, _, _, E :: ... ..... Figure 2.3-10, sheet 11 Revision 18 ooJoo-oc?o:::>:l!cn.:-:: c.oooco-oc:c.oe = =>.: ::.c cpo o o: o: ;-C\1.,'-ooo:iiooo-:.eo:>-Joo ODOOO:>c.-c*oo<<--c o-o 0 O =" """J--':tO<Q -.,,; <<C:. oob o o ... (\. c OC".tCOOC-.i::'!"<G"-c:;,.:.f'l-'0-=>---.=i ..... ,..X..,.\!"CC""'t\ oc: O ;:i * --ra---c ;: 000 0 .;>-< !:"" :0.M ... (\. ... ;;: THE DISTRIBUTION OF WIND SPEED, DIRECTION, AND CJ8 PASQUILL CLASSES DURING fi.T PASQUILL CLASS F CONDITION IN THE FINAL ANNUAL RECORD DAMES & MOORE WINO FREQUENCY DISTRIBUTION (FREQUENCY IN NUMBER OF OCCURRENCES} CALVERT CLIFFS PLANT SITE: STATION 2(IS) DATA PERIOD: NOVEMBER l, 1969 THROUGH OCTOBER 31, 1970 -I (WITH 1971 DATA SUBSTITUTIONS) ffl 0 ..... (/) (Rev. 3/3/72) -I PUQUILI. r (,ROH AtCID[LfA r CRITtRIAo50*18 rtUI r lrHOK AtCIDEl.tll r CRIU.Hll\t50*1e fUfl ;xi .... PllSQUILL I': (Fl\OK AECISIOHll THUii cnltERll\1 RllllOE UbF.01 PASQUILL r ( rROH /l!CISIOKA ?HUA CHI ltl\llOE Ubf:tol CD c ncroK UPPEll Cl.ASS ltlTERVAl.S or VltfD SPF.ED CHl'IO HEAii Sl:CtOH UPPER CLASS IHTr.llV/11.5 or lll!ID SPttO (!IPlll 11£1\1( '"O -i :I> 0 I g 3 4 5 6 1 8 9 10 ll *II TOTAi. SPrED I g 3 4 5 6 1 8 9 10 II *II TOTAi. SPEED VI .0 :z: NHt I I 0 0 8 I 0 I 0 I 0 0 1 4,71 NNE 0 0 0 0 0 0 0 0 I 0 0 0 I 11*30 c .... 0 NE I 0 g 0 0 a 0 D 0 0 0 0 5 3, 34 Ht 0 I 0 0 0 0 0 0 0 0 0 0 I lo60 r-.,, !NE I I 0 0 0 0 0 0 0 0 0 0 a *80 ENE I 0 0 0 0 0 0 0 0 0 0 0 I 1*00 r-IC I 0 II 3 0 0 0 0 0 0 0 0 0 ' l!.18 E 0 I 0 I 0 0 0 0 0 0 D 0 e 2*55 ('") ..... !51. 0 0 I 0 0 0 0 0 0 0 0 0 I e.io ESE 0 0 0 0 0 0 0 0 0 I 0 0 I 9*10 r-:z: SE 0 I I 0 0 0 0 0 0 0 0 u e 1.75 st 0 0 0 0 0 o* 0 0 0 0 0 0 0 I :I> 0 VI Sst 0 II 1 e 4 e 0 0 0 0 0 u 17 3, 49 SSP; 8 0 0 0 0 0 0 u 0 0 0 0 g *95 VI t/) s 0 10 5 a I 0 0 0 0 0 I 0 IY 0*71 s I 0 I 0 0 0 0 0 0 0 0 0 e 1*70 '"CJ .,, r<l ISi/ I 4 Q 0 0 0 Q u 0 0 0 0 9 11.69 5511 I I 0 0 0 0 0 0 0 0 0 0 ' .eo r<l SV I 9 a 0 0 0 0 0 0 0 0 0 5 t.eo Sii I a 0 0 0 0 0 u 0 0 0 0 3 lolO ('") 0 11511 3 I I I 0 0 0 0 0 0 0 0 6 lo6Q VSll I 0 0 0 0 0 0 0 0 0 0 0 I *80 0 . :z II 4 9 0 a 0 I I 0 0 0 0 0 10 2*46 v 0 0 .o 0 0 I 0 0 0 0 0 0 I 6*00 0 0 VIII/ 0 I a G I I 0 0 I 0 0 18 VNll 0 I I 0 0 0 u 0 0 0 0 0 a 8*15 .... en .., :; ;xi -* (t> NV I II 0 A 1 7 1 3 I e 0 4 36 6*44 NII 0 I 0 0 0 I 0 0 0 I 0 0 3 9*17 0 rn g !:V NNll 0 I I 0 a I 0 I 0 e Q I II 7*04 lllfll 0 u 0 0 0 0 0 0 I 0 I 0 8 9*50 :z H I 3 0 I! 0 0 0 e 0 0 I 0 9 .. 2 II 0 0 I 0 0 0 I I I I 0 0 5 6*74 .... w .... 0 ..... I TOTAL 14 33 27 19 QO 15 II 7 I 6

• 162 .. I TOTAi. 1 7 3 I 0 D I I 3 3 I 0 0 3,99 :z 00 ....... -i ,,. ::i: rn :z: en 0 .,, ::r' (bq (t> (t> ....... r--0 ..... :I> ,,. N :z (/) :z a c c :I> .... PASQUILL r C tHOrt AtCIUtl.tA T CRlr£RIAt50* I 8 Fr.ET! PASOUll.t. P /lt:CIDD.TA T cnl tERIA. 50*IR "" r-r r-PASQUll.t. 0 crnoH Ar!C/510HA TlltTI\ CHITF.Al/11 RtvlOt VINOS AT 33 PEET Al!OVE ORADt ;;<;] .. rn (") SECT OH UPPEll CLASS o* VINU Hu:o (HrlO sr.cton urrr.n CLASS llltr.nv111.s OP VHID ftPKF.D it1m1 KtJ\11 n r-,,. I a 3 4 5 6 7 e 9 to II *II IOTA!. Bvrr.o I R 3 4 ft 6 7 8 9 10 II *II TOTl\I. SPUD (/) 0 V> 0 0 0 \ \ 0 0 0 0 0 \) 0 e h20 HHE 3 II I I 3 I 0 I l I 0 0 3o96 r<l (/) 0 0 0 0 0 0 0 0 0 0 0 0 0 I Ill I I 3 4 0 3 0 0 0 0 0 0 ID 3,37 0 ENE 0 0 0 0 0 0 0 0 0 0 0 0 0 I !llF. a I 0 a 0 0 I 0 0 0 0 0 6 a. 65 c E 0 0 0 0 0 0 0 0 0 0 " 0 0 I E I 3 3 I 0 0 0 0 0 0 0 0 8 fo06 ES! I 0 0 0 0 0 0 0 0 0 u u I *RO E't. I a a 0 I 0 0 0 0 3 0 0 9 "'* 44 :z ,, ,,,, 0 0 0 0 0 0 0 0 0 I) 0 0 0 I St I e 4 n I 0 0 0 0 I I D I I* ... ,. m 'IH I 0 0 0 0 0 0 0 0 0 0 0 I **O BU
• 17 Q7 II 6 ' Q Q Q I 0 I 7Q 3o03 s 8 0 0 0 0 0 0 0 0 'll 0 0 e *30 5 ' 41 "9 17 DD 3 6 a 0 B 0 0 153 p,97 I 0 0 0 0 Q 0 u u 0 0 0 I **O SSll II 39 ee 9 ts I a 0 0 0 0 0 99 g, 39 I 0 I 0 0 u 0 0 CJ 0 0 0 a I *SD Sii II a* a3 a e 0 0 I 0 0 0 0 63 g.oJ 0 vsv 0 0 0 0 0 u u 0 0 0 0 u 0 I 11511 6 e 9 3 3 I 0 I 0 0 0 0 31 2** > v 0 (I u 0 0 Q 0 0 0 u 0 u 0 I II 3 P.O 23 16 3 !I 3 0 0 0 0 0 73 B*89 3: rn nv 0 0 u 0 0 0 0 0 u 0 0 0 0 I VllV a 1 BO t:r ' 3 I 11 I 0 0 re 3. 78 Vl NV 0 0 (J I I u u 0 l' 0 0 u a ... oo tlV g 3 4 5 18 15 8 3 4 3 e T 74 6*06 RO 0 0 0 (I I 0 0 u 0 0 0 0 I 5*00 HllV 3 3 e I 7
• 0 e I D 3 I 99 s.2e N 0 0 0 0 0 I q Cl ., (I Cl I 4 ll*Oq H I " 3 I I 3 3 3 I I I R6 s.te 3: 8 0 I v 3 I A 0 II ,, 0 I 3, ::0 I rri totfli. se 183 191 97 99 .. 26 14 10 15 9 II 156 ,, 40 0 l"T1 U1 12" 3: 0 0 ;o l"T1 -i :c l"T1 tn [> g -i en c:: -i .,, )> tn '° ;:; r r n r )> U1 U1 en 0 :z 0 .,, ::z: 0 tn .,, ,.,, l"T1 CJ n 0 :z :::: -i ;o ..... ,.,, o n :z ::::! ..... :z -i :c l"T1 .,, 0 :z )> ::z: 0 r .,, )> )> :z VI :z '° c:: c:: )> ..... r r r ;o m n n r 0 )> ;o U1 0 tn ,.,, tn 0 c:: ;o ..... ::z: en >-:r'j cfCi" -* c:: Cll ....., Ci' (]) ::s w ,._. I 00 ...... i:n g' (]) ...... ...... w WIND FREQUENCY DISTRIBUTION (FREQUENCY IN NUMBER OF OCCURRENCES) CALVERT CLIFFS PLANT SITE: STATION 2(IS) DATA PERIOD: NOVEMBER l, 1969 THROUGH OCTOBER 31, 1970 (WITH 1971 DATA SUBSTITUTIONS) Pll!Q(JILL a <rROH AECID!LTll T CR1TE11h.so-1e FUT> Pll!QIJILL A HllOH AECl&IO"ll TllU" C"lttllllll usrn> UC TOR or VINO SPEED D 3
• 5 6 T 8 9 10 NNI: l EU " Ut s h nv v H TOTAL 0 0 0 0 0 0 0 I I 3 0 I 0 0 0 0 0 0 0 0 0 0 I 4 6 5 ' 6 a 0 I 0 I 0 0 0 0 0 0
• 8 3 3 '
• 0 0 0 0 0 0 0 0 0 0
• 3 I 0 I I 0 0 I 39 U II 0 0 0 I 0 0 0 I 0 0 I 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 I 0 0 0 0 0 0 PASQUILL Q <rROH "EC/Ott.TA T CRlt!HIAo50*1ll rttll PASQUILL C I rROff '-!:CIUGHI\ Tlllf" CRltEttlAI UCTOll UPPER Cl.ASS INUllVlll.S 01" SPttlJ IHl'lll 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 R 3 4 5 6 7 H 9 10 NME NE vu: E SF. ,$U UV LV v v 0 0 0 0 u 0 0
• D I I 0 0 (J 0 o I 0 0 0 0 e o 0 0 I 11 5 I r.3 tn 99 ri IS ! 12 , 7 IR 4 o e U I o n 0 0 I a 0 u I 6 J I (J 2 I u 0 a a I) 0 a I I 0 0 II I 3 (1 I 96 5* " In 0 u 0 0 I) 0 0 0 ,, IJ (I ,, ,, () 0 0 0 0 0 0 II (J 0 0 0 ti 0 0 I 0 0 0 0 0 0 0 0 0 0 0 0 u 0 u 0 u 0 0 u 0 ti 0 0 0 u I) ,, II I 0 0 0 0 0 0 0 0 0 I) 0 0 ll II 0 II u II 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 *II 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 HEii/i TOTAi. SPrtO a.eo I o I I 0 I o I I 1"90 I* D*60 1e a.ea 13 g. 39 11 Do 15 13 o.oa 1 R*61 0 I I I* 10 I 3*RO 1' A* 35 HEllH II *II TOTl\l. srrr.o o o I 2*70 0 0 O I o o 1 3.au II O 4 1*72 O O O I 0 o I l*SO o o e g, o o e. U U o o 23 1. ee o o ao 1.ea u e6 0 l) 2. 0 0 9 Q, OJ 0 0 3 -* Q3 ti ri 3 3o IO 2v"t eo P"SQl/11.1. a P"IQUILL ncro" I Ht 0 rnr. o r. 0 r.sr. 0 SE 0 ssr. o s 0 SSll 5 sv 3 VSll I v Q Vl/11 I Hit O* Ill/II 0 fl 0 TOTA!. 13 PASQUILL 0 PASQUILL 0 11111: E tb! ST. 5Sr. b bit WSV v wHq NW 1mu 0 I 0 0 0 0 A 3 A 3 3 I I 0 0 IFllOH T fEUI lrR011 IHtTll CRITFlll"' RA'IOE US!Dl urrr.n CLASS INTERVALS or VIND SPEEO <HPlll e 3
• s 4 1 e 9 10 0 0 0 0 0 I R 10 " 5 10 T " 0 0 0 I b I 0 0 0 0
• e g I 5 s 0 0 I 47 ee 0 () I 0 0 0 0 0 a 0 I I I 0 0 0 b 0 0 0 0 0 0 3 I 0 0 0 0 I I 0 0 0 0 0 u 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 u 0 cl 0 0 0 0 D 0 0 0 0 0 0 0 0 0 0 u IFROH AECIDtl.TA T CRITElllAo50*12 Fr.!TI 0 0 0 cl 0 0 0 0 0 0 0 0 0 0 0 0 C1 0 0 0 0 0 D 0 0 0 0 0 0 0 0 0 0 0 IFHOH AEC/SIOKA TKETI\ CRtn'1llC\I USEDI UPPrn Cl.l\fi.S INn:.1VAl.S, 0, WHID sr-u::o *HPH) 0 3 -5 6 , 6 y. 10 o n I I 0 0 0 3 I A I 0 I I o I O O -3 3 93 35 II RI 3 J7 I Y 0 10 I 14 13 I 3 I O 0 0 3 0 0 0 0 I I 0 0 (I 0 .. 0 (). 0 0 0 3 I 0 0 I 0 I 0 ll 0 0 I 0 0 I a 0 I 0 0 0 0 0 0 I) 0 0 0 0 0 I 0 0 I 0 0 0 0 0 0 (I u I 0 0 0 0 0 0 I 0 0 I 0 0 0 0 0 0 0 " 0 fl I) (I n o 0 0 0 0 0 u u 1) 0 0 " u 0 0 0 0 ror111. 91 IM !t* tn '* II 0 0 0 cl 0 0 0 0 0 0 0 0 0 0 0 0 0 (Rev. 3/3/72) H!NI *11 TOTA!. SPttD o n 1.u 0 0 I o n 1.so o 0 I 0 0 I 0 I l*TO 0 9 1*65 0 IT t**I 0 IS t.79 0 10 1.55 0 13 lo65 0 15 t.89 O II I* 16 o I 5*00 0 I 4**0 o I 3*00 94 e.01 I t >I I ll'Ht\L SPF.f:O 0 0 :i 2.97 ' 0 0
• g,yg o o s 0 0 4 9;97 o t1 e a .so 0 Cl, I lt"O,' 11 ,, 1a a.oo 0 0 71 V* 0 0 I *d!* 0 0 60 1 u u 19 I *77 o o e. O 0 IP. 3**M o o 7 3.3u 0 " 1 ** 93 o o
• 1. " e.eo

z

0 ,_ :::> a:> ex: ,_ (/) 0 u ;z: w :::> = w ex: lJ... 0 ;z: :i U) L.J u ::z:: ..._, c:: c:: ::::> u u 0 '-0 u ::z:: UJ ::::> = E "' ::z:: 0 :::> I-;:: Ul ro :::> Ul ..... .... > Q) s 0000000000000000 0 = 0 0 OOOOOOOOOOOOOQOO 0 ;:; i-el z: 0000000000000000 0 H ..... _,_, ,.J,.J cc .., ....... 0000000000000000 0 0000000000000000 0 00000000000-0-0-" 0000000000000--0 000000000--00000 ! o ... : .. ... 0 ... c E .. a: .., .... c<> .. , gs ..... ti" ci ' " ..... _ ... .. .,., ..... ,.J"' 0000000000000-00 ooooooooocoo-ooo -2 0000000000000--0 ii: :c -* Q .. .,. :; ., " ? .. .. ... o,. :s., .. .. ., .. "' .. .. !5 .. oooooooa--00-0-0 o-*-0 o-c. 0 c. °" q -f .,.,. " :; : en ffi : lJ... ro lJ... ::£ .,_, _, ::> u 0 ;z: ,_ ex: Ul >Cl _, 0 < u ex .,_, Cl.. < < Cl -.. :.:o _.,,, .. :: .. .. to * <: 7< ,,,_ r:z: . ., .... O::c.> .. -c -= .. ..... :c .... << .. ., i:l::! uu ..... cc >!% 00 ,.J _, ,.J,.J "' ::i ! .. "' e .... D .:c w ..... .. 000000000000-000 ooooooooooeooo-o ooooocoooooooooo 0 oOoooooocoo--ooo 0000000000--0-00

• ca..,n--* Figure 2.3-10, sheet 14 Revision 18 "'"' ,.J,.J == IC f .. ., D cc .. ...... e 000000000000.:;riooo 0000000000000000 oooooooco:ioooc.:.o 0000000-=.::iococco 0000000000-00000 ooooooooco.cr-oo::-.:;i 000000000:>0000:::>0 0000000-00000:>:.o
• o oo-THE DISTRIBUTION OF WIND SPEED, DIRECTION, AND <J0 PASQUILL CLASSES DURING ti.T PASQUILL CLASS G CONDITION IN THE FINAL ANNUAL RECORD 0 ... "' "' "' t 0 .. DAMES & MOORE :: .. 0 ..,

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(_) :z: li.J :::> = LLJ if 0:: li.J N CO 0 :z: 1-88 I-<: ::r:: I-'-" "' => 0 0:: ::r:: li.J I-en "' <D I-:z: <: _J c... 0:: "' li.J LL. co LL. ::E li.J _J > u 0 :z: <: <: 0 > GJ 0:: c ... _, _, gg § ...... "' : -m c ; ! a---00000000000--o-oooooooooooon 0000000000000000 0 00000000--000000 0000000-00000000 " !: : ; .... c .. _, _, _, _, ; ; -0000000000-0000 0000000-00-00000 000000000-000000 Figure 2.3-10, sheet 16 Revision 18 ... 0: " --.. .. g; "'-"' ;';,... > --"' ""' ir:WWW>>>> ..,_ a:i -oo-.;. ci""" -ooooooooooooooc 000000000000000-0 ;:"' .. -: < o-".!5 <'" -< """ "'"' .. j: <<<< .. 5 r;i _ :! v" .. .., <C ,, ... 00 ::::; co _, _, _, _, << ........ 00000000000000-a 0000000000000000 000000000000000-0000000000-00000 ooooooooooocoooc THE DISTRIBUTION OF SPEED, DIRECTION. AND CY8 PASQUILL CLASSES DURING t..T PASQUILL CLASS A CONDITIONS IN THE PRIMARY YEAR OF RECORD ,. _, " " ;:; 0 0 .. ... ... ... : DAMES & HOORE

z: 0 => a:i. e::: ,_ (/") c; u :z: w => CT w e::: LL. C> :z: :i '-' :z: .._, ::::i = '-'-' 0:: LI... e::: w ro 0 :z: l-o u 0 I-<>: :c I-'-"' (/") ::::::> 0 e::: :c w I-m (/") to m I-.... :z: <>: _, c... e::: (/") w LL. ro w _, > u 0 :z: e::: w > Cl _, 0 <>: u e::: w c... <>: <>: Cl > OJ : ..,11"1 *e> .... m!"' ;, 000--0-00000-oi--0000000000000--0 0000000000000000 0 OOODDOOODDOOOOOO 0 0 f-0000000000000000 0 "' .. .. .. 0 ! .. "' ' 000000000000-000 000000-000000 .. 00 000-000000000000 0000-0000000000-G 0000000000000000 0 oooooocooooooooo 0 0000000000000000 0 0000000000000000 0 : g 00000000000000-0 00000000000000-0 0 000000-000000000 0 it 'I: -* 0000-00000000000 00000-0000000000 00000-0000000000 0000-0000000-000 --000000000--ooa-00000000-00-000-0000000000000000 0 0000000000000000 0000000000000000 0 W M M > Figure 2.3-10, sheet 17 Revision 18 ! g 0000000000000-no 00000-0000000000 0 00000000000000-0 ii: "' 00000000000000-0 -o "' "' ..... "' 0 2: _ .. "' ... 0.0 "' .. :: c .... <.>., -00000000000-0--oo-*-000000-ooon 000000000000-00---000000-00-0000 0-0000000000000-000-000000000000 00000000000-ooeo 0000005000000000 .,., _, _, .... :;:; 2: g ... '°f""Cll°'o-4'> -c'tftf')G ... g 0000000000000000 0-00000000000000 0000-00000000000 ooooooooooocoooo 0000000000000-0--00000000000000-000000000000-00-o--ocoo-000000-0 0000000-00000-00 oooooooooocooooo 0000000000000000 THE DISTRIBUTION OF WINO SPEED, DIRECTION, AND o-8 PASQUILL CLASSES DURING t..T PASQUILL CLASS B CONDITION IN THE PRIMARY YEAR OF RECORD ., 0 0 DAMES & MOORE 0 ,..,, Vl RO 3: 8 ;o ,..,, -i :r: ,..,, 0 ...... Vl t> -i ...... °' .,, c ):> -i Vl ...... .0 0 s :z r o r ,, n c r ,..... > :z Vl Cl Vl Vl °' .,, ,,, n 0 :z 0 ,..,, 0 ...... 0 -i ...... ...... :;a 0 ,,, :z n ..... ;z: 0 ;z: -i * :i: ,..,, > :z "'O 0 ;o ..... q <b -< .,, > -< Vl ,..,, .0 s r o r ,, n £: n Vl a Vl ;o ,,, 0 Vl 0 ra ,_. :z Gl ;S. Vl '"1 -* (D g !V w ,_. I 00 ....... Vl ('J) (D ..... ....... 00 WIND FREQUENCY O!STRIBUT!ON (FREQUENCY IN NUMBER OF OCCURRENCES) CALVERT CLIFFS PLANT SITE: STATION 2(IS) DATA PERIOD: NOVEMBER 1, 1969 THROUGH OCTOBER 31, 1970 PASQlllLL ! cmO!f AIC/Dn.T.11 T CRITLRl.fl.!.O*U ruu PASQOILL I ( 111014 .flltC/SI Ot!A ml TA ClllTElllAI MNOE US!Ol Ste TOR HNI HI nu: I !Ult SI .SU s SSV SV vsv v HV HllV H roTA!. 0 0 0 D 0 0 0 0 0 0 0 0 0 0 0 0 0 UPPIR Cl.AU IHTLRVlll.S or VINO SPUD (HP!!>

• 3 .. & 6 7 8 9 10 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 Q 0 0 0 0 o I 0 0 0 I I 0 0 0 0 0 0 0 0 0 0 0 0 0 a 4 I 0 0 0 O I O I 0 I o I 0 0 0 0 0 0 0 0 8 0 0 0 0 0 0 0 0 0 0 II I 0 0 0 0 0 I 0 0 0 0 0 0 0 0 I I 0 0 0 0 0 0 0 0 0 O I 0 I 0 0 0 0 0 0 0 0 0 0 0 0 0 I o I 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 I 0 0 PASQUILL II (FROlf AtCIDD.TA T CRITUIA,SQ*IS rtETI PASQUILL 0 CFROPf AIX:ISIG.'111 n!ETA CRIT!RIAI 111\NOE USED) UCTOR NII! HE r:NE 1 !Sil JIC .SSE .s sv vsv v NV HHV ll TOTAi. 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 llPPLR Ct.ASS IHT!RVlll.S or VltlD SPHD ( 1PHI e 3 ' 6 1 s 9 10 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 I 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 Q 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 I 0 0 0 0 0 0 0 0 0 . 0 0 0 0 0 0 0 0 0 I 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 I 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 II 0 0 0 0 0 0 0 0 0 0 0 0 0 I I 0 II 0 0 0 0 0 0 o* 0 0 0 0 0 0 0 0 0 0 HtA'I
• 11 TOTA!. SP!tD 0 4 5, 17 0 I o I 6000 0 I So IO O I 6*00 0 R 7* 10 o e 1.90 0 0 I O O I O O I 0 II 4. SS 0 O I 0 o I s e u.u II 5 11.tO 0 7 s.u 3* e.oo 'tEM *I I TOTAi. SPUD 0 0 o 1 a.ea o o I o o I o I 6olO 0 0 I 0 0 I O 0 I O 0 I 0 0 I o o I 0 0 I O O I 0 0 I I e 9, 60 0 I e.10 7, 36 PASOUILI. II (rROtf AltCIOit.tll t CRIUJl!A.SO*IA run PllSQUILL r HROtf 1110151 Ot1A mttll CRltlRI Al RNIDt O!l!:DJ SECTOR llllt Ht .mt I £St St SU s SSW SV llSV *V NV llllV N TIJ T /lL 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 UPPl:R CLASS IHT!llVAl.S or VIHD !PEED (tfPll> a 3 .. s s 1 a 9 10 0 0 0 0 0 0 0 0 I 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 b 0 0 0 0 0 I 0 0 0 0 0 0 0 0 0 0 0 0 0 I 0 0 0 0 I 0 0 0 0 0 0 0 0 0 0 0 0 I 0 0 0 0 0 0 0 0 0 0 0 0 0 0 I I 0 I 0 0 0 0 0 0 0 0 I 0 0 e I 0 0 0 0 0 0 0 0 0 0 0 0 I 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 I 0 0 0 0 0 0 0 0 0 0 0 0 0 I PASO Ult.I. 9 ( fRO'f A!CIDIL TA T CRI TV!I Ao 50* I e rJ, I lllllDS AT 33 FEET ABOVE ORADE SECTOR NNE NE ENt E !SE SE SSC s ssv 511 VSV v VNV NV N'lll N Clll.'f 101/u. 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 UPPF.11 CLASS INTERVALS 0, VIND SPEED (:'!Pill Q 3 4 s 6 1 8 9 10 0 0 0 0 0 0 0 0 I 0 0 I 0 0 0 0 0 I 0 I 0 0 0 I 0 0 0 0 0 I I 0 A I 0 0 0 0 I I 0 0 I 0 0 I s 3 0 p I 0 I 0 3 e e O I 0 0 0 0 I 0 0 0 e o 8 0 R R 0 I R I 3 ' s 3 I I 3 e I I! 0 0 0 0 I I I 0 1 3 '0 0 0 0 I 0 0 0 0 0 D O I I I I 0 0 I II 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 8 0 0 11 I I I I 8 I 2 I I 15 U 17 R3 II (Rev. 3/3/72) HUii 11 *II TO TllL SPUD 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 I 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 II 11 *I I 0 0 I o 0 0 0 0 0 0 I 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 I 1 R 8 I B IT a 6.90 4 6040 0 I I 4o IO I 0 I o I o I I l*TO 0 I 0 I o I I 6* 50 I 7, 10 0 I 1 9,113 1e 1. 21 HlJ\11 TOTA.I. SPF.F.D 10 6*01 II 5, 30 3 s.o 1 s 6046 s 7,89 4 s.oe
• 3*40 3 3,93 0 o. a ss, s ... ,4 7 6* 09 ., 10.00 19 10034 B7 60 60 0 I Pft Toi I

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• r., ..... -'-' .,, _, 55 g ... :: = C>C -' _, _, _, :;s .... -f i J? 0000000000000000 oooooooooooocooo 0000000000000000 0 0000000000000000 0 0000000000000000 c 000000000000-000 oooooco-oooooaooo 0000000-00000000 -00000000000-0000 000000000-00-000 m 0000000000000000 0000000000000000 0 0000000000000000 0000000000000000 0000000000000000 0000000000000000 0000000000-D-000 G 00000000000-0000 ooooc::i-0.-0-00-000 ooooooon-n-0-000 oo-oooooocsoceoooo 0000000-00-00000 a 0000000000000000 0 0000000000000000 0 S: : g 00000-:>-ocoooo-oo 0 l-"' ... :! -w ooowooo-0000-00-N ... D .. ,,.,. :: "'" .... :er 00 000-000000000000 0000000000000000 0 ooooooocoooooooo 0 :. .:.:.!...o --o-ao-oonmow-m-o-_,-' g " w .. 0.., .. " % .... " ... o., fi"' .... % :* .. tl .. fi 0. .... :> _, _, "' 0 cc ...... 000000000000000-0000000000000000 000000000000-000 0000000000000000 00000000000-0000 0000000000000-00 o-aooao-0000-000 0000000--0-00000 0000-00000000000 0000000--0000000 0000000000000000 0 Figure 2.3-10, sheet 19 Revision 18 THE DISTRIBUTION OF SPEED, DIRECTION, AND o-0 PASQUILL CLASSES DURING ll.T PASQUILL CLASS C CONDITION IN THE PRIMARY YEAR OF RECORD DAMES & MOORE z 0 => tD er: I-"' 0 u z w => c::r w er: u.. 0 z ;; en LL.I CJ z LL.I er:: er:: :::> u u D u.. D er:: 1..LJ CD :::.: :::> z u z LL.I => c::r LU 0::: ..... "" UJ ..... to 0 u 0 <x: <( 0 > QI 2 : .. u ... ...... ...... 0 i-"' " .. ... .... .. " :. .. .. .. o., "' ... .. .. ?: .. .. < d .. fi .. SS .... .. f i 0-000000000000-0 OOOOODOOOOOOOOOG 0 0000000000000-00 *00000000000000--00000000000000-0000000000000000 0 0000000000000000 0000000000000000 0 n :=c g 0 ow :l:l .... <> cc ... .... .. .. ! 0-000-0000000000 000000-000000-0--o-ooa-00000000-000-0000000000---co*-oooooooocQo a*-000000000000-ooo-oocooooooooo 000000-000000000 Figure 2.3-10; sheet 20 Revision 18 = o i-., .. .. :; ., -o-atooi:--oo--orooc. ooo"oa,,00000---n-a 00001-00-0--0-000 oocooo---0000000 M 00000000000000-0 ., ... ...... .... <: ;'-'\ :-; (JQ .. : -00000000000000-000000000000000--000000000000000 0000000000000000 0 ooocoooooooooooo 0000000000000000 0 ooooooooooooooooa ooooooooooooocgo c -000000000000000 O oooooooooocooooo 0 0000000000000000 0 THE DISTRIBUTION OF SPEED, DIRECTION, AND <J8 PASQUILL CLASSES DURING PASQUILL CLASS C CONDITION IN THE PRIMARY YEAR OF RECORD 0 "' .. ., ... "' .., ., "' DAMES & MOORE

z: 0 => co °" I-V) 0 u :z: w :;:, er w °" LI... C> :z: ;:;; (/) LL.I u :z: LL.I = = ::::> u LI 0 LI... 0 = LL.I co :>:: ::::> :z: u :z: LL.I ::::> er w °" LI... 0 ...... O'l ..... °" w C\J co 0 :z: 1-8 g 1-<C ::i::: I-"' U) ::::> 0 w I-O'l U) <D O'l I-..... :z: <C _, CL = U) w LI... co LI... ::E: w _, > u 0 :z: I-ce: w > C> _, Cl' <C u °" w a.. <C 1-<C C> > Q) !::. c !E l:O* '"' _, _, _, .J :;:; 00 .. ff ! i : .. 0 .. oooooco-00000-00 0000000000000000 0 000000000000--00 m 0000000-00000-00 0000-00000000000 00000-0-00000000 m 000000-000-00000 m : -a n g coooooo-00000000 00000000-000-000 0000000000000000 0 000000000-000000 000000000000-000 -0000-cooooo*oooo -000000--4:-noooo 0000000-00-00000 Figure 2.3-10, sheet 21 Revision 18 "" .J .J ... _, 0--0-0-00000--00 0000000-00000000 Wti::WW>>:t> ! g 00000-mooooo-ooo-ooooooooooooociaoo 0-00000000000.nno 0000---000000.-ow o-oocroaioooo-ct")OO o* ooocawtt'l-o o-(:1; o; o----oca01.,..lf1C-oa l'lca .-o-oo-ooo--o---e1aa10 o*oooo-nr-oocr:1ooeo C> "' ... ;:; "' ..., 0 .. .., ... ;; ;; .. ... ., .. ;. .. "' ., "' 0000--0--0000000 THE DISTRIBUTION OF WIND SPEED, DIRECTION, AND CJ8 PASQUILL CLASSES DURING t..T PASQUILL CLASS D CONDITIONS IN THE PRIMARY YEAR OF RECORD DAMES & HOORE

z: 0 :::> a:i 0:: 1-Vl Cl u :z: w ::::> er w 0:: u.. Cl :z: :;; u :z: w ::::> er w 0:: u._ o:: w :> Cl _J 0 < u 0:: w Cl.. -.: -.: Cl > OJ " """' i5: " t c ! ! . .. c:c: ..... " .. ...1...1 ...J...I "' c cl .. :;; .... = "" ..J..J ..J...I SS Od ,_ 000000-000000000 0000000000000000 0 g e q .., ""-'° ao-0000-0000000-0-00000--oooooco Figure 2.3-10, sheet 22 Revision 18 .... ... ... "'"' ,., .. ...IC ! :: r;i-0 00-o-O. 000 OP3CO : ----m ----aii:Jl.:WW>>>> f i = n--o-o-----__ _,..,.., OIO r> <1Z °"n.-* . . . . .. °'., 4' ., -0000000000000--E --0000000000000-fl " -00000000000000= -aooooocoooooomm -oo-oocooooooo-1n -00000000000000-oooooocooocooooo 0000000000000000 0000000000000000 THE DISTRIBUTION OF SPEED, DIRECTION, AND CY0 PASQUILL CLASSES DURING PASQUILL CLASS D CONDITIONS IN THE PRIMARY YEAR OF RECORD c ., .. 0 C> .. .. DAMES & MOORE WIND FREQUENCY DISTRIBUTION (FREQUENCY IN NUMBER OF OCCURRENCES) CALVERT CLIFFS PLANT SITE: STATION 2(1S) DATA PERIOD: NOVEMBER l, 1969 THROUGH OCTOBER 31, 1970 -I :r: 1"11 (Rev. 3/3/72) 8 (/) PllSUUILL E CfMll ACC/DUTll T CRIURIA*50*19 FEETI PASQUILL E ITnoH AECIDr.LTA r CRlrERlll*50*1P n:tn §! PASQUILL A (fROH AEC/SlOHA TKITA CRITERIA! A/\110£ USF.Dl PASQUILI. D I FnOH Ar.r,ISIOHA THETA CR I TERI Al MNOE US!:DI CD Sf.CTOR UPPER CLASS ltlTERVllLS 0' VIND SptED IHPlll HEAH SECTOR UPP£1t CLASS INTERVALS O' llfND SPttD CHPH> HEA'i > c: I Q 3 ... s 6 1 8 9 10 II *II TOTAL SP!:r.o I e 3 4 5 1 a " 10 II *II tOtN. SPF.\!n (/) -I ,CJ 0 c: 1111£ 0 0 0 I 0 0 0 0 0 0 0 0 I 3o90 NNE 0 0 0 0 I 0 0 0 0 0 0 0 I "do P z HK 0 0 0 0 0 0 0 0 0 0 0 0 0 l '1£ 0 0 0 I 0 0 0 0 0 0 0 0 I r-0 !/IE 0 0 0 0 I 0 0 0 0 0 0 0 I ihl&O 0 0 0 0 0 0 0 0 0 0

• 0 0 " I "'Tl E 0 0 0 0 0 0 0 0 0 0 0 0 0 I E 0 I I I 0 0 0 0 0 0 0 0 3 g,57 n r-:ic: IYE 0 I 0 0 0 0 0 0 0 0 0 0 I l*BO ESE 0 I 0 0 I 0 0 0 0 0 0 0 a 3*R5 > ...... SE 0 I 0 I 0 0 0 0 0 0 0 0 2 e.50 SE 0 0 I I I I 0 0 0 0 0 0 4 **00 (/) z SSE I 0 4 1 4 0 I 0 0 0 0 0 17 3,45 SSf: 0 3 8 1 3 8 I B I 0 f) 0 Bl 3,90 Vl 0 :;d '.!! B I 5 96 IS 23 1 I 0 0 0 0 85 3*"6 s I 13 IB 19 ') 11 P. 0 0 0 I 61 3*99 1"11 Vl UV 0 7 25 15 14 9 0 I 0 0 0 0 71 3oS7 SS\/ 0 ') 1 s 1 3 B I 0 0 0 0 30 3* 63 ('D (JO SV 0 8 IQ IS 16 3 I a 0 0 0 0 57 3*70 sv 0 4 3 I I I 0 0 0 0 0 0 10 g,93 ("') m ;S. i:::: 0 1"11 Cll "'1 vsv 0 8 9 II 9 5 0 0 0 0 0 0 42 3*41 \IS\/ a 9 5 I e 0 0 0 0 0 0 0 12 e * .-.1 z :=' -* ('D v I 3 9 18 II 5 a , I 0 0 0 45 3,97 v 2 2 g I ,, I I u u u I I) 10 3*61 0 0 N VNV 0 2 4 R 5 5 R 0 p 2 0 I es 5, 32 11>/V 0 (J I 3 3 5 R I e I I u 19 5,95 .... 0 ::s . NV 0 0 R R I 0 0 I 0 1' 0 0 1 4.90 NII 0 0 e 0 9 3 I I 3 0 0 I 13 6*40 0 ;;; (.;.) 0 0 0 0 0 0 0 0 0 0 0 0 0 I NHll 0 3 0 0 0 0 0 0 0 0 I 0 4 3*1B z m ........ I N 0 0 0 0 0 0 u 0 0 0 0 0 0 I N I 0 0 " I 0 0 0 0 0 0 0 2 9*45 Vl g 00 ..... ...... TOTAL 3 35 91 e1 84 34 13 6 3 3 0 I 354 3*63 TOTAL ?.1 43 '*I P.I 9 1 & I 3 a 199 4...05 z 0 z Cll -I . ::l"' :r: > (!) ,.,, z (!) "O 0 ....... ;o N (.;.) -< PASQUILL E OROH A!C/DD.TA T CRIT!lllA* SO* I a rttTl PASQUILL E <rAOH llECIDELtll T CRITERIAo50*12 n:i:n -< Vl PASQUILL C UROH AECISIGHA THETA CRITERIAI RANDr. USED> PASO\llLL D lfROH AEC/SIGHA TllETll CRITERIAI HANOF. U)>l'lll ,.,, ,CJ UPPEH INrtRVl\l.S OF srr.i:'o CHPlll 3; s SECTOR UPPrn CLASS INTEllVALS Of 1111m spr.£0 (HPIO HENI SEC TO It H&ktl r-I a 3
• 6 7 8 9 10 II
• IJ TOTllL SP££0 I e 3 ... 5 6 1 e 9 10 II > 11 TOTI\!. SPEED 0 r-"T1 ("') Hiit 0 0 9 I I I 0 0 0 I 0 0 6 4*4B NNE I 0 I I I 0 0 0 I. 0 I 0 6 5*08 ;o r-llE e 0 0 e 0 I 0 0 0 B 0 0 7 5*00 NE u I I 3 I a I g 0 I 0 0 12 S*U ,.,, > !NE 0 0 I I I 0 0 I 0 0 0 4 4, 57 EllE 0 0 5 a 3 0 0 0 I I 0 0 0 II 3.94 n (/) ! 0 I 0 I 3 0 I 0 l1 0 0 0 a 3*N F. I 4 I 3 a 6 5 3 0 0 (J 0 P., 4,74 0 (/) ;o 1"11 ESr. 0 3 a e I e I (J u 0 0 I 3 '" 00 0 2
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• 4 Q 2 0 0 0 0 eJ 4, 75 0 Vl SE u 0 I J 0 I a 0 I 0 0 0 8 4,9, St 0 2 e 3 " 1 4 0 I 0 0 0 RJ 0 SSC I 3 lo ., , 1 3 I e 0 2 Q. 66 5St 0 3 21 RT H 13 II 6 I n
• I J2 5,77 c: s 3 H IY IV 17 " IJ 4 u I I 0 88 lt.96 s n 0 I 3 9 B e I P. 0 I 5P. **32 ,l SSll I 6 J 6 J J *I 0 0 0 JO 4. 17 ssu 0 6 3 0 0 I a 0 0 0 0 0 19 3*02 z 0 4 0 3 0 0 0 0 0 0 0 a. 76 Sii 2 3 3 I I, 0 0 (/ ll. 0 0 0 10 2.1 e "' VSll 0 0 I 3 0 I 0 *) 0 0 0 10 J, 61 llSll 0 I I I 0 0 0 0 0 0 0 5 3. 24 'u I 5 4
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• I (l 0 II I I 15 5*20 1"11 Vl TOTAi, 3* M "' ., JV 20
• 12 7 ,, 371 RO ** 59 7 33 66 90 )0 55 OI 1'7 Pt. Jn :i: 0 0 ;o ,.,,

'-' z: L.J 5-w 0:: LL 0 z: :;; "' w <..) z: w er: 0:: :::> <..) <..) 0 LL 0 er: w CD ::E :::> :z: <..) :z: w :::> CY w er: LL er: "' w LL CO LL ::E w __, > '-' 0 :z: er: w > 0 __, 0 q; '-' er: w a_ > Q) er: "' .. :c _, 0 ..,_ '"" ........ uu <:<: :c5 ...... Q i: 0 :c .. e ; < n--000-oooooonn-m 00000--0000-0000 n 00000000--0-00-0 e 0 "' "' .. ., ., 0 :: .. .. ... o.., .. _, g :; q :: c _, "'" a: ., .. .... " nooooonooooonm--OI c-J-"° m-onooo-000-0000 Figure 2.3-10, sheet 24 Revision 18 ... "' " .. .., fi -.n.,,,., ---oi c: --.n-_,,.. __ -o:ll')Ol c.:c---.cr"'"'°'oa-o.("')c::--0. §"' -OIOJ01--w lo1 :a :a > :> O"M C "'"' :: -:---e --000000=. c:. :)C: -=-.. -c-OJOOOOOOOC-::, OC:: = -n cu-oooooo::c:::i:::::-::.":"'o.. .... oocooooc-coc: OJooc:io-oeQooo-o:---ooooooc-:-o.;, :--::o-00-0000::-*.: ::.-.:i.= 0:00:.: 000000000000.::io::;cc. -THE DISTRIBUTION OF WIND SPEED, DIRECTION, AND CJ8 PASQUILL CLASSES DURING l'i.T PASQUILL CLASS E CONDITIONS IN THE PRIMARY YEAR OF RECORD DAMES & MOORE V> ll" 3: 0 0 ;a rn -! :c rn t> -! ;o V> .0 s r-C"'l r-e;; V> .,, C"'l 0 :z 8 § 0 :z 0 "Tl :z 0 V> "'C rn rn 0 0 ;o :z rn V> n -! ...... -:z 0 :z -I :c rn "'C ;o > :z 0 (bq -< "'C -< ,..., CJ 5 0 .,, ;a 8 ;o 0 ,.... C"'l > V> V> ,..., V> 0 :z en 'T.I ("D -* < CJq -* c Cl> "'"1 5* ("D ;:s w ,_. I 00 ...... Cl> ::r' ("D g. N !Jr WIND fREQUENCY DISTRIDUTIDN (FREQUENCY IN NUMBER OF OCCURRENCES) CALVERT CLIFFS PLANT SITE: STATION 2(IS) DATA PERIOD: NOVEMBER 1, 1969 THROUGH OCTOBER 31, 1970 PASQUILL r <FllOH AtC:/DEl.TA T C:RITERIAoGO*IR fUTI PASQUILL A CFllOH AZC/$lllHA THETll CRIT£RIAI RAllOt USEDI llECTDll UPPIR CLASS INTl:llUl\LS or VlllD SPUD IHPlll a 3 " s 6 1 a 9 10 Niii NE ENE E ISE BC SSE s 1$11 av llSll II 111111 NW ,llflV II TOTAL 0 0 0 0 0 I 0 0 I 0 0 0 0 0 I 0 0 0 0 0 0 0 0 0 0 0 D o R 3 6 10 a 6 5 3 a a " a R 8 D O 0 0 0 0 B3 34 0 0 0 0 0 0 0 I I 0 I " 0 0 0 0 0 0 0 0 0 0 I 1 3 0 0 0 3 0 D 0 0 0 0 D D D 0 R 0 0 0 0 0 0 0 0 0 D D 0 0 0 0 I 0 0 0 0 0 0 0 0 D 0 0 D 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 I 0 I 0 0 0 0 0 D PASQ\llLL P' CfROK AEC/DELTA T CRITERIA,$0*18 FEET> PA&QUILL C I FJ\OH llECISIOHA lHl:TA CRITEAIAJ HAllO£ UC7011 UPPJ:,L I *)f WUIO :iPl::V..IJ CH.>11> 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 a 3 " 6 1 e 9 10 llllE NE lllE E SE s IV vsv II llllV NV TOTAL 0 0 0 0 0 0 0 0 0 0 0 0 0 l* 0 0 fl 0 4 I 111 1:1 I II I 6 Q I u 5 ,, I I 3 I 0 0 I I I I 0 u 0 0 0 0 0 0 3 K 0 0 " 3 3 I n 0 0 0 0 0 0 0 0 0 I U I D 0 I 1

• 3 4 I 0 ,, I u 0 .. .q, :1 I J 0 0 I U 41 0 1$ R l I 0 0 n ,, u 0 0 II 0 0 " " 0 0 I) HEAN TOTAL SPEED O I 0 I 0 I 0 I 0 I I *90 6 8067 RB 3* 36 13 a. 10 9 R*7B 5 B*36 16 R*58 1 3, 13 0 I I *90 o I 86 8086 TOTAL o I o I o I o I 3 R*,O R 7,45 Ill O*dfl 43 a. ')I \1 I 0* J'* h ,,,, B*H ll*'l'J 5 lq u 3 u a .... .,, 5* Ill l* 13 g. 30 I 3*01 PASQUILL r PASllUILL B SECTOR NME 0 llK 0 m& o ' 0 ESlt 0 SE '0 SSE 0 s 0 3 SV R VSll 0 v 0 111111 0 NII 0 NllV 0 II o TOTAL PASllUILL f PASQUILL D S&CTOli NHt O 111: 0 tNl I t () i:sr. o SF: 0 5SV. 0 S I ssv g S\I 3 VS\/ 0 v 0 o NU 0 I N (I TOT/II. !FOO" Al:C/DEl.TI\ T CllltClllll*lll*\I FEttl OllOH AECISIOl1A TllE A CHIUlllAI llAllOE UHDI UPPER CLASS INTP:llUALS OJ VIND BPICD (HPlll B 3 4 & 6 l II 9 10 0 0 0 0 0 0 3 9 6 0 0 0 0 0 0 0 0 0 0 0 0 0 a 7 5 a a a a I 0 0 0 0 0 0 0 I I 3 I 0 0 3 I 0 0 0 1e e3 10 0 0 0 0 0 0 0 3 a I 0 0 a I 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 I 0 0 0 0 0 0 0 0 0 u 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 D 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 C Fr\OH AECIDEI. TA t Cfll TElllA* 50* IR P'ECTI <FROH AECISIOl1A THETll llMOE UliEDI UPPEu Cl.US 1111>:.tll*\l .. 1 O* VUID SPUD jHPlll a 3 4 5 6 7 11 .9 10 O I I 0 I I f'l fl I o I er (J 0 ,, I I 0 a 111 H ff T 3 3 6 $ I 0 0 . 0 4 10 a 3 $ (, L 0 Q I) I n p,. , ... 0 I) ,, 0 0 .() I 3 I 0 ,, I :i -rr ,, 0 I l'r 0 0 0 :* I 0 0 (J 0 I 0 .. ,, 0 0 u " I 0 0 I) D I I 0 () " 0 0 u " I) 0 0 ti (I I r: " '* ' D D ll 11 fl ,, 0 0 ll a U I f II II " ti 0 0 u \) ,, 0 0 11 II " a '* II II 0 0 0 0 0 0 0 0 0 0 0 0 0 0 D D II 0 " I> " I " ,, 0 '3 >II 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 *II 0 0 I* I " "' (I " ,. (Rev. 3/3/72) HSN! TOTAL SPIED 0 l O I o I o I 0 I I 1*10 6 a.aa Ra t*H 11 ** 11 a.oa * **35 5 3oll 5 3o0 3 0 I 0 I 66 a.u Mt.,. TOTAL. SPU*I a 3 a I ll*lil 2: lh8u 1*08 Ill 3*19 !l*l*l I ... l'i .. 1'1 "' I Al*'' 111 "'*'" 't *** ., ,, J. 111 ).52 WIND FREQUENCY DISTRIBUTION (FREQUENCY IN NUMBER OF OCCURRENCES) CALVERT CLIFFS PLANT SITE: STATION 2(IS) DATA PERIOD: NOVEMBER 1, 1969 THROUGH OCTOBER 31, 1970 --f :i: rrl (Rev. 3/3/72) c ..... PASOUILL r HROl1 AECfDELTA T CRITERIA* 50* I e rtETI PASQUILL r <tROH AECIO!:l.TA T CRITf.1\IA*SO*IB rttt> "' PASQUILL ! HnQ11 A!CISIOHA TllEfl\ CRITERIA! RJ\NOE US!D) PASQUll.L r C tllOH AtCISIOt!A THr.tA CRITtRIAJ RJINO! USED> --f ;o ..... S!CtOR UPPER CLASS INfERVAl.S or VIND SPtr.D IHPlll HEAN S!CTOR UPPF.R CLASS ltlTERVAl.S or VUIO SP£r.D (HP<<> lltMI -c o:l I e 3
• s 6 1 8 9 10 II >I I TOTAL SP&to I D 3 4 5 6 7 8 9 10 II *II TOTN. SPEF.O > c: "' ;:::! JO 11'/E I 0 0 0 I 0 0 0 0 0 0 0 e 9*50 NNF. 0 0 0 0 I 0 0 0 0 0 I 0 D c: 0 1. eo ..... :z: ff! I 0 0 0 0 I 0 0 0 0 0 0 e g. 75 Nr. 0 I 0 0 I 0 0 0 0 0 0 0 e 3.g5 r 0 ENE 0 I I 0 0 0 0 0 0 0 0 0 e I *60 DIE I 0 0 0 0 0 0 0 0 0 0 0 I 1.00 r "Tl t 0 0 I 0 0 0 0 I 0 0 0 0 e E 0 0 0 0 0 0 0 0 0 0 0 0 0 I n nr. 0 0 I 0 0 0 0 0 0 0 0 0 I e. 10 !SJ: 0 0 0 0 0 0 0 0 0 I 0 0 I 9olO r :r:: !!: 0 I 0 0 0 0 0 0 0 0 0 0 I ldO SE 0 0 0 0 0 0 0 0 0 0 0 0 0 I > ..... V> :z: SSE 0 I I 0 I 0 0 0 0 0 0 1 3* II SSt e 0 0 0 0 0 0 0 0 0 0 0 e *es "' 0 s 0 1 4 e I 0 0 0 0 0 1 0 15 e. 97 s 1 0 1 0 0 0 0 0 0 0 0 0 e 1. 70 "Tl "' ('!) fJCi SSll 0 4 I 0 0 0 0 0 0 0 0 0 5 1. 60 ssv I I 0 0 0 0 0 0 0 0 0 0 n *00 -c ;S, s:: Sii I e 0 0 0 0 0 0 0 0 0 0 3 I ol l Sii I P. 0 0 0 0 0 0 0 0 0 0 3 I .SO n rn VSV 3 I 0 I 0 0 0 0 0 0 0 0 5 1.50 \/SV I 0 0 0 0 0 0 0 0 0 0 0 I *BO 0 rn en .., :z: 0 ....... ('!) v 3 a 0 e 0 I 0 0 0 0 0 0 e P.o21 v 0 0 0 0 0 0 0 0 0 0 0 0 0 I ;:: -0 N 111111 0 I e 8 4 0 0 0 0 1 0 0 16 .'.)t 69 I/NV 0 I I u 0 0 0 0 0 0 0 0 Q a.es --f c ::s . H\I I I 0 I s 8 3 0 I e 0 3 25 NII 0 I 0 0 0 I I 0 0 I 0 0 4 5,55 0 ..... w NNV 0 I 0 0 I 1 2 I 0 0 I I 8 6* 79 NN\/ 0 0 0 0 0 I 0 0 0 I 0 0 a 7, 45 ;o ....... I N 0 4 0 1 I 0 0 I 0 0 0 I e 4, 74 N 0 0 I 0 0 0 I I I 0 0 0 4 6. ts :z: rrl "' n 00 ....... --f TOTAL 10 96 14 16 13 19 3 I 3 e 5 110 4, 05 TOTA!. l 6 3 0 2 R 2 1 I 3 I 0 P.8 3, 81 ..... 0 :z: :z: en -i ::r :i: > ('!) rrl ('!) :z: ...... -c 0 ;o N <l>q 0\ -< -c > -< "' rrl JO PASQUILL t HROH AtCIDEl.TA T CRITERIA*SO*IQ fEETI Pl\SQUlt.L r (FROM AEC/00.T" T CRT TERI A. so-1 a Ff,) ;;; s PASQUILL 0 ( rno11 AtCfS IOHI\ TH&TA CAI tW1 Al USED> VINOS AT 33 nu l\OOV& QRADE r 0 r stcron CLASS INTF.llUM.s or lllND SPEF.1> <HPll> HF.I\!! SF.CTOll UPPER CJ.ASS llltr.RV/\LS OF VIND SPF.ED 'CtfPlll HEJ\11 ,., n I a 3 4 s 6 1 8 9 10 11
• 11 TOT/\L SPF.ED I e 3 4 ' 6 7 e 9 10 II >11 TOTAi. SPEl'.D ;o r rn ,,. Ntlr: 0 0 0 I 0 0 0 0 0 0 0 0 I 3*60 NllE I 0 t e 2 0 0 o* 0 0 I 0 T 4ol6 n "' HE 0 0 0 0 0 0 0 0 0 0 0 0 0 I NE I I I I I a 0 0 *o 0 0 0 T 3,44 0 "' ;o rrl 0 0 0 0 0 0 0 0 0 0 0 0 0 I mt a I I 1 'O 0 0 0 0 0 0 0 ' l*Sd 0 tn E 0 0 0 0 0 0 0 0 0 0 0 0 0 I t 0 0 e 0 0 D 0 I 0 0 0 0 3 3,53 0 UY: 0 0 0 0 u 0 0 0 0 0 0 0 0 I !St 0 e 1 0 I 0 0 0 *o 3 0 0 1 ,, 30 c: SE 0 0 0 0 0 0 0 0 0 0 u 0 0 I SE I n I I I 0 0 0 0 I e I 10 ,, SSF. I 0 0 0 0 0 0 0 0 0 0 0 I
• 40 SSt 3 I e P.3 1 e s 0 0 o' 0 0 0 52 1*16 % s t 0 0 0 0 0 0 0 0 0 0 0 I *50 5 4 37 19 DI 6 6 0 I 0 I 0 139 3*0!1 "' SSV 0 0 u (I 0 0 0 0 0 0 0 0 0 r SSll 8 31 eo 5 10 I 0 0 0 0 0 0 15 e.u sv I 0 n 0 u 0 0 0 I) 0 0 0 1 . "' Sii 9 9.1 18 I I 0 0 0 1 0 0 0 51 1*91 VSll 0 0 (/ u 0 0 (I 0 0 0 0 0 0 I W'5\I 4 4 7 e I 0 0 0 0 0 0 0 18 Q.1.* 0 II u 0 IJ 0 () () u 0 u u u 0 0 I II 3 26 14 I I 0 0 0 0 0 0 60 a. si 0 11 () u u 0 u 0 (/ (I I) 0 0 r VNV r 5 13 I 7 16 4 1 I 1' I 0 0 60 3, 19 NV 0 0 II I I 0 0 " n 0 0 0 ?. " 00 NII e p I 3 le 19 6 0 3 3 0 s 56 6*00 rrl 0 0 (\ 0 (/ (I 0 II (/ Cl u 0 0 I NNV 3 e I 0 3 e e I 0 I I I 11 4,95 "' ' u 0 " (/ 0 1 0 0 ,, " u (/ 1 5*60 ti 0 5 2 I 'P. 1 I 2 I 0 0 1 16 RO 3: TOT/\L 3 u u e I I " (/ u ti 0 (/ 7 9, 63 0 101111. 140 I 68 74 41 16 5 1 9 8 !>114 3, 30 rrl WIND FREQUENCY DISTRIBUTION (FREQUENCY IN NUMBER OF OCCURRENCES) CALVERT CLIFFS PLANT SITE: STATION 2(IS) DATA PERIOD: NOVEMBER 1, 1969 THROUGH OCTOBER 31, 1970 -! ::i: rn 0 (Rev. 3/3/72) ..... !:'.i VJ ctROll AtCIDn.TA t CRltERIA,50*10 fEttl -! PASQUILL 0 PASQIJILL 0 <rnoH ll!CIDELTA T CRITr;HIA*SO*IR ,![Tl ;:>;l PASQUILL A CrROH llEC/5101111 tHl:TA CRITERllll RAHOt USED' PA!GUILL II IPROH AECISIOHI\ THr.TA CRITERIAI AMOE USEOI -a ;; > c:: U:CtOR UPPtR ct.AH lNttlWN.S or VlHD $rl:!D (HP\0 lltllll $ECTOR UPPtn CLASS lllURVALS or lllllD !PEED (lfl'Jll "!All Vl :::! I e 3 5 6 1 s 9 10 II >11 TOTAi. SPEED I e 3 4 5 6 7 s 9 10 II >II TOTl\L SPEED .0 c:: 0 ..... :z 1111£ 0 0 I 0 0 0 0 0 0 0 0 0 I e.eo HHE I 0 I 0 0 0 0 0 0 0 (I 0 e 1,55 r-0 !IE 0 0 0 0 0 0 0 0 0 0 0 0 0 I H& 0 0 () 0 0 0 0 0 0 0 0 0 0 I r-.,, ID'IE 0 0 0 0 0 0 0 0 0 0 0 0 0 I Vlr. 0 0 I I 0 0 0 0 0 0 0 0 t e.so n E 0 0 0 0 I 0 0 0 0 0 0 0 I I: 0 0 0 0 0 0 0 0 0 0 0 0 0 I r-!St 0 0 0 0 0 I) 0 0 0 0 0 0 0 I ESr. 0 0 0 0 0 0 0 0 0 0 0 0 0 I > Vl :z H 0 0 0 I) 0 0 0 0 0 0 0 0 0 I SE 0 I 0 0 0 Q 0 Q Q 0 0 0 I I *TO Vl 0 Ht!! 0 I 0 0 0 I) 0 Q 0 0 0 0 I I o90 SSE 0 e 0 0 0 0 0 Q 0 0 0 0 II 1*65 en Vl 5 I 4 3 4 I 0 0 0 0 0 0 0 13 0*60 s 0 9 0 3 0 0 0 0 0 0 0 I& 9,44 -a 21 ssv 0 s 8 3 0 0 0 0 tll rn :z ::r -a 0 (1) A' ('!) <bq ...... N -< -a '--l PllSQl)ILL 0 CFROH A!CIDELTA T ClllTEAIA*S0*10 FEET> PllSQUILL HROH llECIDELTll T CAITEAIA*SO*IQ FEETI ;t> -< Vl PASOUILL C !FROH AtCISIOllll TlltrA CRIURIAI RllNOE PASQUJLl. D (FllOH THETA CAITERIAI ftl\llO,. usr.01 rn .0 '* ?i; 5 sr.cron UPPER CLASS OF VINO SPttll IHl'IO Hf'Afl Sl:CTIJR UPl'r.lt or lllN!J CHPlll 111';#\H r-I II 3 A 6 , A 9 10 II *II TOTlll. SPr.ro I a 3 4 s 6 7 II . 9 10 II "1 I TOTlll. SPP.Y.ll 0 r-.,, n NHE 0 0 I 0 0 0 0 0 0 0 0 0 I *t. 70 NH!'; 0 0 a I 0 0 0 0 ' 0 0 " 0 3 1*97 ;:>;l "t 0 0 0 0 0 0 0 0 0 0 0 0 0 I NE 1 I 0 0 1 0 0 0 0 0 *> u 3 9*01 rn !:Ht: 0 0 0 I 0 0 0 0 0 0 0 0 I 3.ao mr. 0 0 3 I I 0 0 0 0 u ,, 0 s n 0 Vl I 0 I 0 I 0 0 0 0 0 0 0 0 e a. 60 t 0 A I 0 0 I 0 0 0 0 0 0 2*97 ;:>;l rn ES!: 0 0 0 0 0 0 0 0 (* 0 0 0 0 I ESF: 0 I I 0 o* 0 0 0 0 0 0 0 II l*AO 0 Vl n 0 I 0 0 0 (J 0 0 0 0 0 0 I 1*50 St 0 0 0 II 0 0 0 0 ll 0 0 II 6 I 0 " 0 0 " II 0 0 Pl I* Ra SI.' a 3*, 19 p 0 Q 0 Q ,, 0 0 0 51 I *87 GI 0 In 7 Q 0 0 0 0 II (\ (\ 0 17 I *?I VSV 0 10 I 0 0 0 0 0 0 0 0 16 I ;,1,1 II 0 2 0 0 Q (\ " 0 0 0 ?.U 2* t'j(I v I 13 13 3 0 u II 0 0 0 0 0 30 e. eo 'JNV 0 0 3 e " I) u (/ ti 0 0 10 3. 21 VNV I I I l) .. .. I l) 'J . u 10 3oll 0 NV 0 0 a I I (J u " 1 0 0 0 5 l.111(1 NV I I (I I " l* I/ u u " " (' ' llo9B 0 0 I I 0 " 1 c u 0 0 0 3 4,43 0 " " u ) 0 0 (I 0 0 0 3 4,QJ N 0 0 P. 0 I Q 0 0 (J " 0 0 JdO N 3 (\ 0 0 0 0 0 0 D 0 0 2 I* ,0 rn Vl TOTAL 3 83 IY 10 l' ' 3 I " 0 0 I T1 R" 101111. 12 I 41 IOI ,,, ,, 4 0 0 \) 0 \: o) :i: 0 0 ;:>;l rn

= 0 :::> m 0::: ._ (/) 0 u = UJ => = UJ 0::: LL. = = t/) w u :z: UJ °" °" ::::> u u 0 LL. 0 °" UJ co :i;: => :z: ;::; u :z: w ::::> Cf UJ °" .... 0::: Ul UJ LL. co LL. :::E: UJ _, > u 0 :z: ._ 0::: w > = _, 0 <.: u 0::: w a_ > "' c a: -:I:"' o .. oooooooooooocooo 0000000000000000 0 0000000000000000 : i: "' lC 0000000000000000 .... WC W2: ..... :0: c: 7C .. c: .w .... ::!" -c c: .. ""' = .... "" '"" tl !: c., ,, '"' "'"' cc ::c:z: .... ...... ...... c t; 0000000000000000 .. .. 0000000000000000 .. .. 0000000000000-0-0000000000000--0 ::: = 000000000-000000 : "' :i ;; w .., Ir: =-:a :. > C>O ,_ "" " :: : ., ... 0 "' "' 0 .. -.,., :: 0 00000000000000-0 0000000000000000 ... -;:-... !!"' .... "' 7< ..-.:: .. :,.: <: .... "'" ... -.:: o:.-""' ... <:"' .. :c ,,., uu .... cc 0::.:: ..... ""' .J..; ...... 0 .., *., Q = _,.. .. ... § "'"' = :: .. c ;:i., IC .. .. .... :> ;; g; :: : 0000000000000000 ooeoocoooooooooo 0000000000-co-oo "' ... "' 0 .. "' QI 0 .. Figure 2.3-10, sheet 28 Revision 18 .. ... o"' ..>< "' .. " -* .... ... r. , .. , ,, '"' ..,.., .... c:c: ..... "c _, ... 0 £: :z: -c-c "' .., '""" ., 0 .. .. ... 0 ... a: .. "' .. = :::"' "' ... '"' c: w a. ... .. :> _, _, 0: e .... u : 0000000000000-00 000000000000-000 00000000000000-0 0000000000000000 00000000000000-0 000-000000-omo;i--o:..er-oo-oiC11 o er n n'Qa--o"-00*::"°4; -"' o ... cw--m-.,<io .,-n t--iin., an-w '"1 Ml .., :a ;II :a > ---0--000:0.,0----ooo:toooooooooooo 0000000000000000 00000000000;>000.:i OCOOOOOO.Q>oooc.ooo 0000000000000000 00000 o ... :-;.o o <r D' ooo ;::> THE DISTRIBUTION OF WIND SPEED, DIRECTION, AND CY0 PASQUILL CLASSES DURING PASQUILL CLASS G CONDITIONS IN THE PRIHARY YEAR OF RECORD c c = "' "' .. ., ... 0 c "' DAMES & MOORE PENNSYLVANIA ... / ................ . ', I PIEDMONT ................................................................. ............................. ', .... ************* MARYLAND *:: *. '.JEPA'lT"IE'l\!T :JF ,

• f\:ES Ar,.'.) , srArE :J:=-".'AQ .... ._,:.r..:c 0J __ E-"'.! 1,g, 11r;7 PROVINCE MILES 10 0 10 20 REGIONAL PHYSIOGRAPHIC MAP DAM*8BMOOR* FIGURE 2.4-1 VIRGINIA REGIONAL GEOLOGIC MILES 10 0 20 K £ v: I sED1VENT.ARY :::>EPosirs -T=:?*Ass*::-sE.: \'!:"-TA:;-y .:<:c:t<s :====::;TERTIARY J l'\::LUDES SCf .. F J ... ,::.=v :::;:c:"'5 SE011YENTS I cr<ETACE:Jus SEJit/E"\JrA:; ...
• tr>>-* c Al-0 I) KINl\'EY# o .. :.,* .. , "JAT!C!l.AI... GEOL:JGY; Ul\:ITEO STATES ".iE:::L. su.::;* ., PA?. 2) J:l32, 'AP CF 3TA.,.E.51 JCE:*'.). ,_ \ ! ! \ -< * **** 9 *....................................... MAP ;' 1 if.:<T *.;.;:; * .a.r-*c .=4:...20ZG I:-.__ _ G/\.EGVS ::?QC>< 0 a: .. .. < GEOLOGIC AGf: 0 RECENT ?LE I STOCENE 100-200-Ml :::CENE 3<1()-400-soo-EOCENE 600-700--PALEOCENE aoo-9<10-IG\'0-1100-120(;-1300 -1400-1500-1600-1700.-CRETACECUS 1800-1900-2000-2100-2200-2300-2400-2600-PRECAMBRIAN 2700-=.: .... ..,,.,.,Yi'sl't'i -*-*---" .. ..... -.... .... ,_ _ _ ,.., .._ _ ...... ,,., ::: -...,. -------_,.,., --..,_ -*-'!"" ....... ,.,. ..,,:,,._._ .. _,_ -...... .... .. ,.,.,. --.. ..... ,.,., -PT'> 1-t _..,...:..;....,...* OESCR/ PT/ ON ALLUVIUM ANO BEACH OEPOS I TS*CONS I STS OF 51 LT ANO SANO TERRACE OEPOs1rs-coNSISTS PRIMARILY OF SILT AND SANO WITH MINOR GRAVEL AND Cla,AY CHESAPEAKE GROUP-CONS I TS OF I NTERBEODEO CLAY, S 1 l.. T AND SAND WI TH SHELL LAYERS IN THE UPPER PORTION (INCLUDES UNOl-Ff:'ERENTIATEO PORTIONS OF THE ST1* MAR"l'S, THE CHOPTANK AND THE CALVERT F'OR"1'.ATIONS) PI NEY PO 1 NT FORMAT t :JN CONS 1 S TS OF MAR I NE Gl.AlJCON IT IC SANO AND C1..AY NANJEMOY FORaM.TION*cONSISTS OF fJIARINE GLAUCONITIC SANO ANO SILT WITH CLAYEY .LA'VERS AQU IA FORMAT I ON-CONSISTS OF MAR I NE GLAUCON IT IC SANO AND S 1 l. T WI TH CLAYEY LAYERS BR I GHTSEAT FORMAT I Ot11-cONS I STS OF GRAY MAR I NE 5 i L TY AND SANDY CL.AY MONMOUTH AND MATAWAN FORMATIONs-cON515T5 OF GRAY TO BLACK MARINE SANDY c .... AY ANO SANO ..VI TH SOJ.AE GLAUCON I TE MAGOTHY FORMATION-CONSISTS OF GRAY TO 1'iH*TE COl\ITINENTAL SANO ANO GRAVEL WITH INTERBEOOED Cl.AYSj CONTAINS PYRITE A,.,,_D LIGNITE RARITAN FORMATION-CONSIST CF INTERBEOOED CONTJNENTAl. SANOS ANO CLAYS Wl TH IRON STOl'\IE l\IOOUl.ES POTOMAC GROUP-COl'llS I STS OF I NTERBEODEO i:i:EO, GRAY 1 YELLOW ANO BROWN CONT I NENTAL. SANQ ANO CL.Av;: CONTAti'llS IRON STONE NODULES ANO t.tGNITE (PATAPSCO FORMATiON) ' ' (ARUl"\IOP:l. I ON) PATUXENT FORMAT ION'"'"COr..S I STS OF GRAY ANO YEL1..0W SANO 'NI TH I NTEqBEOOEO CLAY I GN!OUS AND METAMORPHIC ROCK GRCUND-tJA'l'EJ:t RESCURCES CF SOUTHERN MAPtYLAND COASTAL ;>LAIN1 MD. GEOLOGICAL SUR.VEY BULLETIN 15, 1955, GEOLOGIC COLUMNAR SECTION-SITE AREA DAMllB 8 MOOR* APPLIED EARTH SCIENCES FIGURE 2.4-4

..

.:. .. __ ... /:_._; ... .. *_;: :.-. .:.:. NO.T:E: THIS MAP WAS PREPARED FROM A PORTION OF TECTONIC MAP OF THE , : . :"-. *-UNITED STATES 9.y THE umTE1D STATES GEOLOGICAi.. .SURVEY AND THIE . -. . : -;36°. AMERICAN ASSOCIATION OF PET:ROLEUM GEOLOGISTS\ 1962. . * .. /. ; *,. ; . * ... "e_;, 'f/=f't .. ll! zf;; qJt'.'. ;-" .. ;,-* .=f ,_ LEGEND: t ANTICLINAL AXIS ---t* SYNCLINAL AXIS + DOME 'I I I I I I NORMAL FAULT REGIONAL TECTONIC MAP 10 0 bent IO ltO SO 40 SO e I I I I * * * *

• THRUST FAULT ---CONTOURS ON TOP OF BASEMENT ROCKS .............. l:A .. TM SC:ll'.HCl8 .FIGURE 2.4-S
• + L.0CAT*0' :,AS LOGGE:J AS SH;jfi!\ 01', _, -[ !

REFERENCE:

SITE THIS MAP WAS PREPARED FROM A PORTION IJSGS*, COVE P01Ni7 VARYLAND

• Q\JADRANGLEJ Jg43. 100 FEET 0 100 1200 DAl" .. *98MlNH.l* 4flflLllO IEA"TH 8Cil:NCI* FIGURE 2.4-6

! I * + t.OCATION WHERE CL.IFF WAS LOGGED AS SHO'NN FIGURE 2 * ...i.-j I FEET + l..OCATJON OF MICROTREMOR STATION 600 0 600 1200 R E F r R E N C E: aa.M*aBMOOR* THIS MAP 'NAS pqEPAREO FROM A PORTION OF usas., COVE POINT, -QUADRANGLE1 1ga3. t:AftTH ac1ENCl8 FIGURE 2.4-7

DEPTH IN FEET BLOW COUNT BORING [}M SURFACE ELEVATION+ 120.81 SYMBOLS *DESCRIPTIONS 0 REDD l SH-BROWN SIL TY Fl NE SAND 8 .. 0%-111 WITH GRAVEL (DENSE) 13. 10 12.0%-120 GRADING WITH COARSE SAND 40

• ANO GRAVEL .. 20 13-0%-112 I -80. I 16.0%-111 30 40 . GRAY SILTY CLAY (VERY STIFF) -50 60 I---GRAY CLAYEY SILT (DENSE) 70 2* 80 CLAY GRADING OUT -28.5%-90 74* I 90 I r----GREEN FINE SANDY SILT WITH SHELLS (DENSE) 100 22.0%-'l.02 ,118* 110 GREEN SILTY FINE SAND WITH -220 SHELLS (DENSE) ro 120 22.4%-105 GRAD I NG WI TH MORE SI LT 72. 130 32°8%-86 108* 140 --GREEN CLAYEY SILT (DENSE) -55. 1%-66 50* 150 160 35. 7%-82 36* l?O LOG OF BORING CONTINUED 17 v o-42. 3%-78 35
• 18 19 0 65.0%-59 47. 48.0%-70 60. 210 *-v 30. 23 0 w o* 49.0%-70 32. 24 25 0 r----50.0%-70 50. 26 0 o-51.0%-62 54. 271 28 IV 98.0%-44 73. 290 -132. 0%-37 30. 300 31 0 79-0%-55 GO* *O 32 33 0--150* ) y: 35. %-95 GREEN CLAYEY FI NE SAND WITH SHELLS (DENSE) GREEN FI NE SANDY SI LT (DENSE) GREEN CLAYEY SILT (DENSE) GREEN SILTY FINE TO MEDIUM GLAUCON IT IC SAND-(DENSE) BORING COMPLETED ON 8-27-67 NO CAS l NG USED WATER LEVEL AT 311 ON 9-12-67 500 LB. HAMMER WITH 1811 FALL l 2u7: II NNDD II CCAATTEESS DDEEePT' HH UOf"F DAMES & MOORE UNO I STURBED SAMPLE 1£>1 DAMES & MOORE 0 I STURBED 0 INOl(;ATES DEPTH OF SAMPLING ATTEMPT WI TH NO RECOVERY OF SPL.1 r-sPOON SAMPLE 2" a.o. FIELD f\/IOISTURE CONTENT IN PERCENT DRY DENSITY" LRs./cu.FT. DAM*S& MOORE .4.PPLl£0 EARTH SCIENCES DEPTH IN FEET t: :i; BLOW COUNT 20-----ll 40 50*----38.7%-84 17. 100 19.4%*!03 37. 110 130 37.2%-83 38. 140 150 52.8%-68.S 48. 16 39.6%-79 30 * /'l'D BORING OM 2 SURFACE ELEVATION+99.21 SYMBOLS OESCHIPTIONS YELLOWISH-BROWN FINE SANDY SILT (MEDIUM DENSE) YELLOWISH-QROWN SILTY FINE SAND (DENSE) GRAY SILTY CLAY (STIFF) GRAY FINE SANDY SILT WITH SHELLS (DENSE) GREEN SILTY CLAY (VERY STIFF) GREEN FI NE SANDY SILT WI TH SHELLS (DENSE) GREEN SILTY FINE SANO WITH SHELLS (DENSE) GREEN FI NE SANDY SILT WITH SHELLS (DENSE) I GREEN CLAYEY s I LT (DENSE) LOG 170 41 190 55.7%-65 38* 210 36°4%*83.5 --is** 220 51.0%-68 Z..\* 230 58.0%-61 --22* 240 250 56.0%-64 26. 260 98% 270 --30. 280 23111 290 --300 CONTINUED I i '--I, GREEN SANDY SI LT (DENSE) F I NE SAND GRAD I NG OUT GRADING TO CLAYEY SILT GRADING WITH SAND Gf'EEN CLAYEY GLAUCONITIC SAND (DENSE) 52 u 310 40 1111 320 330 340 BORING BORING COMPLETED 8-30-67 NO CA5 ING USED WATER LEVEL AT 191 ON 9-12-67 500 LB, WITH 1811 FALL DAM*a&MOOR* APPL.ll!D EARTH SCIENCES FIGURE 2.4-98 DEPTH IN FEET !:l ii! BORING OM 3 SURFACE ELEVATION+94.21 BLOW COUNT SYMBOLS DESCRIPTIONS 0 I o-3 I! 20 2 I! J 0-,.I 0 4 50 154 I! 60 -'0-31 I! *-...., 10 I! 90 -163 I! 100 II 0 18 I! 12 'G 0-52 I! 14 v 34 II 150----LOG ---._.__ --BROWN SILTY FINE SAND (MEDIUM DENSE) GRAD I NG WI TH MED I UM SAND YELLOW I SH-BROWN CLAYEY SANO (MEDIUM DENSE) GRAY SILTY CLAY (STJFF) GREEN S 1 L TY FI NE SAND (DENSE) GRADING WITH SHELLS BREEN FINE SANDY SILT (DENSE) GREEN SILTY FINE SAND WITH SHELLS (DENSE) GREEN FINE SANDY SILT (DENSE) GRAD ING WITH SHELLS SOR I NG COMPLETED 8-30-67 NO CASING USED WATER LEVEL AT 231 ON 9-12-67 350 LB, HAMMER WITH 2411 FALL OF BORING --***llllOOm* APP'UED EA"1'H FIGURE 2.4-9C DEPTH IN FEET BLOW COIJNT .., I 0 12 C'l 0--3 0 231!1 4 0 3gl21 '() '() 54 70 27121 0-8 90 38. 10 0 33
• 110 12 0 56. 130 52. 140 ----BORING OM 4 Sl//IFACE ELEVATION+ 44.51 SYMBOLS OcSCR!PT/ONS BROWN F J NE SANDY SJ l T (MEDIUM FI NE SANDY S 1 LT WI TH SHELLS 'DE'<SE) --GREEtl,I SILTY FINE SANO WITH SHELLS (DENSE) I GREEN F' I !\IE SANDY SILT ','/!TH SHELLS (DENSE) GREEN SI LT (DENSE) GREEN CLAYEY SJLT (DENSE) i / LOG OF CONTINUED 40 150--ii. 16<; 36* 170 180----j JO* 1901----37* 200----J I */. 210 I 48* w 220----BORING -_ _, GREEN SILTY FIN( SAND WITH s**ELLS (oEl\:SE) GREEN F !NE SANDY S l LT {DEl\ISE) GREEN SILT (DE'\SE' GREEl\I CLAYE.,.. SI LT !oENSE) BOR\f\IG CCf...,='LtTED 3-3'-67 NO CASll\IG vSED 51'*} LB. W1""1-! 8' F'ALL DAMES 8 MOOR* APPLIEO EARTH SCIENCES FIGURE 2.4-90 DEPTH IN FEET BLOW COUNT 0 I I !!I 10 20 5 I! 30 60 I! 40 50 51!1 60 151! 70 80 61'!1 90 100 541'!1 110 11 l!I 120 681!1 130 140 36"' 150 23!!1 160 170 BORING OM 5 SURFAC£ £L£VATION +118.91 SYMBOLS 0£SCRIPTIONS I 7r ..... BROWN FI NE SANDY SI l T (MEDIUM DENSE) /Sr 'O BROWN SILTY FINE TO MED I UM SAND (MED I UM DENSE) 190 200 (GRAD I NG MORE DENSE) 210 BROWN ANO GRAY FI NE SANDY CLAY (ST I FF) 'O GRAY CLAYEY SlLT (MEDIUM DENSE) 23 0 24 u 250 26 0 GREEN F J NE SANO Y IL T WI TH SHE l LS (DENSE) 211 'O 28 0 --GREEN SILTY FINE SAND WITH SHELLS (DENSE) 290 30 0 -GREEN Ft NE SANDY SILT WITH SHELLS (DENSE) 31 0 GRAD ING LESS DENSE 32 0 330 340 LOG OF BORING 14 !I 15 L'I 13 L'I 16 !!I 50 L'I :15 27 L'I 45 !!I 16 !!I 14 I! 106 l!I CONTINUED >---GREEN CLAYEY SILT (DENSE) GREEN FI NE SANDY SILT (DENSE) GREEN CLAYEY SILT (DENSE) GREEN CLAYEY FINE TO ME6 iUM GLAUCONITIC SAND BORING COMPLETED ON 9-1-67 NO CAS I NG USED WATER LEVEL AT 821 ON-9-12-67 350 lB. HAMMER WITH 2411 FALL DAM*aa MOO** APPLIED EARTH SCIENCES 0£PTH IN F££T 0 BLOW COUNr 34.i%-72 14. I 52.6%-70,7 * ;7.2%-76 0-----21.4%-96 32. 28.4%-99 19 1'!11 f 0-40-10 II 20. ii o-1tl-32-2%-72 ! 4 50 _ 85L 17 37. 26.2%-97 o-40.0%-70 54* 24"' 0 55. 41,0%-77 19 38. -*v 9 1'!11 q,Q%-78 29. 90 10"' 42.Q%1a5* 13 i! 44.0%-75. 100 43. II 0 69.0%-'58 56. *-.., 12 63-8%-62 13 o-57 ..-irJ 14 66.6%-58 52. *-v IS -v 16 44.9%-78 34. /'?V----LOG BORING OM 6 SIJRFACE £'L£'VArtON+48.0' SYMBOLS OESCR/PrtONS 1 ! I YELLOWISH-BROWN FINE SANDY S.ILT (MEDIUM DENSE) BROWN CLAYEY SI LT (MEO IUM OENSE) REDDISH-BROWN FINE SANDY SILT (MEDIUM DENSE) BROWN SILTY FINE SANO (DENSE) GREEN FINE SANDY SILT (DENSE) GRADING WITH SHELLS GREEN SILTY FINE SAND WITH SHELLS (DENSE) GREEN FI NE SANDY SILi WJiH SHELLS (DENSE) GREEN CLAYEY SILT (DENSE) GREEN F I NE SANDY SILT (DENSE) BORING COMPLETED 9-5-67 NO CASI NG USE;:D WATER LEVEL AT 301 ON 9-12-67 500 LB, HAMMER WI TH 1811 FALL OF BORING DAM*B&MDOll* APPLIED EARTH SCIENCES FIGURE 2.4-9F OEPTH IN FEET i:i BLOW ;;\ COUN7 0 81!1 10 171! 20 3CF I! 40 41!1 50 60-I I I! 70 131! 80 9()-311!1 :;,: I I 10 '20-I. 951! *-*v 13 461! *-*v 14 15 o-591! 16 *(/
• 11 I! /'!() BORING OM 7 SURF'ACc cLcVArlON +99.01 SYMBOLS OESCIUPTIONS BROWN SILTY FINE TO COARSE SANO (MEDIUM DENSE) -BROWN AND GRAY F I NE SANDY 5 I l T WITH TRACE OF CLAY (MEDIUM DENSE) I I I I ! GRAD I NG TO GREEN IN COLOR ' . I GREEN SILTY FINE SANO WITH SHELLS (DENSE) GREEN FINE SANDY SILT (DENSE) GREEN CLAYEY SILT (DENSE) LOG OF 17 0 18 0-* 15 I! () 19
• 15 I! 200 -210
• 10. '0-*9 I! 23 0 24 o-*251!1 0 25
• 151!1 26 0 27 o-*s l!I 28 u *3 !! 290 -*g I! 300 310----BORING CONTINUED H I : -I I I ' I I ' ' : i I I I i GREEN FI NE SANDY SILT (DENSE) GREEN SILTY FINE SAND WITH SHELLS (DENSE) GREEN FINE SANDY SI LT (DENSE) GREEN CLAYEY GLAUCONIT IC SAND (DENSE) BORING COMPLETED 9-8-67 NO CAS 1 NG USED WATER LEVEL AT 151 ON 9-12-67 350 LB, HAMMER WITH 2411FALL
• 800 LB. HAt.MER WITH 12"FALL DAM*&& MOOR* APPLl!:C E:ARTH SCIENCES FIGURE 2.4-9G DEPTH IN FEET ::i ii! ::;; BLOW (':I COUNT .... v 11* /"" -14* 20 30--618 40 488 50 60-2911 70 328 a: go-488 I --408 I 10 I. '20-238 I. '30 348 4,; I 50-I 48* 160----LOG BORING OMS SURFACE ELEVArtON +18.31 SYMBOLS DESCRIPTIONS BROWN SILTY FINE SAND (MEDIUM DENSE) ---->-----GREEN FINE SANDY SILT WITH SHELLS (MEDIUM DENSE) GRAD ING TO DENSE GREEN CLAYEY SILT WITH SHELLS GREEN FI NE SANDY SI l T (DENSE) GREEN CLAYEY SILT (DENSE) GREEN FI NE SANDY SI LT (DENSE) GREEN CLAYEY SILT (DENSE) BORING COMPLETED 9-6-67 NO CASI NG USED WATER LEVEL Ai .)1 ON ,..-12-67 500 LB, HA'-MER WITH 1811 FALL (DENSE) OF BORING DAM*e8MOOR* APP'Ll!:D EARTH SCIENCES FIGURE 2.4-9H 0£PTH IN F££T l;'l 0 20 30 40 BLOW COUNT 70
• 21.()%-8'} 64. 50----60 70'----14. BORING OM 9 SURFACE ELEVATION +124.61 SYMBOLS DESCR!Pr/ONS YELLOW-BROWN SILTY FINE TC COARSE SAND (DENSE) SANO GRAD I NG LESS COARSE GRAY CLAYEY SILT (DENSE) I GRADING WITH MORE CLAY 80'-----111111 100----13. ' 110*----ri ! 120-1a* 13:' ... 24. 140*---"' 150--42 911.1.1..1,a__j LOG GREEN FINE SANDY SI LT WITH SHELLS (DENSE) GREEN SILTY FINE SANO WITH SHELLS (DENSE) GREEN FINE SANDY SILT WITH SHELLS (DENSE) BORING COMPLETED ON 9-6-67 NO CAS I NG USED WATER LEVEL AT 401 C>I 9-12-67 500 LB. HA.e.IER AT 18" FALL OF BORING DAM** B llllOOR* Afl'PL.11!0 EARTH SCIENCES FIGURE 2.4-91 0£PTH IN F££T BORING OM 10 SURFACE ELEVA rlON +56.61 BLOW COUNr o----18
• SYMBOLS 10---12-=*,,..I ....,.,"1----l 9. ......... ---! 20---15-*"'" 24
• I 30-__;_9.....,.*wu 62
• 29 40--24. 20 " DESCRIP710N$ REDO I SH-BROWN FI NE SANDY SILT (MEDIUM DENSE) GRAY SILTY CLAY (STIFF) GREEN CLAVEY SI LT (MEDIUM DENSE) GREEN SIL TY FI NE SANO WI TH SHELLS (DENSE) 0 14 44
• 40
• 150 40
• 75-6%-55 245
• 160 16
• 24
• 170 27* 20. 180 30. 24. 0%-<J2 11.j.j;j.l___J 21
• 50---6-1'.'1-I GRAY FINE SANDY SILT (MEDIUM DENSE) 23
• 8 I! 60 200
• 33. 3%-9213 I! 37
• 70 --10_0+_11 3" 102
• 80----'2""0-" 42. 13 l'I 34 l'.'I "'"'"Ul---l 34. 100------£4 I!. ............ ___,, 40. 18 II II 0--"';\q"'-'"""2%"--8""-'l.& I 35. 32. 120--34-.-1 34. /30---3-8-.--1111 38. 14..,ri-,_ ___ GREEN FINE SANDY £iJ..T WITH SHELLS (DENSE) GRAD I NG OUT SHELLS GREEN CLAYEY SILT (DENSE) GREEN CLAYEY FINE SANO WITH SHELLS (DENSE) GREEN FI NE SANDY SILT (DENSE) GREEN CLAYEY SILT (DENSE) LOG OF 190 29
• 54-5%-67 32. 210 220 28. 230 240 28. 250 86.7%-49 32. 260 270
• 280---BORING CONTINUED BREEN FINE SANDY SIL.T WITH FINE SANO (DENSE) GREEN CLAYEY SILT (DENSE) GREEN CLAYEY FINE TO MED !UM GLAUCONIT IC SAND (DENSE) BORING*COMPLETED 9-11-67 NO CAS I NG USED '
• WATER LEVEL AT 131 ON 9-11-67 500 LB. HAMMER AT 1811 FALL DAM*ea APPL.I EC l!:AR'Tl-t SCIENCE. J Fl6URE 2.4-9J

..... lai i::: i.. ..... .... ,.. ! Gi I c !! ::0 I m GI N I '.Pi I I 0 I J> I "-I Cl-----*-----------* z er 0 "' "' --------.. z +100 -= .,. -*--* ----==---=----...:;..-PLEISTOCENE TERRACE DEPOSITS -------___ ....&... _______ _,... -----_ --=:----__ -.... PRIMARILYS1Li -+50 ...... ;; .... ruiinr.w.;.;--SAND WI rH SOME GRAVEC-:.:---_\ ,/_. .. --__ ..,,. ... 0 ------------=-----_,.I-..=:---::;:-;::;=... _-_--=-----........,...... ' H ------== ----=-o----=..,...CHESAPEAKE-GROUP. ----PR I MAR IL y SANDY AND CLAYElr' SILT . _ _.:.. ...... ---!::::..--_:::;;;.._--_.:_-'-':---50----::;--__ -_--.. WITH SOME INTERBEODEO SANO -_ -----*--1l: _-.__-_ -: _-_-=-.!_HE, -_-:::: -100 __ :;; _ ----=-----=----150 --.... --_._ _--:-_, -:-:::------= ----=----................. _ ................. _ .._ __ -....... _ ....... --=------=.._..::+/--* -u -..i....:-:_ ""r--=----. ----. --___. ............ ._..._ -250---*--------***--*--*------* -JOO -------------*--------------------GEOLOGIC SECTION A-A HORIZONTAL SCALE IN FEET 500 0 500 1000 I II I I I "11 a; c ::u m N I 0 ID I I I * (II I 8 I ' .... "' i:: +150------=-............. --=--o --50---"' N -------.!:;------------------------------(!I z "' =----,.._-------PLEISTOCENE TERRACE DEPOSITS " z L-_PRTMAifi LYSIT"fANo --=:sANO w* TH SOME GRAVEL -----:u.--::'>-:_ ....... ... --=* p:..:.-..-=:-... !.= ... : p----::::=...::::::::::-. ... ...... -"'=--=--.=:::5:.;--_ ,., -Vf -----_-; -__::::_ --=---=.. -LAYfHS IN THE lJPPEH ---------=-=--.......=--..... ,...___;::::;:-__ 1<1 -JOO ---:;-: .. i;J -150 ---+--*=--:.::-=--...... -:-:-;-__.:::._ ...... . -----* -200 >= '/.'-; GEOLOGIC SECTION B-B HORIZONTAL SCALE IN FEf."T 500 0 500 1000 "'II Ci) c ;:u m I\) I i5 0 I ! I *

• I g I I .... a.: j:::: 0 " z ---I 25 "' z "' ?=< -=--= -=;;.-w.L -:: _-----,.,;;;\.L----=--'""""' -..:r,.-............................. !!""""" ----=---U!!!> :._ --::: ...::::::.... -::. ::: -:.::.-::..-:-=-:..,_ ---=--..... --= ---=------=-----=-----:"". -_-:-CHESAPEAKE GROUP" -_--:_ -,PRIMARILY SAND't ANO CLAYEY SILT ----50 .... ---TH SOME I NTEABEDOED SANO --... _ _ SHEL.L LAYERS IN THE UPPER PORT ION -----------.:..----=--.._ .,....._ ij_ ,...._ --100 _ -....._--r--......;,;__;_ ""-'""--* --*---* ........ -----150 -=-----==--... : . .-:..._ -...::. .--* --=-... _ --= --"""':"'"----=----::::: ','.;.}.{*:*: .. :*:_--:,:'-(;t:::; .. : *-:> *. * * '._.-,'._:): X-' .. _ _,:.;:./i-GEOLOGIC SECTION C-C HOtllZONTAL, SCALE IN Ff[f SO 0 SO 100 150 200 2 50 tHJiliid I I

"' ... "' 0 :!; c ., z tl ., .... Q. ,...... Q. " 0 "' "' ., " < "' .. i'i "' ... "' 0 CJ 0 :I CHESAPEAKE BAY BROWN SANDY SILT YELLOWISH-BROWN CLAYEY SANO BROWN SANDY SI l T REDO I SH-BROWN FI NE SANO 1 '#ELL CEMENTED YELLOW I SH-BROWN A"'O GRAY SANDY SILT GREEN SANDY ANO CLAYEY SILT GREEN SANDY SILT CEMENTED GREEN CLAYEY SILT CEMENTED LAYER GREEN SILTY SAND WI JH SHELLS GREEN SANDY SILT SCHEMATIC CLIFF PLANT AREA SECTION -+80 -+60 -+40 -+20 -0 ... ., ., "-'= z 0 ... < DAM*ea MOGA* APPL.ll!:D EARTH SC:IENC:ES FIGURE 2.4-11 CLIFF FACE PHOTOGRAPH PL.A.NT SITE VICINITY ............. EA"'TH SCIENCl:9 FIGURI=' ? 4-1?

FIGURE 2.5-2 PIEZOMETRIC SURFACES IN CALVERT COUNTY ..,. I J ' \ I \ " \ \ \ \ ' ' \ I I I I I ' ' \ \ \ ', '\ ' ' PIEZOMETRIC ::::i, ' f>US MAP ... ,9 .... ,,.ou A POltTIDH D' usns. *A,HIHOTON, D*C*; "'""n,.,t.ND: t/IRGIHIA ICJ5?, NJ 18-lt, lllf1rln,,cr '0" .. IP'!OMPUIC DATA IS l MAlll"l.AND M otnLDQ'f' IJINCS AHO ** uc.. -ftCSOUACts DULL.CT IN NO* 15* J /t-'rJ'I \;II .. , "" ,. ..... *,. ""1t*t*tt1, ..

• liERR/".'t; .!iAl" " ... I I I 'il&H.UI 111111['.lo 0 2 3 4 I I I I 1.)' \ 7\ \ .:, -\ ... *' '" COUNTY I' I 5 ' ' I '1-t 1 r:> \ r v*t -*8BMOO** Revision 0

 ;,::..' ', ' ,, ' \ RE: FERENCE: THIS IMP W/\S FROM /\ PORTION USBS QU/\OR/\NGLE COVE POINT> IMR'l'L/\N9* 1943* MAP OF AREA SHOWING SURFACE HYDROLOGY 1000 0 FEET 2000 4000 *!*!*!*!*!== ==--1 DAM*e&MOOR* APPLIED EARTH !ICIENCE!I FIGURE 2.S-1 PUBLIC l' ,f ./ I *; 6 "' BEACH ,, JI !i if .;. ', WATER SUPPLIES ------IN ISLAND ( > ..J ' DAM*a&MOOR* FIGURE 2.5-3 KEY: 010 ";. ; ,. 1.._; . .. / ... '-* APPROX !MATE LOCATION OF WATER 'NELLS CATALOGED BY THE GEOLOGICAL SURVEY , MAP OF AREA SHOWING KNOWN WATER WELLS ADDITIONAL WELL.S LOCATED BY DAMES I!: MOORE REFERENCE' TH IS MAP WAS PREPARED FROM A PORl I ON OF USGS QUADRANGLE COVE POINT> .. AR¥LAN9, 1943* 1000 0 FEET 2000 4000 ' ., N 0 T E: NUMBERS REFER TO TABLE 2C-4 SHAl..1..0W WEt.1..S LESS THAN 50 FEET DEEP ARE NOT SHOWN. ...... ft....,..p APPLIEO EAll'IM SC:IENCES FtC3URE 2.5-4 U, S. SrANtM/fD SIEVE SIZE $.JN. Vil.JN. Na. 4 NO. /0 NO., 40 NO. 100 ..... ... ........... !' ! ; : *; I 11 l i 80 ,1-i,-,...i:i-i-i:il-* ....... ---+-,-+, .J.+-H-f w ........ Ii 11 \ ' I ; 'i ;1 l"INE SILr Olf CLAY U, S. SrAN/MMI SIEVE SIZE $/ii. IN. ND. <I NO. 10 NO., <10 NO. 100 ... $IN. I ! !l ! \ i Iii! ; J,' QI 60 , I 1 ! \ l' * * *; I so ,-t\rt-+-:t--t--r-___,,.-t*1.-.,,_;,-+-1 -t-t--r*--r--... 40 'i ;1 I --I ill ti $0 I l/ll I / Ill :illllll"""BORING 451+--f---t+t-HH--+--i ,; I 1 ! 0._.._ __ ..&,j,.l;.l.U.._1.-..._ __ ____ ..... ...., __ ..... ....,..,...._. $00 100 10 1.0

• 0.1 0.01 ,.AIN 5F£ IN COBBLES ;.,_ __ __ ___,, __ -,-___ C°"RSE FINE 1c;oARS!j MEDIUM FINE SILr Olr CLAY PARTICLE SIZE ANALYSES DAMI 18 .U MOOR* APPLll!:C E:ARTH SCIENC£S FIGURE 2.5-5A U.S. STAN/MlfD SIEVE SIZE /J JN. !J/4, IN. NO. 4 NO. IO NO., 40 NO. ?<JO *O :L. I: ::;I! i J ; t-' --11\+--b-' NG1 OM-I 2251 -++++-+-+-........, i.. 110 ' ll '1:: I I i i i i *,., * : \ * * .;. 1 I , 70 i 1 ....,"ti,. / i/ I t i,i: / l--+---H-iH-t-+-1-+--+--BOR' NB ** @' 330, ,.."'--++++l-!'l1 rl-<!+--il!_1+--+h.,_l I_. r-+1_,___+1"++1++-+--il 60 I: ; I I :: ;1 ; ' i ,;, . I ' I * ! ; '!'_ i .\ '.;. * ' I I, , I ' :*;;I "' :1 ! I I :, i ";; ' Ii I : '* ! :; I I ' I I : 0.1 I 0.()/ 0.00/J COiia.ES __ COAlfSE FINE MEDIUM FINE SILT 011 CLAY I/. S. STANIMllO SIEVE SIZE !J IN. !J/11, IN. NO. 4 NO. 10 NO. 40 NO. too .* --.,.,.'l'T"!.,..,_, 10 I-.1 ! 1 , "' 11'1. I : :;1 , I I E *,: i i ' . ! I'\ Ill' ; . i I ... 11:-..... .. .. _ _,_:=_-==--+:=H1::.= ..... :_:+i_-;+_::._-1: !' i I!, ; ; : :'Ii 1 \ ** ....... it , ' * : I !l I \ I \ ; :ji 1 '
• i;: 1 * '; i ; ,\ ! \. \ ; !* i ! I; I
• iS
• r ! i I I\ . \. ' '*1 l ! i I ; ljj I I I. I ' ..... ! I I I i 1--+---H-i:++-++-+--E---f+_ -rlHl-1'-+-:*,f--+--+Hi IHi-+:-+i;r-,f---!-'-...-'-+M' ,-<t,.,.-t,.-..a.oR: NG ;r:J.<*5 t--0._.._ __ __ .._ .... .-.,l'-......... __ 300 100 10 1.0 ! 0./ ::i'. 0.01 O.OO!J c>>rAIN 5FE IN 11/IL'f"'ETElfS GRAVEL SAND SILT 011 CLAY COllllLES *,*. COAlfSE i FINE MEDIUM j FINE PARTICLE SIZE ANALYSES DAMES 8 MOOR* APPL * .IEO EARTH SCIENCES FIGURE 2.5-58

.. '" '" 25: to.. s:: 2::: 225: DESCR I: i * : .... PLEISTOCENE SIL l1 CLA "I ANO RAVEL ! **t:::CS"'lE, ..

• j =A ... ECCE'\iE ...... ... ) SANDI Sil.. T ANO : ..... ---CLAY* *. . I .. *-:*::****1.*: j I ! CRETACEOUS SAN01 SI LT ANO CLAY I :NA:.. .,.,.tivE vn::::c * ':" .. (FT /sEc) (\4EASUREO) :: t.:: _ ,s:: AA* I: .* 32 1EST",'A";'E0' :.L.-: JEL:-C: T .. .... : ::**::\,, T:::' '.'EA51.=REO' -Ao<;E .... FR:t! '}, 9L5' Ex=:..:S SCu"IOS 5'"1AL ... ::. r'tA"'ER, scc1E-v :F A'IERICA ,_.E'YtC i R 2-COLUMNAR SECTION SHOWING GEOPHYSICAL DATA UP>t "' WE' Gi-t* ( .. es.' cu.n.) 3: -c *35 'Es* *V.A"'ED' DAMBS 8 MOOR* APPL! EU EARTH SCtE:NCES FIGURE 2.6-1 r3 !'l NOTE: THIS MAP WAS PREPARED FROM A PORTION OF TECTONIC MAP OF THE UNITED STATES BY THE UNITED STATES GEOLOGICAL SURVEY ANO THE AMERICAN ASSOCIATION OF PETROLEUM GEOLOGISTS, 1962. _ ' '-V'-'1' ;-\-,/, Z (: (/ *l# '"7; W',< ****1-*1-*-1
• I // '/ 0 : )""',, 7 I 1*i LEGEND: EPICENTRAL LOCATION J ANTICLINAL AXIS ---t-SYNCLINAL AXIS + DOME I I I I I I I NORMAL FAULT 10 0 MILES 10 20 30 40 -------* * * * ' THRUST FAULT ---ON TOP OF BASEMENT MAP DAM*e B MOO** EARTH SC11ENCl:S F'IGURE 2.6-2 0.01 0.02 o.oo 0 <o 20 C'I';-)-' )' & C' 40 60 100 1------------120 140 160 TIM£ IN SECONDS 0.03 0.04 '
• I -\ \ L .I.. ,, '" -<\ ""' -0 UPHOLE SEISMIC SURVEY 0.05 0.06 DAM*Sa MOGA* APPLIED EAs;fTH SCIENCES FIGURE 2.6-3

.., .. g g g d Cl) Q z 0 (.) CD l&I d Cl) 9 Q 0 a: l&I a. N .., .. in co CD S2 S!OICDP.. CO In 'I' .., N a ONO:>lS/S::IH:>NI NI A.l.10013A RESPONSE SPECTRA OPERATING BASIS EARTHQUAKE N d 0 S! 0 CD 0 co ii i 0 .., 2 Q z 0 (.) Lii Cl) .... Cl) Lii (.) > (.) > (.) z l&I ::::> 0 l&I a: ... DAM*88MOOR* FIGURE 2.6-4

a am =en ; 1J *o *Z t en ;m l'l > en = 1J 111 m = () > -I = ::a )> "'II m i c ::0 m *

• N Cll a, I I (JI I
• a !50 40 30 20 ' 6 z 5 0 4 It) .... : 3 % u 3!: ::! 2 0.2 3 PERIOD IN SECONDS 1.0 0.8 0.6 0.5 0.4 0.3 0.2 0.1 .oe .()6 0.04 0,03 0.02 0.01 z 3 4 8 10 20 30 40 50 60 80 100 FREQUENCY IN CYCLES/SECOND

-_ ... "..:. . ; i..-** .. . -r_ . .. : . :.*_.,, ' ,. '"** "'* r. )' .. *" .. J *.f ... I i--. 1 ** " ' c E s !" "* A p E A K E .. qs" "*'? GA.A.fHIC. *' 8 A 'j/ t)'; ._,,.. ... ,,_ y ,. . _, !I l -----.. I \ "'*, -* ' '-'*. . 1;.* l I I .. I I I l ' ', ,, y *

• B*2l?7P t: L EGENOj -..-: . .... -B *I B*f> DRILLED JIV BROWNi ' .l-*'-*oRlllED BY GIRQ:t.ER iout.I0-'1101-i :' Etc.fLOR-.TION p -DENOTES PIEZOMpER . ll<HAL"\o :
• f f S -pw1, ewi t ews OR***U> ev l!.ROWNt
• I f *t l > lkolll*** ........ TltU[ . "::-.: UOllTH
• 1 * * ,:: 411:.* -' -.. :.. ..... :J/ .-/ GENER-.L NOTES I F1'0M 5UtlV'EY 0, JU'<[, 19!07 PREPARED 11V JR i TRVt ANO GIUD * ..: _,. l.AAR fLAMO !>TA'TE , a_ PLANT A.NO E.A.$,. GftlO E5TA&Ll.!."ID A'!. PR:l..!£.C:T GRID .t 4. .. .. ae_:. FOR Ri.VltW. SOIL ** A.NO 'lTll(R SOIL DA1'A. ***w et INSPECTEb ev A.N*"OIN"fM ENT AT -atcnn.L A\Soc.b.Tll _ OR f!hLTIMORE GA>;. ANO EL£CTRI<; *:c .... \P-.*IY' '.'\'".l'J :)rFICE.$. * ' REFERENCE OWGS BORING PLOT

-, / ,. \ _, .. J ,_ , J ***5. :--.::,---.,....___, ** ... " I 000 -t -l 1; /j I I E.1t.OOO I ' '* / "'.-.---...... ei + \ I *------i ... : 0 ____ ---t i "' J-.. )-----1 -_;..-** ., ,--1 I \ /* i r I {. '!' '/ *'* .. '.* ' ,* .. ' .* z I ; l ;--4 'I ----! \_ + --!----* --"--r--* .. ,It-' I .---1-/ ----'. ! \ " * .... '!> ., + I ,/ ' i { .. *.---+--/, --/"'" ,(. ',:0wo2l / "-,, u;--.. / _,,.* . ,' .... J * *-3'" ,. _ .... ,. I + 0 .Q --g ;! / . I .. ... 91 *s.o O=> "' '2, 1 !-=>.oo ,.,..,!> " 10,.(!0 00 .JOw ... 0 00 ,..: "' o, > :.o )._Fl; 8 "' .0,43'0. 00 " "'!-" 10,7!-000 ,.-e ... 101 00 ,, e I I.: oo "'" N . ""' h ... .. .. .. *O, c2:} 00 ... 10, 00 ., " .. I I "";J UJ ..i:"J*':' IJ .. 70() 00 ,_ ., "' " I 00 N"!-., l ,.:.z)(.1 O> .,, ZJ .. t., c" 2 .. c, ".;:::! oo ,, .. n. ' _N. 61 760_00_ l.J *o; OD ""' ,,., :-:. ... :::.,.:.°'.::.co ,!> "' o, -,7i;:) .JO N 5 .., c .. '2;1' 00 v .. 2d " 10, *7!>, co "'"' 'H ... ... ' ,_ "' c, :"!-.oc. -;z ,.,., :.;. ,, .. , I I I i,. I E. I i. I I i,. I L " i-.. § !I .. )lrioo 41o41:DO 4ao !!!' 21 -im;' 1l!!!iimi!!!1 !!!!!!!!!!!i1 M:ALI!. * § ; .. ! § = i;; .. ... .. .tJoTE..S l -\.I l\.OCO l R£F'ERE.UCE. ORAWl'-JGS I Sk.<*171' OFFSt-eORE. '.)FFSHORE ShEET 2 .JF-2 ! .. : : i GRAPHIC BORING LOGS BALTIMORE GAS AND ELECTRIC COMPANY CALVERT CLIFFS NUCLEAR POWER PLANT SUPPLEMENTARY SITE BORINGS GENERAL NOTES 1. ELEVATIONS REFERENCED TO MEAN SEA LEVEL. 2. NUMBER ADJACENT TO BORING LOG IS STANDARD TION RESISTANCE . . 3. STANDARD PENETRATION IN BLOWS/FOOT OF A. 140 LB. WEIGHT FREE FALLING 30 INCHES. 4. BORINGS BY GIRDLER FOUNDATION AND EXPLORATION CO., JUNE 1969. LEGEND Greenish Gray Silty SAND Greenish Gray Sandy SILT * * * * * * * * * * * * * * * * * * * * * * * * * * * * * *

• FIGURE 2. 7-4 IAlTIMORE GAS AND ELECTRIC COMPANY !ELEV. DEPn DESCRIPTION -VISUAL 2.7 0 o-Greenish gray silty sand with shells 10--10 --m"een!Sh gny ST!?)>" Siffill' __ --20-Greenish gray silty sand -20. . 30--30 ---------------Greenish gray silty sand with 40-trace of shells so_ Greenish gray slightly sandy silt -so -60 Greenish gray slightly sandy slightly clayey silt 80-i..--80 -Greenish gray sandy silt 90 -*90 -100 -100 ----GRAPHIC IORING LOG SUPPLEMENTAL sin IORINGS CALVERT CLIFFS NUCLEAR POWiR PUNT BORING NO. 210 GRAPHIC rcNC 111ATION RESl:s I ANl.;C LAB BLOWS/FOOT N LOG 0 10 30 50 70 100 +10C TESTS * * * * * --.... _ *1:1. * * * * * * * *
• i-. .... * * * * .... .., 100 * * * *
• j * * * * * * * *
• I * * * * " * * * *
• 74 * * * * * * * * * * * * * * * * * * * * *
• 66 * * * * * * * * * * * * * * " * * *
• I'
• 100 * . * * * * * *
• IJ * * * * * * * * * * * * *
• LI 90 * * * * * * * * * * * * * * * * * * * * *
• 63 * * * * * * * * * * * * *
• I\ * * * * * * * * * )I 84 . * *
• 4111/[/ * * * * * * * * * * * * *
• 56 * * * . * * * * * * * * * * * * * * * * *
• 68 ' 69 73 / I' 54 65 79 81 11 74 78 [\ 93 91 rnn FIGURE 2. 7-1 I I I HLTIMORE GAS AND ELECTRIC COMPANY -'ELEV
• DEPTI DESCRIPTION -VISUAL. .<C,b u o-Greenish gray slightly clayey silty sand with shells 10--10-Greenish gray silty sand with shells 20--20-Greenish gray silty sand 30-Greenish gray slightly clayey slightly sandy silt 40--40-Greenish gray clayey silty Greenish gray slightly clayey 50-silty sand 60-Greenish gray slightly clayey slightly sandy silt 70--70-80--so-90--90 -100--100-Greenish gray clayey silty 110-sand -110-Greenish gray slightly clayey ' sandy silt -120 -120-Greenish gray slightly clayey silty sand --GRAPHIC IORING LOG SUPPLEMENTAL Sin IORINGS CALVERT CUFFS NUCLEAR POWER PLANT BORING NO. 211 GRAPHIC l'l:Nl:INAllUN Hl::::ilSlAN!.iE LAB BLOWS/FOOT N LOG 0 10 20 30 50 70 100 +101 TESTS * * * * * ""i--, 33 * * * * * * * * * ""r-... * * * *
• It 70 * * * * * * * *
• J * * * * * * * * * ** * * * ...... 60 * * * * * * * * * ""' * * * * * * * *
• r--* * * *
• 100+ * * * * ' * * * * * * * * * * * * * *
• 150 * * * * * * * * * * * * * * * * * * * . . . -I 150
• 150 * * * *
• I * * . * * * * .
• 100+ * * * * * * * * * \ * * . * * * * * * * * .
• 150 * * * * * * * * * * * * * * ' 150 I 150 I/ii' t.-i 150
• 73 77 11 66 } 75 65 71 11 75 .-* * *
• v 71 * * * * * * * * * .. *.*-*-* .. 57 61 . . . . * * * *
• I\ * * * * * * * *
• 82 * * * * * * * * * * * * * * * * * * *
• 84 . * *
• FIC URE 2. 7-6 I I° I I lllTIMORE GAS AND ELECTRIC COMPANY 'ELEV .DEPTt' DESCRIPTION -3.1 0 0-Greenish gray slightly clayey silty sand with shells --------Greenish gray silty sand with 10-trace of shells . 20--20-. 30-Greenish gray slightly sandy clayey silt Greenish gray very silty sand ----------40-Greenish gray slightly clayey silty sand with shells so-Greenish gray slightly sandy -so-silt 60-Greenish gray slightly clayey slightly sandy silt 70--70-Greenish gray very slightly sandy silt 80--so-90--90-' 100--100--. -. -. GRAPHK IOllNG LOG SUPPLEMENTAL SITE IORINGS CALVERT CUFFS NUCLEAR POWER PLANT BORING NO. 212 GRAPHIC rcru: I nA 1 IUN Ht::il;:, I ANi.t: LOG BLOWS/FOOT N LAB 0 10 20 30 50 70 100 +1M TESTS * * * . ' ,....., * *
• e e I I--.. J!> * * * * * * * *** I'-. ..... * * * ' 81 * * * * * / *
• e e I / ** * * * ** / * * * * * * *
• e: I--.. 4S *** *
• e I * * * ** r--.. .... * * * * * * *
• 110 * ** * *
• I * * *** * * * * * * **** ***
• 160 * * * *
• e e I .... * * * * * ** v * * * .v * * * *
• 59 * * * * * ** * *
• 67 * . . . -' * * *
• 55 * * * * * * * * * " * * * * * * * * * 'l * * * *
• 85 * * * * * * * * * * * * * * * * * * * * *
• 86 * * * * * * * * * ' = _a 160 --160
• I./ vi/ 57 \ /l) 79 ,/ 49 I\ 68 , FIGURE 2. 7-7 I I I IAlTIMOIE GAS AND ILICJIK COMPANY !ELEV. IDEPT* DESCRIPTION -VISUAL 2.8 0 Greenish gray sandy silt 0-Greenish gray silty sand with shells GRAPHIC IOllNG LOG SUPPUIENTAl SITE IOllNGS CAlYERT CllffS NUClfAI POWEi PUNT BORING NO. 213 GRAPHIC . .. ,." .. iluN Ht:l>ll>IANt;t: BLOWS/FOOT N LAB LOG 0 10 20 30 50 70 100 +1nl TESTS 27 ............ '""-i-.. . . . . . 150 * * * * * * * * * * * * * * * * * * *a 10-Gr;;;;rsh'"'8r'aY'fine7and7ith -* * *
• 150 * * * * * :..-i--.. * * *
• shells i...-i.--" * * * *
• i.-i,..-Greenish gray sandy silt with a:: 25 shells ........ , 20-65 60 30-56 51 40-Greenish gray slightly silty * * *
• 44 sand with trace of shells * * * * * . . . . 48 I\ Greenish gray slightly clayey 50-11andy silt ' 63 52 60-56 65 70-72 . i1 75 80-78 -so-I\ 78 "' 90-150 *90-150 100-I/'/ 69 -100-65 ..... -. Greenish gray clayey silty .. . . . . .... ' 76 no-sand ........ *, --------........ *, -110-Greenish gray slightly clayey ........ *, 77 silty sand ........ *, 120-.. * ..... *, 150 .. . . . . . . . . ' * * * * * -120-* * *
• t 200 * * * * * -FIC URE 2. 7-8 I I I HlTllORE GAS AND ELECTRIC COMPANY GRAPHIC IOllNG LOG SUPPLEMENTAL SITE IORINGS CALVERT CLIFFS NUCLEAR POWER PLANT I ELEV. IDEPTt u ... u-0 10--20-20--30--4o-40--so-so--60-60 ... 70--so-80-100--100----BORING NO. 214 DESCRIPTION -VISUAL GRAPHIC LOG Brown silty sand (fill for * * * * * * * * * * -------i.*.*.*.*.* * * * *
• Greenish gray silty sand with * * * * * * * * *
• shells II******** 1 * * * * * * * * *
• 11.*.*.*.*. lo ********* ------------11.*.*.*.*. Greenish gray silty sand lo********* Greenish gray sandy silt Greenish gray clayey silty sand with shells Greenish gray sandy silt *-*-*-*-* * * * * * * * * * * * * * * * . . . . ' It * * * * . . . . ' It **** . . . . ' la * * *
• l"t:Nt:l MA I IUN Mt::OI;:) I ANl.#1:: BLOWS/FOOT 0 10 20 30 50 70 100 +1rv \ N 150 44 150 150 73 65 69 78 83 45 52 47 50 47 56 59 55 62 65 70 LAB TESTS FI iURE 2. 7-9 I I I IAlTIMOIE GAS AND ELECTRIC COMPANY GRAPHIC IORING 106 SUPPLEMENTAL SIR *IORINGS BORING NO. 215 I nH 1 IUN CALVERT CLIFFS NUCLEAR POWER PUNT IELEV. DEPT* DESCRIPTION-VISUAL GRAPHIC LOG o BLOWS/FOOT N LAB TESTS 0.2 0 u 10.. -20. 20.. 30.. 40.. 50.. 60_ 70-' -so-80--90-90-. -100-100----Greenish gray silty sand with shells -----------Greenish gray silty sand with trace of shells Greenish gray sandy silt with trace of shells Greenish gray quite silty sand ********* .... ' ........ *, * * * * * ......... ' .... ' * * * *
• It ******** ' .... ' * * * * * ********** ***** * * * * * * ********* ***** * * * * * ***** * * * * * ***** . . . . ********** ***** e e e e I ---------, .... e e e e I Greenish gray silty sand with I******** 1 shells , * * *
• Greenish gray fine sandy silt I 10 30 -i-9::: r 50 70 i-.. ,_ c,..l.---i..--i.,.. t--..t-. i--...i-... i.,.-v I\ I ) * ) 100 +10C ... t-i.,..i.. t>> """"' a l/i.--,u, 150 31 150 200 67 36 36 52 57 61 63 53 57 63 52 57 67 65 62 FIGURE 2. 7-I 0 Rejv. 9-}0-71 I IAlYIMORE GAS AND ElECTllC COMPANY IELEV
• DESCRIPTION -VISUAL 0.4 o-0 Greenish gray silty sand with . shells 10-20-Greenish gray sandy clayey silt with shells Greenish gray silty sand 30--40-40-Greenish gray silty sand with shells *SO-SO-Greenish gray sandy silt *60-60--10-70--so-80--90-90--100-100 ---. GRAPHIC IORING lOG SUPPlEIENTAl SITE IORINGS CALVERT CLIFS NUCLEAR POWER PLANT BORING NO. 216 GRAPHIC I""' ttt::>lo:>IANl..C LAB BLOWS/FOOT N LOG 0 10 20 30 sn 70 100 +1rv TESTS . . . . , S2 ********** S6 ********** I e e e e ***** ...... ********** 61 ***** ........, * * * * * * * * * * * * * *
• 150 *** * * ,.,,.,,,. \/ 65 .-.-.-.-. ***** l 69 lo * * * * ***** It **** ) ***** ********** 78 ii **** * * * * * * * * * * * * * *
• 72 ********* * * * * * * * * * *
• t 66 ********** ********** * * * *
• 62 ***** I ***
• 63 v 60 4 48 \ 51 \ 69 I I ; 67 I 59 I ' 62 67 63 FIGURE 2. 7-11 I I I

. IALllMORE GAS AND ELECTRIC COMPANY IELEV. DEPn DESCRIPTION -VISUAL z.z 0 o-Greenish gray silty sand with shells io--10-20-Greenish gray sandy silt wlth shells 30--30-40-Greenish gray silty sand with shells so--so-Greenish gray sandy silt 60--60-70--10-80--so-90--90-100--100-. -. --' GRAPHIC IOllNG LOG SUPPLEMENTAL sm IORINGS CAL YEH CUFFS NUCLEAR POWER Pl.ANT BORING NO. 217 GRAPHIC _,... *-** LAB BLOWS/FOOT N LOG 0 10 20 30 60 70 100 +11V TESTS 11* ***. -r--17 * * *

• r---**** t -* * * *
• r--_ ;:c. 125 **** t * * * * * **** t ,.,., * * * * * ,....v **** t
• 39 * * * * * 'r-**** 4 * * * * * ........ 1"-r--. **** t . . . . 125 **** t * * *
• iecv ,.,., 43 r--.. 87 t 79 73 . . . . tv !I **** 58 **** t * * * * * * * *
• t * * * * * . . . . ' ' 62 * * * * * . _._._._, 4 57 ' 60 ll ,, 45 I 48 [\ 62 t 59 4 58 , 63 t 67 I 69 Ill 73 FIGURE 2. 7-1 2 I I I IALYIMORE GAS AND ELECTRK COMPANY IELEV. DEPn DESCRIPTION -VISUAL S.2 0 Green1sn gray very sugnc1y ** ( H 11 .. 0-' Greenish gray silty sand with shells 10---------------10-Greenish gray slightly silty sand 20--------------20-Greenish gray slightly silty sand with shells 30--30-------------40-Greenish gray silty sand 50-Greenish gray slightly fine sandy silt -so-60-Greenish gray very slightly clayey very silty fine sandy silt 70-Greenish gray very slightly fine sandy silt so--so-90--90-lOO--100---GRAPHIC IOllNG lOG SUPPLEMENTAL SIB IORINGS CALVERT CUFFS NUClUI POWER PLANT BORING NO. 218 GRAPHIC LOG 0 .-.. ,,.._, "" *-** ... ,.,. -**w-LAB BLOWS/FOOT N 10 20 30 60 70 100 +1nl: TESTS * -* -. ' * * * * * * .... ' * * * * * * * * . ' * * * * * . . . . ' ***** * * * . ' * * * * * . . . . ' ***** . . . . ' ***** .... ' ' . . . . * * * . ' --r--. 16 r-r-_ r--1-r--a 125 150 !/ ..... ...... !-' I .... ...... 41 * * * * * * . * . ' * * * * * * * * . ' * * * * * * * * ** * * * * * ,_ 27 -r---r-... l""'-r--250 e e e e I ***** e e e e I ' * * * * * * * . ' /,/I,/ 75 * * * * * * * * . ' * * * * * * *
• e I ' 67 * * * * * * * * . ' ' * * * * * *
• e I * * * *
• 11 66 * * * . ' * . * * . . . . ' It ****
• a e .. e -..... I ' 68 [\ \t 83 73 67 68 63 66 125 125 67 73 250 FIGURE 2. 7-13 1 I I IALTIMORE GAS AND ELECTRK COMPANY GRAPHIC IOllNG LOG SUPPLEMENTAL Sift IOllNGS CALVERT CLIFS NUCLEAR POWER PLANT BORING NO. 219 leuv. DEPn DESCRIPTION -VISUAL GRAPHIC LOG 6. 7 0--20--30--40--so--60--so--90--100--110--120--130-140 -o lrreenisli gray suty ** ** ** *
• i----=--'----'---\.!r:!!o!!a.!!.dLJ--1' * * *
• 1 Greenish gray silty sand with * * * * * * *
• 1 10-* * * * * * *
• 1 . . . . e e e e I li°;;e;i';h g"";;y fTr;' s;;;d ---* * * * * * *
• 1 * * *
• e
• e e I I--------------********* Gre'!!nish gray silty a and with * * *
• 20-shells e e e e e e e e I gray sandy silt with llreenish gray s i1 ty send 30-* * * * ...... * * *
• I le ******** I ...... . . . . ' Iii **** * * *
• I "' ******** I 40-It ******* *, ...... -----------..... lo **** i>reenish gray silty sand so-.... ' lite e e e e e eel lite e e e e e e. I lo **** **** l lo **** e e e e I It ******* ** It ********* -------------* .... gray clayey silty sand 19 * * * * * * * *' 60-th shells * * * *
• 70_ t;reenish gray sandy silt 80_ reenish gray slightly sandy silt 90-100-reenish gray silty aand with 110 -race of shells reenish gray silty sand 120-130-reenish gray fine sandy silt . 140 * * * *
• e e e e I It ******* *, '"-*-*-*-* * * * * * .... ' * * * * ***** I e e e e ***** I e e e e ***** . . . . . .... ' I e e e e .... ' I e e e e .... ' I e e e
• e e e e I * * *
• e e e e I ***** . . . . ' . . . . . ...... 0 BLOWS/FOOT 10 20 30 60 70 100 +toe ..... , i-...i-... ..... t-. ... I:. "" I.; ... << ........ ....... .... i...., ) ,,, 1/' t I-.. t-."' [,.. t> vi/ ' i.... .... , IJ II ' J 'r-.. r-.. v r i. * \ 11 N LAB TESTS 17 39 150 55 100 150 l 71 67 125 70 69 65 250 84 I ! 84 I i I I 80 I I I i i 71 i I ! j I 77 I ! I I ' i 250 I I I i 84 I i I I I I 82 I I I 74 I I I 71 75 71 68 87 84 83 87 I FIGURE 2. 7 87 I I 14 / -----

IAlTIMORE GAS AND ELECTRIC COMPANY IELEV. DEPTI DESCRIPTION -VISUAL '.9 0 !:;reenilh gray silty *and (roadway' !Greenish gray silty aand 0-I 10.. *10-20-!Greenish gray quite aandy silt *20-!Greenish gray silty sand 30-*30-40-*40-50-Greenish gray fine sandy silt -so-60-*60-70---*70-<;reenish gray slightly sandy silt so-*80-90-*90-100* -100-Greenish gray silty a and 110 --110_ 120 --120_ 130. -130-Greeni*h gray aandy silt 140 *140 -150-GRAPHIC IOllll6 LOG SUPPLEMENTAL sm IORINGS CALVERT ClfFS NUCLEAR POWEi PUllT BORING NO. 220 GRAPHIC ""l;l"llM n,..,, "",. LAB BLOWS/FOOT N LOG In 10 20 30 50 70 100 +uv TESTS -f\ 7 lo ** -. *, r-19 ***** "'"""'t--. * * * * .... ' It **** t:o. 300 .... ' It **** .._..i..-8 8 8 8 I (.....-II e e * * .._..i-. . . . 1-= i-24 II * * *

• i--..'""'...._ .... ' lo **** .... ' 100 II * * * * "'t;:a .... ' I * * *
• i...-1..-i.. 60 I * * *
• lo **** 62
• 8 8
• I lo **** * * *
• j It **** 68 It **** ' * * * * .... ' It **** 8 8 8 8 I 60 It **** I .... ' It **** * * *
• t * * * *
• 61 .... ' * * * * * * * *
• 4 64 63 I 58
• 63 I J 54 I I I I 57 I I I I I 51 I I I ' 55 ! 1\1"1 ! 79 I 1.1\ 78 I j 150 I I I 1, ! i. 84 I i II i I * * * * ! i . . . . 66 I * * * *
• 1 * * *
• I It **** . . . . ' It **** 67 .... ' "r--. ...... ! .... ' ,a ...... 144 **** 1 . . . . .... ********** 300 It.*.*.*.*, i,...i... i.. . *.*. *.*, l\L--********* 54 **** ** ** * . *. *.*. *, 60 ...... \ 8 8 8 8 I 73
• 70 I FIGURE 2. 61 I I I 7-15 IALTllOll GAS AND llECTllC COMPANY (euv. DEPT* DESCRIPTION -VISUAL 1.0 0 Greenhh sny silty Hnd o-------------10.. Creenilh gray silty Hnd with *hells 20-Greeniah gray fine sandy clayey silt with shells 30--30-40-Greenish gray fine sandy silt wit shells -50-i;reenish gray fine sandy s i l c -so-60-gray silty sand 70-. PreeniBh gray slightly sandy silt 80--so-90--90-100--100-110-gray silty aand -no-120--------------gray ailty fine sand -120-gray slightly fine sandy 130-*Ut *130. 140-Greenish gray slightly fine sandy dlt *140-150-GRAPHIC IOllNG LOG SUPPLEMENTAL sm IORINGS CALYEIT CllfS IUCUll POWEi ft.ANT BORING NO. 221 GRAPHIC BLOWS/FOOT N LAB LOG lo 10 20 30 50 70 100 +11111 TESTS I e e e e 9 .... ' .. * ... *. *, ['-.. 29 .. * .* ... *, ,....... * * * * * ....r-. .... ' r"'" -125 .. *. *. *. *, ==-f .. * .* .*. *, .......... I e e e e r:::: .......... 27 i I .... '
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• I ' I I' 200 300 I 200 I 500 FIGURE 2. I I I ' 144 7-16 I IALTIMOIE GAS AND ELECTIK COMPANY GUPHK IOllNG LOG SUPPLEMENTAL Sin IORINGS CAlYHT CllfS llUCLUR POWEi PUNT BORING NO. 222 I ELEV. OEPTI DESCRIPTION -VISUAL 0 1. 5 0-u .. reyyish gray silty eand with * * * * * * * * *, _11_ --------!Greenish gray slightly clayey ; * * * * * * * *, silty sand with shells * * * * * * * *, ****** ** ** lO-------------_1t.*.*.*.*. -30--40--so--60--70--so-*90--100-Greenish gray slightly silty ssnd It********, with shells * * * * * * * * * !Greenish gray fine sandy silt wit 20.. trace shells Greenish gray silty sand IQ.. !Greenish gray sandy silt with 3 slight trace shells !Greenish gray silty sand with 40.. shells 60-* * * * **** t It * * * * **** t * * * * * . . . . ' Ill **** e e e e I It * * * * .... ' II * * * * .... ' II * * * * !Greenish gray slightly clayey fine lsandy silt 70_ breenish gray sandy silt 80-90----,llUl'I BLOWS/FOOT 10 -20 30 50 70 I 100 +100 N 16 150 170 l!JO 150 61 89 79 67 71 61 60 53 54 56 59 79 82 74 71 71 LAB TESTS FIGURE 2. 7-17 I I I Ill TIMOIE GAS AND ELECTRIC COMPANY !ELEV. DEPn DESCRIPTION -VISUAL 4.2 0 Greenish gray silty sand o-Gree"niiiii"' silty sand with shells . 20_ --------------20-Greenish grsy silty sand with shells 30--30-------------40-Greenish gray silty sand Greenish gray slightly sandy silt 50--so-60--60-70--10-Greenish gray slightly clayey silty sand 80--so-Greenish gray slightly sandy clayey silt 90--90-Green!';h'"gray sandy silt 100--100--. --GRAPHIC IORING LOG SUPPLEMENTAL Sin IORINGS BORING NO. 223 rcNc 1 ""I IVl'll CAlYHT CLIFFS NUCLEAR POWER PLANT GRAPHIC LAB BLOWS/FOOT N LOG 0 10 'D 30 50 70 100 +100 TESTS ***** ........ 20 .... I'... ***** ******** ** 56 ****** 1. ... }' .... ' * * * *
• 28 ***** I * * * * * . . . . ' It * * * * .... ' 24 It **** .... ' It * * * * ['.. .... ' It **** r-47 . . . . ' It * * * * .... ' * * * * * **** 58 * * * * * . . . . ' * * * * * . . . . ' * * * *
• 44 ***** * * * * * * * * *
• It **** . . . . ' t---.. 47 * * * * * "'!'-. .... ' It **** !'-. 150 .... ' 'Q * * * * * .... ' It **** 200 /v 72 ' 65 62 \ , 73 I 77 I ! .-.-.-. 78 I ***** I I 8 8 e 8 I ***** I ***** 73 I I ***** I I e e e e * * * * * --... -* 69 I 73 77 I 81 I I I FIGURE 2. 7-18 I I l lllllMORE GAS AND ELECTRIC COMPANY IELEV. DEPT> DESCRIPTION -VISUAL 3.7 0 Greenish gray ai1ty aand o---------------gray silty sand with lshella 10--10--20-------------gray very slightly clayey 30-sand with trace of ahells .30-40--40-greenish gray fine sandy silt 50--so--60-gray very slightly clayey sandy silt 70-gray sandy ailt so--so-90-*90-100--100----. GRAPHIC IOllNG lOG SUPPlEMENTAl Sift IORINGS CALVERT CUFFS NUCLUI PoWEI PLAIT BORING NO. 224 GRAPHIC l'l:Nl:IHA11UN HUISlA""'i: LAB BLOWS/FOOT N LOG 0 10 20 30 50 70 100 +1N' TESTS 8 8 8 8 I -"-i-17 ........ *, ........ r-...., ""'"" II e * *
• 1::9 150 e 8 e e I i,..L.. * * * * * .... ' i...-1,,.o II * * * * (.,...-.... ' 42 ........ *, II * * * * .... ' r--, It il * * * *
• 8 8 8 8 I I/ II * * * * .... ' *.*.*.*.*, k 51 II * * *
• B e 8 8 I *.*.*.*.*, 71 II * * * * .... ' * ....... *, J II * * *
• 64 .... ' II * * * * .... ' II * * * * .... ' II * * *
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• 70 II 74 68 -... 71 I.I[> 150 I/ 77
• I'\ I\ G 200 150 IJ i. 83 I\ 150 I 250 FI,URi 2. 7-11 9 HL llMORE GAS AllD ELECTRIC COMPANY I ELEV. DEPT" DESCRIPTION -VISUAL lS.O 0 )l_rown 11ilty aand*fill tor road BrOWn Brltyaandwitii""ibelli --. -----------gray silty eand with 20-trace of ehells 0-20--10-30--20-40-r;-;;;ni-;;-;r-;;-sii't; ;;;;d -----30-50-gray sandy silt gray slightly clayey silt with trace of shells 60--50-gray sandy silt 70--60-gray slightly clayey 80-silt 90--80-100--90-. ---. -. GUPHIC IOllNG LOG SUPPLEMENTAL SltE IORINGS CALVERY CLIFFS NUCLEAR POWER PUNY BORING NO. 226 GRAPHIC ....... " ........... *--*------LAS BLOWS/FOOT N LOG 0 1[1 20 30 50 70 100 +1M TESTS ..... ........ 12 * * * *
• r---... * . . . . ' , . . . . 33 .... ' * * * *
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• 70
• 68 80 81 79 I 82 87 \ > 100 'I 89 1 85 FIGURE Z. 7-I I 20 I HLTIMORE GAS AND ELECTRIC COMPANY !ELEV. DEPTI DESCRIPTION -VISUAL z.o 0 !Brown sU!htly 111ty aand with o---gray silty sand -------*----10-gray slightly clayey milty sand with trace of shells 20--20-30-treenish gray slightly clayey !fine sandy silt 40-gray quite silty sand 50-treenish gray fine sandy silt -so-60--60-Greenish gray very slightly claye sandy silt 70--70-80---. ----. GRAPHIC IOllNG lOG SUPPLEMENTAL Sin IORINGS BORING NO. 227 ;1u,.. CALVERT CUFFS NUClUI POWER PLAIT GRAPHIC LAB BLOWS/FOOT N LOG TESTS 0 10 n 30 50 70 100 +10< .-.-.-.-. 21 e
• e e I It ******* *, IL 18 It * * * * .... ' It ******* *, 13 ........ *, t--It * * *
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• lfi...-64 68 11 .... ' J 50 a * * * * .... ' ********** \ a * * * * ***** 57 * * *
• I\ 70 \ 75 81 82 86 FIGURE Z. 7-Z I I I 1 HlTIMORE GAS AllD ELECTRIC COMPANY IELEV. DEPn DESCRIPTION -VISUAL 1.50-u mea1U111 aana \De&cn 1Bna1 claye_y _ isilty sand with shells 10--10-gr;y-;utY'Sa;.rw"i t-h-*ishells ---------20-gray silty sand 30--30 -Greenish gray slightly clayey 40--40-llandy silt -so--so -60-70 --70 -' 80 ----. -----. GRAPHIC IOllNG LOG SUPPLEMENTAL SITE IOllNGS CALVERT CLIFFS NUCLEAR POWEi PLANT BORING NO. 228 GRAPHIC _,._ , , ,_,. nc;:,I;:,' ""'"c LAB BLOWS/FOOT N LOG 0 10 20 30 60 70 100 +10!: TESTS .... ' L 14 It **** e e e e I ***** 13 e e e e I -***** r-e e e e I ,...... -r-***** ,.........,. e e e e I "CT 100 It * * * *
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• GENERAL NOTES l. tu:V.t..flOt.tSl.£FP.tt<io10MfANSfAUYn. 2. NOMM.llADJACENf TO IOllNG LOG IS StAHDAID fl:Nnto\* TION mrstANCE.. 1.
• l. StANDoUO KNETaATION usiriANC:E IN 11..0WS /FOOT OF A MO u. W(tGHf Fl:fE FALLfNG 30 INCK?S._ -41. GaoL.fl FOUNOATION AND EXPLCbTIQt'! CO. ., OFF*SHOkt GRAPHIC BORING SHEET NO. I ELECTRIC PRODUCTION Pl.ANT . ...

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• 1, EUVA110NSRfff'l:ENCEOTOM!ANSEAlfVEL,. ,... : ;-, ... 2. M.llo6Et ADJACENT TO LOG IS STAND.Ip \ J, PESISTANCE IN BLOWS /FOOT 1<40 IJ. WEIGHT FREE FALUNG INCHES,
• _ 4. 8021:-IGs Bl" GC.DlEl FOUNOATK>N ANO EXPLORA.TION CO.-'"i'" JUNE 19.:,9. ... '"'. _ '"-. ' OFF-SllORE GRAPlllC CAL.VERT CLIFFs

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• IOllNG Wl-22 -"" IOllt-IG we-27 \ IOllNG Wl-23 tOtlNG Wl-28 . *.** .. .. *:** IM Siity Sil.ND c;..,.....,,c.,.v G,.,. SM6,-Sll.T GENERAL llOTES OFF*SHORE GRAPHIC BORltjG SHEET NO. 3 CALVERT CLIFFS

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• fil] D *-* . ',i : .. -; .< GENERAl NOTES * -. ; l, tuv>.TK>NStS:EltNCto10MEANSfAltvEl. . ... 2. 10 IOflNG LOG IS STANO.I.ID l. STANO.U.D flENETtA.TION IESISTANCEIN .,. --4. JUN£1969. a.: Off*SHORE GRAPHIC BORING N<l. 4 <L \ .. j'< l :..-* -_.-' . */ .-. .... .

.... r.uv.fn. ISO 100 so llSL 0 -so -100 -150 -too SCALE 1* *so* o* ELEY./P'T. 190 100 50 llSL 0 -so -100 -ISO -!00 0 too 200 MIO cl c._J PROFILE A" cl 100 TOO c.....J PROFILE 11811 so 2!1 0 50 i-:w I 100 I GRAPHIC SCALE DI D._.J DI *OO 900 o.._J 1000 100 1100 For S*c1fon locat,Qdl He 8orln9 Plat Plan IJOO GENERALIZED SOIL PROFILES SHEET IOF 2 FIGURE 2.7-27 ELEV./fT. IOO 100 50 llSL 0 too -150 -200 0 100 200 300 400 500 ELEV./ FT. 150 100 50 llSL 0 100 -150 -zoo 0 100 200 300 400 500 600 f; .l!' I 7u:* * &; 700 PROFILE TOO I l 800 CONTAIN .. ENT HltUCTUll[ : 'i ; : L,_B llcll 18 800 L,_B PROFILE 11011 50 25 0 r-:LJ I GRAPHIC SCALE 100 I , .. 900 900 AUXILIARY *LDO. t : l 1000 TURBINE 11..DQ.. 0 .. 1000 CONTAINlll[NT ITRUCTUfll[ r-: 1100 L_A IA .. 1100 1300 120<.* 1300 For Section tocatlona all lorl11g Plot Pion 1400 1500 GENERALIZED SOIL PROFILES SHEET 2 OF 2 FIGURE , ., .. {' \. *I . .:*-_ \ I /._ "-... ) ,;_ _ ... J' *, \ \ I (, I ---I ., { ' I \ _ ... --.. ..->-----__ .. i I h .. ..L... . ' '1 "-Ii _. ____ . *..

• Lr -J, Ji-. j* ii. ; "' ! -,1-lfl " ----..,, 1*' *=-1 -----------. : j., . . : fl! I .---.u....-"4..'\' ! : *j.-I -----: -, :----+..-. I ' ,.., ,:.?1('1.00 J c:-e. t,Sl\.-:..0 -1 t 0: J' I . -. -1 . I I . \ . lJ... I , I \ . ' ... ,_,,., ... --.!.,-. *---r----**-' t' I I. I -I \-. 1-'t.Q, . !>,. ---,. .._+---* I I ---'* ' ' -* _::;_: ------p -*.-'!:l) .. ; .. ---*: -,,,i. -,*. . ..... * . _..---! .*.. .-o.>} i/ . .-.:;,_ } * --"l ** " J 3-.... &LEV. 140 11t0 100 ... M> ao to 0 'TA. 1*$0 W*OO SECTION A*A ELEV. no 1 TURalNE ... OG. IT OtOO 1*00 l*OO ,.oo 4*00 !>*00 G+OO '400 e*oo '; SECTION B*B ... -souTHWUT h51!iit6" JJliiiiJ4" JJQIW 0 STA. 0-00 IOOO 1*00 7*00 SECTION C*C PR0,.0$10 Ao..O 100 CONTAINMtNT ao ... 'Z.O 20 STA. O 0*00 1*00 '!1*00 SECTION O*O SECTION E*E '!IOtolOO 4;0 !Ml;;K;;;O!-' M SLOPE IOO f I 000 f !I soo 8 .. 2 CROSS-SECTIONS AT PLANT FIGURE 2.7-30

..j .... ::: .... i::: § '41 I ""r I .Gi a * .c :u * ""r I 1ai I I :: I I CHANCc OF OCCURRcNCc IN PcRCcNT ,:f ,, r r r r r i i r i i i r i T i 1 T i i 13 12 II 10 9 8 7 6 5 4 3 2 I 0 .... Jf i ',I I "' "' ---!--GENERALIZED TIME FREQUENCY CURVE NOTE: REPRCOUCEO FROM Pl.ATE 3 OF H.ooc. NO* 88 CONGRESS, 2No. SESS* 11TIOEWATER PORTIONS OF PATU>cENT, POTOMAC ANO RAPPAHANNOCK RI VE'RS, INCLUDING AOJACE:NT CHESAPE:AKE BAY SHOREL INE:S" ( SKGro/El> /:"IOH ) C.( ti.S. MAP 1100 r..l "' 1' I , iy < u '3*8'-C) \J ... .q: ..J i... '{, 37"- .-, ;-'*) -:-. :-T-:-*::t *:: _ ____.....,* Jr:4: J -"i-*-c-:.::i'-._:::;_,--_. '.. .;-.-_-_ ... .,. .; "t-.--. --* .:..../. -.. .---.: ... -*--. .-tJ-'H. . f-8 :r:p: tL -l:+..;-1. :-.-1.J :-L . _* T+. *..l-t:..1p-1. -:4-::"""f :_+,+t-H"iH-ti' . .,.. -. --!-,-.--'*!--, r-r, --7* 1:d:Z*iit.:: +f. . .,,: . . --.:+/-l :j::j:t:. IL;:' r *f, 1-i !+.'-.* "-j-:-f .:l:l:f 7l s* rffi-+1..:-J *, -. JI * '--'-1--f-1-,-:-H +l---I-'-. I--!rt .. :p: :p_ t-r-f **-:-.--** ,+*-*-* ,-; 7-+-1'. i+' 7 r:.,t: +H :f+/-!:i. . .. *: f 'I -...! -'r41 * .. }+t-r+--j-' -f + -.. -. +.; *-, ::+f:t I l-1,:i:?.

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WATER SURFACE ELEVATION (FT.) . ,.1-,::::' -1,-/ A===;_:..,--,.2! .,...i 0 <D '° 0 N '° 0 CD v 0 v v 0 0 v 0 <D ,., 0 N I') r,-._ r.r.i 0 Q co ZJ N 0 u 0 r.r.i '-' v N ,u r.r.i 0 0 N 1-E-< 0 0 0 co 0 1C) v 0 ! .: . [ p I FIGURE 2. 9 -I REV. II

CALVERT CLIFFS UFSAR 3.1-1 Rev. 47 REACTOR 3.03.1 GENERAL DESIGN SUMMARY Both of the Calvert Cliffs reactors are of identical design. Consequently, reference throughout this section is made to a single reactor and, unless otherwise noted, implies either Unit 1 or 2. The reactor is of the pressurized water type, using two reactor coolant loops. A vertical cross-section of the reactor is shown in Figure 3.1-1. The reactor core is composed of 217 fuel assemblies and 77 Control Element Assemblies (CEAs). The fuel assemblies are arranged to approximate a right circular cylinder with an equivalent diameter of 136" and an active height of 136.7".

The fuel assembly consists of 176 rods (pins) and 5 guide tubes. The pins may contain fuel and/or a neutron poison. The assembly is held together by spacer grids and is closed at the top and bottom by end fittings. Lateral support and positioning of the fuel rods within an assembly is provided by spacer grids. The spacer grids are welded to five full-length guide tubes. The guide tubes provide channels which guide the CEAs over their entire length of travel and form the longitudinal structure of the assembly. In selected fuel assemblies the central guide tube houses incore instrumentation (ICI). Design characteristics of demonstration or lead fuel assemblies are discussed in Section 3.7.

The fuel is low enrichment uranium dioxide (UO2) in the form of ceramic pellets clad in Zircaloy or ZIRLO for Westinghouse fuel (as part of an advanced cladding test program, some fuel pins in Batches 2NT, 1RT, 2TF, and 2TW utilize cladding other than Zircaloy or ZIRLO) tubes which are welded into a hermetic enclosure. Starting with Unit 2 Cycle 19 and Unit 1 Cycle 21, AREVA fuel uses M5 alloy cladding. Initially the fuel was managed in a three-cycle, mixed central zone, fuel management plan (Figure 3.4-3). Starting with Unit 2 Cycle 8 and Unit 1 Cycle 10, the 24-month cycle core utilized low leakage fuel management. Starting with Unit 1 Cycle 11 and Unit 2 Cycle 10, low fluence fuel management is employed to reduce the fluence on the critical vessel weld. Low fluence fuel management includes replacement of fresh fuel located on the core flats with once or twice-burned fuel. In Unit 1 Cycle 11 and Cycle 12 low fluence fuel management also included the addition of Guide Tube Flux Suppressors (GTFSs) in selected assemblies near the periphery. Sufficient margin is provided to ensure that power peaks are minimized. The reactor coolant enters the upper section of the reactor vessel, flows downward between the reactor vessel wall and the core barrel, passes through the flow skirt where the flow distribution is equalized and into the lower plenum. The coolant then flows upward through the core, removing heat from the fuel rods, exits from the reactor vessel and passes through the tube side of the vertical U-tube steam generators where heat is transferred to the secondary system. The reactor coolant pumps return the coolant to the reactor vessel.

The reactor internals support and orient the fuel assemblies, the CEAs, and the incore instrumentation and guide the reactor coolant through the reactor vessel. The reactor internals also absorb static and dynamic loads and transmit the loads to the reactor vessel flange. They will safely perform their functions during normal operating, upset, emergency, and faulted conditions. The internals are designed to safely withstand forces due to deadweight, handling, temperature and pressure differentials, flow impingement, vibration, and seismic acceleration.

Reactivity control is provided by two independent systems: (1) the Control Element Drive System (CEDS) which controls CEA motion, and (2) the Chemical and Volume Control System (CVCS) which is used to control the Reactor Coolant System (RCS) boric acid concentration.

CALVERT CLIFFS UFSAR 3.1-2 Rev. 47 Boric acid dissolved in the coolant is used as a neutron absorber to provide long-term reactivity control. In order to reduce the boric acid concentration required at Beginning of Life (BOL) operating conditions and lower power peaking, mechanically fixed burnable poison rods (BPRs) may be provided in certain fuel assemblies. Originally, the neutron poison was boron carbide which is dispersed in alumina pellets; the pellets are clad in Zirconium alloy to form rods which are similar to the fuel rods. Gadolinia and erbium oxide, mixed into fuel pellets, are also being used as a neutron burnable absorber. Beginning with Unit 2 Cycle 16 and Unit 1 Cycle 18 Zirc Diboride (ZrB2), applied as a coating on the fuel pellets, was used as a neutron burnable absorber. Poison rods are also called shims. Beginning in Unit 2 Cycle 19 and Unit 1 Cycle 21 with AREVA fuel, Gadolinia (Gd2O3) mixed into the fuel is used as the neutron burnable absorber. The CEAs consist of five Inconel tubes filled with neutron absorbers. Four tubes are assembled in a square array around the central fifth tube. A spider joins the tubes at the upper end. The hub of the spider couples the CEA to the drive assembly. The CEAs are activated by magnetic jack Control Element Drive Mechanisms (CEDMs) mounted on the reactor vessel head. The maximum reactivity worth of the CEAs and the associated reactivity addition rate are limited by system design to prevent sudden large reactivity increases. The design restraints are such that reactivity increases do not result in violation of the fuel damage limits, rupture of the reactor coolant pressure boundary, nor disruption of the core or other internals sufficient to impair the effectiveness of emergency cooling.

Control Element Assemblies are moved in groups to satisfy the requirements of shutdown, power level changes, and operational maneuvering. The control system is designed to produce power distributions that are within the acceptable limits of overall nuclear heat flux factor and Departure from Nucleate Boiling Ratio (DNBR). The Reactor Protective System (RPS) and administrative controls ensure that these limits are not exceeded.

In order to assure control of axial power distribution (APD), particularly in the event of axial xenon oscillation, eight CEAs designated as Part Length CEAs (PLCEA) were initially installed. They have since proved unnecessary and were removed along with their extension shafts. Control Element Assembly guide tube plugs were inserted into the locations previously occupied by the PLCEAs. They have also proved unnecessary and were removed before Unit 1 Cycle 8 and Unit 2 Cycle 7.

CALVERT CLIFFS UFSAR 3.6-1 Rev. 47 3.6 ORIGINAL FUEL DESIGN EVALUATION 3.6.1 FUEL DESIGN AND ANALYSIS The fuel rod cladding is designed to satisfy the design bases given in Section 3.2.3.5. The effects of irradiation on UO2 and cladding materials are considered in the design calculations. The predicted effects of anticipated transients are also considered in the design process.

As stated in Section 3.2.3.5, the fuel rod cladding is designed to the following bases: Basis 1 Maximum primary stress during steady state operation, expected transients, and depressurization is limited to two-thirds of the minimum yield strength of the material at operating temperature.

Basis 2 Predicted permanent hoop strain of the cladding at the end of fuel life is less than 1.0%. These bases are conservative and the calculations used to demonstrate their satisfaction were conducted for limiting cases using limiting assumptions. This is considered advisable in the prediction of long-term fuel behavior under irradiation. Maximum tensile stress in the fuel cladding occurs during a depressurization transient near EOL when internal gas pressure is highest. Clad thickness is such that under the anticipated transient conditions, this stress does not exceed two-thirds of the unirradiated value of yield stress of the clad material at its operating temperature. An unirradiated value is used for conservatism. The satisfaction of Basis 2, the long-term total strain limit, was demonstrated as follows: a. Clad stress-strain behavior was based on a stress analysis which includes the effect of creep. The loads considered were those due to fuel thermal and fission growth, fission gas pressure and external coolant pressure. b. The fuel thermal and fission growth was calculated considering the fuel as a solid rod with unrestrained thermal expansion and a volumetric growth rate of 0.16% for 1020 fissions/cm3 (Reference 1), and a LHR of 17.5 kW/ft. The fission gas pressure was calculated for a 31% fission gas release (which was based on the derivation of Lewis (Reference 2) considering the change in plenum volume due to the thermal expansion and growth of the rod). c. The analysis was based upon an incremental approach, which divided the three-year fuel life span into discrete time intervals and evaluated the clad stress and strain, including the effect of creep, during these intervals. The relation between the incremental creep and the actual stress state is expressed by the Prandtl-Reuss formulae. The basis for creep is given by the von Mises criterion and the relation between creep rate and generalized stress is that given by Holmes (Reference 3). The rapidly convergent iterative technique was employed to solve the resulting non-linear equations. d. For the nominal fuel-to-clad gap, at about 775 hours after BOL, the fuel has expanded to completely fill the fuel-to-clad gap and to restore the clad to a circular shape after its initial collapse onto the fuel. The fuel was subsequently assumed to swell unrestrained with the clad following. Based upon this conservative assumption, the final strain after three years service was 0.5%. That is, for CALVERT CLIFFS UFSAR 3.6-2 Rev. 47 average fuel-to-clad gap at peak power density, Basis 2 was satisfied without credit for fuel strain under load. e. For the most adverse initial condition, i.e., minimum clad ID, maximum pellet OD coincident with the point of maximum power density which was assumed to be sustained over lifetime, application of the unrestrained fuel growth model resulted in a computed strain at the end of the third cycle (EOC3) of 0.8%. However, as is well known (References 4, 5, and 6), the effect of restraint from the exterior, cooler regions of the fuel pellet, the clad, and the external pressure result in a significant limitation on radial swelling with corresponding flow of pellet material into the dish provided. These analyses were conducted throughout with design BOL power density, although it was known that for fuel in its third burnup cycle, LPD would be substantially below these values. Thus, the LPD increase which might be associated with overpower transients near end of fuel life was conservatively considered. Further consideration of EOL power density is provided in subsequent paragraphs together with a summary of data justifying the maximum linear heat ratings and peak burnups. Table 3.6-1 contains typical maximum linear heat ratings as a function of burnup. The maximum linear heat rating for the first core was 17.5 kW/ft at BOL. The maximum heat rating near EOC3 was 14.9 kW/ft, resulting in a BOL/EOC3 ratio of 1.18. This was greater than the value of 1.12 for the ratio of maximum transient to steady state heat ratings. Thus, use of BOL power densities in these calculations for EOC3 transients provided considerable margin.

Studies by Notely, et al (References 5 and 6) in which 27 fuel elements were irradiated without failure, reported measured clad strains up to 3.33%. In a series of experimental element irradiations, Westinghouse (Reference 4) reported strain values at failure for Zr-4 clad fuel elements of 0.78 to 2.6% depending on the fuel properties assumed. Also, Lustman (Reference 7) noted that failures in pile have occurred at strain values between 0.5 to 1.0%. However, these results are based on relatively low Zr-4 cladding temperatures as compared to contemporary, large, commercial PWRs. It is known (Reference 8) that permissible strain values for Zircaloy increase above 650°F. In the zone of interest, the average Zr-4 cladding temperature is about 720°F; this should result in increased ductility and thus a higher strain limit to failure. For the AREVA design, compliance was demonstrated using the NRC-approved methodology using the RODEX2 code.

3.6.2 ANALYSIS OF BURNUP AND LINEAR HEAT RATINGS Prior to a discussion of the experimental bases for justifying the initial maximum linear heat ratings and burnups, it is necessary to relate these parameters so that they may be viewed in the proper perspective. The maximum linear heat rating was reached but not exceeded only during approximately the first 28,000 MWD/MTU of peak burnup. The maximum linear heat rating decreased with additional burnup beyond this value. Typical values at the time of initial design are shown in Table 3.6-1, which contains an analysis of burnup, total nuclear peaking factors, and the corresponding maximum linear heat rating (including consideration of the combination of total nuclear and mechanical peaking factors), for the most adverse equilibrium core. Table 3.6-2 contains a comparison of maximum heat ratings for a number of plants of that period. Peak linear heat ratings for this plant were consistent with current practice and were considered as slightly conservative with respect to a number of the designs. CALVERT CLIFFS UFSAR 3.6-3 Rev. 47 Although it was believed that fuel rods could operate satisfactorily with a small amount of fuel melting, the initial design did not permit fuel melting even under conditions imposed by anticipated transients. Cycle 1 design offered considerable margin with respect to the core linear heat rating of 24 kW/ft for melting (BOL value; typical EOC3 value was about 23 kW/ft), even when expected transients (112%) were considered. 3.6.3 SUMMARY OF PERTINENT FUELS IRRADIATION INFORMATION The LHRs specified in this section are as they appeared in the referenced literature and represent total core heat rates. 3.6.3.1 High Linear Heat Rating Irradiations The determination of the effect of linear heat rating and fuel-cladding gap on the performance of Zircaloy-clad UO2 fuel rods was the object of two experimental capsule irradiation programs conducted in the Westinghouse Test Reactor (WTR) (Reference 9). In the first program, 18 rods containing 94% TD UO2 pellets were irradiated at 11, 16, 18 and 24 kW/ft with cold diametral gaps of 0.006", 0.012" and 0.025". The wall thickness to diameter ratio (t/OD) of the Zircaloy-cladding was 0.064 which is slightly higher than the value of 0.059 of Cycle 1. Although these irradiations were short duration (about 40 hours), significant results applicable to Cycle 1 design were obtained. No significant dimensional changes were found in any of the fuel rods. Only one rod, which operated at 24 kW/ft with an initial diametral gap of 0.025", experienced center melting. Rods which operated at 24 kW/ft with cold gaps of 0.006" and 0.012" did not exhibit center melting on these bases. The initial gap of 0.0085" and the maximum linear heat ratings for this design (Table 3.6-1) provided adequate margin against center melting even when 12% overpower conditions were considered. These results also indicated that an initial diametral gap of 0.0085" was adequate to accommodate radial thermal expansion without inducing cladding dimensional changes, even at 24 kW/ft. This margin with respect to thermal expansion, decreased with increasing burnup at a rate of 0.16% V per 1020 fissions/cm3. However, the linear heat rating also diminished with burnup (Table 3.6-1). Since the diametral thermal expansion (assuming BOL maximum heat ratings) is almost twice as great as the swelling diametral growth (on the EOC3 burnup), these data added considerable weight to the conservative treatment of the influence of transients on fuel element integrity. Further substantiation of the capability of operation at maximum linear heat ratings in excess of those in the first cycle design was obtained from later irradiation tests in WTR (Reference 9). Thirty-eight-inch long and 6" long fuel rods were irradiated at linear heat ratings of 19 kW/ft and 22.2 kW/ft to burnups of 3450 and 6250 MWD/MTU. The cold diametral gaps in these Zircaloy-clad rods containing 94% dense UO2 were 0.002", 0.006" and 0.012". The cladding t/OD was 0.064. No measurable diameter changes were noted for the 0.006" or 0.012" diametral gap. Only small changes were observed for the rods with a 0.002" diametral gap. Additional successful radiations had been performed with SS cladding in Saxton at 23 kW/ft and in Plum Point at 22 to 25 kW/ft.

3.6.3.2 Shippingport Blanket Irradiations Zircaloy-clad fuel rods operated successfully (three defects had been observed which were a result of fabrication defects) in the Shippingport blanket with burnups of about 37,000 MWD/MTU and maximum linear power ratings of about 13 kW/ft (References 9, 10, and 11). Although higher linear heat ratings at lower burnups would be experienced, swelling (primarily burnup-dependent) and thermal CALVERT CLIFFS UFSAR 3.6-4 Rev. 47 expansion (linear heat rating dependent) provide the primary forces for fuel cladding strain at the damage limit. Thus, Shippingport irradiations demonstrated that Zircaloy-clad rods with a cladding t/OD comparable to that for this plant (0.059) should successfully contain the swelling associated with 37,000 MWD/MTU burnup, while at the same time containing the radial thermal expansion associated with heat ratings of the time. Irradiation test programs in support of Shippingport in in-reactor loads demonstrated successful operation of burnups of 40,000 MWD/MTU and linear heat ratings of about 11 kW/ft with cladding t/OD ratios as low as 0.053 (compared with 0.059 for this plant) (Reference 12). 3.6.3.3 NRX Irradiations (AECL - Canada) Eleven Zircaloy-clad, large diameter fuel elements (approximately .750" OD) with clad thicknesses of .016", .024", and 0.037" (t/OD = .021, .031, and .047 corresponding to TD percentages of 94.3, 94.3 and 93.7, respectively) were irradiated in the NRX pressurized loop facility of AECL, Canada (Reference 13) at loop pressures of 2000 to 3000 psi. The cold diametral gaps for the test elements were .0035" and .0040", and the fuel was UO2 sintered pellets (0.700" diameter) loaded in an argon atmosphere. The elements were operated for 535 full power days to an average burnup of 10,280 MWD/MTU at a maximum linear power output of 14.8 kW/ft. These elements experienced 308 power cycles. No failures were reported for these elements, and the final dimensions of the rods were reported to be virtually unchanged from pre-irradiation values. The successful operation of these elements with considerable lower clad-to-diameter ratios than those for Cycle 1 demonstrated the capability of safe operation of Zircaloy-clad elements with thin cladding for many power cycles. Additional tests on similar elements were then in progress at NRX involving test elements with UO2 and (U, Pu) O2 (PuO2 = 2.4 wt%) at average linear heat ratings of 11.4 and 17.2 kW/ft. Those elements had accumulated burnups of 6,400 and 28,700 MWD/MTU without failure.

3.6.3.4 Saxton Irradiations UO2-PuO2 fuel rods containing pellets of 94% TD and clad with Zircaloy-4 had been successfully irradiated in Saxton to burnups approaching 25,000 MWD/MTU at 16 kW/ft under USAEC Contract AT(30-1)-3385 (Reference 14). The t/OD of the cladding was 0.059 which is equivalent to the Cycle 1 design. The amount of PuO2, 6.6%, was considered as insignificant with respect to providing any differences in performance when compared with that for UO2. In fact, the higher thermal expansion coefficient for this PuO2-UO2 composition than that for UO2 would induce greater cladding strain under equivalent irradiation conditions. Subsequent tests on two of the above rods (18,600 MWD/MTU at 10.5 kW/ft) successfully demonstrated the capability of these rods to undergo power transients from 16.8 kW/ft to 18.7 kW/ft. 3.6.3.5 Vallecitos Boiling Water Reactor - Dresden The combined Vallecitos Boiling Water Reactor (VBWR) - Dresden irradiation of Zircaloy-clad oxide pellets (Reference 15 and 16) provided additional confidence with respect to the design conditions for the fuel rods for Cycle 1 core. Ninety-eight rods irradiated in VBWR to an average burnup of about 10,700 MWD/MTU were assembled in fuel assemblies and irradiated in Dresden to a peak burnup CALVERT CLIFFS UFSAR 3.6-5 Rev. 47 greater than 48,000 MWD/MTU. The reported maximum heat ratings for these rods was 17.3 kW/ft, which occurred in VBWR. The t/OD cladding ratio of 0.052, pellet TD of 95%, and the external pressure of about 100 psi are conditions which are all in the direction of less conservatism with respect to fuel rod integrity when compared with the design values of 0.059 cladding t/OD ratio and an external pressure of 2250 psi. Ten of these VBWR - Dresden rods representing maximum combinations of burnup, heat rating and pellet density had been selected for detailed destructive examinations as part of an AEC program. The remaining 88 rods were returned to Dresden and successfully irradiated to the termination of the program. 3.6.3.6 Large Seed Blanket Reactor Rods Two rods operated in the B-4 loop at the Materials Testing Reactor provided a very interesting simulation for contemporary PWR designs (Reference 4, 17, and 18). Both rods were comprised of 95% TD pellets with dished ends clad in Zircaloy. The first of these, No. 79-2, was operated successfully to a burnup of 12.41x1020 f/cc (approximately 48,000 MWD/MTU) through several power cycles which included linear power from 5.6 to 13.6 kW/ft. The second fuel pin, No. 79-25, operated successfully to 15.26x1020 f/cc (approximately 60,000 MWD/MTU). The basic difference in this rod was the 0.028" wall thickness, as compared to 0.016" (t/OD 0.058) in the first rod. All other parameters were essentially identical. The linear heat rating ranged from 7.1 to 16.0 kW/ft. After the seventh interim examination, the rod operated at a peak linear power of 12.9 kW/ft at a time when the peak burnup was 49,500 MWD/MTU. These high burnups were achieved with fuel elements which were assembled by shrinking the cladding onto the fuel. This indicated that a comparable irradiation of the fuel elements for this reactor would allow a considerable increase in swelling life at a given clad strain.

One additional rod irradiated in Materials Testing Reactor as part of the Large Seed Blanket Reactor (LSBR) series (rod 79-18) demonstrated the effect of clad restraint on the swelling behavior of a UO2-Zircaloy-clad rod (Reference 19). A starting fuel density of 81.4% of theoretical was used in conjunction with a zero cold gap and a 0.060 cladding t/OD ratio. The rod was irradiated to 49,000 MWD/MTU with no measurable change in rod diameter. 3.6.3.7 Central Melting in Big Rock As part of a Joint U.S. - Euratom Research and Development Program, Zircaloy-clad UO2 pellet rods with 95% of TD had been irradiated under conditions designed to induce central melting in the Consumers Big Rock Point Reactor (Reference 20). The test included 0.7" diameter fuel rods (cladding t/OD = 0.061, fuel-to-clad gap of about 0.011") at maximum linear heat ratings of about 27 kW/ft and 22 kW/ft with peak burnups up to 20,000 MWD/MTU. Results of these irradiations provided a basis for incorporating linear heat ratings well in excess of those calculated for this reactor (Reference 21). These results showed that the presence of localized regions of fuel melting were not catastrophic to the fuel assembly. 3.6.3.8 Peach Bottom 2 General Electric (GE) had successfully irradiated fuel pins of the Peach Bottom 2 design to burnups in excess of 42,000 MWD/MTU at peak linear heat ratings of 23 kW/ft. An interim examination at 32,500 MWD/MTU indicated a satisfactory condition (Reference 22). CALVERT CLIFFS UFSAR 3.6-6 Rev. 47 3.6.4 EVALUATION It was concluded from the above information that heat ratings as high as 23 to 24 kW/ft could be achieved in the fuel elements without fuel centerline melting. Linear heat ratings in the Cycle 1 core design fell significantly below this limit even at the 112% overpower condition. Heating ratings and burnups for this design were well demonstrated by the existing technology. Nevertheless, it was felt fruitful to consider the question of what constitutes a fuel element failure. For one, the cladding must be violated. On the subject of the influence of expected transients, a conservative analysis had been presented of the factors which influence cladding performance during such transients. The fuel rod cladding was designed on a conservative basis and the calculations considered limiting cases and limiting assumptions. Consideration of peaking factor reductions shown in Table 3.6-1 increased the conservatism of these analyses. The analyses had been conducted throughout with design BOL power density, although it was known that for fuel in its third burnup cycle, LPD would be substantially below these values. Thus, the LPD increase which might be associated with overpower transients near end of fuel life had been conservatively considered. Cladding integrity had been demonstrated even under these adverse conditions. Consideration of peaking factor decreases noted in Table 3.6-1 made this analysis even more conservative.

Present heat rating limits are based on LOCA/Emergency Core Cooling System stored energy considerations and are included in Section 14.17.

3.

6.5 REFERENCES

1. M.L. Bleiberg, R.M. Berman, and B. Lustman, "Effects of High Burnup on Oxide Ceramic Fuel," WAPD-T-1455, March 1962 2. B. Lewis, "Engineering for the Fission Gas in UO2 Fuel," Nuclear Applications, Vol. 2, No. 2, April 1966 3. J.J. Holmes, J.A. Williams, D.H. Nyman, and J.C. Tobin, "In-Reactor Creep of Cold Worked Zircaloy Z," Flow and Fracture of Metals and Alloys in Nuclear Environments, ASTM-STP-380, 1965 4. E. Duncombe, J.E. Meyer, and W.A. Coffman, "Comparison with Experiment of Calculated Dimensional Changes and Failure Analysis of Irradiated Bulk Oxide Fuel Test Rods Using the CYGRO-I Computer Program," WAPD-TM-583, September 1966 5. Notely, Bain, and Robertson, "The Longitudinal and Diametral Expansion of UO2 Fuel Elements," AECL-2143, November 1964 6. M.J.F. Notely and J.R. MacEwan, "The Effect of UO2 Density on Fission Product Gas Release and Sheath Expansion," AECL-2230, March 1965 7. B. Lustman, "Fuel Clad Design Basis for Thermal Reactors," Bettis Atomic Power Laboratory, May 1966 8. P.J. Pankaskie, "Creep Properties of Zircaloy-2 for Design Application," HW-75267, October 1962 9. Indian Point Nuclear Generating Unit No. 2, Preliminary Safety Analysis Report, Appendix A 10. J.T. Stiefel, H. Feinroth, and G.M. Oldham, "Shippingport Atomic Power Station Operating Experience, Developments and Future Plans," WAPD-TM-390, April 1963 CALVERT CLIFFS UFSAR 3.6-7 Rev. 47 11. Question V.B.2, Prairie Island Nuclear Generating Plant, Preliminary Safety Analysis Report, (Docket No. 50-306) 12. T.D. Anderson, "Effects of High Burnup on Bulk UO2 Fuel Elements," Nuclear Safety, Vol. 6, No. 2 Winter 1964-1965, p. 164-169 13. R.D. MacDonald, et al, "Zircaloy-2 Clad Fuel Elements Irradiated to a Burnup of 10,000 MWd/MTU," AECL-1952, 1964 14. R.S. Miller, et al, "Operating Experience with the Saxton Reactor Partial Plutonium Core-II" paper presented at AEC Plutonium Meeting in Phoenix, August 1967 15. C.J. Baroch, J.P. Hoffmann, H.E. Williamson, and T.J. Pashos, "Comparative Performance of Zircaloy and Stainless Steel Clad Fuel Rods Operated to 10,000 Mwd/T in the VBWR," GEAP-4849, April 1966 16. F.H. Megerth, "Zircaloy-Clad UO2 Fuel Rod Evaluation Program, Quarterly Progress Report No. 2, February 1968 - April 1968," GEAP-5624 (May 1968) 17. R.M. Berman, H.B. Meieran, and P. Patterson, "Irradiation Behavior of Zircaloy-Clad Fuel Rods Containing Dished-End UO2 Pellets," (LWBR-LSBR Development Program), WAPD-TM-629, July 1967 18. J.T. Engel, et al, "Performance of Fuel Rods Having 97 Percent Theoretical Density UO2 Pellets Sheathed in Zircaloy-4 and Irradiated at Low Thermal Ratings," (LSBR/LWBR Development Program), WAPD-TM-631, July 1968 19. J.E. McCauley, et al, "Evaluation of the Irradiation Behavior of a Zircaloy-4 Clad Fuel Rod Containing for Density UO2 Fuel Pellets," LWBR-LSBR Development Program, WAPD-TM-596, January 1968 20. J.P. Blakely, "Action on Reactor and Other Projects Undergoing Regulatory Review of Consideration" Nuclear Safety, Vol. 9, No. 4, p. 326 (July-August 1968) 21. S.Y. Ogawa, Final Report, "Power Reactor High Performance UO2 Program," Joint US-Euratom Research and Development Report, GEAP-10042, June 1969 22. Summary description of Peach Bottom Atomic Power Station Units No. 2 and No. 3 and Review of Considerations Important to Safety, Docket No. 50-277 and 50-278 CALVERT CLIFFS UFSAR 3.6-8 Rev. 47 TABLE 3.6-1 TYPICAL PEAK BURNUP - MAXIMUM HEAT RELATIONSHIP MAXIMUM LOCAL EXPOSURE MWD/MTU TOTAL NUCLEAR PEAKING Factor MAXIMUM HEAT RATING kW/ft 24,200 2.86 17.5 24,200 - 36,000 2.86 17.5 36,000 - 48,500 2.42 14.9 CALVERT CLIFFS UFSAR 3.6-9 Rev. 47 TABLE 3.6-2 COMPARISON OF MAXIMUM HEAT RATINGS REACTOR kW/ft Maine Yankee 16.7 Fort Calhoun 17.1 Calvert Cliffs, Unit 1 17.5 Calvert Cliffs, Unit 2 17.5 Hutchinson Island, Unit 1 17.8 Millstone Unit 2 17.8 Turkey Point 17.3 Surrey 17.5 Prairie Island 17.4 Three Mile Island 17.5 Oconee 17.5 Indian Point, Unit 2 18.5 Diablo Canyon 18.9 Browns Ferry 18.5 Sequoyah 18.8 San Onofre, Units 2 and 3 18.5

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Calvert Cliffs Nuclear Power Plant UNIT 1 QUARTER-CORE ASSEMBLY MAP Figure 3.4-4 Revision 49 Calvert Cliffs Nuclear Power Plant UNIT 2 QUARTER-CORE ASSEMBLY MAP Figure 3.4-5 Revision 49

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Calvert Cliffs Nuclear Power Plant UNIT 1 ASSEMBLY RELATIVE POWER DENSITY AT 10,000 MWd/MTU, HFP, ARO, EQUILIBRIUM XENON Figure 3.4-10 Revision 49 Calvert Cliffs Nuclear Power Plant UNIT 2 ASSEMBLY RELATIVE POWER DENSITY AT 10,000 MWd/MTU, HFP, ARO, EQUILIBRIUM XENON Figure 3.4-11 Revision 49

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CALVERT CLIFFS UFSAR 4.4-1 Rev. 47 4.4 LOOSE PARTS DETECTION SYSTEM 4.4.1 DESIGN BASIS The loose parts monitoring system monitors the RCS for internal loose parts. The system is designed to detect a loose part striking the internal surface of RCS components with an energy level of one-half foot pound or more. 4.4.2 SYSTEM DESCRIPTION The loose parts monitoring system consists of transducers, preamplifiers, amplifiers and an analyzer to record the occurrence of a loose part within the RCS. Eight piezoelectric accelerometers are attached to the RCS boundary, where loose parts are most likely to become entrapped, as follows: - 2 on the reactor vessel lower head, diametrically opposed - 2 on the reactor vessel studs, diametrically opposed - 2 on the primary head of each SG. In addition, the RSGs are provided with two accelerometer attachment locations on the lower secondary shell adjacent to the tubesheet on the SG. These accelerometer attachment locations are provided for future use. The signals from the transducers are amplified inside the Containment and are then directed to the data acquisition and analysis system in the Control Room. Signals that exceed a fixed and floating setpoint value actuate an alarm. Additional information may be obtained from an analysis module that records and analyzes the signals from all eight channels. The audio output of any one of the eight channels can also be monitored through the use of a loudspeaker and a selector switch. 4.4.3 DESIGN ANALYSIS Experience in laboratories and on operating plants has shown that the signal-to-noise ratio resulting from a loose part with an energy level of one-half foot-pound or more is of sufficient magnitude that the loose part will be detected by the loose parts monitoring system. CALVERT CLIFFS UFSAR 4.5-1 Rev. 47 4.5 COASTDOWN OPERATION AT END OF CYCLE Toward the very end of a fuel cycle, the reactor core may reach a point at which it no longer has sufficient nuclear fuel to allow full power operation under normal operating conditions. Operation beyond that point is called coastdown operation. During coastdown operation, the reactor thermal power gradually decreases. However, it is possible to minimize (or even delay the start of) the loss of thermal power during the coastdown operation by reducing the RCS inlet temperature. Reducing the RCS inlet temperature at end of cycle adds positive reactivity to the core by taking advantage of the negative moderator temperature coefficient. The reload safety analyses support both a RCS bulk inlet temperature coastdown and a thermal power coastdown. The inlet temperature may be reduced to 537°F. As long as the inlet temperature remains 537°F, the thermal power level shall be 100% but the inlet temperature/power program limit. To operate below 537°F or below 31.25% power, the plant must be on the inlet temperature/power program shown in Figure 4-9. The coastdown must end when the burnup reaches the cycle specific burnup limit. Figure 4-18 shows the allowed combination of thermal power and RCS inlet temperature during coastdown operation.

TYPE 24" x 17 VENTURI BARTON MOD 200 ..._ _ __. No. I, 0-20 psi t 0.5% F.S. BARTON MOD 2 0-20 psi -0.5% F.S. BARTON MOD 200 No.4, 0-15 psi ___ _. +/-0.5% F.S. BARTON MOD 200 No.33898, 0-15 psi..._ _ __. +/-0.5% F.S. ROYLYN No. 5 0-3000 psi

• O. 25o/o .. . -BARTON MOD 227 1o----i 150 psid i: 0.5% F.S. CROSBY No. 4 0-3000 psi t 0.5% F.S. BALTIMORE GAS & ELECTRIC CO. Calvert Cliffs Nuclear Power Plant REACTOR COOLANT PUMP FLOW TEST INSTRUMENTATION Figure A Rev.18 TRUE PUMP CASE AP MEASURED TEST LOOP AP + 0.5% FS = + 0. 75 PSI AP* 85 PSI AP .. 85 + 0. 75 PSI AP max 85. 75 PSI 6. p INDIVIDUAL PUMP CURVE I : I I I I I ' I I I I 'f \ 92 500 .) ' 94, 940 I !::. p = 85 -(Q.005) 85 PSI TRUE TEST LOOP FLOW" TRUE. REACTOR COOLANT SYSTEM FLOW VENTURI t::.P + 1%0F THEORETICAL ' MEASURED VENTURI AP :!: 2.5% OF READING t . CALCULATED TEST LOOP FLOW ALLOWANCE FOR PUMP CASE AREA VARIATION 840 GPM APPARENT REACTOR COOLANT SYSTEM FLOW, WORST CASE ERRORS MEASURED PUMP A P IN REACTOR COOLANT SYSTEM :!: 0.5% OF READING t::.Pmin 84.57 Q = 92,500 t::,. Pc( 8. 64 x 109 + (. 02518. 64 xl09 A B. 86 x 109 Q=./4E T . Q .)s.86x109 qr.ax 94, 100 GPM Offiax94, 940 GPM 0niax = 96,100 GPM Figure B BALTIMORE GAS & ELECTRIC CO. Calvert Cliffs Nuclear Power Plant ERROR IN ESTABLISHMENT OF FLOW Rev. 18

Closure Stud Cooling Water STUFFING BOX COOLING WATER OUTLET Drive Mount Heat Exchanger Pump Case SALT IM ORE Figure GAS & ELECTRIC Co. Reactor Coolant Pump 4_4 Calvert Cliffs Nuclear Pt>Wer Plant FOR SEAL AREA SEE FIG, 4-6 AND FIG.4.6A) REV. I I 1/91 SEAL LEAKAGE CONNECTION CONTROLLED BLEEDOFF CONNECTION PRESSURE. TAP CONNECTIONS ._+-r11-t+ COOLING WATER CONNECTION BALTL'ltORE GAS & ELECTRIC Co. Calvert Cliffs Nuclear Pcwer Plant Figure Reactor Coolant Pump -Sea,......:.I A....:..r...:.e...;,_a __ -f 4-6 REV. II 1/91 TM BALANCED STATOR SEAL DESIGN FEATURES STATIONARY SEAL 'GLAND' ANTI---. ROTATION LUG SEAL SETTINGl GAP --_ COMPRESSION SPRING ROT AT ING ..... SEAL RING ROTATING SUPPORT RING SECONDARY SEAL SLEEVE SECONDARY *sEAL' STATIONARY SEAL RING: CARRIER "CENTERING' "0-RJNGS' SHAFT SEAL SLEEVE BALTIMORE GAS & ELECTRIC Co. Calvert Cliffs Nuclear Pc<Wer Plant Reactor Coolant Pump -Seal Area ------t Figure 4-6A REV.I I 1/91 z 0 ::: > ("J tll §"Q(<:Ocv ::;::til> r r 0 -['Il....., :E (') (') §§: ::;-:r. (') r:; 5 (") 9 :::::0 CD Q,) !l. 0 ..., (""') 8 Q) :::J --0 c:: 3 "'C -0 CD ..., -0 ..., 3 Q,) :::J n CD .s:::.. :!:1 I l.C ........ c:: ..., CD NPSH in Feet of Water Total Head in Feet 500 400 300 200 100 0 0 ;/ Head Cap. B.H.P. at SP.GR.= 1.0 20 NPSH Req'd (Cold) 300 Efficiency Percent 200 r 90 100 0 80 70 60 50 40 30 20 10 0

600 590 L1... 580 a .. Vl 570 LIJ a:: ::i I-560 <( a:: LIJ Taverage CL 550 :a LIJ I-I-540 z Teo Id <( ...J a 532 a u 530 a:: a I-u 520 <( w a:: 510 500 0 10 20 30 40 50 60 70 NSSS Power. Percent 100% = 2700 Mwt NOTE: THIS FIGURE SHOWS MAXIMUMThot ANDTcvercge ASSUMING MINIMUM RCS FLOW. BALTIMORE GAS & ELECTRIC CO. Colver-t Cliffs Nuclear Power Plont TEMPERATURE CONTROL PROGRAM 601 OF 574.50F 548 OF 80 90 100 REV. Figure 24 4-9 (/) Q) I:. 0 c +-c *-0 a. +-Q) V> Q) > Q) ...J L Q) N L :::::> (/) (/) Q) L a.. 230 220 210 200 190 180 170 160 150 140 530 NOTE: 1510.0° F. 216.0"l <54.3.3° F, 158.4 ") 535 540 545 550 555 560 565 Measured RCS Average Terrperature -F THE PRESSURIZER LEVEL SET POINT PROGRAM MAY BE ADJUSTED FOR THE INCREASE IN Tove CAUSED BY SG TUBE PLUGGING. BALTIMORE GAS & ELECTRIC CO. Ca I vert CI i ffs Nuclear Power Plant Pressurizer Level Set Point Program 570 575 REV.24 Figure 4-10 c 0 Q. Q) (/) (/) Q) -I:. Q) () > c Q) -__J

• E L 0 0 L L L w_ -Q) Q) > > (!) (!) __J ...J "'O L Q) Q) L N :J (lj *-0 L Q) (/) <f--o a... c 0 +-0 > (!) Cl BALTIMORE L 0 L L w Q) > Q) _J .c. QI :c L 0 L L UJ Cl> > (!) ...J !t 0 ...J GAS & ELECTRIC CO. Colvert Cliffs Nuclear Power Plant 40 30 20 10 20 Notes: High level alarm---*-ON C+39"l I , +37"l OFF 1. Nominal letdown is 38 gpm to maintain pressurizer level at setpoint with one charging purrp operating continuosly. 2. Mox. letdown is 126 gpm. 3. Min. letdown is 30 gpm. ON Energizes a I I pressur i zer ----11..... < +1 3") heaters and backup signal I to step backup charging purrps c +S" 1 OFF C-6" l OFF IC-10"> ( -15 It) OIFF( -4 "l C-9" l ON OFF Storts charging .. *--pump chosen as first backup Starts charging purrp chosen as second backup ON Low level olorm ond b<Jckup signal to start all charging pumps Revision 24 Pressurizer Level Control Program Figure 4-11 Nil-Ductility Transition Temp Increase, of 400 300 200 100 Legend o -A302B Plate * -A533B Plate x -A533B Welds * -A533B Heat Affected Zone 5 1018 1019. 1o20 G/\S & ELECTRIC CO. Fluence, nlcm2 > 1 MeV C-E Design Curve of NDTT Increase (55D°F I rradiati0n) Figure 4-12 Calvert Cliffs Nuclear Power Plant .___ ______ -.Jo... ________ --*---* __________ __...._ __ _

z Cl i:: > ri en -n .. <'C"l > i:irorr 0 1:1 :::::: a:l ;;v ;-Ill () [Tl () 9 b n 9l. 180° (-------... t I / *,/outlet Nozzle I ' t ' I , 7 I ' ........__ ' _, ___ ,"'"'\ \Inlet \Nozzle \ ) -, 0 :::J 114......J Core Support Barrel I I ,1 0 -Vl I 97° < CD Ql :::J £ jg30 (""') Ql "C "' c: CD )> "' "' C1> 3 CT CD "' J::ir, '° .,!.. c: w ..., CD ,..,, I I ,.J \ ' \ \ \ \ ........ Enlarged Plan View Reactor Vessel I .... ' ' ' I I I -. _______ .,,. 00 I I I I , ____ ....... ,,, Core Midplane Vessel Capsule Assembly1 ' ,.... r4 Core Reactor Vessel Elevation View Support Barrel Tensile -Monitor----... ..... , Compartment Tensile -Monitor Compartment Tensile -Monitor----1 Compartment BAL ---Lock Assembly } Wedge Coupling Assembly Charpy Impact Compartmen1s Charpy Impact Compartments GAS & ELECTRIC CO. Calvert Cliffs Nuclciu Power Pinnt Typical Surveillance Capsule Assembly 1gure 4-14 Wedge Coupling -End Cap Charpy Impact Specimens Rectangular Tubing Wedge Coupling -End Cap BALTIMO!m GAS & ELECTRIC CO. Calvert Cliffs Nudear Powc-r Pinnt Typical Charpy Impact Compartment Assembly F1gu re 4-15 Wedge Coupling -End Stainless Steel Tubing Threshold Detector Flux Spectrum Monitor Temperature Monitor Temperature Housing Tensile Specimen -Split Spacer -Tensile Specimen BALTIMORE Flux SP.ectrum Monitor Caam iu m shielded Stainless Steel Tubing Cadmium Shield Threshold Detector Weight Low Melting Alloy Rectangular Tubing Wedge Coupling -End Cap GAS & ELECTRIC CO. Calvert Cliffs Nuclear Plant Typical Tensile-Monitor Compartment Assembly 1gure 4-16

Calvert Cliffs Nuclear Power Plant CALVERT CLIFFS COASTDOWN OPERATING REGION Figure 4-18 Revision 49 CHAPTER 5 STRUCTURES TABLE OF CONTENTS PAGE5.0 STRUCTURES 5.1 CONTAINMENT STRUCTURE 5.2 ISOLATION SYSTEM 5.3 EXTERNAL MISSILES, SNUBBERS, AND WATERTIGHT DOORS 5.4 SYSTEM DESIGN EVALUATION 5.5 TESTS AND INSPECTION CHAPTER 5 STRUCTURES TABLE OF CONTENTS PAGE 5.6 OTHER STRUCTURES 5.7 CONTROL OF HEAVY LOADS APPENDIX 5A - STRUCTURAL DESIGN BASIS APPENDIX 5B - QUALITY CONTROLS CHAPTER 5 STRUCTURES TABLE OF CONTENTS PAGEAPPENDIX 5CSTUDY OF THE EFFECTS OF MISLOCATED VERTICAL TENDONSAPPENDIX 5DSTUDY OF UPPER VERTICAL TENDON BEARING PLATESAPPENDIX 5EREDUCTION IN CONTAINMENT PRESTRESS AND LONG-TERM CORRECTIVE ACTIONS FOR VERTICAL TENDON CORROSION(NOTE: Appendices 5A, 5B, 5C, 5D, and 5E are referenced here, but located in UFSAR Volume III.) CHAPTER 5 STRUCTURES LIST OF TABLES TITLEPAGE CHAPTER 5 STRUCTURES LIST OF FIGURES FIGURE CHAPTER 5 STRUCTURES LIST OF FIGURES FIGURE CHAPTER 5 STRUCTURES LIST OF ACRONYMS CALVERT CLIFFS UFSAR 5.1-1 Rev. 47 STRUCTURES 5.05.1 CONTAINMENT STRUCTURE 5.1.1 DESIGN BASIS General plans at various elevations and sections through the Containment Structure interior are shown in Figures 1-6 through 1-15 and show the general arrangement of various equipment such as reactor, steam generators, pressurizer and reactor coolant pumps. The support and anchorage details of these components are shown in Figures 5-11 through 5-15. At an Elevation 177'0", a landing platform, to the ladder on the vent stack provides an access route to the top of the Containment Structure. A steel ladder, provided inside the Containment Structure provides an access up to the polar crane. The Containment Structure is a Seismic Category I structure and is designed for all loading combinations described in Section 5A.3.

The Containment Structure completely encloses the Reactor Coolant System (RCS) to minimize release of radioactive material to the environment should a serious failure of the RCS occur. The structure provides adequate biological shielding for both normal operation and accident situations. The Containment Structure is designed for maximum of 0.16%/day leakage by weight of the original content of air at a design pressure of 50 psig and a concrete surface temperature of 276°F. The principal design basis for the structure is that it be capable of withstanding the internal pressure resulting from a loss-of-coolant accident (LOCA) with no loss of integrity. In this event, the total energy contained in the water of the RCS is assumed to be released into the Containment Structure through a break in the reactor coolant piping. Subsequent pressure behavior is determined by the building volume, engineered safety features, and the combined influence of energy sources and heat sinks.

Energy is available for release into the Containment Structure from the following sources: RCS Stored Heat Reactor Stored Heat Reactor Decay Heat Metal-Water Reactions The energy release and the containment pressure transient curves are shown in Section 14.20. The design of the engineered safety features systems and their operation is discussed more fully in Chapter 6; only their relation to the basis of Containment Structure design is discussed below. The engineered safety features systems are provided to limit the consequences of an accident. Their energy removal capabilities limit the internal pressure so that Containment Structure design limits are not exceeded and the potential for release of fission products is minimized.

The safety injection systems inject borated water into the reactor vessel to remove core decay heat and to minimize metal-water reactions and the associated release of heat and fission products. Flashed primary coolant, RCS sensible heat, and core decay heat transferred to the Containment Structure are removed by the Containment Cooling System which is comprised of two subsystems; the containment spray subsystem and the air recirculation subsystem.

The containment spray subsystem reduces pressure in the containment by condensing the Containment Structure steam and removing heat from the containment atmosphere by recirculation of the spray water through the shutdown cooling heat exchangers. CALVERT CLIFFS UFSAR 5.1-2 Rev. 47 The air recirculation subsystem reduces pressure and removes heat directly from the Containment Structure atmosphere to the Service Water System with recirculating fans and cooling coils. 5.1.2 DESIGN CRITERIA 5.1.2.1 General Description The Containment Structure houses the RCS. Its purpose is to contain any accidental release of radioactivity from the RCS. It is designated as a Seismic Category I structure. The basic design criteria are that the integrity of the liner plate be guaranteed under all loading conditions and the structure shall have a low-strain elastic response such that its behavior will be predictable under all design loadings. The structure consists of a post-tensioned reinforced concrete cylinder and dome connected to and supported by a massive reinforced concrete foundation slab as shown in Figure 5-1. The entire interior surface of the structure was lined with a 1/4" thick welded American Society for Testing and Materials (ASTM) A36 steel plate to assure a high degree of leak tightness. Numerous mechanical and electrical systems penetrate the Containment Structure wall through welded steel penetrations as shown in Figures 5-2 and 5-3. The penetrations and access openings were designed, fabricated, inspected, and installed in accordance with the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel (B&PV) Code, Section III, Class B. Principal dimensions of the Containment Structure are: Inside Diameter 130' Inside Height (including Dome) 181-2/3' Vertical Wall Thickness 3-3/4' Dome thickness 3-1/4' Foundation Slab Thickness 10' Liner Plate Thickness 1/4" Internal Free Volume 1,989,000 ft3 The Containment Structure is shown in Figures 1-6 through 1-16. In the concept of a post-tensioned Containment Structure, the internal pressure load is balanced by the application of an opposing external force on the structure. Sufficient post-tensioning was used on the cylinder and dome to more than balance the internal pressure so that a margin of external pressure exists beyond that required to resist the LOCA pressure. See Appendix 5E for an evaluation that reduced the original containment minimum design prestress. Nominal, bonded reinforcing steel was also provided to distribute strains due to shrinkage and temperature. Additional bonded reinforcing steel was used at penetrations and discontinuities to resist local moments and shears. The internal pressure loads on the foundation slab are resisted by both the external bearing pressure due to dead load and the strength of the reinforced concrete slab. Thus, post-tensioning was not required to exert an external pressure for this portion of the structure.

CALVERT CLIFFS UFSAR 5.1-3 Rev. 47 The post-tensioning system consists of: a. Three groups of 68 dome tendons oriented at 60° to each other for a total of 204 tendons anchored at the vertical face of the dome ring girder. b. Two hundred four vertical tendons anchored at the top surface of the ring girder and at the bottom of the base slab. c. Six groups of 78 hoop tendons, each enclosing 120° of arc, for a total of 468 tendons anchored at the 6 vertical buttresses. Each tendon consists of approximately 90 1/4" diameter wires with button-headed BBRV-type anchorages, furnished by the Prescon Corporation. The tendons are housed in spiral wrapped, corrugated, thin-wall, carbon steel sheathing. After fabrication, the tendon was shop dipped in a petrolatum corrosion protection material, bagged and shipped. After installation, the tendon sheathing was filled with a corrosion preventive grease (Viconorust 2090P). The ends of all tendons were covered with pressure-tight, grease-filled caps for corrosion protection. All the vertical tendons for each Unit received new corrosion preventive grease between 1997 and the end of 2002. Some original vertical tendons for each Unit were restressed or replaced with a new tendon between 2001 and 2002. See Appendix 5E for details.

American Society for Testing and Materials A615, Grade 60 reinforcing steel, mechanically spliced as needed with B- and T-series CADWELDS, was used throughout the foundation slab and around the large penetrations. The same type of steel was used for the bonded reinforcing throughout the cylinder and dome as crack control reinforcing and at areas of discontinuities to provide an additional margin of elastic strain capability.

The 1/4"-thick liner plate was attached to the concrete by means of an angle grid system stitch welded to the liner plate and embedded in the concrete. The details of the anchoring system are provided in Figure 5-1. The frequent anchoring is designed to prevent significant distortion of the liner plate during accident conditions and to insure that the liner maintains its leak-tight integrity. The design of the liner anchoring system also considers the various erection tolerances and their effect on its performance. The liner plate was protected from corrosion on the inside with 3 mils of inorganic zinc primer topped with 6 mils of an organic epoxy up to Elevation 75'0", and 3 mils of an inorganic topcoat above that elevation. There is no paint on the side in contact with concrete.

The aggregate used in the structure produced an excellent high-strength, dense, sound concrete. The 28-day design strengths were 5000 psi for the shell and 4000 psi for the foundation slab.

Personnel and equipment access to the structure is provided by a two-door personnel lock with double seals on both doors and by a 19'0" clear diameter, double gasketed, single-door equipment hatch as shown in Figure 5-3. A two-door emergency personnel escape lock is also provided. These locks and hatch were designed and fabricated from A516, Grade 70 firebox quality steel made to A300 specification and Charpy V-notch impact tested to 0°F in accordance with the ASME, B&PV Code, Section III. All piping penetrations furnished adhered to the same requirements.

A containment outage door on the exterior of each Containment Structure at the equipment hatch opening serves as a substitute for the equipment hatch when setting containment closure during Modes 5 and 6 conditions. The containment CALVERT CLIFFS UFSAR 5.1-4 Rev. 47 outage door was designed, fabricated, examined, inspected and tested to the requirements of the 1995 Edition of ASME Section VIII, Rules for Construction of Pressure Vessels, Division 1. However, the containment outage door cannot be credited for severe weather. If Emergency Response Plan Implementation Procedure 3.0, Preparing for Severe Weather, is implemented and requires containment closure to be established, then the equipment hatch must be used.

Structural brackets provided for the Containment Structure polar crane runway were fabricated from ASTM A36 steel shapes and ASTM A516, Grade 70 insert plates (Figure 5-1). Like the penetration assemblies, structural brackets and thickened plates were shop fabricated, stress relieved and shipped to the job site for welding to the 1/4" liner plate.

The strength of the Containment Structure at working stress and overall yielding was compared to various loading combinations to assure safety. The Containment Structure was examined with respect to strength, the nature and the amount of cracking, the magnitude of deformation, and the extent of corrosion to assure proper performance. The structure was designed and constructed in accordance with design criteria based upon American Concrete Institute (ACI) 318-63, ACI 301-66, and ASME, B&PV Code, Sections III, VIII, and IX to meet the performance and strength requirements prior to prestressing, at transfer of prestress, under sustained prestress, at design loads, and at yield loads.

The structure was originally analyzed using Bechtel's Finite Element Program for Cracking Analysis CE 316-4, for individual and various combinations of loading cases of dead load, live load, prestress, temperature and pressure. The computer output included direct stresses, shear stresses, principal stresses and displacements of each nodal point. See Appendix 5E for an evaluation that reduced the original containment minimum design prestress. Stress plots which showed the total stresses from appropriate combinations of loading cases were made and areas of high stress were identified. The modulus of elasticity was corrected to account for the nonlinear stress-strain relationship at high compression in these areas and stresses were recomputed.

In order to consider creep deformation, the modulus of elasticity of concrete under sustained loads such as dead load and prestress was differentiated from the modulus of elasticity of concrete under instantaneous loads such as internal pressure and earthquake loads.

The forces and shears were added over the cross-section and the total moment, axial force and shear were determined. From these values, the straight-line elastic stresses were computed and compared to the allowable values. The ACI 318-63 design methods and allowable stresses were used for concrete and prestressed and unprestressed reinforcing steel except as noted in the design criteria.

It is the intent of the criteria to provide a structure of unquestionable integrity that will meet the postulated design conditions with a low strain elastic response. The Calvert Cliffs Containment Structure meets these criteria because: (See Appendix 5E for an evaluation that reduced the original containment minimum design prestress.)

CALVERT CLIFFS UFSAR 5.1-5 Rev. 47 a. The design criteria are in general based on the proven stress and strain to meet the ACI or ASME codes. Departures from or additions to these codes have been made in the following manner: 1. The environmental conditions of severity of load cycling, weather, corrosion conditions, maintenance, and inspection for this structure have been compared and evaluated with those for code structures to determine the appropriateness of the modifications. 2. The consultant firm of T.Y. Lin, Kulka, Yang and Associate was retained on earlier projects to assist in the development of design criteria. In addition to assisting with the criteria submitted in the Preliminary Safety Analysis Report, they were involved in the review of design methods to assure that the criteria were implemented as intended. 3. Dr. Alan H. Mattock of the University of Washington was retained on earlier projects to assist in developing the proper design criteria for combined shear, bending and axial load. 4. All criteria, specifications and details relating to liner plate and penetrations and corrosion protection have been referred to Bechtel's Metallurgy and Quality Control Department. This department maintains a staff to advise the corporation on problems of welding, quality control, metallurgy and corrosion protection. 5. The design of the Calvert Cliffs Containment Structure was continually reviewed as the criteria were revised for successive license applications. b. The primary membrane integrity of the structure is provided by the unbonded post-tensioning tendons, each one of which is stressed to 80% of ultimate strength during installation and performs at approximately 50 to 60% during the life of the structure. Thus, the main strength elements are individually proof-tested prior to operation of the plant. c. Eight-hundred-seventy-six such post-tensioning elements have been provided, 204 in the dome, and 204 vertical and 468 hoop tendons in the cylinder. Any three adjacent tendons in any of these groups can be lost without significantly affecting the strength of the structure due to the load redistribution capabilities of the shell structure. The bonded reinforcing steel provides for crack control assures that this redistribution capability exists. d. The unbonded tendons are continuous from anchorage to anchorage, being deflected around penetrations and isolated from secondary strains of the shell. Thus, the membrane integrity of the shell can be assured regardless of conditions of high local strains. e. The unbonded tendons exist in the structure at a slightly ever-decreasing stress due to relaxation of the tendon and creep of the concrete and, even during pressurization, are subject to a stress change of very small magnitude (2% to 3% of ultimate strength). Thus, the main structural system is never subjected to large changes in load, even during accident conditions. f. The concrete portion of the structure, similar to the tendons, was subject to the highest state of stress during the initial post-tensioning. During pressurization, it is subject to a large change in load (or state of stress) but the change is, in general, a decrease in load. The large membrane compressive forces are diminished and/or replaced by relatively small radial pressures and stresses. CALVERT CLIFFS UFSAR 5.1-6 Rev. 47 g. The deformations of the structure during plant operation, or due to accident conditions, are relatively minor due to the low-strain behavior of the concrete. The largest deformations occurred at the time of initial post-tensioning and shortly thereafter, prior to operation. This low strain behavior and the inherent strength of the structure permit the anchoring of all piping penetrating the structure directly to the shell. Such details (Figure 5-2) eliminate the use of expansion bellow seals and significantly reduce the likelihood of leaks developing at the penetrations. The exception to this is when the fuel transfer tube is in use, requiring use of the transfer tube bellows (Section 5.1.4.4.d). 5.1.2.2 Loads Prior to prestressing, the structure was designed as a conventionally reinforced concrete structure. It is designed for dead load, live loads and a reduced-wind load. Allowable stresses are computed in accordance with ACI 318-63. Loads at Transfer of Prestress See Appendix 5E for an evaluation that reduced the original containment minimum design prestress. The Containment Structure is checked for prestress loads and the stresses compared with those allowed by ACI 318-63 with the following exceptions: ACI 318-63, Chapter 26, allows concrete stress of 0.60 at initial transfer. In order to limit creep deformations, the membrane compression stress is limited to 0.30 whereas in combination with flexural compression the maximum allowable stress will be limited to 0.60 per ACI 318-63. For local stress concentrations with nonlinear stress distribution as predicted by the finite element analysis, 0.75 is permitted when local bonded reinforcing is included to distribute and control the localized strains. These high local stresses are present in every structure but they are seldom identified because of simplifications made in design analysis. These high stresses are allowed because they occur in a very small percentage of the cross-section, are confined by material at lower stress and would have to be considerably greater than the values allowed before significant local plastic yielding would result. Bonded reinforcing was added to distribute and control these local strains.

Membrane tension and flexural tension are permitted provided they do not jeopardize the integrity of the liner plate. Membrane tension is permitted to occur during the post-tensioning sequence but will be limited to . When there is flexural tension but no membrane tension, the section is designed in accordance with the ACI Code, Section 2605(a). The stress in the liner plate due to combined membrane tension and flexural tension is limited to 0.5 fy. Shear criteria are in accordance with the ACI 318-63 Code, Chapter 26, as modified by the equations in the structural yielding subsection of this section using a load factor of 1.5 for shear loads. Loads Under Sustained Prestress See Appendix 5E for an evaluation that reduced the original containment minimum design prestress.

The conditions for design and the allowable stresses for this case are the same as above except that the allowable tensile stress in unprestressed reinforcing is CALVERT CLIFFS UFSAR 5.1-7 Rev. 47 limited to 0.5 fy. ACI 318-63 limits the concrete compression to 0.45 for sustained prestress load. Values of 0.30 and 0.60 are used as described above, which bracket the ACI allowable value. However, with these same limits for concrete stress at transfer of prestress, the stresses under sustained load are reduced due to creep, shrinkage, relaxation, and possible tendon wire breakage. See Appendix 5E for a discussion on possible tendon wire breakage. At Design Loads This loading case is the basic "working stress" design. The Containment Structure is designed for the following loading cases: a. D + F + L (Construction case) b. D + F + L + To + E (Operating case) c. D + F + L + P + TA (Design incident case) d. D + F + L + 1.15P (Test case)

e. D + F + L + Ts + E (Prolonged shutdown case) D = Dead Load L = Appropriate Live Load F = Appropriate Prestressing Load P = Design Pressure To = Thermal Loads Due to Operating Temperature TA = Thermal Loads Corresponding to Pressure P E = Operating Basis Earthquake (OBE) of 0.08g Ts = Thermal Loads Due to Transient Wall Temperatures Over a Prolonged Shutdown (20°F outside face, 50°F inside face) Sufficient prestressing is provided in the cylindrical and dome portions of the vessel to eliminate membrane tensile stress (tensile stress across the entire wall thickness) under design loads. Flexural tensile cracking is permitted but is controlled by bonded reinforcing steel.

Under the design loads, the same performance limits given for loads at transfer of prestress apply with the following exceptions: a. If the net membrane compression is below 100 psi, it is neglected and a cracked section is assumed in the computation of flexural bonded reinforcing steel. The allowable tensile stress in bonded reinforcing is 0.5 fy. b. When the maximum flexural stress does not exceed and the extent of the tension zone is not more than 1/3 the depth of the section, bonded reinforcing steel is provided to carry the entire tension in the tension block. Otherwise, the bonded reinforcing steel is designed assuming a cracked section. When the bending moment tension is additive to the thermal tension, the allowable tensile stress in the bonded reinforcing steel is 0.5 fy minus the stress in reinforcing due to the thermal gradient as determined in accordance with the method of ACI-505. c. The problem of shear and diagonal tension in a prestressed concrete structure should be considered in two parts: membrane principal tension and flexural principal tension. Since sufficient prestressing is used to eliminate membrane tensile stress, membrane principal tension is not critical at design loads. Membrane principal tension due to combined CALVERT CLIFFS UFSAR 5.1-8 Rev. 47 membrane tension and membrane shear is considered in the next subsection. Flexural principal tension is the tension associated with bending in planes perpendicular to the surface of the shell and shear stress normal to the shell (radial shear stress). The present provisions of ACI 318-63, Chapter 26 for shear are adequate for design purposes with proper modifications as discussed in the next subsection using a load factor 1.5 for shear loads. Crack control in the concrete is accomplished by adhering to the ACI-ASCE Code Committee standards for the use of reinforcing steel. These criteria are based upon a recommendation of the Prestressed Concrete Institute and are as follows: 0.25 percent reinforcing shall be provided at the tension face for small members 0.20 percent for medium size members 0.15 percent for large members A minimum of 0.20% bonded steel reinforcing is provided in two perpendicular directions on the exterior faces of the wall and dome for proper crack control. The liner plate is attached on the inside faces of the wall and dome. Since, in general, there is no tensile stress due to temperature on the inside faces, bonded reinforcing steel is not necessary there.

Loads Necessary to Cause Structural Yielding The structure is checked for the factored loads and load combinations that will cause structural yielding.

The load factors are the ratio by which loads will be multiplied for design purposes to assure that the load/deformation behavior of the structure is one of elastic, low-strain behavior. The load factor approach was used in this design as a means of making a rational evaluation of the isolated factors which must be considered in assuring an adequate safety margin for the structure. This approach permits the designer to place the greatest conservatism on those loads most subject to variation and which most directly control the overall safety of the structure. It also places minimum emphasis on the fixed gravity loads and maximum emphasis on accident and earthquake or wind loads. The final design of the structure satisfies the load combinations and factors shown in Appendix 5A.

The load combinations, considering load factors referenced above, are less than the yield strength of the structure. The yield strength of the structure is defined as the upper limit of elastic behavior of the effective load carrying structural materials. For steels (both prestressed and unprestressed) this limit is taken to be the guaranteed minimum yield given in the appropriate ASTM specification. For concrete, it is the ultimate values of shear (as a measure of diagonal tension) and bond per ACI 318-63 and the 28-day ultimate compressive strength for concrete in flexure . The ultimate strength assumptions of the ACI Code for concrete beams in flexure are not allowed; that is, the concrete stress is not allowed to go beyond yield.

The maximum strain due to secondary moments, membrane loads and local loads exclusive of thermal loads is limited to that corresponding to the ultimate stress divided by the modulus of elasticity and a straight-line distribution from there to the neutral axis assumed. CALVERT CLIFFS UFSAR 5.1-9 Rev. 47 For the loads combined with thermal loads the peak strain is limited to 0.003 in./in. For concrete membrane compression, the yield strength is assumed to be 0.85 to allow for local irregularities, in accordance with the ACI approach. The reinforcing steel forming part of the load carrying system is allowed to go to, but not to exceed, yield as is allowed for ACI ultimate strength design.

A further definition of yielding is the deformation of the structure which causes strains in the steel liner plate to exceed 0.005 in./in. The yielding of unprestressed reinforcing steel is allowed, either in tension or compression, if the above restrictions are not violated. Yielding of the prestressed tendons is not allowed under any circumstances.

Principal concrete tension due to combined membrane tension and membrane shear, excluding flexural tension due to bending moments or thermal gradients, is limited to . Principal concrete tension due to combined membrane tension, membrane shear, and flexural tension due to bending moments or thermal gradients is limited to . When the principal concrete tension exceeds the limit of , bonded reinforcing steel is provided in the following manner: a. Thermal Flexural Tension - Bonded reinforcing steel is provided in accordance with the methods of ACI-505. The minimum area of steel provided is 0.20% in each direction. b. Bending Moment Tension - Sufficient bonded reinforcing steel is provided to resist the bending moment on the basis of cracked section theory using the yield stresses stated above with the following exception: When the bending moment tension is additive to the thermal tension, the allowable tensile stress in the reinforcing steel is fy minus the stress in reinforcing due to the thermal gradient, as determined in accordance with the methods of ACI-505. Shear stress limits and shear reinforcing for radial shear are in accordance with ACI 318-63, Chapter 26 with the following exceptions:

Formula 26-12 of the Code was replaced by: where but not less than 0.6 for p' 0.003. For p' >0.003, the value of K shall be zero. fpe = Compressive stress in concrete due to prestress applied normal to the cross-section after all losses (including the stress due to any secondary moment) at the extreme fiber of the section at which tension stresses are caused by live loads. fn = Stress due to axial applied loads (fn shall be negative for tension stress and positive for compression stress). CALVERT CLIFFS UFSAR 5.1-10 Rev. 47 fi = Stress due to initial loads at the extreme fiber of a section at which tension stresses are caused by applied loads including the stress due to any secondary moment (fi shall be negative for tension stress and positive for compression stress). n = , constant in value of K above p' = ratio of compression steel area to area concrete V = Shear at the section under consideration due to the applied loads. M' = Moment at a distance d/2 from the section under consideration, measured in the direction of decreasing moment, due to applied loads. Vi = Shear due to initial loads (positive when initial shear is in the same direction as the shear due to applied loads). The lower limit placed by ACI 318-63 on Vci of 1.7 b'd is not applied. Formula 26 -13 of the Code was replaced by: Where Vp = radial shear component of effective prestress due to tendon curvature at the section considered, and the term fn is as defined above. All other notations are in accordance with ACI 318-63, Chapter 26. It should be noted that this formula is based on the tests and work done by Dr. A. H. Mattock of the University of Washington, and has been included in ACI 318-77, Section 11.5.2. When the above-mentioned equations show that allowable shear in concrete is zero, radial horizontal shear ties are provided to resist all the calculated shear. Other Design Loads The Containment Structure shell is designed for the following loads: a. Dead load b. Prestress forces

c. Live load including allowances for piping, ductwork and cable trays

d. Wind, including tornado

e. Earthquake f. Thermal expansion of pipes attached to the Containment Structure wall g. Uplift due to buoyant forces Transients resulting from the LOCA and other lesser incidents are presented in Chapter 14 and serve as the basis for the Containment Structure design pressure of 50 psig and a design concrete surface temperature of 276°F. The external design pressure of the Containment Structure shell is 3 psig. This value is approximately 0.5 psig beyond the maximum external pressure that could be developed if the Containment Structure were sealed during a period of low barometric pressure and high temperature and, subsequently, the Containment Structure atmosphere was cooled with a concurrent rise in barometric pressure.

Vacuum breakers are not provided.

CALVERT CLIFFS UFSAR 5.1-11 Rev. 47 5.1.2.3 Equipment Supports a. Reactor Vessel Supports 1. Restrain the vessel to maintain the integrity of emergency core cooling systems and to prevent the rupture of additional primary pipes should LOCA occur due to single pipe rupture; 2. Permit slow radial thermal expansion of the vessel under normal operation; and 3. Restrain the vessel against seismic and LOCA jet forces. b. Steam Generator Supports 1. Restrain the vessel to prevent simultaneous rupture of the primary coolant pipe, and the steam or the feedwater pipes; 2. Permit slow thermal growth of the loops and the vessel; and 3. Restrain all motion under seismic or LOCA loads. Calvert Cliffs Nuclear Power Plant Units 1 and 2 are approved for leak-before-break based on References 6 and 8, and compliance with Regulatory Guide 1.45 for leak detection as documented in UFSAR Section 4.3.1. As a result of the application of leak-before-break, the mechanical/structural loads associated with the dynamic effects of a large break LOCA in the RCS hot leg or cold legs are no longer considered part of the plant design basis. Accordingly, leak-before-break is credited in the design of the steam generator sliding base supports for the replacement steam generators.

c. Pressurizer Support 1. Support normal operating loads; and 2. Restrain the vessel under seismic loads. d. Main Coolant Pumps Supports 1. Permit slow thermal movements of the pump; and

2. Restrain the pump under seismic loads. e. Safety Injection Tank Support 1. Support normal operating loads; and 2. Restrain the tank under seismic loads. Materials for Equipment Supports a. Reactor Vessel 1. Plate material ASTM A302, Gr B 2. Structural shapes ASTM A441 3. Anchor bolts ASTM A354, Gr BC 4. Welding electrodes ASTM A233, E 7018 b. Steam Generator 1. Plate material ASTM A302, Gr B 2. Anchor bolts ASTM A490 3. Welding electrodes ASTM A233, E 7018 CALVERT CLIFFS UFSAR 5.1-12 Rev. 47 c. Pressurizer 1. Anchor bolts ASTM A354, Gr BC d. Main Coolant Pump 1. Structural shapes ASTM A36 2. Machine bolts ASTM A325 3. Steel plate (S. hangers) ASTM A515, Gr 65 4. Pipe section (S. hangers) ASTM A53 5. Machined rod (S. hangers) AISI C-1015 6. Spring coils (S. hangers) AISI 4161 H e. Safety Injection Tank 1. Anchor bolts ASTM A354, Gr BC Allowable Stresses and Strain for Equipment Supports Excluding Supports for Piping and Vessels: a. Stresses under normal operating loads and the OBE are within the applicable code allowable limits. b. Stresses under combined normal, Safe Shutdown Earthquake (SSE) and accident loads are kept below the yield strength of the material. c. Allowable local and average strains are determined by the maximum permissible deformation for the support. 5.1.3 CONTAINMENT STRUCTURE DESIGN ANALYSIS The analysis for the Containment Structure shell falls into two parts, axisymmetric and non-axisymmetric. The axisymmetric analysis is performed through the use of a finite element computer program for the individual loading cases of dead load, live load, temperature, prestress and pressure, as described in Section 5.1.3.1. The axisymmetric finite element representation of the Containment Structure shell does not include the buttresses, penetrations, brackets, and anchors. These items of configuration, the lateral loads due to earthquakes or winds, and various concentrated loads are considered in the non-axisymmetric analysis described in Section 5.1.3.2.

This section includes only analytical techniques, references and design philosophy. The results of these analyses are shown in Section 5.1.4. The design criteria and analysis have been reviewed by Bechtel's consultants, T.Y. Lin, Kulka, Yang and Associate. 5.1.3.1 Axisymmetric Techniques The finite element technique is a general method of structural analysis in which the continuous structure is replaced by a system of elements (members) connected at a finite number of nodal points (joints). Conventional analysis of frames and trusses can be considered an example of the finite element method. In the application of the method to an axisymmetric solid structure such as the Containment Structure shell, the continuous structure is replaced by a system of rings of quadrilateral cross-section which are interconnected along circumferential joints. Based on energy principles, sets of work equilibrium equations are formed in which the radial and axial displacements at the circumferential joints are the CALVERT CLIFFS UFSAR 5.1-13 Rev. 47 unknowns. The results of the solution of this set of equations are the deformations of the structure under the given loading conditions. For the output, the stresses are computed knowing the strain and stiffness of each element. The original finite element mesh used to describe the structure is shown in Figure 5-4 (see Sheets 1 and 2). See Appendix 5E for an evaluation that reduced the original containment minimum design prestress. The upper and lower portions of the structure were analyzed independently to permit the use of a greater number of elements for those areas of the structure of major concern, e.g., the ring girder area and the base of the cylinder. The finite element mesh of the base slab was extended down into the foundation to take into consideration the elastic nature of the foundation material and its effect upon the behavior of the base slab. The tendon access gallery was designed as a separate structure. The finite element mesh for the Containment Structure does not include the interior structures. The interior structures were included in the finite element input as a lumped mass. The finite element analysis produces stresses due to axisymmetric loads. The stresses from interior structure loads and earthquake loads are superimposed on the finite element stresses. The final summation of all stresses was used to design the base slab, exterior shell and dome. The use of Bechtel's finite element computer program, CE 316-4, permitted an accurate estimate of the stress pattern at various locations of the structure. The major benefit of the program is the capability to predict shears, normal forces and moments due to internal restraint and the interaction of the foundation base slab with the subgrade. The forces and moments were applied to all directions. The following material properties were used in the program for the various loading conditions: Econcrete' Foundation 3.64x106 psi Econcrete' Shell 4.07x106 psi concrete (Poisson's Ratio) 0.17 concrete (Coefficient of Expansion) 0.55x10-5 Eliner 29x106 psi Fy liner 34,000 psi Esoil (Construction and Operating Case, Figure 5-4, Sheet 1) 1st Layer E = 6,200 psi 2nd Layer E = 9,600 psi 3rd Layer E = 12,000 psi Esoil (Testing and Accident Case) 1st Layer E = 9,600 psi 2nd Layer E = 14,400 psi 3rd Layer E = 18,000 psi Esoil (Factored Load-Yield Stress) 1st Layer E = 10,000 psi 2nd Layer E = 20,000 psi 3rd Layer E = 30,000 psi CALVERT CLIFFS UFSAR 5.1-14 Rev. 47 In arriving at the above-tabulated values of E, the effect of creep is included by using the following equation for long-term loads such as thermal load, dead load and prestress: Ecs = Eci [i/(s + i)], where: Ecs = sustained modulus of elasticity of concrete, Eci = instantaneous modulus of elasticity of concrete, i = instantaneous strain, in./in. per psi, and s = creep strain, in./in. per psi The thermal gradients used in the design are shown in Figure 5-5. The design pressure and concrete surface temperature of 50 psig and 276°F became 75 psig and 276°F at factored conditions. The compressive stress and strain level is the highest (after the LOCA when the temperature is still relatively high, 200°F, and the pressure is dropping rapidly) at the inside face of the concrete at the edge of openings and also near the liner plate anchors. Neither concentration is a result of what may be considered a real load. In the case of an opening, the real stress is a result of prestress, reduced pressure and dead load. Applying stress concentration factors to these stresses maintains the concrete stress essentially in the elastic range. When the strain and resulting stress from the thermal gradient are also multiplied by a stress concentration factor, the total strain and resulting stress will be above the linear stress range determined by a uniaxial compression test. The relatively high stress level is not of real concern due to the following: a. The concrete affected is completely surrounded by either other concrete or the penetration nozzle and liner reinforcing plate. This confinement puts the concrete in triaxial compression and gives it the ability to resist forces far in excess of that indicated by a uniaxial compression test. b. The high state of stress and strain exist at a very localized area and have no effect on the overall containment integrity. However, to be conservative, reinforcing steel was placed in these areas. The penetration nozzle will also function as compressive reinforcement. The concrete under the liner plate anchors has some limited yielding in order to get the necessary stress distribution required to resist the liner plate self-relieving loads.

By criteria, yielding was only permitted in the design of the liner plate and of the bonded reinforcing steel for the Containment Structure. Subsequent design analyses, as tabulated in Table 5-1, indicate that the stresses in the liner plate and bonded reinforcing steel will not exceed the allowable yield stresses.

The thermal loads are a result of the temperature gradient across the structure wall. In the finite element analysis, when temperatures are given at every nodal point, stresses are obtained at the center of each element.

The liner plate was handled as an integral part of the structure and was included in the finite element mesh of the Containment Structure, but having different material properties (Figure 5-4, Sheet 1). CALVERT CLIFFS UFSAR 5.1-15 Rev. 47 Under the LOCA condition or factored load condition, cracking of the concrete at the outside face would be expected. The value of the sustained modulus of elasticity of concrete, Ecs, was used in ACI Code 505-54 to find the stresses in concrete, reinforcing steel and liner plate from the predicted design incident thermal loads and factored incident loads.

The method of determining stresses in the concrete and reinforcement required the evaluation of the stress blocks of the cross-section being analyzed. Stress values were taken from the computer output in the case of axisymmetric loading and from analytical solutions in case of non-axisymmetric loading. Both computations were based on homogeneous materials; therefore, some adjustment was necessary to evaluate the true stress-strain conditions when cracks develop in the tensile zone of the concrete. An equilibrium equation was written considering the tension force in the reinforcement, the compressive force in the concrete and the axial force acting on the section. In this manner, the neutral axis was shifted from the position defined by the computer analyses to a position which is a function of the amount of reinforcement, the modulus ratio, and the acting axial forces. The thermal stresses in the containment wall are comparable to those developed in a reinforced concrete slab which is restrained from rotation. The temperature varies linearly across the slab. The concrete will crack in tension and the neutral axis will be shifted toward the compressive extreme fiber. The cracking will reduce the compression at the extreme fiber and increase the tensile stress in reinforcing steel.

The following analysis is based on the equilibrium of normal forces; therefore, any normal force acting on the section must be added to the normal forces resulting from the stress diagram. The effects of Poisson's ratio are considered assuming the reinforcement to be identical in both directions. Stress-strain relationship in compressed region of concrete: Ec x = x - vc y (1) Ec y = - vc x+ y (2) From the above equations (1) and (2): (3) (4) Substituting, x = y = c and x = y = c into equations (3) and (4) (if vc = 0.17) The reinforcement is acting in one direction, independently from the reinforcement in the perpendicular direction. CALVERT CLIFFS UFSAR 5.1-16 Rev. 47 Example: If Ec = 4.07x106 and Es = 29x106 The liner plate is acting in two directions, similar to the concrete except for the difference caused by the Poisson's ratios and elastic modulus:

The concrete and reinforcement stresses, due to a moment caused by a loading other than thermal, are calculated by conventional methods. The analysis assumes homogeneous concrete sections. Those concrete and reinforcing steel stresses are then added to the thermal stresses as obtained by the method described above. Notation: Ec Modulus of elasticity of concrete. Es Modulus of elasticity of steel. nL Modular ratio of liner plate-concrete. nR Modular ratio of reinforcement-concrete. c Concrete strain. s Steel strain. x Concrete strain in the X direction. y Concrete strain in the Y direction. vc Poisson's ratio of concrete. vL Poisson's ratio of liner plate. c Stress in concrete. L Stress in liner plate. x Stress in concrete in the X direction. y Stress in concrete in the Y direction. 5.1.3.2 Non-axisymmetric Techniques The non-axisymmetric aspects of configuration or loading required various methods of analysis. The descriptions of the methods used, as applied to different parts of the containment, are given below.

Buttresses The buttresses and tendon anchorage zones are defined as Seismic Category I elements and were designed in accordance with the general design criteria for the Containment Structure and with the applicable provisions of ACI 318-63, Chapter 26.

The buttresses were analyzed for two effects, non-axisymmetric and anchorage zone stresses. At each buttress, two out of three hoop tendons are spliced by being mutually anchored on the opposite faces of the buttresses; the third tendon is continuous through the buttress. (The anchors are located on 21-3/4" centers, and are CALVERT CLIFFS UFSAR 5.1-17 Rev. 47 staggered every 7-1/4" to the opposite face of the buttress. Combined with the continuous tendon, this results in the hoop tendons being positioned at every 7-1/4" along the vertical cross-section of the wall.) Between the opposite anchorages, the compressive force exerted by the spliced tendon is twice as much as elsewhere. This value, combined with the effect of the tendon which is not spliced, will be 1.5 times the prestressing force acting outside of the buttresses. The cross-sectional area at the buttress is about 1.5 times that of the wall, thus the hoop stresses, as well as the hoop strains and radial displacements, can be considered as being nearly constant all around the structure. The vertical stresses and strains, caused by the vertical post-tensioning, become constant at a short distance away from the anchorages because of the stiffness of the cylindrical shell. The stresses and strains remain nearly axisymmetric despite the presence of the buttresses. The effect of the buttresses on the overall analysis is negligible when the structure is under dead load or prestressing loads. When an increasing internal pressure acts upon the structure, combined with a thermal gradient such as at the design incident condition, the resultant forces are axisymmetric. The stiffness variation caused by the buttresses will decrease as the concrete develops cracks. The structure will then tend to shape itself to follow the direction of the acting axisymmetric resultant forces even more closely. Thus, the buttress effect is more axisymmetric at yield loads (which include factored pressure) than at design loads including pressure. This fact, combined with the design provision that alternate horizontal tendons terminate in a single buttress, indicates that the buttresses will not reduce the margins of safety available in the structure. The analysis of the anchorage zone stresses at the buttresses has been determined to be the most critical of all the various types of anchorage areas of the shell. The local stress distribution in the immediate vicinity of the bearing plates has been derived by the following two analysis procedures: a. The Guyon equivalent prism method: This method is based on experimental photoelastic results as well as on equilibrium considerations of homogeneous and continuous media. It should be noted that the relative bearing plate dimensions are considered. b. In order to include biaxial stress effects, use has been made of the experimental test results presented by S. J. Taylor at the March 1967 London Conference of the Institution of Civil Engineers (Group H, Paper 49). This paper compares test results with most of the currently used approaches (such as Guyon's equivalent prism method). It also investigates the effect of the rigid trumpet welded to the bearing plate. The Guyon method yields the following results: Maximum compressive stress under the bearing plate, c = -2400 psi Maximum tensile stress in spalling zone, spalling = +2400 psi = - c Maximum tensile stress in bursting zones, maximum bursting = (0.04) x avg. stress = +96 psi S. J. Taylor's experimental results indicate that the anchor plate will give rise to a similar stress distribution pattern as Guyon's method; the main difference lies in CALVERT CLIFFS UFSAR 5.1-18 Rev. 47 the fact that the central bursting zone has a tensile stress peak of twice Guyon's value: maximum bursting = +192 psi A state of biaxial tension in the concrete will appear on the outside face under the loading case 1.05D + 1.5P + 1.0TA + 1.0F. The superposition of the corresponding state of stress with the local anchor stresses reduces the load carrying capacity of the anchorage unit and causes a reduction in the maximum tensile strain to cracking.

On the other hand, the uniform compressive state of stress (vertical prestress) applied to the anchorage zone increases the load carrying capacity of the anchorage unit, with the maximum tensile strain to cracking being increased. The design of the buttress anchor zones considered such additional vertical stresses, leading to a state of pseudo biaxial stress, the second direction being radial through the thickness. For the above-mentioned case, i.e., 1.05D + 1.5P + 1.0TA + 1.0F, the averaged vertical (meridional) stress component is: fa ~ + 400 psi The compressive bearing plate stress at 10" depth below the bearing plate is: fc ~ - 1500 psi Thus, the two values introduced in the biaxial stress envelopes proposed in S. J. Taylor's article are: fc/ = 1500/5000 = 0.3 fc// = 400/5000 = 0.08 These values show that failure could occur if vertical reinforcing was not provided. In fact, the maximum allowable vertical average tensile stress according to Taylor's interaction curve is fa/ = 0.03, therefore fa = +150 psi. The three-dimensional stress distribution in the anchor zones was analyzed in sufficient detail to permit the rational evaluation of stress concentrations. A conical wedge segment was used as the basic design element and the radial splitting tension was determined as a tangential distribution function. The summation of splitting stresses through the entire volume of the lead-in zone established the value of the splitting force. This force is a function of the a/b ratio and the cone angle and/or, a/b and h. Several different combinations of the values were analyzed and the most critical values selected. A system analysis for the vertical splitting force was carried out based on statics. The magnitude of vertical and spalling forces were also determined.

The most unfavorable loads and load combinations were considered in the analysis of the anchorage zone. Stresses based on transient thermal gradients were used in all cases where the use of a steady state gradient underestimated the stresses and strains and were superimposed on the bursting stresses determined from the triaxial stress calculations. The computed stresses are less than the ACI allowable values. The design of the concrete reinforcement is based on this conservative analysis to provide a margin of safety similar to the other components of the reactor building structure and to control cracking in the CALVERT CLIFFS UFSAR 5.1-19 Rev. 47 anchorage zone. As a result, there is no danger of delayed rupture of the concrete under sustained load due to local overstress and microcracking. The reinforcing details, including the method for anchoring and splicing the reinforcing, are shown on Figure 5-1.

The reinforcement required has been designed primarily to resist tensile forces, and has been located such that it will efficiently do so. The reinforcement was provided for load cases which create the maximum tensile forces and for other load cases the relevant shear forces or stresses were superimposed. The amount of reinforcing steel was computed manually for all Seismic Category I structures, except the exterior shell and the base slab of the Containment Structure, using conventional reinforced concrete design methods for "Working Stress Design" and "Ultimate Strength Design" depending on the governing load combination.

The seismic analysis was conducted as described in the following subsection, providing values for lateral accelerations, shears, moments and displacements at specified locations. The lateral acceleration values are applied to the axisymmetric Containment Structure as non-axisymmetric static loading in Bechtel's Analysis of Axisymmetric Shell and Solid Subject to Non-Axisymmetric Static Loading, Dynamic Loading or Base Acceleration Program, CE 771. The Containment Structure is idealized as an assemblage of a series of discretized elements. Resultant shear forces, longitudinal, circular and cross moments are calculated from stresses obtained as an output from program CE 771 at 15° increments, from 0° to 180° and at 270°. The combined shear forces, normal forces, and moments are applied to the specified section and, using crack section analysis and compressive and tensile stresses in concrete, liner plate interior and exterior reinforcement are determined. Allowable working case stress and yield case stress determine the required area of reinforcement in the specified section.

The possibility of the concrete breaking along shear planes was considered at the intersection of (1) the buttress with the cylinder and (2) the cylinder with the base slab.

a. Buttress - Cylinder Intersection An increase in the compression force at the buttress corresponds to an increase in the concrete area of the same magnitude. b. Cylinder - Base Slab Intersection An analysis for the most critical radial shear conditions was performed. The difference in shear stiffness between the shell and the buttress and the remainder of the shell was included as a shear amplification factor. The reinforcing required was less then the reinforcing provided. The possibility of concrete breaking along a shear plane is excluded by providing ample reinforcing. In other locations, breakage along the shear plane has been excluded by the opposition of prestressing and anchor forces.

For this reason, special anchorage zone reinforcing is based on the following considerations: a. Full-scale load tests and final designs of similar anchorages. CALVERT CLIFFS UFSAR 5.1-20 Rev. 47 b. The post-tensioning supplier's recommendations of anchorage reinforcing requirements. c. Review of the final details of the combined reinforcing on earlier projects by the consulting firm of T.Y. Lin, Kulka, Yang and Associate. Seismic or Wind Loading Seismic loading of the structure is higher in all cases than that of tornado or wind loading. The seismic analysis was conducted in the following manner: The loads on the Containment Structure caused by earthquake were determined by a dynamic analysis of the structure. The dynamic analysis was made on an idealized structural model of lumped masses and weightless elastic columns acting as springs. The analysis consists of three steps: (1) The determination of the natural frequencies of the structure and its mode shapes, (2) the response of these modes to the earthquake by the response spectrum technique, and (3) combining modal responses to obtain structural response. The natural frequencies and mode shapes were computed using the matrix equation of motion shown below for a lumped mass system. The form of the equation is: , where [K] = matrix of stiffness coefficients including the combined effects of shear and flexure in the structure and the rotation, and horizontal translation of the base slab on soil. [m] = matrix of concentrated masses. {n} = matrix of mode shapes n = angular frequency of vibration in the n-th mode. The results of this computation are the several values of n for n and mode shapes n = 1, 2, 3,.p, where p is the total number of degrees of freedom (i.e., lumped masses) in the idealized model. The response of the structure to the specified earthquake was then computed by the response spectrum technique as follows: a. Using mode frequencies and respective damping values, a response acceleration, Sn, is read from the spectrum curves for each mode. The modal acceleration, An, is given by: An = Sn The acceleration per point i and per mode n is given by: Ain = in An CALVERT CLIFFS UFSAR 5.1-21 Rev. 47 The inertial force per point per mode, Fin' is given by: Fin = mi An where mi is the mass lumped at point i. b. Using the inertial force per point per mode, Fin, shears and moments are computed per point per mode. The "Root Mean Square" method is used for combining modal responses (shears, moments, stresses, deflections, and/or accelerations) in the response spectrum modal analysis of Seismic Category I structures. In this method, responses are combined using the square root of the sum of the squares of responses of each mode. This method was for combining all predominant modal responses including closely spaced modal frequencies. We have examined the effect of adding closely spaced modes linearly, and find that allowable stresses have not been exceeded. c. Seismic and wind shears are transferred across construction joints either by friction, by bond, by shear keys or by a combination of these. Large Openings (Equipment Hatch and Personnel Lock Opening) The primary loads considered in the design of the equipment hatch and personnel lock opening, as in any other part of the structure, were dead load, prestress, pressure, earthquake, and thermal loads. The secondary loads considered were the following effects caused by the above primary loads: The deflection of tendons around the opening; The curvature of the shell at the opening; and The thickening around the opening. The primary loads listed are mainly membrane loads with the exception of the thermal loads. In addition to membrane loads, incident pressure also produces punching shear around the edge of the opening. The magnitudes of these loads for design purposes were the magnitudes at the center of the opening. These are fairly simple to establish knowing the values of hoop and vertical prestressing, incident pressure, and the geometry and location of the opening.

Secondary loads were computed by the following methods: a. The membrane stress concentration factors and effect of the deflection of the tendons around the equipment hatch were analyzed for a flat plate by the finite element method. The stresses computed by conventional stress concentration factors, compared with those values found from the above-mentioned finite element computer program, demonstrated that the deflection of the tendons does not significantly affect the stress concentrations. This is a plane stress analysis and does not include the effect of the curvature of the shell. However, it gives assurance of the correctness of the assumed membrane stress pattern caused by the prestressing around the opening. b. With the help of Reference 1, stress resultants around the large opening were found for various loading cases. Comparison of the results found from this reference with the results of a flat plate of uniform thickness with a circular hole, showed the effect of the cylindrical curvature on stress concentrations around the opening.

CALVERT CLIFFS UFSAR 5.1-22 Rev. 47 Normal shear forces (relative to the opening) were modified to account for the effect of twisting moments (Reference 1). These modified shear forces are called Kirschoff's shear forces. Horizontal wall ties were provided to resist a portion of these shear forces. c. The effect of the thickening on the outside face around the large opening was investigated using several methods. Reference 2 was used to evaluate the effect of thickening on the stress concentration factors for membrane stress. A separate axisymmetric finite element computer analysis, for a flat plate with thickening on the outside face, was prepared to handle both axisymmetric and non-axisymmetric loads. This program predicts the effect of the concentration of hoop tendons with respect to the Containment Structure at the top and bottom of the opening. For the analysis of the thermal stresses around the opening, the same method was used as for the other loadings. At the edge of the opening, a uniformly distributed moment, equal but opposite to the thermal moment existing on the rest of the shell, was applied and evaluated using the methods of the preceding Reference 1. The stresses were then superimposed on the stresses calculated for the other loads. In the case of LOCA temperature, after the incident pressure has already decreased, very little or no tension develops on the outside, so thermal strains will exist without the relieving effect of the cracks. However, the liner plate will reach a high strain level and so will the concrete at the inside corner of the penetrations, thereby relieving the very high stresses, but still carrying a high moment in the state of redistribution stresses.

In the case of 1.5P (prestress fully neutralized) + 1.0TA (accident temperature), the cracked concrete, with highly strained tension reinforcement, constitutes a shell with stiffness decreased but still essentially constant in all directions. See Appendix 5E for an evaluation that reduced the original containment minimum design prestress. In order to control the increased hoop moment around the opening, the hoop reinforcement is about twice that of the radial reinforcement (Figure 5-3). The equipment hatch opening was thus thickened for the following reasons: 1. To reduce the predicted high membrane stresses around the opening; 2. To facilitate tendon placement;

3. To facilitate steel reinforcing placement; and

4. To compensate for the reduction in the overall shell stiffness due to the opening. Since the resultant forces on any part of the containment exterior structure produce compressive stresses in the concrete, and sufficient unprestressed reinforcement is provided to control any local cracking, no significant cracks are expected. In the absence of any significant cracks, it is believed that the concrete will provide adequate corrosion protection for the liner plate. Cathodic protection for the Containment Structure is described in Section 5.1.7. Thus, no surveillance measures are necessary to detect the corrosion of the liner plate in the Containment Structures.

The working stress method, i.e., elastic analysis, was applied to both load combinations for design loads, as well as for yield loads, using the analytical CALVERT CLIFFS UFSAR 5.1-23 Rev. 47 procedures described above. The only difference is the higher allowable stresses under yield conditions. The various factored load combinations and capacity reduction factors are specified in Appendix 5A and were used for the yield load combinations using the working stress design method. The design assumption of straight line variation of stresses was maintained under yield conditions.

The governing design condition for the edges of the equipment hatch opening at the outside face is the LOCA. Under this condition, approximately 60% of the total bonded reinforcing steel needed at the edge of the opening at the outside face, is required to resist the thermal load. Excluding the thermal load, the remaining stress, equivalent to approximately 40% of the total stress including thermal, at the edge of the outside face, is the sum of the following stresses: Normal stresses, resulting from membrane forces, including the effect of thickening, contribute approximately minus 45% (minus 18% of the total). Flexural stresses, resulting from the moments caused by thickening on the outside face, contribute approximately 155% (62% of the total). Normal and flexural stresses, resulting from membrane forces and moments caused by the effect of cylindrical curvature, contribute approximately minus 10% (minus 4% of total). Penetrations Analysis of the Containment Structure penetrations was divided into three parts: (1) the concrete shell, (2) the liner plate reinforcement and closure to the pipe, and (3) the thermal gradients and protection requirements at the high-temperature penetrations. The three parts will be discussed separately.

a. Concrete Shell In general, special design consideration was given to all openings in the Containment Structure. Analysis of the various openings has indicated that the degree of attention required depends upon the penetration size. Small penetrations are considered to be those with a diameter smaller that 2-1/2 times the shell thickness, i.e., approximately 8' in diameter or less. For openings of 8' diameter or less, the curvature effect of the shell is negligible (Reference 1). In general, the typical concrete wall thickness has been found to be capable of taking the imposed stresses using bonded reinforcement, and the thickness is increased only as required to provide space requirements for radially deflected tendons. The induced stresses, due to normal thermal gradients and postulated rupture conditions, distribute rapidly and are of a minor nature compared to the numerous loading conditions for which the shell must be designed. The small penetrations are analyzed as holes in a flat plate. Applied piping restraint loads due to thermal expansion or accident forces are assumed to distribute in the cylinder as stated in Reference 3. Typical details associated with these opening are indicated in Figure 5-2.

b. Liner Plate Closure The stress concentrations around openings in the liner plate were calculated using the theory of elasticity. The stress concentrations were then reduced by the use of a thickened plate around the opening. In the case of a penetration with no appreciable external load, shear connectors were used to maintain deformation compatibility between the liner plate CALVERT CLIFFS UFSAR 5.1-24 Rev. 47 and the concrete. Inward displacement of the liner plate at the penetration was also controlled by shear connectors.

In case significant external operating loads are imposed upon the pipe penetration, the stress level from the external loads is limited to the design stress intensity values, Sm, given in the ASME, B&PV Code, Section III, Article 4. The stress level in the shear connectors from external loads is in accordance with American Institute of Steel Construction (AISC) Code for A-36 steel. The combining of stresses from all effects was performed using the methods outlined in the ASME, B&PV Code, Section III, Article 4, Figure N-414. Figure 5-9 shows a typical penetration and the applied loads. Stresses due to the effects of pipe loads, pressure loads, dead loads, and earthquake loads were calculated and the stress intensity was kept below Sm. The stresses from the remaining effects were combined with the above-calculated stresses and the resultant stress intensity kept below Sa. c. Thermal Gradient The only high temperature pipes penetrating the Containment Structure shell are the main steam and feedwater pipes, steam generator blowdown line, and the reactor coolant letdown and sampling lines. Cooling was provided to maintain the temperature in these penetrations below 150°F. Liner Plate The primary purpose of the liner plate (including welds) is to provide leak tightness integrity to the post-tensioned concrete containment. Structural integrity of the structure is provided by the post-tensioned concrete and not by the liner plate.

The design, construction, inspection, and testing of the liner plate, which acts as a membrane and is not a pressure vessel, was not covered by any recognized code or specification. All components of the liner that must resist the full design pressure, such as penetrations, were selected to meet the requirements of ASME, B&PV Code, Section III, Nuclear Vessels, Paragraph N-1211. ASTM A516, Grade 60 or 70 made to ASTM A300 is a steel which meets these requirements and thus was used as a plate material for penetrations. This material has excellent weldability characteristics. There are no design conditions under which the liner plate is relied upon to assist the concrete in maintaining the integrity of the structure, even though the liner may provide assistance in order to maintain deformation compatibility. Loads are transmitted to the liner plate through the anchorage system and through direct contact with the concrete and vice versa. At times, loads may also be transmitted by bond and/or friction with the concrete. These loads cause, or are caused by, liner strain. The liner was designed to withstand the predicted strains.

Possible cracking of concrete has been considered and reinforcing steel is provided to control the width and spacing of the cracks. In addition, the design is such that total structural deformation remains small during the loading conditions, and that any cracking will be orders of magnitude less than that sustained in the repeated attempts to fail the prestressed concrete reactor vessel "Model 1," and CALVERT CLIFFS UFSAR 5.1-25 Rev. 47 even smaller than the concrete strains of overpressure tests of "Model 2" (both at Gulf-General Atomic) (References 4 and 5). As described, the consequences of concrete cracking on structural integrity are limited by the bonded reinforcing and unbonded tendons that have been provided. The effect of concrete cracking on the liner plate has also been considered. The anchor spacing and other features are such that the liner will sustain orders of magnitude of strain less than did the liner of Model 1 at Gulf-General Atomic (Reference 4) without tensile failure. Liner Plate Anchors The liner plate anchors were designed to preclude failure when subjected to the worst possible loading combinations. The anchors were also designed such that, in the event of a missing or failed anchor, the total integrity of the anchorage system would not be jeopardized. The following loading conditions were considered in the design of the anchorage system: Prestress Internal Pressure Shrinkage and Creep of Concrete Thermal Gradients Dead Load Earthquake Loading Wind or Tornado Loadings Vacuum The following factors were also considered in the design of the anchorage system: Initial inward curvature of the inner plate between anchors due to fabrication and erection inaccuracies. Variation of anchor spacing. Misalignment of liner plate seams. Variation of plate thickness. Variation of liner plate material yield strength. Variation of Poisson's ratio for liner plate material. Cracking of concrete in anchor zone. Variation of the anchor stiffness. The anchorage system satisfies the following conditions: The system has sufficient strength and ductility, with energy absorbing capability sufficient to restrain the maximum force and displacement resulting from the condition where a panel with initial outward curvature is adjacent to a panel with initial inward curvature. The system has sufficient flexural strength to resist the bending moment which would result from the above condition. The system has sufficient strength to resist any radial pull-out forces. CALVERT CLIFFS UFSAR 5.1-26 Rev. 47 Liner Supports In designing for structural bracket loads applied parallel to the plane of the liner plate, or loads transferred through the thickness of the liner plate, the following criteria and methods have been used: The liner plate was thickened to reduce the predicted stress level. The thickened plate, with the corresponding thicker weld attaching the bracket to the plate, will also reduce the probability of the occurrence of a leak at this location. For tensile loads applied perpendicular to the plane of the liner plate, sufficient anchorage is provided.

The allowable stress in the perpendicular direction was calculated using the allowable predicted strain in that direction together with the predicted stresses in the plane of the liner plate. In setting the above criteria, the reduced strength and strain capability of the material, perpendicular to the direction of rolling, was also considered. In this case, the major stress is normal to the plane of the thickened liner plate. The allowable stresses were reduced to 75% of the allowable stress calculated above. 5.1.4 IMPLEMENTATION OF CRITERIA See Appendix 5E for an evaluation that reduced the original containment minimum design prestress. This section documents the manner in which the design criteria were met by the designer. Section 5.1.4.1 discusses original isostress plots and tabulations of predicted stresses for various materials. The isostress plots of the homogeneous cracked concrete structure indicate the general stress pattern for the structure as a whole, under various loading conditions. More specific documentation is made of the predicted stresses for all materials in the structure. In these tabulations, the predicted stresses are compared with the allowable to permit an easy comparison and evaluation of the adequacy of the design.

Sections 5.1.4.3 and 5.1.4.4 illustrate the actual details used in the design to implement the criteria. 5.1.4.1 Results of Analysis See Appendix 5E for an evaluation that reduced the original containment minimum design prestress. The isostress plots, Figures 5-6 and 5-7, show the three principal stresses and the direction of the principal stresses normal to the hoop direction. The principal stresses are the most significant information about the behavior of the structure under various conditions and were a valuable aid for the final design.

The plots were prepared by a cathode-ray tube plotter. The data for plotting were taken from the stress output of the finite element computer program of the following load cases: D + F + L (Construction Case) D + F + L + 1.15P (Test Case) CALVERT CLIFFS UFSAR 5.1-27 Rev. 47 D + F + L + P + TA (Design Incident Case) (1.05 D + F + 1.5P + TA) (Factored Load Case) The above axisymmetric loading conditions have been found to be governing in the design since they result in the highest stresses at various locations in the structure. The containment stress analysis results for structural concrete and liner plate, including shear stresses, are shown here. 5.1.4.2 Prestress Losses See Appendix 5E for a discussion on possible prestress losses due to tendon wire breakage.

In accordance with ACI Code 318-63, the original design provided for prestress losses caused by the following effects: a. Seating of anchorage; b. Elastic shortening of concrete;

c. Creep of concrete;

d. Shrinkage of concrete;

e. Relaxation of prestressing steel stress; and f. Frictional loss due to intended or unintended curvature in the tendons. All of the above losses can be predicted with sufficient accuracy.

In this case, the environment of the prestress system and concrete is not appreciably different from that found in numerous bridge and building applications. Considerable research has been done to evaluate the above items and is available to designers in assigning the allowances. Building code authorities consider it acceptable practice to develop permanent designs based on these allowances.

The following categories and values of prestress losses have been considered in the design: Type of Loss(a) Assumed Value Seating of Anchorage None Elastic Shortening of Concrete 2 ksi Shrinkage and Creep of Concrete for Dome tendons (test data) 146.7x10-6 in./in. Hoop tendons (test data) 248x10-6 in./in. Vertical tendons (test data) 137.5x10-6 in./in. Relaxation of Prestressing Steel(b) 9.5% of 0.70 = 15.96 ksi Frictional Loss K = 0.0003, µ = 0.158 _______________________ (a) See Appendix 5E for a discussion on possible prestress losses due to tendon wire breakage. (b) See Appendix 5E for a discussion on the relaxation value of the prestressing steel used for vertical tendons replaced in 2001 through 2002 on both Units. CALVERT CLIFFS UFSAR 5.1-28 Rev. 47 There is no allowance for the seating of the BBRV anchor since no slippage occurs in the anchor during transfer of the tendon load into the structure. Sample lift-off readings will be taken to confirm that any seating loss is negligible. The loss of tendon stress due to elastic shortening was based on the change in the initial tendon relative to the last tendon stressed.

The value used for shrinkage and creep loss represents only that which could occur after stressing. In general, since the concrete is well aged at the time of stress, little shrinkage and creep is left to occur and add to prestress loss. The value of relaxation loss is based on the information furnished by the tendon system vendor, The Prescon Corporation. See Appendix 5E for a discussion on the relaxation value of the prestressing steel used for vertical tendons replaced in 2001 through 2002 on both Units. Frictional loss parameters for unintentional curvature (K) and intentional curvature (µ) are based on full-scale friction test data. This data indicates actual values of K = 0.0003 and µ = 0.125 versus the design values of K = 0.0003 and µ = 0.158. Assuming that the jacking stress for the tendons is 0.80 or 192,000 psi and using the above prestress loss parameters, the following tabulation shows the magnitude of the design losses and the final effective prestress at the end of 40 years for a typical dome, hoop, and vertical tendon. See Appendix 5E for a discussion on the relaxation value of the prestressing steel used for vertical tendons replaced in 2002 through 2002 on both Units, and a discussion on the final effective prestress at the end of 60 nominal years due to license extension. Dome(c) (ksi) Hoop(c) (ksi) Vertical(c) (ksi) Jacking Stress 192 192 192 Friction Loss 20.86 29.6(a) 21.7 Seating Loss 0 0 0 171.14 162.4 170.3 Elastic Loss 2.0 2.0 2.0 Creep and Shrinkage Loss 4.26 7.19 3.98 Relaxation Loss 15.96 15.96 15.96 Final Effective Prestress(b) 148.92 137.25 148.36 ______________________ (a) Average of crossing tendons. (b) This force does not include the effect of pressurization, which increases the prestress force. (c) Losses shown are for original nominal 40-year license. See Appendix 5E for a discussion of losses at the end of 60 nominal years due to the license extension. To provide assurance of achievement of the desired level of final effective prestress and that ACI 318-63 requirements were met, a written procedure was prepared for guidance of post-tensioning work. The procedure provided nominal values for end anchor forces in terms of pressure gauge readings for calibrated jack-gauge combinations. Force measurements were made at the end anchor since that is the only practical location of such measurements.

CALVERT CLIFFS UFSAR 5.1-29 Rev. 47 The procedure required the measured temporary jacking force for a single tendon to approach, but not exceed 850 kips (0.8 ). Thus the limits set by ACI 318-63 2606(a)1, and of the prestressing system supplier, were observed. Additionally, benefits were obtained by in-place testing of the tendon to provide final assurance that the force capability exceeded that required by design. During the increase in force, measurements were required of elongation changes and force changes in order to allow documentation of compliance with ACI 318-63 2621(a). The procedure required that the prestressing steel be installed in the sheath before stressing for a sufficient time period that the temperatures of the prestressing steel and concrete reach essential equilibrium, to establish conformance with ACI 318-63 2621(e). The jacking force of 0.8 further provided for a means of equalizing the force in individual wires of a tendon to establish compliance with ACI 318-63 2621(b). The procedures required compliance with ACI 318-63 such that, if broken wires resulted from the post-tensioning sequence, compliance with Section 2621(d) was documented. Each of the above procedures contributed to assurance that the desired level of final effective prestress would be achieved. Paragraph 2606(a)2 of ACI 318-63 refers to "tendons" rather than to an individual tendon. Further, the paragraph does not refer to the location to be considered for the determination of 0.8 , for example, "temporary jacking force" referred in paragraph 2606(a)1. Two interpretations were required; both had to consider the effect of the resultant actions on both the prestressing system and structure.

The first interpretation was that the location for measurement of the seating force, used in calculating the percentage of , was at the end anchor and just subsequent to the measurement of the "temporary jacking force" referred in ACI 2606(a)1. The advantages of this location are several. One is that it is a practical one and thus the possibility for achieving valid measurements is greater. The second is that it is the same location used for measuring the "temporary jacking force" and measurements could be made without the added complexity of additional measuring devices. The third advantage is that measurements at this location provide assurance that the calculated percentage of at seating does not anywhere exceed the maximum percentage of to which that tendon has been subjected. Several possible cases were considered for the second interpretation so as to allow anchoring of an individual tendon without exceeding the requirement stated for "tendons" collectively in ACI 318-63 2606(a)2. One such case assumed that the anchoring force for the typical tendon was that for a tendon anchored midway through the prestressing sequence. It further assumed that the losses to be assumed were one-half of the sum of elastic losses, and of the creep, shrinkage, and relaxation predicted to occur during the entire prestressing sequence. This interpretation, however, was considered to be neither practical nor enforceable, since it resulted in changing the seating forces as the actual (as compared to the schedule) time length of the prestressing period was dictated by weather and manpower availability.

In another case, the stressing is done by jacking each tendon to the required force of 850 kips (0.8 ), and placing shims of predetermined thickness, corresponding to the calculated elongation, between the bearing plate assembly and the washer. Proper tendon stress is assured by comparing the jack pressure and tendon elongation with previously calculated values.

CALVERT CLIFFS UFSAR 5.1-30 Rev. 47 The case adopted was to seat each tendon with a measured "pressure" reading for the jack, at "lift-off" of the end anchor, of 775 kips (between 0.72 and 0.73 ,). This procedure had several advantages. One advantage was that the force on the containment and the tendon was within the bounds of those for which it had been tested and resulted in no known detrimental effects. The second advantage was that the stressing procedure was simplified, since the stressing crews did not have to accommodate a large number of different anchoring force requirements. The third advantage was that, at the completion of stressing the last tendon, the expected losses were such that the average percentage of at the end anchors of the tendons would be less than 0.7 thus establishing compliance with ACI 318-63 2606(a)1 and 2. The fourth advantage was that the percentage loss of prestressing force was less than would be the case if the tendons were anchored in such a manner that the calculated percentage of nowhere exceeded 0.7 . The latter advantage deserves special mention since it plays a strong role in assuring that the final effective prestress equaled or exceeded the desired value. For example, if the percentage of at the anchorage of the tendons were 0.1 , creep and shrinkage of concrete could result in the loss of almost all of the prestressing force. Assuming that the total losses due to creep, shrinkage and elastic shortening equal 0.1 , then the final effective prestress would be 20% of an initial prestress equivalent to 0.5 . If the initial prestress was equivalent to 0.7 , the final effective prestress, neglecting relaxation for the moment, would be about 86% of the initial prestress. Clearly, the assurance (that the concrete creep and shrinkage losses have been properly accounted for) increases as the percentage of for the anchored tendon(s) increases. However, this design was committed to meeting the ACI 318-63 requirement and the anchorage force for the tendons was kept at or below 0.7 in accordance with the interpretation described. 5.1.4.3 Liner Plate The following design bases were applied to the containment liner so that it meets the specified leak-rate under LOCA conditions: a. The liner is protected against damage by missiles. (Section 5.1.5.3) b. The liner plate strains are limited to allowable values that have been shown to result in leak-tight vessels or pressure piping. c. The liner plate is prevented from development of significant distortion. d. All discontinuities and openings are properly anchored to accommodate the forces exerted by the restrained liner plate, and careful attention is paid to details of corners and connections to minimize the effects of discontinuities. Pressure vessels, pressure piping, high pressure hydraulic tubing, and similar containers are made by cold forming, drawing, and dishing operations, where strains may approach the elongation capacity of the material. (For mild steel at failure, this elongation varies from 15 to 30%.) These forming operations result in high strains both in tension and compression. Vessels and piping components manufactured by these methods have a history of high leak-tight integrity, proving that subjecting the steel material to high strain does not affect its leak-tight integrity.

The best basis for establishing allowable liner plate strains is considered to be that portion of the ASME, B&PV Code, Section III, Nuclear Vessels, Article 4. CALVERT CLIFFS UFSAR 5.1-31 Rev. 47 Specifically, the following sections have been adopted as guides in establishing allowable strain limits: Paragraph N-412(m) Thermal Stress Paragraph N-414.5 Peak Stress Intensity Table N-413 Classification of Stresses for Some Typical Cases Figure N-414 Stress Categories and Limits of Stress Intensity Figure N-415(A) Design Fatigue Curves Paragraph N-412(n) Operational Cycle Paragraph N-415.1 Vessels Not Requiring Analysis for Cyclic Operation American Society of Mechanical Engineers design codes require that the liner material be prevented from experiencing significant distortion due to the thermal load and that the stresses be considered from a fatigue standpoint [Paragraph N-412(m)(2)]. The following fatigue loads were considered in the design of the liner plate: a. Thermal cycling due to annual outdoor temperature variations. Daily temperature variations do not penetrate a significant distance into the concrete shell to appreciably change the average temperature of the shell relative to the liner plate. The number of cycles for this loading is 40 cycles for the plant life of 40 years. b. Thermal cycling due to interior temperature variations during the startup and shutdown of the reactor system. The number of cycles for this loading was assumed to be 500. c. Thermal cycling due to the LOCA was taken very conservatively to occur only once during plant life. Thermal load cycles in the piping systems are somewhat isolated from the liner plate penetrations by the concentric sleeves between the pipe and the concrete. The attachment sleeve was designed in accordance with ASME, B&PV Code, Section III fatigue considerations. All penetrations were reviewed for a conservative number of cycles to be expected during plant life. Thermal stresses in the liner plate fall into the categories considered in Article 4, Section III, Nuclear Vessels of the ASME, B&PV Code. The allowable stress in Figure N-415(A) is for alternating stress intensity for carbon steels and temperatures not exceeding 700°F. In addition, the ASME code further requires that significant distortion of the material be prevented. In accordance with ASME, B&PV Code, Paragraph 412(m)(2), the liner plate is restrained against significant distortion by continuous angle anchors and never exceeds the temperature limitation of 700°F. It also satisfies the requirements for limiting strains on the basis of fatigue consideration. A typical section showing the anchors is included in Figure 5-1. American Society of Mechanical Engineers, B&PV Code, Paragraph 412(n), Figure N-415(A), has been developed as a result of research, industry experience, and the proven performance of code vessels. Because of the conservative factors it contains on both stress intensity and stress cycles, and its being a part of a recognized design code, Figure N-415(A) and its appropriate limitations have been used as a basis for establishing allowable liner plate strains. Since the graph in Figure N-415(A) does not extend below ten cycles, ten cycles was used for the LOCA instead of one cycle mentioned above.

CALVERT CLIFFS UFSAR 5.1-32 Rev. 47 Establishing an allowable strain based on ten significant thermal cycles of the LOCA condition would permit an allowable strain [from Figure N-415(A)] of approximately 2%. Maximum allowable tensile or compressive strain has been conservatively set at 0.5% (compared to 2% shown above). The maximum predicted strain in the liner plate during LOCA conditions has been found to be 0.25% compression.

At the design LOCA condition, there will be no tensile stress anywhere in the liner plate membrane. This is true both at the time of initial pressure release and under any later pressure and temperature condition. The purpose of specifying a non-destructive examination temperature requirement is to provide protection against a brittle fracture or cleavage mode of failure. However, this type of failure is precluded by the absence of tensile stresses. No allowable compressive strain value has been set for the test condition because the value will be less than that experienced under the LOCA condition. The maximum allowable tensile strain will be 0.2% under test conditions; the predicted value is much smaller. The maximum compressive strains are caused by LOCA pressure, thermal loading prestress, shrinkage, and creep. The maximum calculated strains do not exceed 0.0025 in./in. and the liner plate will always remain in a stable condition.

The stability of the liner plate has been studied for the loading cases and deformations to which it may reasonably be subjected. The critical loading cases that were considered included the LOCA condition and the operating condition during the winter.

Two separate solutions to the plate stability problem were made: a. As a compressed panel under biaxial compression, assuming that the channel and angle stiffeners are rigid in their attachment to the prestressed concrete Containment Structure and the liner. b. As a compressed panel under biaxial compression, assuming the panel to be a portion of a large cylinder with a flexible stiffener system. Figure 5-1 illustrates the actual physical configuration of the stiffening system for the liner plate. The channels function as horizontal stiffeners and the angles as vertical stiffeners.

For the solution, an initial deflected form for the liner plate is expressed in terms of a Fourier Series of the form: where: defines the central angle in a plan view of the cylinder from a point on the circumference where there is zero radial deflection to the point on the circumference where there is maximum radial deflection; x defines the radial deflection and defines the unsupported length in the vertical direction. The stability analysis based on the above assumptions indicates that the critical buckling stress of the liner will be approximately 29,000 psi.

CALVERT CLIFFS UFSAR 5.1-33 Rev. 47 Under normal operating conditions, the maximum compressive stress in the liner will be approximately 23,000 psi. This reflects a buckling margin of 20%. Under LOCA conditions the expected stress in the liner plate will be 35,000 psi. Under this stress the overall structural stability of the liner plate can be maintained. Also of concern is the nature of the state of stress and its behavior at the point of attachment between the stiffeners and the liner plate. Special tests have been conducted on simulated models of the liner plate and vertical stiffener assembly to determine the shear capacity of the angle and vertical stiffener assembly in order to determine the shear capacity of the angle anchorage. In the tests, two different configurations of support were used for the simulated continuous anchor. One case simulates the expected condition that will exist in the Containment Structure. A second case simulates the condition that might exist at an isolated location if the concrete were not in continuous contact with the anchor. Being guided in the proportioning of the liner plate stiffeners by the more critical values of shear transfer for the case of non-continuous contact will, in general, result in a margin of safety for progressive failure of anchors of approximately 2.7. The weld configuration shown in Figure 5-1 is felt to be adequate to transfer all loads that are considered in the design of the structure between the liner plate and the stiffener-anchors. The conservative design approach of the stiffening system used in the liner plate to prevent significant distortions at accident conditions, and the stringent welding and weld inspection requirements ensure that the leak tightness of the liner plate at the LOCA condition will not change from that at the test condition. In isolated areas the liner plate may have initial inward curvature due to construction. The anchors are designed to resist the forces and moments induced when a section of the liner plate between anchors has initial inward curvature. The liner plate is anchored at all discontinuities to eliminate excessive strains. The forces in the liner plate at the discontinuities were evaluated and the anchors designed to resist these forces.

The liner adjacent to the penetrations is backed up by concrete. Refer to Figure 5-2 for a description of the anchoring arrangement at discontinuities and typical details. The containment penetrations behave, partly or fully depending on their size, as elements of a pressure vessel. The sizes range from the small closure pieces for small pipe penetrations to larger components such as air locks and the equipment hatch. Those portions that are not backed up by concrete (the penetrations themselves) are treated as a pressure vessel and comply with ASME, B&PV Code, Section III, Nuclear Vessels, Subsection B, and are designed in accordance with the requirements of Section VIII, Paragraphs UG-27 through UG-33 for the service conditions outlined in Section 5.1. The pressure part to thickened liner plate transition areas are designed and reinforced in accordance with the requirements of Paragraphs UG-36 through UG-41 as a welded assembly conforming to ASME, B&PV Code, Section VIII, Paragraph UW-13. The air locks to thickened liner attachment rings were analyzed by finite-element methods; the membrane meridional, membrane hoop and the average radial stresses were investigated and the adequacy of design was verified.

CALVERT CLIFFS UFSAR 5.1-34 Rev. 47 The liner plate is considered as a composite part of the containment shell and as such, it is investigated by utilizing the strain and deformation data obtained by the finite-element analysis of the containment shell as tabulated in Table 5-1. See Appendix 5E for later tables associated with an evaluation to reduce the original amount of required containment prestress. The ASME B&PV Code is used only as a guideline to establish allowable strain limits, and not to verify the structural and leaktight adequacy of the liner.

In general, where the liner is firmly anchored and bonded to the concrete and maintains a strain compatibility with the concrete shell, the thickening of the liner will neither reduce nor increase the stresses. However at the openings, and particularly in the transition boundaries where the pressure elements of the penetrations are welded to the liner, the increase in the thickness of the steel plate will: a. reduce the possible local stress concentrations, both in the steel plate and the adjacent concrete; b. increase the stiffness of the plate and hence its buckling limit; and c. increase the leaktight integrity of the transition welds. At all penetrations the liner plate is thickened to reduce stress concentrations in accordance with the ASME, B&PV Code 1968, Section III, Nuclear Vessels. The thickened portion of the liner plate is anchored to the concrete by use of anchor studs all around the penetrations. For details of the penetrations see Figure 5-9. The sleeves, pipe cap and all welds associated with the penetrations are designed to resist all loads previously mentioned and also the prestress forces and internal design pressure.

At each location where a load is to be delivered to the walls, slab, or dome of the Containment Structure, an insert plate is provided to transmit the load through the liner. The insert plate is thickened and stiffened as required to deliver the load and to reduce stress concentrations in the liner. The insert plate is anchored to the concrete by appropriate anchors and shear connections. Typical examples of such insert plates are the polar crane brackets, and the floor beam brackets at the operating deck; typical details are shown in Figure 5-1.

5.1.4.4 Penetrations All penetrations are pressure resistant, leak-tight, welded assemblies designed, fabricated, and tested in accordance with the ASME, B&PV Code, Section III, Nuclear Vessel Code.

Types of Penetrations a. Electrical Penetrations Two types of electrical penetration assemblies are used - canister and unitized header. All electrical penetration assemblies are fabricated and tested in accordance with the ASME, B&PV Code, Section III, Nuclear Vessel Code. The canister type is inserted in a nozzle of suitable diameter integral with the Containment Structure and field welded on the inside end. The unitized header type is welded to the nozzle on the outside end. All penetration assemblies are provided with a means to pressurize for monitoring of leakage. Any abnormal depressurization of an assembly is annunciated locally and in the Control Room. There are three different types of electrical penetrations, used as follows: CALVERT CLIFFS UFSAR 5.1-35 Rev. 47 1. Type 1 - 15 kV medium voltage power penetration canisters. These are 30" diameter cylinders constructed from steel plate (SA516, GR 70, 0.652" thick). Stainless steel header plates are then welded into each end containing epoxy bushings welded into the header plates. 2. Type 2 - Low voltage power and control penetration: a. Canister design: These are constructed of 10" diameter schedule 40 seamless steel pipe (SA 106B). Conductors are sealed with glass hermetics either direct fired into header plates or as assemblies welded into header plates. b. Unitized Header Design: Header is constructed of stainless steel, Type 304 (SA 240). Feedthroughs consist of conductors with resilient seal, are mounted and sealed to a header. 3. Type 3 - Instrumentation, thermocouple and coaxial penetrations: a. Canister design: See type 2.a above. b. Unitized Header design: See type 2.b above. Figure 5-2 shows the different types of electrical penetration assemblies.

b. Piping Penetrations Single barrier piping penetrations are provided for all piping passing through the containment walls. The closure of the pipe to the liner plate is accomplished with a pipe cap welded to the pipe and to the liner plate reinforcement. In the case of piping carrying hot fluid, the pipe is insulated and cooling is provided to reduce the concrete temperature to 150°F. Figure 5-2 shows typical hot and cold pipe penetrations. The modes of containment isolation are covered in Section 5.2.

The anchorage of penetration closure connecting pipes to the containment wall were designed as Seismic Category I structures to resist all forces and moments caused by a postulated pipe rupture. The design conditions include the maximum pipe reactions and pipe rupture forces.

The following design basis for typical piping penetrations was used to ensure the integrity of the liner penetration junction at the piping.

1. The penetration assembly, consisting of pipe cap and the assembly welds and welds to the liner plate, utilizes full penetration welds. The assembly is anchored into the wall concrete and designed to accommodate all forces and moments due to pipe rupture and thermal expansion. 2. The design basis is that the pipe penetration is the strongest point in the system when a pipe break is postulated. Pipe stops, increased pipe thickness or other means are used to attain this. Part of this basis also is that the operation of closure valves will not be impaired by any postulated pipe break. c. Large Penetrations (Equipment and Personnel Access Hatches) An equipment hatch opening, 19' in diameter, is provided as shown on Figure 5-3. The equipment hatch opening is covered on the inside of the Containment by a dished hatch, fabricated from welded steel, furnished with a double gasketed flange and bolted in place. The equipment hatch opening also has an outage use door at the exterior opening. The containment outage door is comprised of three assemblies: (1) a transition CALVERT CLIFFS UFSAR 5.1-36 Rev. 47 ring assembly welded directly to the equipment hatch nozzle steel liner plate; (2) a door frame assembly attached to the transition ring by means of bolts and locating pins that support all dead weight and seismic loads; and (3) a hinged door assembly. The frame of the containment outage door is equipped with fittings or penetrations that will allow electricity, compressed air, water, etc., to be run through the equipment hatch opening without going through the door. Flanges, cable penetrations, and isolation valves at these penetrations will be designed to withstand containment pressure resulting from postulated boiling in the RCS, as might occur during a loss of shutdown cooling. The design load to pass through the containment outage door is a reactor coolant pump motor on its transport trailer. For larger equipment, like replacement steam generator components, the bolted-on door frame assembly and door assembly may be removed from the transition ring assembly to restore the equipment hatch opening to its full diameter. Internal lugs welded to the nozzle liner plate and designed to carry the reactions due to the axial pressure loads and seismic loads to the nozzle liner plate may be removed and replaced when the door is restored.

Two personnel locks are provided. One of these is for emergency access only. Each personnel lock is a double door, welded-steel assembly. A quick-acting equalizing valve connects the personnel lock with the interior and exterior of the Containment Structure for the purposes of equalizing pressure in the two systems when entering or leaving. Typical details of the personnel access lock are shown in Figure 5-3. The two doors in each personnel lock are interlocked to prevent both from being opened simultaneously and to ensure that one door is completely closed before the opposite door can be opened. Remote indicating lights and annunciators situated in the Control Room indicate the operational status of the door. Provision is made to permit bypassing the door interlocking system to allow doors to be left open during plant cold shutdown. An emergency lighting and communication system operating from an external emergency supply is provided in the lock interior.

d. Special Penetrations 1. Refueling Tube Penetration A refueling tube penetration is provided for fuel movement between the refueling pool in the Containment Structure and the spent fuel pool in the Auxiliary Building. The penetration consists of a 36" stainless steel pipe installed inside a 42" pipe sleeve. The inner pipe acts as the refueling tube and is fitted with a gate valve in the spent fuel pool and an encapsulating pipe sleeve which is welded to the refueling pool liner and sealed off from the containment with a testable double O-ring blind flange in the refueling pool. This arrangement prevents leakage through the refueling tube in the event of a LOCA. The 42" pipe sleeve is welded to the containment liner.

CALVERT CLIFFS UFSAR 5.1-37 Rev. 47 Bellows expansion joints are provided on the transfer tube to compensate for any differential movement between the tube and the building structures. Figure 9-14 is a drawing of the refueling tube installation. The design basis for each expansion joint is listed below: AUX. BUILDING FUEL POOL EXPANSION JOINT CONTAINMENT REFUELING POOL EXPANSION JOINT External design pressure 28 psig External design pressure 29 psig Internal design pressure 28 psig Internal design pressure 29 psig Design temperature 150°F Design temperature 273°F Lateral movement 1-1/2" Lateral movement 1/2" Axial movement 1-1/2" (expansion or contraction) Displacements were selected to accommodate the maximum differential building settlements. The expansion joint in the refueling pool may be visually inspected from the outside with a periscope when the refueling pool is empty. Repair would probably involve removing the 54" OD pipe that forms the containment boundary. The outside of the expansion joint in the spent fuel pool may be visually inspected by draining the pool or by remote means. A test connection on the bellows provides a means of testing bellows integrity. Repair would require draining the spent fuel pool.

2. Containment Supply and Exhaust Purge Penetrations The ventilation system purge penetrations are equipped with a testable double O-ring blind flange in the penetration room and a tight-seating butterfly valve inside containment used for isolation purposes. The blind flange is used to provide containment integrity during Modes 1-4.

The valve is manually opened for containment purging in Modes 5 and 6 as described in Section 9.8.2.2. 3. Containment Vent Penetration This system is equipped with two valves to be used for isolation purposes. These valves are opened by a handswitch in the Control Room to vent containment pressure during power operation. Although control of hydrogen in Containment following an accident is not required, this penetration may be used as a hydrogen purge. Design of Penetrations a. Design Basis Penetrations conform to the applicable sections of American Standards Association (ASA) N6.2-1965, "Safety Standard for the Design, Fabrication, and Maintenance of Steel Containment Structures for Stationary Nuclear Power Reactors" which has since been withdrawn. All personnel locks and any portion of the equipment access door extending beyond the concrete shell conform in all respects to the requirements of ASME, B&PV Code, Section III, Code for Nuclear Vessels. All future penetrations will conform to ASME, B&PV Code, Section III, Division 1 for fluted head analysis, and Division 2 for those integrated with steel in concrete openings.

CALVERT CLIFFS UFSAR 5.1-38 Rev. 47 Each line which penetrates the containment and contains high-pressure or high-temperature fluids (steam, feedwater, and steam generator blowdown) passes through a structural steel sleeve mounted on the containment wall. This sleeve acts as a positive stop to prevent whipping associated with fracture of a line containing high internal energy and thereby prevents damage to the penetration and breaching of the containment.

Further protection of each line, necessary to preclude pipe rupture between the penetration and the first valve, is accomplished by shortening the exposed length of pipe and installing the first valve as close as possible to the internal or external wall of the structure, dependent upon valve operating and maintenance clearances. Design bases which apply to the provision of automatic and manual isolation valves in the penetration lines are contained in Section 5.2. All penetrations, except the equipment hatch and emergency personnel lock, are inside the Auxiliary Building; therefore, the temperature of Auxiliary Building penetration material will not fall below 30°F. Using the assumption of +60°F temperature inside the containment and outside air temperature of +20°F, the temperature of the equipment hatch and emergency personnel door will not be below +30°F, during startup, operation, or cooldown of the reactor. An investigation was made to determine necessity and/or feasibility of protecting the exterior surface of these penetrations. None was required. Two temperature indicators are provided to monitor the temperature inside the containment. A technical specification is not considered to be necessary for the above reasons. b. High-Temperature Penetrations The main high-temperature piping consists of two penetrations for feedwater, two penetrations for main steam, two penetrations for steam generator blowdown, one for the reactor coolant letdown line, and one for the reactor coolant sampling line. These have a maximum operating temperature range between 435°F and 653°F. Thermal insulation is provided on the outside diameter of each line and separate coolant circulation, with instrumentation suitable for flow monitoring, is provided in the air gap between the insulation and the penetration liner sleeve. The combination of insulation and coolant circulation is designed to restrict the maximum temperature in the concrete to 150°F. For the condition of loss of penetration coolant circulation, the maximum steady state temperature in the concrete will be 300°F at the penetration surface and decreases to 120°F at a maximum radial depth of 48" in the containment wall. Actual peak temperatures in the penetrations resulting from LOCA conditions are expected to subside within six hours. A maximum temperature of 390°F may be tolerated for 120 days (Reference 5) without appreciable loss of concrete strength. The basis for limiting strains in the penetration steel is the ASME, B&PV Code, Section III, Nuclear Vessels, Article 4, 1965, and therefore, the penetration structural and leak-tightness integrity will be maintained. Local heating of the concrete immediately around the penetration will develop compressive stress in the concrete adjacent to the penetration and a negligible amount of tensile stress over a large area. The mild steel CALVERT CLIFFS UFSAR 5.1-39 Rev. 47 reinforcing added around penetrations will distribute local compressive stresses for overall structural integrity.

c. Penetration Materials The material for the penetrations, including the personnel and equipment access hatches together with the mechanical and electrical penetrations, is carbon steel and conforms with the requirements of the ASME, B&PV Code, Nuclear Vessel Code. As required by the Nuclear Vessel Code, the penetration materials which form the pressure boundary meet the necessary Charpy V-notch impact values at a temperature 30°F below the lowest service temperature.

1. Piping Penetration Materials Materials specifications are listed below: Piping Penetration Material Specification Penetration Sleeve ASTM A155 Penetration Reinforcing Rings ASTM A516 Penetration Sleeve Reinforcing ASTM A516 Bar Anchoring Rings and Plates ASTM A516 Rolled Shapes ASTM A36 (nonpressure boundaries) 2. Electrical Penetration Materials The penetration sleeves to accommodate the electrical penetration assemblies are Schedule 40 carbon steel pipe, except where otherwise noted.

3. Access Penetration Materials The equipment hatch and personnel access locks materials are all ASTM A516 made to ASTM A300.

Installation of Penetrations The qualification of welding procedures and welders is in accordance with Section IX, "Welding Qualifications" of the ASME, B&PV Code. The repair of defective welds is in accordance with paragraph N-528 of Section III, "Nuclear Vessels" and Section VIII of the Code.

Testability of Penetrations Only the following penetrations are ASME, B&PV Code, Section III, Class B penetrations as classified in the Atomic Energy Commission (AEC) Technical Safety Guide, "Reactor Containment Leakage Testing and Surveillance Requirements," December 15, 1966. As Class B penetrations, they are subject to individual periodic leak-rate tests separate from the integrated Class A tests. a. Equipment Hatch b. Personnel Lock c. Emergency Personnel Escape Lock

d. Refueling Tube

e. Purge Line Inlet and Outlet CALVERT CLIFFS UFSAR 5.1-40 Rev. 47 5.1.5 INTERIOR STRUCTURE For the original compartment designs, the occurrence of a LOCA was postulated to result from the rupture of the RCS piping, including the main loop piping, either within the reactor cavity or the SG compartments of the containment. For the current compartment designs, breaks are not postulated to occur in the main loop piping based on the Leak-Before-Break Evaluation (References 6 and 7). The design of the Babcock & Wilcox, Canada replacement steam generators also incorporates the application of leak-before-break, which included dynamic analysis of steam generator internals (U-tubes, divider plate, lattice grids, and shroud) for the effects of the most limiting postulated break (12" pressurizer surge line break). American Society of Mechanical Engineers Code-reconciled design and methods were used to install the replacement steam generators to ensure the equivalency of the replacement steam generators-to-RCS piping connections to the original steam generator-to-RCS piping connections, and thus the continued validity of the assumptions made in References 6 and 7. Since only main loop piping is present in the reactor cavity, no LOCA is currently postulated to occur in this compartment. However, smaller bore piping connected to the RCS is present in both steam generator compartments and the pressurizer compartment. The following discussion regarding a LOCA in the reactor compartment is provided as historical design information only. 5.1.5.1 Design Basis Design of the containment interior structures evolves around four basic systems: The primary coolant system, the turbine steam system, the engineered safety system, and the fuel handling system.

The structures which house or support the basic systems are designed to sustain the factored loads of Appendix 5A.

The design bases applied are: a. The structures will sustain all operating dead and live loads, thermal loads, and design seismic loads, without exceeding code allowable stresses. b. Loads and deformations resulting from a LOCA failure and its associated effects in any one of the basic systems will be sustained and restricted such that propagation of the failure to any other system is prohibited. In addition, a failure in one loop of the Nuclear Steam Supply System will be restricted such that propagation of the failure to the other loop is prohibited. Loss-of-coolant accident loads and associated effects include: a. Thrust loads resulting from rapid mass release from a pipe break in any system; b. Pressure buildup in locally confined areas such as the reactor cavity or the secondary shielding compartments; and c. Jet forces resulting from the impingement of the escaping mass upon adjacent structures. The following method was employed to compute the jet forces: 1. Rupture may occur anywhere on portions of the main coolant pipes that are near to the structure or component under investigation. The jet force due to a double-ended break acts along the pipe axis, which may change its direction. The jet force due to a slot break may act in any radial direction normal to the pipe axis. CALVERT CLIFFS UFSAR 5.1-41 Rev. 47 2. Both the double-ended and the slot break expose an area equal to the inside cross-sectional area of the pipe ruptured. The length of a slot break is equal to twice the diameter of the ruptured pipe. 3. The magnitude of the jet force is computed by the following equation, and does not change with distance to target: F = 2.0 x Operating Pressure x Inside Cross-Sectional Area of Pipe 4. The area of the jet plume diverges at a half angle of 10° from the pipe opening. The jet force is evenly distributed to the area of plume at any distance to obtain the impinging pressure on the structure. 5. A spectrum of all possible breaks are considered for a particular structure, and the ones that produce maximum stresses in the structure are used for design. d. Erosion effects of jet spray. e. Pipe whipping following a break in the pipe. f. Rapid rise in ambient temperature to 276°F and accompanying rise in ambient pressure to 50 psig. All Containment Structures have been evaluated for the revised maximum vapor temperature referenced in Section 14.20. g. Missiles as described in Section 5.1.5.3. The containment interior structures were analyzed by utilizing a lumped parameter model which treated the exterior shell, base slab, foundation media and the interior structures as a coupled system, and included details of the interior structures. The ground response spectra, as shown in Section 2.6.3, were used to determine the shears, moments, forces and displacements in the various structural, components, expected to be induced by the SSE and OBEs. The method employed is basically the analysis through the response of the normal modes to spectral accelerations, and is outlined in Section 5.1.3.2 as part of the "Containment Structure Design Analysis." Seismic induced shears, moments and forces are then multiplied with the load factors for Seismic Category I structures as indicated in Section 5A.3.1. When checking each component, particular attention has been paid to tall and slender, or tall and cantilevered portions for possible further amplification. Such components were investigated by utilizing the floor response spectra. Seismic analysis for the interior structures was done using procedures outlined in Section 5.1.3.2.

Where concrete structures such as the primary shield wall are subjected to sustained internal heat buildup, mechanical cooling devices are included to keep the internal temperatures below 150°F. Localized concrete temperatures up to 200°F are acceptable. The total heat generation rate in the primary shield includes contributions from the following radiation sources: fission neutrons and gammas; core secondary gammas; secondary gammas from the core shroud, core support barrel, the reactor vessel, and the intermediate water regions; and secondary gammas in the concrete primary shield. Secondary gamma sources due to neutron scattering and capture were based on neutron distributions calculated using methods described in Chapter 3. The computer program QAD-P5 incorporates a point-kernel numerical integration method for the gamma radiation and a point-kernel moments method for each of the source contributions listed above and the heating rates due to the several contributions summed. The temperature gradient of the primary shield wall resulting from radiation-generated heat is combined with the CALVERT CLIFFS UFSAR 5.1-42 Rev. 47 temperature gradient resulting from convective heat from the reactor to form a combined temperature gradient in the primary shield wall. The combined gradient was used in the thermal stress analysis and the resulting stresses in the wall were determined by using the design methods established by ACI 505-54. These stresses were combined with the stresses resulting from the analysis of loading combinations as described in Section 5.1.5.1. The critical stresses were then used to evaluate required reinforcing in both directions. All stresses are within the allowable as established Appendix 5A. 5.1.5.2 Design Loads and Materials a. Design Loads: The reactor cavity wall, the steam generator compartment walls, and containment interior structures are all Seismic Category I structures, the final design of which satisfied the most severe of the load combination equations of Appendix 5A. b. Materials: The following materials have been used in the construction of the Containment Structure interior: Concrete = 5000 psi @ 28 days Rebar A615, Gr 60 Plate Steel A441 & A533 Structural Steel A36 & A441 Anchor Bolts A325, A354 & A490 5.1.5.3 Missile Protection Inside Containment High pressure RCS components which could be a source of missiles are suitably screened, either by the concrete shield wall enclosing the reactor coolant loops, by the concrete operating floor or by special missile shields, to block any passage of missiles to the containment walls. Potential missile sources are oriented so that the missile will be intercepted by the shields and structures provided. A shield is provided over the control rod drive mechanism to block any missiles generated from postulated fracture of the nozzles.

All internally-generated missiles inside the Containment Structure are listed in Table 5-2. For all other internally-generated missiles outside the Containment Structure, see Section 5.3.1.1. All Seismic Category I structures or parts of structures have been checked for impact of missiles. The modified Petry formula was used for checking the effects of missiles on the structure at the point of impact. Even under extreme assumptions, missiles do not penetrate the concrete barriers completely. Missile protection inside the Containment Structure is provided to comply with the following criteria: a. The Containment Structure and liner are protected from loss-of-function due to damage by such missiles as might be generated in a LOCA for break sizes up to and including the double-ended severance of a reactor coolant pipe. b. The engineered safety features and components required to maintain containment integrity are protected against loss-of-function due to damage by such missiles. CALVERT CLIFFS UFSAR 5.1-43 Rev. 47 Missile protection necessary to meet the above requirements was implemented using the following methods: a. Components of the RCS were examined to identify and to classify missiles according to size, shape, and kinetic energy for purpose of analyzing their effects. b. Missile velocities were calculated considering both fluid and mechanical driving forces which can act during missile generation. c. The RCS is surrounded by reinforced concrete and steel structures designed to withstand the forces associated with the double-ended rupture of a reactor coolant pipe and designed to stop missiles. d. The structural design of the missile shielding takes into account impact loads and is based upon the state of the art of missile penetration calculational techniques. The types of missiles for which protection is provided are: a. Valve stems; b. Valve bonnets;

c. Instrument thimbles; and d. Various types and sizes of nuts and bolts. Protection is not provided for certain types of missiles for which postulated accidents are considered incredible because of the material characteristics, inspections, quality control during fabrication, and conservative design as applied to the particular component. Included in this category are missiles caused by massive, rapid failure of the reactor vessel, steam generator, pressurizer, and main coolant pump flywheels and casings.

It is not expected that the polar crane will become or generate an internal missile. The crane is anchored with seismic lugs and the crane and its components are designed for the OBE and SSE. Additionally, design features and administrative controls are in place to minimize the probability of a load drop from the polar crane. A description of our means of controlling heavy loads is presented in Section 5.7.

5.1.5.4 Thermal Gradients Ventilation and cooling systems maintain thermal gradients at a level low enough to have very small structural effects on the concrete walls. These effects are, nevertheless, considered in the design.

Reactor Cavity Wall Energy is deposited in the concrete of the reactor cavity wall by nuclear radiation emanating from the surface of the reactor vessel. In the equilibrium condition, assuming no axial leakage, all the deposited energy is removed either by the air-cooling system at the inner surface of the cavity wall or by convection from the outer surface. In order to find the temperature distribution within the concrete, an analytical expression for the energy deposition distribution is first derived. This expression is substituted in the heat diffusion equation which is solved for temperature using the temperature boundary conditions at the inner and outer face of the concrete.

CALVERT CLIFFS UFSAR 5.1-44 Rev. 47 Steam Generator Compartments The ventilation system provided eliminates temperature gradients across the secondary shield walls, across the refueling pool walls and across the operating floor. 5.1.5.5 Differential Pressures Generally, the occurrence of a LOCA is postulated to result from a rupture of the primary system piping either within the reactor cavity or the steam generator compartments of the containment. A pipe rupture of this type, within a compartment, results in the expulsion of high enthalpy water that flows out of the ruptured pipe, flashing partly to steam. As the pressure builds up within the compartment, the steam-air-water mixture flows through openings into the main containment. The maximum pressure differential will depend on the number and shape of the openings leading between the compartments, the volume of each compartment, and the blowdown rate from the broken pipe. Differential pressure analyses are made with the Bechtel computer program COPRA, which calculates the transient pressure responses of two containment compartments during a LOCA. COPRA calculates a mass and energy balance of the two-phase, two-component steam-water-air mixture as primary coolant enters the compartment during the LOCA and exits through vents and openings into the main containment building. There is no provision in the code for heat transfer to structures or for engineered safety features. These options generally have a negligible effect on compartment pressures for the short time following the rupture within which peak differential pressures occur. Because of this short time interval between the LOCA and the peak compartment pressure differential, each case analyzed is divided into very small time intervals, generally equal to or less than 0.1 msec. The thermodynamic equations are solved for each time advancement. The steam, water and air throughout each compartment are in thermal equilibrium at all times. Water, steam, and air entering a compartment are mixed homogeneously and instantaneously; no accumulation of water occurs on the walls or in the sump. A COPRA Summary Document which lists and discusses the program assumptions, is contained in Reference 9. This document is on file with the Nuclear Regulatory Commission (NRC).

The discussion of the results of the analyses performed for each interior containment compartment is contained in FSAR [Final Safety Analysis Report] Section 14.16.5, as revised by FSAR Amendment No. 29, Revision 18 (currently Updated Final Safety Analysis Report Section 14.20).

The compartment walls have been designed for jet impingement forces and differential pressures, both assumed to occur simultaneously with the SSE. This design has the following conservative considerations: a. A dynamic load factor of two is applied to the maximum jet load. (The jet load is based on sudden severance of the primary loop piping nearest to the wall.) b. The differential pressure and jet impingements are considered to be in phase. CALVERT CLIFFS UFSAR 5.1-45 Rev. 47 As dictated by design, any transient response of the containment compartment walls is expected to contain the most severe accident postulated. Reactor Cavity There are two types of openings for flow out of the reactor cavity into the main containment volume: a. Openings around the main coolant pipes - 40.0 ft2 b. Openings around the reactor vessel seal ledge - 173 ft2 The current plant design includes a permanent refueling pool seal that further restricts the opening around the reactor vessel. This analysis is provided as historical design information only (Reference 7). The areas around the main coolant pipes extend along the annular space between the pipes and the walls of the pipe tunnels. A calculated nozzle flow coefficient of 0.61, based on the geometry of these flow spaces, is utilized. The opening around the refueling seal ledge is treated as an orifice. Flow coefficients for orifice geometries are supplied automatically by the COPRA computer code. Immediately after a pipe rupture, the volume available to the expanding steam within the reactor cavity is the free volume below the seal ledge.

Both guillotine and slot primary pipe failures have been considered: a. For the guillotine break, the coolant pipes are partially restrained by the tunnel walls. Restraining the primary pipes in this manner gives a flow area of 0.61 ft2 for failure of the 42" line, which results in a cavity peak differential pressure of 23 psi. b. The slot longitudinal break has a length of twice the inside pipe diameter and an area equal to the pipe cross-sectional area. The flow from the part of the break within the pipe tunnel is divided between the reactor cavity and the steam generator compartment, while the flow from the part of the break that is between the primary shield wall and the nozzle of the vessel enters the reactor cavity only. The total flow of mass and energy into the reactor cavity is approximately 27% of the flow from a 42" single-ended rupture. As the reactor cavity pressure builds up, the flow leaving the pipe within the pipe tunnel will enter the steam generator compartment; however, credit for this effect is conservatively ignored to simplify the problem. The peak differential pressure resulting from the slot failure is 31 psi. The following table summarizes the input parameters used for the computer analysis of reactor cavity pressurization: Main containment free volume 2.0x106 ft3 Reactor cavity free volume 5.135x103 ft3 Nozzle-type relief area (around piping) 40.0 ft2 Nozzle coefficient 0.61 Orifice-type relief area (around seal ledge) 173.0 ft2 Initial containment temperature 110°F Initial containment pressure 14.7 psia Initial containment humidity 50% CALVERT CLIFFS UFSAR 5.1-46 Rev. 47 Note that this analysis was redone for the neutron shield with no significant differences. Steam Generator Compartments For the steam generator compartments, only the double-ended guillotine break in the hot leg is considered since it provides the largest rate of energy and mass release. The relief areas are divided into three classes: long, sharp-edged nozzles, sharp-edged orifices and well-rounded orifices. An orifice coefficient of 0.61 is applied to the long nozzles and a coefficient of 0.97 is applied to the well-rounded orifices; coefficients for sharp-edged orifices are supplied by the computer. The compartment volumes and total flow areas are listed below: Compartment Volume Long Sharp -edged Nozzles Sharp -edged Openings Well Rounded Openings P East (a)53,980 ft3 356 ft2 770 ft2 1,072 ft2 16.4 psi West 51,500 ft3 182 ft2 414 ft2 1,072 ft2 18.4 psi (a) East volume is greater than west because east includes pressurizer as well as one steam generator and two reactor coolant pumps. The initial conditions and the main compartment volume are the same as those used in the reactor cavity.

5.1.5.6 Design Bases Temperature and Pressure Differentials Safety factors were not determined or established for pressure and temperature acting alone. The design and analysis methods as applied to the interior structures, accounted for and investigated the effects of temperature and pressure in conjunction with the other loads occurring in those conditions as stated in Appendix 5A. The associated individual safety margins can best be examined by: discussing the contribution of these loads to the total stress intensities, or to the total ultimate load carrying capacity of the structure or component under the two main categories of design loading conditions.

The two main design loadings conditions and the serviceability required of the interior structures in each case are that: a. They provide support during plant operation and prevent the occurrence of a LOCA from: 1. Normal operating loads (including dead loads) 2. Design live loads

3. Hydrostatic loads

4. Thermal loads

5. OBE seismic loads b. They mitigate its consequences, should a LOCA occur, by protecting the containment and all engineered safety features systems from: 1. Blowdown forces (including jet thrust and impingement loads) 2. Whipping pipes

3. Missiles

4. Differential pressures CALVERT CLIFFS UFSAR 5.1-47 Rev. 47 5. SSE seismic loads 6. Loads a.1, and others, if applicable The safety factor of the structure is the ratio of allowable stresses to the computed stresses in case (a) above, and the ratio of the ultimate load resisting capacity of the structure to the total applied loads in case (b).

In general, this ratio is not the same for all components of the interior structure and varies with both the modes of stress or loading and the hazard associated with the structure or component thereof (i.e., some of the components that were governed by radiological shielding requirements have capacities much higher than the design loads; uniaxial, flexural normal, shearing torsional, tensile and compressive stresses were treated differently). Further, the loads above were combined per Appendix 5A load combination equations which have an inherent factor of safety defined by their load coefficients. No credit is taken for this safety factor. For the operating loads, the minimum safety margin is as determined from ACI 318-63 code for reinforced concrete, and AISC Specification, 1963, for structural steel. The contribution of the operating temperatures is negligible in all components except the reactor cavity wall, for which additional reinforcing steel was furnished for the thermal stresses in accordance with the requirements of ACI 505-59. The accident condition controlled the design of most of the interior structures, including the reactor cavity wall, the secondary shield walls which house the RCS, and all major vessel support structures. Where stresses due to blowdown forces were in the same direction and thus combined with the stresses due to other loads including temperature and pressure, the share of the former ranged from 60 to 80%, thus resulting in an increase of 2.5 to 5 times the inherent safety factor for the latter loads acting alone. Where temperature and pressure stresses acted alone, the safety factors inherent in the load combination equations were increased by 3.0 and 1.5 in the operating and accident cases respectively, for both reinforcing and structural steel. In all cases, the safety margin for concrete stresses were kept higher than steel in order to preclude brittle modes of failure. The calculated differential pressures and temperature gradients, and those that have been used as the design bases of the major interior structures are listed below: P (Design) P (Calculated) T (Design) T (Calculated) East Steam Generator Compartment Walls 28 psi 16.4 psi None None West Steam Generator Compartment Walls 28 psi 18.4 psi None None Reactor Cavity Wall 96 psi 31.0 psi(a) 30°F/7' 6.7°F/7' (a) The current plant design includes a permanent refueling pool seal which further restricts the opening around the reactor vessel. This analysis result is provided as historical design information only (Reference 7). 5.1.6 LIGHTNING PROTECTION OF CONTAINMENT Lightning protection is provided over the dome of the containment vessel and consists of the following: CALVERT CLIFFS UFSAR 5.1-48 Rev. 47 One 1/2"x24" copper air terminal is placed in the center of the dome. Eight equally spaced 1/2"x25" air terminals are installed on a 20'-diameter circumference. These are connected to closed loop of 2/0 AWG cable. Four cables of like size connect the loop with the center air terminal. Another 16 equally spaced 1/2"x24" air terminals are installed on 40' circumference. These are connected to a closed loop of 2/0 AWG cable. Another 20 equally spaced 1/2"x24" air terminals are installed on a 69'3" radius circumference.

These are connected to a closed loop of 2/0 AWG cable. Four equally spaced 1/2"x24" air terminals connect the three loops. Four 2/0 AWG down conductors connect the outer loop to the ground grid. Ground guards are provided to protect down conductors to an approximate height of 8' above grade. 5.1.7 CATHODIC PROTECTION The containment cathodic protection system and related environmental or galvanic corrosion protection schemes are explained in the following discussion. Heavy waterproofing (40 mils thickness) membrane is used underneath the containment base slab, within the 4"-thick mud mat and around the cylindrical shell up to finish grade. This membrane is protected to full height by 1/2" (minimum) thick asphalt-impregnated fiber board prior to backfilling. Remotely-located shallow anodes are used in cathodic protection system. Anodes placed approximately 100' from the membrane could protect the structure regardless of the total cross-section of apertures and would provide protection as required for pipes and tank bottoms. Zinc and/or copper sulphate permanent reference electrodes are located in the soil to enable measurement of protective gradients around the foundation. This scheme is based on the concept that the containment building foundation steel is formed with an inhibitive concrete cover, a heavy waterproofing membrane and a relatively low chloride/oxygen environment. Steel embedded in or covered by concrete in the containment sub-structure will be protected by the waterproof membrane. The system is conservatively designed for a 40-year life, derating manufacturer's recommendations for anodes by approximately 50%.

The surface of the exposed liner plate was protected by an initial cleaning by wire brushing or sandblasting, and then coated with Amercoat zinc paint or its equivalent at the fabrication plant. After erection, all exposed areas adjacent to field welds were cleaned or reprimed, and the entire exposed face was painted with a finish coat. The outside of the liner is protected by a minimum of 3-1/4' of prestressed concrete, which is cast against the liner offering a very high degree of corrosion protection. The ACI Building Code Requirements for Reinforced Concrete (ACI 318-63), Section 808, Concrete Protection For Reinforcing, have proved to be effective for preventing chemical corrosion of concrete reinforcement. The reinforcing concrete protection exceeds Section 808 requirements by 50% or more. Tendon sheathing has a minimum of 11" of concrete cover at any location in the Containment Structure. 5.1.8 LEAKAGE MONITORING SYSTEM No continuous leakage monitoring system is provided. The barrier to leakage in the Containment Structure is the 1/4" steel liner plate. All penetrations were continuously welded to the liner plate before the concrete in which they are embedded was placed. These penetrations, shown on Figures 5-2 CALVERT CLIFFS UFSAR 5.1-49 Rev. 47 and 5-3, became an integral part of the liner and were so designed, installed, and tested. The steel liner plate is securely attached to and is an integral part of, the prestressed concrete Containment Structure which is conservatively designed and rigorously analyzed for the extreme loading conditions of the postulated LOCA, as well as for all other types of anticipated loading conditions. Thorough controls were maintained over the quality of all materials and workmanship during all stages of fabrication and erection of the liner plate and penetrations, and during construction of the entire Containment Structure. The comprehensive program for preoperational testing, inspection, and post-operational surveillance is described in detail in Section 5.5 and is summarized in the following paragraphs. During construction, the entire length of every seam weld in the liner plate was leak tested. Individual penetration assemblies were shop tested. Welded connections between penetration assemblies and the liner plate were individually leak tested after installation. Following completion of construction, the entire Containment Structure, the liner and all its penetrations were tested at 115% of the design pressure to establish structural integrity. The initial leak rate test of the entire structure was conducted at 50 and 100% of the calculated peak pressure to demonstrate vapor tightness, and to establish a reference for periodic leak testing for the life of the plant. Penetrations such as the permanent equipment hatch opening and personnel access locks cannot be opened except by deliberate action. The personnel access locks are interlocked and provided with alarm devices so that the Containment Structure cannot be breached unintentionally. The liner plate over the foundation slab is protected by cover concrete. Wherever access to the liner plate welds is not possible, means are provided so that they can be tested for leakage. The liner plate is protected against corrosion by suitable coatings and by cathodic protection. Walls and floors for biological and missile shielding also provide protection for the liner plate.

Once the adequacy of the liner plate was established initially, there is no reason during the life of the plant to anticipate progressive deterioration which would reduce the effectiveness of the liner as a vapor barrier. Inside the Containment Structure, the atmosphere is subject to a high degree of temperature control. The outside of the liner is protected by prestressed concrete which is resistant to all weather conditions. Inspection on a periodic basis, as necessary, may be conducted in all accessible areas during full power operation. Biological shielding is provided to reduce direct radiation from the RCS to acceptable limits. A visual inspection of the Containment Structure, from the inside and outside, will be conducted during regularly scheduled unit shutdowns. All penetrations, except the following, are located in groups, and a penetration room is located at each group. a. Equipment Hatch Opening b. Personnel Access Lock c. Emergency Personnel Access Lock d. Refueling Tube CALVERT CLIFFS UFSAR 5.1-50 Rev. 47 e. Purge Line Inlet and Outlet f. Containment Emergency Sump Any leakage that might occur from these penetrations will be collected and exhausted through vents or drains. In this manner, leakage which might occur from these groups of penetrations will be isolated from leakage which might occur through the Containment Structure, itself.

Provisions are made such that these penetrations may be pressure tested for leakage during normal operation. The containment normal sump penetration is another exception that is not located in a penetration room. However, there are no provisions made to collect and exhaust leakage from this penetration or to enable pressure testing during normal operation. This penetration is further described in Section 5.2.2. Within the penetration rooms, provisions are made such that individual penetration assemblies with resilient seals may be pressure tested for leakage during normal operation. This degree of control over leakage through penetrations greatly reduces the probability of undetected leakage for the Containment Structure as a whole.

A considerable background of operating experience is being accumulated on containment and penetrations. Full advantage of this knowledge is taken in all phases of design, fabrication, installation, inspection, testing, and operation. For the foregoing reasons, it has been concluded that a continuous leakage monitoring system is unnecessary. Since there is no such system provided, there can be no misoperation or malfunction, which in itself might constitute a hazard. The steel-lined Containment Structure is self-sufficient, and other than valves and hatch doors, there are no operating parts. The containment boundary is extended only by specific penetrations which are further described and tabulated in Section 5.2, "Isolation System." 5.

1.9 REFERENCES

1. A.C. Eringen and A.K. Naghdi, "State of Stress in a Circular Cylindrical Shell with a Circular Hole" 2. Levy, Samuel, A.E. McPherson, and F.C. Smith, "Reinforcement of a Small Circular Hole in a Plane Sheet Under Tension," Journal of Applied Mechanics, June 1948 3. K.R. Wickman, A.G. Hopper, and J.L. Mershon, "Local Stress in Spherical and Cylindrical Shells due to External Loadings," Welding Research Council Bulletin No. 107, August 1965 4. HTGR and Laboratory Staff, Prestressed Concrete Reactor Vessel, Model 1, GA7097 5. Advanced HTGR Staff, Prestressed Concrete Reactor Vessel, Model 2, (GA7150) 6. CEN-367-A, "Leak Before Break Evaluation of Primary Coolant Piping in Combustion Engineering Designed NSSS," February 1991 7. Letter from Mr. D. G. McDonald (NRC) to Mr. R. E. Denton (BGE), dated February 3, 1994, Installation of a Neutron Shield/Pool Seal at the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 (TAC Nos. M87176 and M87177)

CALVERT CLIFFS UFSAR 5.1-51 Rev. 47 8. Letter from NRC to Mr. E. C. Sterling, III, Chairman CEOG, dated October 30, 1990, Acceptance for Referencing Topical Report CEN-367, "Leak-Before-Break Evaluation of Primary Coolant Loop Piping in Combustion Engineering Designed Nuclear Steam Supply Systems" 9. Containment Pressure Analysis, NS731 TN, December 1968 CALVERT CLIFFS UFSAR 5.1-52 Rev. 47 TABLE 5-1 STRESS ANALYSIS RESULTS(8) STRUCTURAL DATA LOCATION CONCRETE REINFORCING STEEL - psi t - in Pm - % Ph - % A 5000 39 .260 .094 B 5000 39 .619 .214 C 5000 55.6 .159 .150 D 5000 55.6 .253 .232 E 5000 138.12 .060 .060 F 5000 138.12 .286 .112 G 5000 45 .316 .185 H 5000 45 .741 .289 J 5000 45 .235 .235 K 5000 45 .235 .235 L 5000 70.2 .315 .247 M 5000 70.2 .570 .247 N 4000 120 .223 .185 O 4000 120 .447 .479 P 4000 120 .286 .222 Q 4000 120 1.144 .514 KEY ELEVATION (Showing location of reference sections) CALVERT CLIFFS UFSAR 5.1-53 Rev. 47 TABLE 5-1 STRESS ANALYSIS RESULTS(8) ALLOWABLE STRESSES WORKING STRESS DESIGN YIELD STRESS DESIGN

SHELL CONCRETE: fa = 1500 psi fa = a (fc) = (0.85) (5000) = 4,250 psi fce = 3000 psi fce = ce (fc) = (0.90) (5000) = 4,500 psi BASE CONCRETE: fce = 1800 psi fa = a (fc) = (0.85) (4000) = 3400 psi fce = ce (fc)= (0.90) (4000) = 3600 psi STEEL: A615, GR 60 fs = 30,000 psi fs = (fy) = (0.90) (60,000) = 54,000 psi CALVERT CLIFFS UFSAR 5.1-54 Rev. 47 TABLE 5-1 STRESS ANALYSIS RESULTS(8) D + F + L (Stresses in psi) Construction Case I MERIDIONAL HOOP SHEAR SECTION e OUTSIDE e INSIDE a AXIAL e OUTSIDE e INSIDE a AXIAL vci vcw Shell A - B -827 -1630 -1256-814 -1609 -1207-23105687 C - D +243 -1630 -609+77 -734 -250+3245452 E - F -248 -784 -464+136 -863 -330+1543360 G - H 26 -618-55 -1154 -613-40104513 J - K 8 -664-575 -1645 -1120-869427 L - M 2 499-481 -645 -583698359 Base N - O 113 -251 -41565 -426 -2932138157 P - Q 560 -971 -69565 -470 -1244135123 CALVERT CLIFFS UFSAR 5.1-55 Rev. 47 TABLE 5-1 STRESS ANALYSIS RESULTS(8) CONTAINMENT STRUCTURE - SUMMARY OF CONCRETE AND REINFORCING STEEL STRESSES (psi) CONCRETE COMPUTED COMPUTED VS. ALLOWABLE SECTION LOAD CASE em eh am ah II D+F+L+1.15P -997-935-591 -542-60.3320.3940.017 III D+F+L+TO+E -2141-2363-1307 -1324200.7880.8830.056A-B IV D+F+L+TA+P -2990-2985-545 -51070.9930.3630.020 V 1.05D+F+1.5P+ TA -2610-2213-215 -6120.5800.0510.003 VI 1.05D+F+1.25P+ TA+1.25E -2048-2205-652 -519240.4900.1530.040 VII D+F+P+ TA+E' -2247-2392-804 -732170.5320.1890.030 II D+F+L+1.15P -814-573-240 50680.2710.1600.200 III D+F+L+TO+E -2210-1367-645 -293650.7370.4300.180C-D IV D+F+L+TA+P -2820-2539-224 -35830.9400.1490.230 V 1.05D+F+1.5P+ TA -1970-2058-94 281280.4570.0220.220 VI 1.05D+F+1.25P+ TA+1.25E -2444-2183-145 251710.5430.0340.280 VII D+F+P+ TA+E' -2848-2564-226 -391250.6330.0530.210 II D+F+L+1.15P -357-709-293 -235150.2360.1950.040 III D+F+L+TO+E -232-1409-471 -293150.4690.3140.040E-F IV D+F+L+ TA+P -346-2984-269 -13370.9950.1790.020 V 1.05D+F+1.5P+ TA -451-2626-197 42230.5840.0460.040 VI 1.05D+F+1.25P+ TA+1.25E -374-2843-231 -25200.6320.0540.033 VII D+F+P+ TA+E' -349-3014-272 -135110.6700.0640.018 CALVERT CLIFFS UFSAR 5.1-56 Rev. 47 TABLE 5-1 STRESS ANALYSIS RESULTS(8) CONTAINMENT STRUCTURE - SUMMARY OF CONCRETE AND REINFORCING STEEL STRESSES (psi) REINFORCING STEEL COMPUTED COMPUTED VS. ALLOWABLE SECTION LOAD CASE m h II D+F+L+1.15P -269-16 0.0090.001 III D+F+L+TO+E 2098694 0.0700.023 A-B IV D+F+L+TA+P +35904253 0.1200.142 V 1.05D+F+1.5P+TA 47604512 0.0880.084 VI 1.05D+F+1.25P+TA+1.25E 1444028664 0.2670.531 VII D+F+P+TA+E' 988816483 0.1830.305 II D+F+L+1.15P 17504786 0.0580.160 III D+F+L+TO+E 628210485 0.2090.349 C-D IV D+F+L+TA+P 404015967 0.1340.532 V 1.05D+F+1.5P+TA 14719229 0.0030.356 VI 1.05D+F+1.25P+TA+1.25E 147518794 0.0270.348 VII D+F+P+TA+E' 408416127 0.0760.298 II D+F+L+1.15P 1242148 0.0040.072 III D+F+L+TO+E -2765701 0.010.190 E-F IV D+F+L+TA+P 11108667 0.0370.289 V 1.05D+F+1.5P+TA 408010259 0.0760.190 VI 1.05D+F+1.25P+TA+1.25E 234310036 0.04340.186 VII D+F+P+TA+E' 11218754 0.0210.162 CALVERT CLIFFS UFSAR 5.1-57 Rev. 47 TABLE 5-1 STRESS ANALYSIS RESULTS(8) CONTAINMENT STRUCTURE - SUMMARY OF CONCRETE AND REINFORCING STEEL STRESSES (psi) CONCRETE COMPUTED COMPUTED VS. ALLOWABLE SECTION LOAD CASE em eh am ah II D+F+L+1.15P 746-154 -224-130.2490.1490.037 III D+F+L+TO+E -1415-1770-683 -793610.5900.5290.170G-H IV D+F+L+TA+P -101-2827-68 -195-110.9420.1300.310 V 1.05D+F+1.5P+TA 1766114 23-660.3920.0270.110 VI 1.05D+F+1.25P+TA+1.25E -1453-1496-162 -221-200.3320.0520.033 VII D+F+P+TA+E' -1529-1683-194 -36410.3740.0860.002 II D+F+L+1.15P 653-201 -13310.2180.1340.003 III D+F+L+TO+E -1821-2244-732 -1209-90.7470.8060.250J-K IV D+F+L+TA+P 2629-111 20.8760.0740.006 V 1.05D+F+1.5P+TA 2828169 42020.6280.0990.003 VI 1.05D+F+1.25P+TA+1.25E -1488-957-209 -7550.3310.0490.008 VII D+F+P+TA+E' -1697-1669-311 -447-30.3770.1050.005 II D+F+L+1.15P +10-848-277 -71810.2830.4790.003 III D+F+L+TO+E -621-2012-494 -635-20.6710.4230.006L-M IV D+F+L+TA+P 2914-254 -7311470.9710.4870.415 V 1.05D+F+1.5P+TA 2973-263 -4651630.6610.1090.271 VI 1.05D+F+1.25P+TA+1.25E -1633-3510-197 -914140.7800.2150.023 VII D+F+P+TA+E' -1427-3360-296 -870-60.7450.2050.010 CALVERT CLIFFS UFSAR 5.1-58 Rev. 47 TABLE 5-1 STRESS ANALYSIS RESULTS(8) CONTAINMENT STRUCTURE - SUMMARY OF CONCRETE AND REINFORCING STEEL STRESSES (psi) REINFORCING STEEL COMPUTED COMPUTED VS. ALLOWABLE SECTION LOAD CASE m h II D+F+L+1.15P 28842284 0.0960.076 III D+F+L+TO+E 26163332 0.0870.111 G-H IV D+F+L+TA+P 1664910293 0.5550.343 V 1.05D+F+1.5P+TA 738115437 0.1360.285 VI 1.05D+F+1.25P+TA+1.25E 2372428024 0.4400.520 VII D+F+P+TA+E' 2359520489 0.4370.379 II D+F+L+1.15P 23682882 0.0800.096 III D+F+L+TO+E 5283351 0.1760.012 J-K IV D+F+L+TA+P 1158011410 0.3860.380 V 1.05D+F+1.5P+TA 872010523 0.1610.195 VI 1.05D+F+1.25P+TA+1.25E 2904326964 0.5370.499 VII D+F+P+TA+E' 2547414396 0.4710.266 II D+F+L+1.15P 22007-5482 0.7340.183 III D+F+L+TO+E 33661812 0.1120.0604 L-M IV D+F+L+TA+P 112163168 0.3740.105 V 1.05D+F+1.5P+TA 133693063 0.2470.057 VI 1.05D+F+1.25P+TA+1.25E 199959888 0.3700.183 VII D+F+P+TA+E' 107709651 0.1990.178 CALVERT CLIFFS UFSAR 5.1-59 Rev. 47 TABLE 5-1 STRESS ANALYSIS RESULTS(8) CONTAINMENT STRUCTURE - SUMMARY OF CONCRETE AND REINFORCING STEEL STRESSES (psi) CONCRETE COMPUTED COMPUTED VS. ALLOWABLE SECTION LOAD CASE em eh am ah II D+F+L+1.15P 69-36268 761270.2010.600 III D+F+L+TO+E -911-1204-64 -17480.6690.220N-O IV D+F+L+TA+P -207-59426 581140.330Note0.540 V 1.05D+F+1.5P+TA 7-70420 861490.196(a)0.413 VI 1.05D+F+1.25P+TA+1.25E -861-120354 521640.3340.454 VII D+F+P+TA+E' -1098-119825 521540.3330.427 II D+F+L+1.15P -1338-47356 113740.7430.350 III D+F+L+TO+E -1562-1540-43 -63430.8680.202P-Q IV D+F+L+TA+P -1697-599-45 121650.943Note0.306 V 1.05D+F+1.5P+TA -2253-829-31 681350.626(a)0.374 VI 1.05D+F+1.25P+TA+1.25E -2115-1086-15 75740.5880.205 VII D+F+P+TA+E' -1621-1251-42 46790.4500.220 CALVERT CLIFFS UFSAR 5.1-60 Rev. 47 TABLE 5-1 STRESS ANALYSIS RESULTS(8) CONTAINMENT STRUCTURE - SUMMARY OF CONCRETE AND REINFORCING STEEL STRESSES (psi) REINFORCING STEEL COMPUTED COMPUTED VS. ALLOWABLE SECTION LOAD CASE m h II D+F+L+1.15P 25620957 0.0090.700 III D+F+L+TO+E 1916826123 0.6400.871 N-O IV D+F+L+TA+P 1317226362 0.4400.878 V 1.05D+F+1.5P+TA 1659635106 0.3070.650 VI 1.05D+F+1.25P+TA+1.25E 3031541218 0.5610.763 VII D+F+P+TA+E' 2919640450 0.5420.749 II D+F+L+1.15P 2693324818 0.8970.827 III D+F+L+TO+E 2497030500 0.8321.000 P-Q IV D+F+L+TA+P 2682430012 0.8941.000 V 1.05D+F+1.5P+TA 3698342126 0.6840.780 VI 1.05D+F+1.25P+TA+1.25E 3586240468 0.6640.749 VII D+F+P+TA+E' 2998441033 0.5550.759 _______________________ NOTE (a): fa = 0.3 not applicable to base slab. CALVERT CLIFFS UFSAR 5.1-61 Rev. 47 TABLE 5-1 STRESS ANALYSIS RESULTS(8) CONTAINMENT STRUCTURE - SUMMARY OF CONCRETE AND REINFORCING STEEL STRESSES (psi) CONCRETE REINFORCING STEEL SECTION LOAD CASE em eh m h A-B D + F + L + To -1860 -1845-18-190111D + F + L + To + E -2141-2363202098694D + F + L + Ts -1400-1374-23-4470-4419D + F + L + Ts + E -1466-1526-10-10516-102461.05D + F + L + 1.25P + TA -653-46523370041981.05D + F + L + 1.25P + TA + 1.25E -2048-2205241173125506D + F + L + P + TA -897-794735904253D + F + L + P + TA + E' -2247-239317988916483C-D D + F + L + To 7143-12,200-6074D + E + L + To + E -2281-13936581397912D + F + L + Ts 13722032-7880-1838D + F + L + Ts + E -2283-1392-4810179031.05D + F + L + 1.25P + TA 71114-404073531.05D + F + L + 1.25P + TA + 1.25E -2006-9461711404210321D + F + L + P + TA 8183-4340-1033D + F + L + P + TA + E' -1545-1209125939013818E-F D + L + F + To -242-126610-13965701D + L + F + To + E -836-154215 -728 10234D + F + L + Ts -225-68115-9931337D + F + L + Ts + E -732-72330-11359011.05D + F + L + 1.25P + TA 321-65313-105274271.05D + F + L + 1.25P + TA + 1.25E -193-213220 -75052514D + F + L + P + TA 262-7117-8099653D + F + L + P + TA + E' -336-201211-169149431 CALVERT CLIFFS UFSAR 5.1-62 Rev. 47 TABLE 5-1 STRESS ANALYSIS RESULTS(8) CONTAINMENT STRUCTURE - SUMMARY OF CONCRETE AND REINFORCING STEEL STRESSES (psi) CONCRETE REINFORCING STEEL SECTION LOAD CASE em eh m h G-H D + F + L + To -1321-15224126223475D + F + L + To + E -1415-1770612616 3332D + F + L + Ts -816-1022-40-1187-1466D + F + L + Ts + E -783-109518411449841.05D + F + L + 1.25P + TA -2584-2440-3214927127301.05D + F + L + 1.25P + TA + 1.25E -1452-1495-202372428024D + F + L + P + TA -2660-2712-1114972-10960D + F + L + P + TA + E' -1529-168212359520489J-K D + F + L + To -1595-2115-64228476D + E + L + To + E -1821-2244-95283351D + F + L + Ts -927-1407-8-511-3931D + F + L + Ts + E -990-1453293960-87471.05D + F + L + 1.25P + TA -1029-164857-511177271.05D + F + L + 1.25P + TA + 1.25E -989-9575396026965D + L + F + TA -2655-2628-21158011410D + L + F + TA + E' -1696-1669-32547314396L-M D + F + L + To -304392-15-24223821D + F + L + To + E -621-2012-233661812D + F + L + Ts -1039-6146961-3751D + F + L + Ts + E -1117-788-16833-101881.05D + F + L + 1.25P + TA -2005-29801653713629141.05D + F + L + 1.25P + TA + 1.25E -1633-351014199959888D + F + L + P + TA -1845-2914147281113168D + F + L + P + TA + E' -1427-3360-6107709651 CALVERT CLIFFS UFSAR 5.1-63 Rev. 47 TABLE 5-1 STRESS ANALYSIS RESULTS(8) CONTAINMENT STRUCTURE - SUMMARY OF CONCRETE AND REINFORCING STEEL STRESSES (psi) CONCRETE REINFORCING STEEL SECTION LOAD CASE em eh m h N-O D + F + L + To -708-696241225115290D + F + L + To + E -910-1070481916826123D + F + L + Ts -537-4923287425101D + F + L + Ts + E -930-10803320719272971.05D + F + L + 1.25P + TA -484-7289822724300461.05D + F + L + 1.25P + TA + 1.25E -861-12031643031541218D + F + L + P + TA -763-7431142531628528D + F + L + P + TA + E' -1098-11981542919640450P-Q D + F + L + To -1251-688221716115582D + F + L + To + E -15621355432497030500D + F + L + Ts -911-4464485063814D + F + L + Ts + E -1165-6306017945137201.05D + F + L + 1.25P + TA -1821-6704432542301411.05D + F + L + 1.25P + TA + 1.25E -2115-1086743586240468D + F + L + P + TA -1697-599652682430012D + F + L + P + TA + E' -1621-1251792998441033 CALVERT CLIFFS UFSAR 5.1-64 Rev. 47 TABLE 5-1 STRESS ANALYSIS RESULTS(8) _______________________ NOTES: 1. Loading Cases I, II, III, & IV are working stress analysis whereas Loading Cases V, VI, & VII are yield stress analysis. 2. For notation see next page.

3. All concrete extreme fiber stresses sc are for the inside surface. Outside surfaces stresses are indicated by ( ). The stresses listed are the controlling stresses for that section. 4. Computed vs. allowable ratios for Cases V, VI, and VII include appropriate factors, e.g.,f a/fa = a/a(fc). 5. Allowable shear stresses include stirrups whenever applicable.

6. Deviations in allowable stresses are in accordance with Section 5.1.2.2.

7. Reinforcing steel is Type A615, GR 60.

8. See Appendix 5E for later tables associated with an evaluation to reduce the original amount of required containment prestress. NOTATION D Dead load F Prestress P LOCA pressure load E Earthquake (OBE)

E' Earthquake (SSE) TA Accident temperature fc Concrete compressive strength at age of 28 days fy Steel re-bar yield stress fa Allowable concrete axial stress fce Allowable concrete axial & flexure stress y Allowable concrete shear stress including stirrups if applicable fs Allowable steel stress a Nominal membrane stress e Combined axial & flexure nominal stress Actual shear stress h Subscript indicating hoop direction m Subscript indicating meridional direction ph Hoop steel percentage pm Meridional steel percentage CALVERT CLIFFS UFSAR 5.1-65 Rev. 47 TABLE 5-1 STRESS ANALYSIS RESULTS(8) + Tensile stresses - Compressive stresses Cracked section analysis ci Section 5.1.2.2 (loads necessary to cause structural yielding) cw Section 5.1.2.2 (loads necessary to cause structural yielding) CALVERT CLIFFS UFSAR 5.1-66 Rev. 47 TABLE 5-2 ASSUMED INTERNALLY GENERATED MISSILES IN CONTAINMENT STRUCTURE ITEM AND LOCATION WEIGHT MASS/ AREA SOURCESHAPE IMPACT VELOCITY POINT OF IMPACT lb/in2 (b,c) ft/sec I. REACTOR VESSEL a) Closure Head Nut 116 .0034 1 Cylindrical 35.5 Overhead Shielding Slab b) Closure Head Nut and Stud 710 .0477 1 Cylindrical 21.2 Overhead Shielding Slab c) Instrument Assembly 335 .0445a 2 Rod 49.4 Overhead Shielding Slab d) Instrument 165 .0219a 2 Flat Rod 211 Overhead Shielding Slab II. STEAM GENERATOR a) Primary Manway Cover 1000 6.57a 2 Disk 156.6 Steam Generator Walls b) Primary Manway Cover Stud and Nut 24 7.64 1 Rod 33 Steam Generator Walls c) Secondary Handhole Cover 150 3.76a 2 Disk 181 Steam Generator Walls d) Secondary Handhole Cover Stud and Nut 4 4.04 1 Rod 17.9 Steam Generator Walls e) Secondary Manway Stud 4.67 .0098 1 Rod 506.7 Pressurizer Shield Wall f) Inspection Port Cover 30 1.72 2 Disk 668 Steam Generator Walls g) Inspection Port Cover Stud and Nut 1.33 3.02 1 Rod 1295 Steam Generator Walls III. PRESSURIZER a) Safety Valve & Flange 550/800 .4540 2 Cylindrical 97.2/80.6 Pressurizer Shield Wall b) Valve Flange Bolt 3.75 .0079 1 Rod 16.5 Pressurizer Shield Wall c) Relief Valve & Flange 150 .0311a 2 Cylindrical 150 Pressurizer Shield Wall d) Manway Cover 680 .0150a 2 Disk 369 Pressurizer Shield Wall e) Upper Temp. Element 2.75 .0052a 2 Rod 81.8 Pressurizer Shield Wall f) Lower Temp. Element 3 .0056a 2 Rod 80.2 Pressurizer Shield Wall g) Safety Valve 200 .1648 2 Cylindrical 158 Pressurizer Shield Wall IV. CONTROL ROD DRIVER 1100 .0365 2 Cylindrical 58 Overhead Missile Shield V. MAIN PUMP & PIPE a) Temp. Nozzle & Rtd 11.1 .0209a 2 Cylindrical 81.3 Secondary Shield Wall VI. SURGE & SPRAY LINES a) Thermal Well & Rtd 1.75 .0033a 2 Cylindrical 105 Secondary Shield Wall CALVERT CLIFFS UFSAR 5.1-67 Rev. 47 TABLE 5-2 ASSUMED INTERNALLY GENERATED MISSILES IN CONTAINMENT STRUCTURE ITEM AND LOCATION WEIGHT MASS/ AREA SOURCESHAPE IMPACT VELOCITY POINT OF IMPACT lb/in2 (b,c) ft/sec VII. SHUTDOWN COOLING a) Line Valve Stem 85 .0450 2 Cylindrical 50.4 Secondary Shield Wall _______________________ a Cross-section used is projected edge area (dia x thk) all other areas are minimum projected area. b Stored mechanical strain energy converted to kinetic energy by separation of stud. c Hydrostatic pressure force converted to fluid jet force by separation of mechanical restraint.

CALVERT CLIFFS UFSAR 5.4-1 Rev. 47 5.4 SYSTEM DESIGN EVALUATION This system is not dependent on the operation of a continuous leakage monitoring system for the entire containment, or on a continuous leakage surveillance system for containment penetrations and seals, since neither of these systems is being furnished. Therefore, no analysis of the capability of these systems is necessary. The penetration room ventilation system provides a partial double containment system and is an additional engineered safety feature. A full evaluation of the containment system is included in Section 5.1.8, Leakage Monitoring System. The Containment Structure, with its associated engineered safety systems, prevents uncontrolled release of radioactivity to the plant and surrounding areas during normal operating and accident conditions.

CALVERT CLIFFS UFSAR 5.5-1 Rev. 47 5.5 TESTS AND INSPECTION The quality of both the materials and construction of the Containment Structure was assured by a continuous program of testing and inspection. Qualified field supervisory personnel and inspectors were assigned to the project to carry out the work in accordance with the specifications and drawings. Project design personnel made frequent visits to the job site to coordinate the construction with the design. Inspectors were experienced and thoroughly familiar with the type of work to be inspected, particularly in the field of prestressed concrete. The inspector was given complete access to the work to perform such examinations as were necessary to satisfy himself that the standards set forth in the applicable codes and specifications were met. Where material did not satisfy the standards, he had the authority to stop work until the necessary alterations were made. Appropriate inspection records were maintained. 5.5.1 PREOPERATIONAL TESTING AND INSPECTION 5.5.1.1 During Construction Test, code, and cleanliness requirements accompanied each specification or purchase order for materials and equipment. Hydrostatic, leak, metallurgical, electrical, and other tests to be performed by the supplying manufacturers were enumerated in the specifications together with the requirements, if any, for test witnessing by Bechtel inspectors. Fabrication and cleanliness standards, including final cleaning and sealing, were described together with shipping procedures. Standards and tests were specified in accordance with applicable regulations, recognized technical society codes, and current industrial practices. Inspection was performed in the shops of vendors and subcontractors as necessary to verify compliance with the specifications. The following codes and practices were used to establish standards of construction procedures: ACI 301 - Specification for Structural Concrete for Buildings ACI 318 - Building Code Requirements for Reinforced Concrete ACI 306 - Recommended Practice for Cold Weather Concreting ACI 347 - Recommended Practice for Concrete Formwork ACI 605 - Recommended Practice for Hot Weather Concreting ACI 613 - Recommended Practice for Selecting Proportions for Concrete ACI 614 - Recommended Practice for Measuring, Mixing, and Placing Concrete ACI 315 - Manual of Standard Practice for Detailing Reinforced Concrete Structures Part UW - Requirements for Unfired Pressure Vessels Fabricated by Welding of Section VIII of the ASME, B&PV Code AISC - Steel Manual, Code of Standard Practice ACI - Manual of Concrete Inspection PCI - Inspection Manual AWS - Code for Welding in Building Construction (D 1.0-66 and D 2.0-66) Dimensional tolerances for construction, unless stated otherwise, conform to AISC Code of Standard Practice for erection of steel, and to ACI 301-66 and ACI 318-63 for placing of concrete.

CALVERT CLIFFS UFSAR 5.5-2 Rev. 47 Concrete Concrete work was accomplished basically in accordance with ACI 318-63, "Building Code Requirements for Reinforced Concrete" and ACI-301, "Specifications for Structural Concrete for Buildings." Other codes and specifications are listed above. Concrete is a dense, durable mixture of sound coarse aggregate, fine aggregate, cement, and water. Admixtures were added to improve the quality and workability of the plastic concrete during placement and to retard the set of the concrete. Maximum practical size aggregate, water reducing additives, and a low slump of 2 or 3" were used to minimize shrinkage and creep. Aggregates conformed to "Standard Specifications for Concrete Aggregate," ASTM Designation C33.

Acceptability of aggregates was based on the following ASTM tests. These tests were performed by a qualified commercial testing laboratory. Test L. A. Abrasion ASTM C131 Clay Lumps Natural Aggregate ASTM C142 Material Finer No. 200 Sieve ASTM C117 Mortar Making Properties ASTM C87 Organic Impurities ASTM C40 Potential Reactivity (Chemical) ASTM C289 Sieve Analysis ASTM C136 Soundness ASTM C88 Specific Gravity and Absorption ASTM C127 Specific Gravity and Absorption ASTM C128 Petrographic ASTM C295 Cement was Type II low alkali cement as specified in "Standard Specification for Portland Cement," ASTM Designation C150, and was tested to comply with ASTM C114. Fly ash was not used in the concrete for the Containment Structure, or in any other concrete on the project.

Water used in concrete was clean and free from deleterious amounts of acid, alkali, salts, oil, sediment, or organic matter. Water used in concrete mixing was sampled and analyzed by a qualified testing laboratory to assure conformance with specification. The water-reducing agent, Placewell LS, was selected as the one providing shrinkage similar to that prescribed by ASTM C494, "Specifications for Chemical Admixtures for Concrete." Admixtures containing chlorides were not used. Concrete mixes were designed in accordance with ACI 613, using materials qualified and accepted for this work. Only mixes meeting the design requirements specified for Containment Structure concrete were used. Trial mixes were tested in accordance with applicable ASTM Codes as indicated below: Test Making and curing cylinder in Laboratory ASTM C192 Air Content ASTM C231 CALVERT CLIFFS UFSAR 5.5-3 Rev. 47 Slump ASTM C143 Compressive Strength Tests ASTM C39 The concrete had a design compressive strength of 5000 psi at 28 days for the containment wall and dome, and 4000 psi at 28 days for the containment base slab.

Concrete strength, slump, and temperature tests were performed. The purpose of the tests was to ascertain conformance to specifications. The basis for the inspection procedures was the ACI Manual of Concrete Inspection with modifications as set forth in construction specifications for this application.

Test cylinders were cast from the mix selected for construction and the following concrete properties were determined: Uniaxial creep ASTM C512 Modulus of Elasticity and Poisson's Ratio ASTM C469 Autogenous Shrinkage ASTM C342 Thermal Diffusivity ASTM C34 and CRD-C36-63 Thermal Coefficient of Expansion ASTM C342 and CRD-C124-62 Compressive Strength ASTM C39 An independent laboratory tested the concrete mixes. To maintain the quality of the mix used in the structure, the workability and other characteristics of the mixes were ascertained before placement. A small concrete-control laboratory was set up close to the batch plant. A batch plant inspector was assigned, and testing, as shown below, was performed. Field control was accomplished basically in accordance with the ACI Manual of Concrete Inspection as reported by Committee 611. Aggregate testing was carried out as follows: a. Sand Sample for Gradation (ASTM C33 Fine Agg) b. Organic Test on Sand (ASTM C40)

c. 3/4" Sample for Gradation (ASTM C33 Size No. 67)

d. 1-1/2" Sample for Gradation (ASTM C33 Size No. 4) e. Check for Proportion of Flat and Elongated Particles Concrete samples were taken from the mix according to ASTM C172, "Sampling Fresh Concrete." From these samples, cylinders for compression testing were made in accordance with ASTM C31, "Tentative Method of Making and Curing Concrete Compression and Flexure Test Specimens in the Field."

Samples were taken at the point of truck discharge. In addition, a minimum of five cylinders were taken at the pipe discharge for each 1000 c.y. for each class of concrete placed in Seismic Category I structures. Slump, air content, and temperature measurements were taken when cylinders were cast. Slump tests were performed in accordance with ASTM C143, "Standard Method of Test for Slump of Portland Cement Concrete." Air content tests were performed in accordance with ASTM C231, "Standard Method of Test for Air Content of Freshly Mixed Concrete by the Pressure Method." Compressive CALVERT CLIFFS UFSAR 5.5-4 Rev. 47 strength tests were made in accordance with ASTM C39, "Method of Test for Compressive Strength of Molded Concrete Cylinders." Evaluation of compression tests was in accordance with ACI 214-65. The inspection and testing of cement, in addition to the tests required by the cement manufacturers, included the following: Chemical Analysis ASTM C114 Fineness of Portland Cement ASTM C115 Autoclave Expansion ASTM C151 Time of Set ASTM C191 Compressive Strength ASTM C109 Tensile Strength ASTM C190 The purpose of the above tests was to ascertain conformance with ASTM Specification C150. In addition, tests ASTM C191 and ASTM C109 were repeated periodically during construction to check storage environmental effects on cement characteristics. These tests supplemented visual inspection of material storage procedures. Initial Containment Prestressing After Construction See Appendix 5E for a discussion on the stressing and restressing of new replacement vertical tendons and original construction tendons, respectively, between 2001 and 2002 on both Units.

Testing and inspection of all prestressing materials and special installation equipment is described in Appendix 5B. Full-time supervision of the prestressing operation was provided. The BBRV post-tensioning system furnished by the Prescon Corporation was used.

Each tendon consists of 90 1/4"-diameter wires conforming with ASTM A421-65T, and 2 anchor heads and 2 sets of shims conforming with ASTM A6-66. The tendon sheathing system consists of spirally-wound carbon steel tubing connecting to a trumplate (bearing plate and trumpet) at each end. The bearing plates were fabricated from steel plate conforming with ASTM A6-66 and the trumpets from AISI C1010-C1020 material. Tendons were delivered to the site coated with a rust preventive and specially covered. Each tendon came precut to exact length, with one end unfinished and the other end shop button-headed and threaded through the stressing washer.

Tendons were fabricated by the Prescon Corporation at their Mauldin, South Carolina, plant and shipped to the Calvert Cliffs job site. During combing and twisting operation, the hoop and dome tendons were banded every 8' to 10' with steel banding and twisted one complete 360° turn per each 40' of tendon length by an automatic twister. During tendon fabrication, a wire sampling method was used to ensure that all wires met minimum tensile strength specifications. The wires were shipped in coils, each weighing 800 to 1200 lbs. Two samples per coil were used for wire-sampling inspection to compare their actual breaking strength with minimum wire breaking strength. A 3' minimum wire sample was taken from the start end of each coil and placed in an appropriate storage tube in the wire sample cart for delivery to the tensile testing machine. After the sample broke, a reading was taken from the maximum load pointer and recorded on a wire inspection record under the minimum breaking strength column. If the sample failed, a CALVERT CLIFFS UFSAR 5.5-5 Rev. 47 second supporting inspection was conducted. Wire strength acceptance was based on the final results of two out of three samples if the first sample failed. If the actual breaking strength was greater than the minimum breaking strength, the wire was recorded as being acceptable. This wire sampling operation verified the initial strength of the wires.

One unit of tendon stressing equipment consisted of a 500-ton ram, a jack base (which is bolted on the ram), a pull rod, a nut, 2 hydraulic hoses, and a hydraulic pump with 440 Volt electric power. The jack base was designed to rotate with ease by inserting a rod into one of the eight holes in the base and turning it. This simplified the dome or hoop stressing operation that required many orientations for the jack base. During tendon stressing, the pull rod could rotate a maximum of 90° due to the twist applied to the hoop and dome tendons during fabrication. This pull rod could freely rotate while stressing the tendon without exerting any twisting moments on the anchor.

It is conceivable, however, that the torque remaining after the tendon had seated, would produce insignificant stress in the anchor. The primary bearing and shear stresses (due to twisting), produced during post-tensioning and which were transferred to the concrete, were well below allowable design value. Only the dome and hoop tendons were twisted when fabricated to give helical shape to the wires and equalize their lengths. (See Appendix 5E for a discussion on new vertical tendons installed between 2001 and 2002. These new vertical tendons were twisted when fabricated.) The average maximum eccentricity, (1/2" over the total tendon length) that may exist due to erection inaccuracies, in our judgment, produced no significant increase in compressive stress in the anchor material or concrete.

The tendon installation prestressing procedure was carried out as follows: a. To assure a clear passage for the tendons, a "sheathing rabbit" was run through the sheathing prior to, during, and following placement of the concrete. b. Tendons were uncoiled and pulled through the sheathing unfinished end first. c. The unfinished end of the tendons was pulled out with enough length exposed so that field attachment of the stressing washer and button-heading could be performed. To allow this operation, trumpets on the opposite end have a larger diameter to permit pulling in the shop finished ends with their stressing washers. d. The stressing washers were attached and the tendon wires button-headed. e. The shop finished end of the tendon was pulled back and the stressing jack attached. f. The post-tensioning was done by jacking to the permissible overstressing force to compensate for friction and placing the shims (as required) to lengths corresponding to the calculated elongation. Proper tendon stress was achieved by comparing both jack pressure and tendon elongation against previously calculated values. The vertical tendons were prestressed from either one or both ends, while the horizontal and dome tendons were prestressed from both ends. g. The grease caps were bolted onto anchorages at both ends and made ready for pumping the tendon sheathing filler material. See Appendix 5E for a discussion on new vertical tendon grease caps installed between 2000 and 2002 on both Units. CALVERT CLIFFS UFSAR 5.5-6 Rev. 47 h. The tendon sheaths and grease caps were filled with sheathing filler and sealed. The sheathing filler material had limitations specified for deleterious water soluble salts. During installation of the Unit 1, post-tensioning system two vertical and three horizontal tendons were not installed. These missing tendons are addressed in Section 3.1.4 of the Final Prestressing Report for Unit 1, November 1973. Similarly, one horizontal tendon was abandoned during construction of Unit 2 containment, and is addressed in Section 3.1.4 of the Prestressing Report for Unit 2, June 1977. Reinforcing Steel Reinforcing steel in the base slab of the Containment Structure and around penetrations in the cylinder was of the deformed billet steel bars conforming to ASTM Designation A615-68, Grade 60. This steel had a minimum elongation of 7% in an 8" specimen. Deformed billet steel bars conforming to ASTM A615, Grade 40 or Grade 60, were used in the cylinder wall and the domed roof to control shrinkage and tensile cracks. The Grade 40 steel had a minimum yield strength of 40,000 psi and a minimum tensile strength of 70,000 psi; the Grade 60 steel had a minimum yield strength of 60,000 psi and a minimum tensile strength of 90,000 psi.

Mill test reports were obtained from the reinforcing steel and "Cadweld" suppliers for each heat of steel to show proof that the reinforcing steel and mechanical splice sleeves had the specified composition, strength, and ductility. Welding of reinforcing steel, if required, was performed by qualified welders in accordance with AWS D12.1, "Recommended Practice for Welding Reinforcing Steel, Metal Inserts, and Connections in Reinforced Concrete Construction." For the filling of blockouts in the Auxiliary Building, reinforcing steel was welded using an angle splice as shown in Figure 5-16. The design criteria and quality control is described in Section 5B.1.

Reinforcing steel had not been welded at anytime in the Containment Structure. Number 14S and 18S reinforcing steel was spliced by the Cadweld Process. The design criteria and quality control for Cadweld is described in Section 5B.3. All reinforcing steel was user-tested in accordance with ASTM specifications. Tests include one tension and one bend test per heat for each diameter bar except that no bend tests were performed on #14 and #18 bars. High strength bars were clearly identified prior to shipment to prevent any possibility of mix-up with lower strength reinforcing bars. Visual inspection of fabricated reinforcement was performed to ascertain dimensional conformance with specifications and drawings. Visual inspection of in-place reinforcement was performed by a placing inspector to assure dimensional and location conformance with drawings and specifications. Liner Plate The Containment Structure is lined with a welded steel plate 1/4" thick conforming to ASTM A36 to ensure low leakage. This steel had a minimum yield strength of 36,000 psi and a minimum elongation in an 8" specimen of 20%. Structural steel shapes, bars, and backing strips used in fabrication of the liner also conformed to ASTM A36.

CALVERT CLIFFS UFSAR 5.5-7 Rev. 47 The A-36 material was chosen on the basis that it has sufficient strength as well as ductility to resist the expected stresses from design basis loading and at the same time preserve the required leak tightness of the containment. In addition, A-36 steel is readily weldable by all of the commercially available arc and gas welding processes.

The crane bracket together with the thickened liner plate was a shop fabricated assembly, and all welds have been spot radiographed and magnetic particle inspected. These welds were not considered working welds, since all applied loads were transferred to the concrete and not the liner plate. The liner plate was designed to function only as a leaktight membrane. It does not serve as a structural member to resist the tension loads from internally applied pressure which may result from any credible accident. Structural integrity of the containment is maintained by the prestressed, post-tensioned concrete. Since the principal applied stress to the liner plate membrane is in compression and no significant applied tension stresses were expected from internal pressure loading, there was no need to apply special NDTT requirements to the liner plate material. On the other hand, all material for containment parts which must resist applied internal pressure stresses, such as penetrations, was impact tested in accordance with the requirements of ASME, B&PV Code, Section III, Nuclear Vessels, Paragraph N-1211.

A fundamental requirement for fabrication and erection of the liner plate was that all welding procedures and welding operators be qualified by tests as specified in ASME, B&PV Code, Section IX. This code required testing of welded transverse root and face bend samples in order to verify adequate weld metal ductility. Specifically, Section IX of the Code required that transverse root and face bend samples be capable of being bent cold 180° to an inside radius equal to twice the thickness of the test sample. Satisfactory completion of these bend tests was accepted as adequate evidence of required weld metal and plate material compatibility.

Mill test reports were obtained for the liner plate material. The plate was visually checked for thickness, possible laminations, and pitting. Steel plate was tested at the mill in full conformance to the applicable ASTM Specifications. Certified mill test reports were supplied for review and approval by the design group in the project engineer's office. There was impact testing done on the liner plate material. The purpose of impact testing is to provide protection against brittle failure. The possibility of a brittle fracture of the liner plate is precluded because at the design accident pressure condition, there will be no significant tensile stress anywhere in the liner plate since the principal applied stress is compression. This is true whether there is instantaneous release of pressure or there is some time lag in temperature load application.

Welding inspection conformed to the quality control inspection procedure described in detail by Appendix 5B. All of the welding was visually examined by a technician responsible for welding quality control. The basis for visual quality of welds was as follows:

CALVERT CLIFFS UFSAR 5.5-8 Rev. 47 Each weld was uniform in width and size throughout its full length. Each layer of welding was smooth and free of slag, cracks, pinholes, and undercut, and was completely fused to the adjacent weld beads and base metal. In addition, the cover pass was free of coarse ripples, irregular surface, nonuniform bead pattern, high crown, and deep ridges or valleys between beads. Peening of welds was not permitted.

Butt welds were of multipass construction, slightly convex, of uniform height, and had full penetrations. Fillet welds were of the specified size, with full throat and legs of uniform length.

All welding covered by concrete or otherwise inaccessible after construction was vacuum box soap bubble tested. In this test a leak detector solution was applied to the weld. A vacuum box containing a window was then placed over the area to be tested, and was evacuated to produce at least a 5 psi pressure differential. Leaks were indicated by the appearance of bubbles which were observed through the window in the vacuum box. Welds which were inaccessible for soap bubble testing due to physical limitations or configurations were liquid penetrant inspected. Radiography was not recognized as an effective method for examining welds to assure leak tightness. Therefore, the only benefit that could be expected from radiography in connection with obtaining leak-tight welds was an aid to quality control. Random radiography of each welder's work provided verification that the welding was or was not under control and being done in accordance with the previously established and qualified procedures. In addition, employing random radiography to inspect each welder's work had been demonstrated by past experience to have a positive psychological effect on improving overall welding workmanship. Radiographic techniques were in accordance with ASME, B&PV Code, Section VIII, Paragraph UW-51. At least one 12" spot radiograph was taken in the first 10' of welding completed in the flat, vertical, horizontal, and overhead positions by each welder. Thereafter, approximately 10% of the welding was spot examined on a random basis using 12" film. Dye penetrant and magnetic particle inspections were also used as an aid to quality control. The field welding inspectors used dye penetrant or magnetic particle inspection to closely examine welds judged to be of questionable quality of the basis of the initial visual inspection. Also, dye penetrant inspection was used to confirm the complete removal of all defects from areas which had been prepared for repair welding. Dye penetrant or magnetic particle inspection of liner plate welds were in accordance with ASME, B&PV Code, Section VIII.

The welds for each section of base slab liner plate were vacuum box soap bubble tested immediately upon installation. After successfully passing this leakage test, they were covered with test channels and the particular welds associated with that section of liner plate were pressure tested. Any repairs were carried out utilizing the same high standards and control exercised in the initial construction.

A testing pipe was provided for each continuous segment of the bottom liner plate leak chase channels (equivalent to containment weld channels). The tops of the pipes were above the cover slab and were sealed with caps. These pipes were initially used to test the leak tightness of the bottom liner and can also be used at a later date, if so required. CALVERT CLIFFS UFSAR 5.5-9 Rev. 47 5.5.1.2 Structural Test at Completion of Initial Construction The purpose of instrumenting and testing prestressed concrete Containment Structure is to provide a means for comparing the actual response of the structure to the loads induced both during post-tensioning and pressure testing with the predictions of the design calculations. If the response is as predicted, the design techniques are assumed to have been verified.

The Containment Structure was pressurized to 115% of design pressure for one hour following completion of construction to establish the structural integrity of the building. The structural integrity test was conducted in accordance with a written procedure. Personnel access limitations included in the written procedures designated areas of limited access during specific periods of the test. The test objectives were: a. To provide direct verification that the structural integrity as a whole is equal to or greater than that necessary to sustain the forces imposed by (a) the structural test at 115% of the design pressure and (b) the post tensioning sequence. b. The in-place tendons (the major strength elements) have a strength of at least 80% of guaranteed ultimate tensile strength and that the concrete has the strength needed to sustain a strain range from high initial average concrete compression when unpressurized to low average concrete compression when pressurized. A quality assurance program was instituted as described in Appendix 5B. In addition, each individual tendon was tensioned in place to 80% of the guaranteed ultimate tensile strength and then anchored at a lower load that is still in excess of those predicted to exist at test pressure levels. During pressurization of the structure, the structure's response was observed at selected pressure levels with the highest being 115% the design pressure. An indication that the structure is capable of withstanding internal pressure resulted from these tests. The strain measuring program is described earlier. Individual test values which fall outside the predicted ranges will not be considered as necessarily indicative of a lack of adequate structural integrity. The Calvert Cliffs Units were very similar to the Turkey Point, Oconee, Point Beach, and Palisades structures, differing only in being somewhat larger in diameter. The design and construction are the same. The structures for both Turkey Point and Palisades are completely instrumented. The Turkey Point instruments provide approximately 400 strain measurements at 55 locations throughout the structure and liner. In addition, about 25 optical measurements of structural deformation are made. The Palisades instrumentation is comparable. This amount of data will permit a detailed comparison between design calculations and observed response. The basic structural design and the accuracy of the calculation procedures used by Bechtel was, therefore, verified by these tests. This verification was applicable to the Calvert Cliffs design calculations. Since the detailed confirmation of the design techniques is available, instrumentation of the Calvert Cliffs structure is not required and no additional confirmation of design techniques is necessary. For these reasons, no provisions for strain gauge instrumentation of the structural members of the Calvert Cliffs Containment Structure are made. CALVERT CLIFFS UFSAR 5.5-10 Rev. 47 Prior to reactor fuel loading and operation, the integrity of the Containment Structure was demonstrated by a pressure proof test. The post-tensioning and pressure tests permitted verification that the structural response due to the induced loads is consistent with the predicted behavior. This was accomplished by measuring deflection of Containment Structure using taut wires.

The measurement technique required stretching taut wires across the Containment Structure at appropriate elevations and azimuths and around the equipment hatch openings. These displacements were correlated with measurements made on Turkey Point, Oconee, and Point Beach I Containment Structures for verification of structural behavior.

In analyzing the structures to obtain the calculated displacement, the most probable values of material constants were used rather than the highly conservative design values. For example, values of the elastic modulus for concrete were predicted to provide an estimate of its most probable value at the time of the test.

The use of only two meridians for taking measurements during pressure testing is justified as follows: a. It represents the true cross-section of the cylindrical shell where uniform wall thickness and buttress (thickened wall) sections exist. Other discontinuity areas, such as the equipment hatch, are individually checked for strain measurements. b. Analytical methods are based on an assumption that the structure is axisymmetric and the material properties assumed for calculation purposes are idealized for derivation of the theories of elasticity. The basic method of analysis is Bechtel's Finite Element Program, CE 316-4 as explained in Section 5.1.3.1. This analysis furnished the predicted strain for this test, assuming the actual structure was perfectly cylindrical with no discontinuities such as buttresses or penetrations and that there were no deviations from axisymmetry of applied forces. c. The correctness of the predicted strains versus measured strain will not significantly differ by increasing the number of measurements at more than two meridians because the basic assumptions as mentioned in b would be identical. d. Tests of Containment Structures with similar configuration have demonstrated that the predicted and measured strain values are in good agreement. The applied test procedure and selected points for strain measurements were identical for all tests. Nevertheless, the Calvert Cliffs Nuclear Power Plant test procedure included additional points to the extent possible to obtain measurements as described in AEC Safety Guide 18, Structural Acceptance Test for Concrete Primary Reactor Containments.

From the previous experience and analytical assumptions, it was expected that agreement would have been between test results and analytical predictions in the following range: Cylinder at equator 15% Dome 15% Bottom slab 25% Bottom slab - Wall junction 25% CALVERT CLIFFS UFSAR 5.5-11 Rev. 47 Dome - Wall junction 20% Around opening 30% Localized stress concentration 100% If the measured strains had fallen noticeably beyond the above-mentioned ranges of error, a review and investigation would have been made to determine the cause of such discrepancies.

5.5.1.3 Initial Leakage Test At the time of the initial leakage test, the design leak-rate was 0.20% by weight of the contained atmosphere in 24 hrs at 50 psig. It has been demonstrated that, with good quality during erection, this is a reasonable requirement. The purpose of these tests is to ensure that leakage through the Containment Structure and associated systems is held below the design leakage rate (Reference 1). Initial leak-rate tests of the Containment Structure and its penetrations were conducted at pressures of 50 and 100% of the calculated peak pressure, maintaining each pressure for a sufficient length of time to establish the leak-rate. Values of Containment Structure ambient dry bulb temperature and relative humidity were recorded during the test period for correction of data as required. The preservice leak-rate test equipment consisted of bottled air or nitrogen, pressure regulator and pressure, temperature and flow indicator. Each part's measuring range and accuracy were as follows: a. Pressure Regulator Range: 2000 psig to 50 psig

b. Pressure Indicator Pressure gauge: Readout unit, calibration accuracy of 0.015% of reading, readout 100,000 counts = full scale Range: 0 psia to 100 psia Minimum graduation: 0.1 psia Accuracy: 0.1% of full scale Repeatability: 0.03% of full scale Sensitivity: 0.01% of full scale c. Temperature Indicator Range: 0°F to 125°F Accuracy: 0.5°F Readability: 0.5°F

d. Flow Indicator (Rotameter) Range (dual scale): 87-875 cc/min 23-230 cc/min (air at 70°F and 50 psig) Accuracy: +/- 2% of maximum flow The test established the capability of the Containment Structure to contain the pressure for which it was designed at a leak-rate not exceeding that specified in the license application. These data were plotted to establish initial relationships CALVERT CLIFFS UFSAR 5.5-12 Rev. 47 between internal pressure, leak-rate, external pressure, temperature, relative humidity.

5.5.2 POST-OPERATIONAL SURVEILLANCE 5.5.2.1 Leakage Monitoring The reactor containment and other equipment subjected to containment test conditions are designed to allow periodic leakage rate testing at containment design pressure in compliance with AEC General Design Criteria 52, published in the Federal Register on February 20, 1971. Frequency of the periodic leakage rate test is explained in the Containment Leakage Rate Testing Program. Periodic leakage-rate tests of the Containment Structure will be conducted to verify its continued leak-tight integrity. The post-operational leakage-rate tests are conducted at an internal pressure between 96% of the peak containment accident pressure and 100% of the containment design pressure. The acceptable leakage rate for the test pressure used is given in the Technical Specifications and in the Containment Leakage Rate Testing Program.

The temporary hatch cover plate on the emergency personnel lock shall be seal-tested prior to use during movement of irradiated fuel within the containment. Periodically, in accordance with the Containment Leakage Rate Testing Program, a visual inspection of the exposed accessible interior and exterior surfaces of the containment, including the liner plate will be conducted to assure that no corrosion or other visually apparent deterioration has occurred. The basic steps in conducting leakage-rate tests include the following: a. Measurements of absolute pressure, temperature and moisture content within the Containment Structure. b. Verification of the integrated leakage-rate measurement system by the use of precise measurements of a flow causing a change in the weight of air in the containment that is approximately equal to the measured or permissible 24-hr leak. c. Maintaining pressure between 47.4 and 50 psig for the length of time required by the integrated leakage rate test procedures. d. Controlling containment temperature between 50°F and 120°F. e. Obtaining measurement accuracy tolerances within 95% confidence limits, such that the calculated leakage-rate plus the accuracy tolerance is less than the permissible leakage-rate at the appropriate test conditions. Formulas used in computing the integrated leakage-rate are based on the formulas found in American National Standards Institute/American Nuclear Society 56.8 - 1994, "Containment System Leakage Testing Requirements." The Type A (primary containment overall leakage), Type B (local leakage at penetrations), and Type C (isolation valves) tests for both pre-service and inservice are discussed in Sections 5.1.8, 5.2.3, and 5.5, the Technical Specifications and the Containment Leakage Rate Testing Program, which include acceptance criteria, corrective action to meet the acceptance criteria, test frequency and duration and requirements for reporting test results.

It was expected that the inservice leakage-rate test equipment will be similar to that for the preservice tests. CALVERT CLIFFS UFSAR 5.5-13 Rev. 47 5.5.2.2 Surveillance of Structural Integrity See Appendix 5E for additional surveillances associated with the long-term corrective action plan for addressing vertical tendon corrosion discovered in 1997. The primary objective of the program for Inservice Inspection of the Containment Structure concrete, tendons, and liner during the lifetime of the plant is to ensure the strength and reliability of the post-tensioning steel and other major components such as stressing washers, shims, and bearing plates. The condition of the containments is monitored by a combination of physical testing and visual examinations performed on a regular schedule as called for by ASME Section XI, Subsections IWE/IWL (ASME XI) and 10 CFR 50.55a as it pertains to the Containment Structures.

During construction, 3 tendons of each type, hoop, dome, and vertical, were constructed with 93 wires in lieu of the standard 90 to provide designated surveillance tendons. However, the current ASME XI program requires that a random selection be made with the number examined specified as a percentage of the total population, with a minimum and maximum, of each type of tendon. This percentage varies with plant age and previous surveillance results. Thus, the original surveillance tendons are now a part of the general population. The Code also requires the designation of a common tendon of each type that is examined at each surveillance.

Under the Regulatory Guide 1.35 testing program that followed the initial tendon surveillance program, the Unit 2 Containment did not undergo tendon lift-off testing. The current ASME XI program now requires that both units undergo comparable inspections with an allowance to shift some examinations between units based on the similarity and timing of their construction. Thus, Unit 2 is now subject to the same tendon surveillance requirements as Unit 1. The selection criteria and examination frequencies as specified in ASME XI. Because the tendons were initially strength-tested, the inservice inspection program is conducted to monitor the tendons for corrosion and to verify that the force applied by the tendons meets design assumptions. To achieve those goals, the random selection of tendons is visually examined and the corrosion protection grease is sampled. At alternating surveillances, the tendons are force checked via lift-off testing and one tendon of each type is detensioned and a wire is removed for tensile testing. Those tendons are restressed appropriately immediately after the wire is removed.

The lift-off values are compared to predicted values to determine whether the force required at the end-of-plant life will be met. Any components or values not meeting the acceptance criteria of ASME XI requires scope expansion and/or engineering evaluation.

Visual examinations are conducted over the entire exterior of the containments in accordance with the ASME XI and 10 CFR 50.55a. These examinations are timed and designed to detect any degradation mechanism before it can affect the structural integrity.

Accessible portions of the interior steel liner are visually examined once per inservice inspection period. This examination is to detect any abnormality that could affect the leak tightness of the liner. When necessary, the visual examinations are supplemented with other methods. CALVERT CLIFFS UFSAR 5.5-14 Rev. 47 Since the Unit 2 containment is a duplicate of Unit 1 design, tendon surveillance has been limited to visual inspection without dismantling load bearing components or the anchorage. End anchorages, adjacent concrete surfaces, and the liner plate are inspected. 5.

5.3 REFERENCES

1. Bechtel Corporation, Testing Criteria for Integrated Leakage Rate Testing of Primary Containment Structures for Nuclear Power Plants, BN-TOP-1, Rev. 1, November 1, 1972

CALVERT CLIFFS UFSAR 5.7-1 Rev. 47 5.7 CONTROL OF HEAVY LOADS The NRC issued NUREG-0612, "Control of Heavy Loads at Nuclear Power Facilities" in July 1980 to document the results of Unresolved Safety Issue (USI) A-36. NUREG-0612 provided a set of guidelines intended to minimize the possibility of load drops on safe shutdown or decay heat removal systems. In response to Generic Letters dated December 22, 1980 (Un-numbered), and February 3, 1981 (Generic Letter 81-07), Baltimore Gas and Electric Company submitted a two-phase report reviewing provisions for handling and control of heavy loads at Calvert Cliffs, and evaluating these provisions with respect to the guidelines of NUREG-0612. The NRC accepted our Phase I evaluation (Reference 1). In a safety evaluation report on the implementation of Generic Letter 85-11, the NRC declined to review Phase II responses and released all the respondents from any commitments made in them. However, the NRC stated that while not a requirement, they encourage implementation of actions identified in Phase II. As a result, Calvert Cliffs considered the implementation of these actions to be voluntary. Overhead load handling systems which handle heavy loads (at Calvert Cliffs - loads in excess of 1600 pounds) in the vicinity of the reactor vessel or near spent fuel in the spent fuel pool are subject to the general guidelines of NUREG-0612. Overhead load handling systems that handle any weight in areas where a load drop may damage safe shutdown or decay heat removal systems are subject to the guidelines of NUREG-0612. These guidelines include: a. Definition of safe load paths b. Development of load handling procedures

c. Qualifications, training, and specified conduct of operators

d. Special lifting devices should satisfy the guidelines of American National Standards Institute (ANSI) N14.6-1978 e. Lifting devices that are not specially designed should be installed and used in accordance with the guidelines of ANSI B30.9 f. Periodic inspection and testing of cranes

g. Design of cranes to ANSI B30.2 or CMAA-70 The Calvert Cliffs equipment identified as meeting the criteria of NUREG-0612 are listed in Table 5-10. The remainder of the overhead load handling equipment existing at the time was excluded from NUREG-0612 by the Phase II submittal. The Spent Fuel Cask Handling Crane was designated "single failure proof" using NUREG-0612 and NUREG-0554 criteria. The change in status of the Spent Fuel Cask Handling Crane was acknowledged by the NRC via license amendment (Reference 2). These amendments allow the movement of heavy loads by the Spent Fuel Cask Handling Crane over the spent fuel pool. More information on this crane can be found in Section 9.7.2. The two monorails installed to handle the containment purge system blind flanges, the Containment Auxiliary Crane, and the four containment roof (exterior) jib cranes were evaluated as meeting the criteria of NUREG-0612. Table 5-10 states whether each piece of load handling equipment is subject to NUREG-0612 guidance and what the exemptions are based on. The NRC accepted our method of compliance with NUREG-0612 (Reference 3). A Phase II report was submitted, but the NRC announced a blanket acceptance of Phase II via Generic Letter 85-11 (Reference 4). Based upon our method of compliance with the seven general guidelines of NUREG-0612, fuel damage events resulting from heavy loads incidents will result in offsite doses, due to the release of gap activity, of less than one-fourth of the 10 CFR Part 100 limits. Calvert Cliffs Nuclear Power Plant maintains various controlling procedures providing guidance on load path and lift height restrictions for those cranes subject to compliance with NUREG-0612 and Reference 10.

CALVERT CLIFFS UFSAR 5.7-2 Rev. 47 Restrictions on movement of heavy loads: a. When minimum electrical conditions as defined in Technical Specifications are not met, movement of heavy loads over irradiated fuel is prohibited during shutdown conditions. b. Operations involving movement of recently irradiated fuel in or movement of loads over recently irradiated fuel on a spacer in the spent fuel pool, other than with the single-failure-proof spent fuel cask handling crane, require at least an operable charcoal absorber bank, an operable exhaust fan and an operable high efficiency particulate air filter in the spent fuel pool ventilation system. Additionally, these operations also require that the spent fuel pool water level must be at or above 21 1/2' above the irradiated fuel assemblies. c. When either unit is shut down, the fuel storage oil tank for the emergency diesel generator supporting the shut down unit must be operable for movement of heavy loads over irradiated fuel assemblies. On December 1, 2008, the NRC issued Regulatory Issue Summary 2008-28 (Reference 10), which endorsed Nuclear Energy Institute (NEI) 08-05. The NEI guidance addressed a concern about lifting the reactor vessel head. An Engineering Evaluation was performed to document a reactor vessel head drop analysis that used the NEI guidance. The analysis concluded that the reactor vessels in both Units, attached piping, and support systems are capable of withstanding the impact loads of a hypothetical reactor vessel head drop concentrically onto the reactor vessel flange from a height of 29'. All stresses in steel components would be well within allowable limits; however, there would be significant damage to the concrete beneath the reactor vessel nozzle supports. Due to the damage in the concrete, the reactor vessel would come to rest fully supported by the RCS piping. Procedures are used to control the lift and replacement of the reactor pressure vessel head. These procedures establish limits on load height, load weight, and medium present under the load.

Heavy load equipment is inspected to: Code Equipment B30.2-1976 Overhead and Gantry Cranes (Top Running Bridge, Single or Multiple Girder, Top Running Trolley Hoist) (Reference 5) B30.9-1971 Slings (Reference 5) B30.11-1973 Monorails, Underhung Cranes and Jib Cranes (Reference 6) B30.16 Overhead Hoists (underhung) and Chain Falls (Reference 5) -1987 Same equipment in the Societe Alsacienne De Constructions Mecaniques De Mulhouse Emergency Diesel Generator Building (Reference 7) CMAA-70-1975 Electric Overhead Traveling Cranes (Reference 8) -1983 Same equipment in the Independent Spent Fuel Storage Installation (Reference 9) B30.22-2000 Articulating Boom Cranes 5.

7.1 REFERENCES

1. Letter from R. A. Clark (NRC) to A. E. Lundvall, Jr. (BGE), dated May 27, 1983, Evaluation of Phase I of Control of Heavy Loads 2. Letter from D. G. McDonald, Jr. (NRC) to G. C. Creel (BGE), dated January 17, 1992, Issuance of Amendments for Calvert Cliffs Nuclear Power Plant CALVERT CLIFFS UFSAR 5.7-3 Rev. 47 3. Letter from S. A. McNeil (NRC) to G. C. Creel (BGE), dated August 7, 1989, Supplement to Phase I Safety Evaluation of the Control of Heavy Loads 4. Generic Letter 85-11, NRC to Licensees, dated June 28, 1985, Completion of Phase II of Control of Heavy Loads at Nuclear Power Plants, NUREG-0612 5. Letter from A. E. Lundvall, Jr. (BGE) to D. G. Eisenhut (NRC), dated January 4, 1982, Calvert Cliffs Nuclear Power Plant Unit Nos. 1 & 2; Docket Nos. 50-317 & 50-318, Control of Heavy Loads 6. Letter from R. F. Ash (BGE) to D. G. Eisenhut (NRC), dated August 2, 1982, Calvert Cliffs Nuclear Power Plant Unit Nos. 1 & 2; Docket Nos. 50-317 & 50-318, Control of Heavy Loads 7. Letter from R. E. Denton (BGE) to NRC Document Control Desk, dated July 20, 1993, Calvert Cliffs Nuclear Power Plant Unit Nos. 1 & 2; Docket Nos. 50-317 & 50, Emergency Diesel Generator Project - SACM Diesel Generator and Mechanical Systems Design Report 8. Letter from A. E. Lundvall, Jr. (BGE) to D. G. Eisenhut (NRC), dated March 1, 1982, Calvert Cliffs Nuclear Power Plant Unit Nos. 1 & 2; Docket Nos. 50-317 &

50-318, Control of Heavy Loads 9. Letter from G. C. Creel (BGE) to NRC, Director, Division of Industrial and Medical Nuclear Safety Office of Nuclear Material Safety and Safeguards, dated July 20, 1993, Calvert Cliffs Independent Spent Fuel Storage Installation (ISFSI) Application 10. NRC Regulatory Issue Summary 2008-28, dated December 1, 2008, Endorsement of Nuclear Energy Institute Guidance for Reactor Vessel Head Heavy Load Lifts CALVERT CLIFFS UFSAR 5.7-4 Rev. 47 TABLE 5-10 EQUIPMENT MEETING NUREG-0612 CRITERIA LIFTING EQUIPMENT NUREG-0612 REASON FOR EXCLUSION FROM NUREG-0612 Polar Crane Y Intake Structure Semi-Gantry Crane Y Transfer Machine Jib-Crane Y Purge Flange Monorail (2) Y Spent Fuel Cask Crane Y Containment Roof (Exterior) Jib Crane (4)(a) Y Containment Auxiliary Crane Y Turbine Building Main Crane N Sufficient separation Turbine Building Auxiliary Crane N Sufficient separation Filter Cask Monorail N No floor penetration Solid Waste Disposal Trolley N No safe shutdown or decay heat removal systems endangered Diesel Generator Room Monorail N Sufficient separation Main Steam Room Monorail N Sufficient separation; No safe shutdown or decay heat removal systems endangered Main Steam (MSIV) Room Access Hoist N Sufficient separation; No safe shutdown or decay heat removal systems endangered Machine Shop Monorail N Sufficient separation Containment Equipment Hatch Hoist N Sufficient separation Component Cooling Water Room Hoist N No floor penetration; No safe shutdown or decay heat removal systems endangered Switchgear Room Monorail Hoist N Sufficient separation Chlorine House Monorail N Sufficient separation Condensate Demineralizer Area Monorail N Sufficient separation Condenser Waterbox Removal Monorail N Sufficient separation Vertical Lifting Rail N Sufficient separation Hot Machine Shop Crane N Sufficient separation Decontamination Room Hoist N Sufficient separation _______________________ (a) These cranes have been classified as Augmented Quality and meet the criteria of NUREG-0612. These cranes are unique in that they do not have their own motorized hoisting systems. As such, they are inspected prior to use in lieu of the codes specified in Section 5.7.

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... tr. ;::, t:t. In ... ffi 2 0 u IQ i\j l! ;.. 1\1 8 <<i "' Cl' .... ,;i !ti I w; "' CD ..... ;, b ... t. .. i:: ... r:i 8 "' .. ... 0 l: :z ti: :::! < ..... Ll.J Cl '*"" CONTAINMENT STRUCTURE FINITE £LEMEMT t-4ESH (WALL. AND DOME:) FIG. 5-4 SHEET 2 OF Z Original mesh. See Appendix SE for an evaluation that reduced the original containment minimum design prestress. Revision 33 300 Tl 250 TZ T3 T4 200 -\J... wl50 a :> t-<( er: l.J.J a...100 :I" LU t-50 0 0 Tl= 200 SEC DESIGN THERMAL GRADIENT ACROSS CONTAINME. NT WALL CAL VERT CltF'FS POWE.R F\.ij,! OUT SIDE OF CONC. WALL . T 2 I 000 SEC LINER PL ATE ---T 3 3000 SEC T4 5000 SEC T 5 tOOOO SEC T6 OUTSIDE AMBIENT DESIGN TRANSIENT 10 20 30 40 DISTANCE IN CONCRETE (INCHES) * .. 45 CONTAtNMENT STRUCTURE THERMAL GRADIENT FIG. 5-5 See Appendix 5E for art evaluation that reduced the original containment minimum design prestress. Revision 33 tHt MM.STRESS STRESS INCJIE11f)IT 30 P.S.I. I + ... ... ... ... MERIDICN.ll STRESS STRESS INCREMEljT 200 P.SJ. J NOTES: THE ISOSlRESS CURVES SHOW THE STRESSES IH' THE CONCM:TE ONLY. 2. MAX. STRESS STANOS FOR THE MOllE lEll!At OF THE T\1'0 PRllitl?AL STRE.SSES Ill THE VERTICAi. IUHE. Miii. STRESS STAllDS FOR THE MORE .COHPl!ESSIVE PRINCIPAi. STRESS. THt SPACING Of THE COORDINATE GR() IS INCHES Ill EACH n.\ECTION. 4. Al.I. lfJMERICAI. VALUES REPRESENT /I/If.RIG. STll£SS IN FlllTE El.£HEHTS. CONTAINMENT ISOSTRESS PLOT W"L\. AMO DOME (IHl+Fl FIG 5-6 !llEET I OF4 Original analysis mesh. See. Appendix 5E for an evaluation that reduced the orziginal containment minimum design Revision 33 '8 ;, MAX. STRESS i + + iOP.St. l .... ... ... + + + DIRECTION OF MAX. ANO MIN. STRESS g; "' ... ;r; MIN. STRESS STRESS 1NCRE14£NT 200P.!>J. ;:: .; i' .. mi? ri 'l" 1:1 .,; !i "' "' ::! "I . 'll HOOP STRESS STRESS INCREMENT 200P.SI. '1 -t ;; 1 '4 f '$ MERIDIONAL SiRESS STRESS mR£MOIT s9 P.SJ. ... ... ... ... ... .. r if "' :!: .. 7 ><'f x1 NOTES: l Tl£ ISOSTRESS CUllVES SHOW THE STRESstS IN lHE CONCRETE 0"'-Y. 2. MAX. STRESS ST""'°5 FOR Tt£ MORE TENSl\.E OF TIE. TWO PllliCJ>AI. S'lll£5!.ES IN THE VERTIC.41. Pl.AHE , MH. STRESS STAl<OS FOR THE MORE Cl:M'AESslVE STRESS. !. THE SPAciHG OF THE COORDINATE 6l1IO IS 75 INC>(S 1H E lCH DIRECTION. 4. "1.1. NUMERICAi.. V.11.UES REl'RESENT STRESS IN FINITE E\.£14ENTS. CONTAINMENT STRUCTURE ISOS"TRESS Pl.OT WALi. AHD D<lME (0+1.+F *U!SP} FIG. 5-6 SHEET Z OF4 ,.; d .. "' .. ';" .. Original analysis mesh. See J,\ppendix 5E for an evaluation that reduced the original containment minimum design prestress. Revision 33 ' MAX STRESS 1 60 P.S.I. + t .... DIRECTION OF MAX. ANO MIN. STRESS : :'x. MIN. STRESS ltlCRtMENl 6COP.51. a .... d ... HOOP STRESS STRESS IWCAEMENT 600 P.SJ. .... a i f f ... 1 : . ... . Cl ><a ? ;j. "" ' "' I, THE ISOSTl\fSS CUl\VES liHOW THE STRtSSES II; THE tONti;(TE Otl\.Y. 2. MAX, ST!ifSS STANOS Tt1E "'<>RE TEHSll..£ Of THE TWO Pl<INCIF4l. STR£SS£S IN THE VERTICAi. l'\.ANE. MIN, S.TllESS STANDS FOR THE MOI\£ COW'llESSIVE PfUNOPAI. STUSS. 3. lHE Sl'ACING er THE tooPJ)INATE GRID IS 7' rtCHES 1H EAOI ".Al.I. NUMEIHCAI.. VAl.1£5 R£PRESENT AVERAGE STl!ESS IN FlllTE CONTAINMENT STRUCTURE ISOSTRESS PLOT WAl.L AND DOMt {0+1.+f'"+F'+TA} FIG. 5-6 Sllt£T l OF ot .1 Ci 4 Original analysis mesh. See Appendix 5E for an evaluation that reduced the original containment minimum design prestress. ;; .,. .. .... 1 :t Revision 33 MAX. STRESS S'IRESS INCllf.MtNT DIRECTION OF MAX. AND MIN. STRESS -£0111£CTION Of 1111. STl\(SS OIRFC:TION Of MAX. SlllESS t. MN. STRESS i STRESS INCREMENT a .. a ; 500P.SJ. d .. ;;; .. f HOOP STRESS STRESS INCREMENT 500 P.S,I, a STRESS I 50 P.S.I. .. ,. I fl! a " T 'f E 7 a xa f' >< '* (j I. THE ISOSTRESS 0,1111/ES SHOW THE . STR£SS£S IN THE CONCRETE ONLY. 2.MAX. STAESS STANDS FOR THE HORE TEMSU. OF 1Hf: lWO STRESSES 11 ll£ VERTICM. l'UllE. MIN. STRESS STANDS FOR THE tiORE tc:M'l£SSIYE PlllNClPAL STRESS, ll. THE sPACING

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o.* "' "' .... z w a

0

:: V'I V'I w a: I--G. *. -15. Original analysis mesh. See Appendix SE for an evaluation that reduced the original containment minimum design -33. -6. -IO. -17. .... z -6:36. a: u -984. 1n . -972. -1297. . -1633. -1402. STRESS INCREMENT 50 ai iri er I ...; STRESS INCREMENT 40 ... ... DIRECTION OF MAX. AND MIN. STRESS _£""DIRECTION OF MIN. STRESS '-DRECTION OF MAX. STRESS . \ MIN. ST RESS -STRESS INCREMENT 200 P.S.I. -466. -217. -470. ;j. I HOOP-STRESS

• J STRESS INCREMENT 50 P.S.I. t/'I' 'I' P.S.I. 1 'f'" ' I 'if! -=/ + 1-.,, ........ .,, 12 0: 1-.,, NOTES: I, THE ISOSTRESS CURVES SHOW THE STRESSES IN THE CONCRETE ONLY. 2.MAX, STRESS STAt.IOS FOR THE MORE TENSILE OF THE TWO PRINCIPAL STRESSES IN THE VERTICAL Pl.Al'£. HIN. STRESS SiANOS FOR THE MORE. COl'f'RESSIVE STRESS. 3. THE' SPACING OF THE COORDINATE GRID IS 100 INCHES IN EACH OIRECTIOI( NUMERICAL VALUES REl=-f.tESENT STRESS IN FINITE ELEt4ENT_S CONTAINMENT STRUCTURE ISOSTRESS PLOT WALL ANO BASE lO+L+Fl FIG. 5-7 SHEET I OF 4 Revision 33 Original analysis mesh. See Appendix SE for an evaluation that reduced the original minimum desiQ_fl prestress. :&: -66. -11. -.... ---59. -60. -120. -216:--212. -464 x-0.1 . "' !!j re . l\:i 1 I I \ \ :AX. S TkESS \_STRESS INCltMENT 1 ... ... . DIF\ECTION OF MAX. AND. MIN. STRESS _,,-DlkECfJON MIN. STltESS "-DlkECTION OF MAX STKl:.SS ill "' 'f f ' MIN. STRESS TRESS INCREME. NT 200P.SJ. 311. "'v I I . ' I HOOP STRESS _/ STRESS INCREMENT 100 F.S.1. ti) ti) w a: .... ti) MERIDIONAL STRESS y STRESS INCREMENT 50 P.S.I. q: NOTES: I. TH"El$0sTRESS CURVES SHOW THE. STRESSES IN THE CONCRETE ONLY. 2.MAX,STRE.SS STANDS FOR THE MORE TENSILE OF THE TWO PRINCIPAL STRESSES IN THE VERTICAi. PLANE. MIN. STRESS STANDS FOR THE MORE COMPRESSlVE PRINCIPAL *sT1'ESS. 3. THE. SPACING OF THE COORDINATE GRID 1$ 100 IN EACH DIRECTION. 4.ALL VALUES REPRESENT AVERAGE STRESS IN FINITE ELEMENJS. CONTAINMENT STRUCTURE ISOSTRESS PLOT WALL AND BASE <D+L +F *USP) FIG. 5-7 SHEET 2 OF 4 Revision 33

"' N I-f;) :a }S! (J w 0: t; -76. Original analysis mesh. See Appendix 5E for an evaluation that reduced the original containment minimum design prestress. .-1ss. -181. g I -217. -683. I ' *j HOOP -STRESS STRESS INCREMENT 100 P.SJ. MERIDIONAL STRESS 1 / STRESS INCREMENT 50. P.S.I._/ \cl { \ ro o; i:; "f* IO '° I I MAX-STRESS STRESS INCREMENT 100 P.S.I. DIRECTION OF MAX. AND MIN. STRESS -!(DIRECTION OF MIN, STRESS DIRECTION OF MAX. STF\ESS er 0 ! "" " .,. ,. " ,, JC :c I ! "' .. + -.... ... ...... NOTES: I. THE ISOSTRESS CURVES SHOW THE STRESSES IN THE CONCRETE ONLY *. 2.MAX. STRESS STANDS FOR Tl£ MORE TENSILE*OF THE TWO PRINCIPAL STRESSES IN THE VERTICAL PLAI'£. MIN. STRESS STANDS THE MORE COMPRESSIVE PRl!>.ICIPAL STRESS. 3. THE SPACING OF Tl-* COORDINATE GRID IS 100 INCl-t:S IN EAOi DIRECTION. 4. ALL l'U!ERICAL VALUES REPRESENT AVERAGE STfiESS IN FINITE ELEMENTS. CONTAINMENT STRUCTURE ISOSTRESS PLOT WALL ANO BASE (0+L+F+P+TA) FIG. 5-7 SHEET 3 OF 4 Revision 33 ( .. { -17. '* Vi (l.* &() -42. N '-: ;j :5 0.1 ...... ;: Original analysis mesh. See Appendix 5E for an evaluation that reduced the original containment minimum design prestress. -69. cs ..; !'> "q' I ' ! I lri T I MAX.-STRESS STRESS INCREMENT 100 RSJ. DIRECTION OF MAX. AND MIN. STRESs _.rllRECTION OF MIN. STRES5 * '\....olRECTION OF MAX. STRESS ./ ...: ..., T \ i-: *T MIM. STRESS STRESS INCREt>'f:NT 200P.SJ. )( 0. -?GS:. -396. X-QI "'I HOOP-STRESS "'-STRESS !NCREMENT 100 P.SI. MERIDIONAL

• STRESS y STRESS INCREMENT 50 P.S.I._/ ..... z w u *!!: .fil2lli.: l THE ISOSTRESS CURVES SHOW THE STRESSES IN THE CONCRETE ONLY. ....... -......,.,.......+.,.-r-2.MAX, STRESS STANDS F9R TH£ MORE .:TENSILE OF THE TWO PR!NCFAL STRESSES IN THE VERTICAL Fl.At£. MIN ** STRESS STANDS FOR THE MORE COMPRESSIVE PRINCIPAL STRESS. 3. THE SPACING Tl£ COORDINATE GRID IS 100 IN EACH . DIRECTION. ALL HJ!.ERICAL VAt.l.ES l\'EPRESENT AVERAGE STRESS FINITF CONTAINMENT ' STRUC1'URE ISOSTRESS PLOT WALL AND BASE LOSlO+U-+f"+L5P+ TA FIG. 5-7 SHEET 4 OF 4 Revision 33

,L/.?OM CO;VCRETc (PRcSTRESS, OE.40 LOAD, CREEP, SJIRINKAGE, E/.RTf-,IQU.LJJ./E, PRC:.S..5URE' TEA1PcR.4T()RE) ACCIDENT PRESSUR£ '. * * <5 t .. 1 .. * $ .!!-.. -'_ .. ' * . .j' . . I ** * '. ' . '.Jlf!... . .;_ .... '** ... : .. .. ALSO //v'CtUDE (ACC/DDVT TEMPCR4TUPE EFFECl.S') '. ', ' ' ' I ' ..!(* > -I * \ I _ __..____ -----+-g.__ -------------,MOMENT -A TORS/O;V PIPE LOAOS CONTAINMENT STRUCTURE PENETRATION LOADS FIGURE 5-9 / t r I\ . " .l\ PENE. NO. IA ORIG. FSAR SERVICE PENE. NO. REACTOR COOLANT 21 ANO PRE:SSUR llER SAMPLING -PENE. VALVE PE NETR AT ION PENE. TYPE ARR GT, LOCATION LINE SIZE I 19 WEST PIPING 3/4" PENETRATION R\4, DWG. !SOLA TION POST-NUMBER VALVE: ISi INCIDENT POSIT ION OM-66 CV-5464 CLOSED SH.I OF 3 CV-5465 CLOSED CV-5466 CLOSED CV-5467 CLOSED DIAGRAM OUTS I GE c nn. s rnuc Tu RE IN 5 IDE c TM T. 5 TR u c T P,_R_F ____ ___,,___ __ I *--*------------' ... f--L, -(ff};" '.'.0' \ v c ; ff S'4G5 *1i;:; AO l NOTES I I 5lAS -j CL.J'St::3 T--' --I. f!ON ON' 1:i.LL SHEETS SUCH AS OWG."JOS., ISOLA f 10'11 VALVE NOS. AND!fHE DIAGRAM AR1=: GIVF::N UNIT I O'l._Y. REFER TO THE FCR INF'OR\tATION 0111 UN!T 2. 2. fHE OIAGRA\45 TAKEN FROM THE REFER*:-.lCED OM DRAWINGS. ALL VALVES AR::'. SHOWN IN THEIR DURING !\IOR\4AL PLANT OPERATIO l\iOOE II AS SHOWN ON THE 0\.1 DRAWINGS. 3. SY"-4BOLS ON TIHf DIAGRAMS ARE !DEN TIF IEO ON FSAR FIG. '9 -1. ALL SOLE NO ID Y'ALVE_S ARE SHOWN IN THE IR DEE: NERGIZE 0 POSIT ION PER F",AR FIG. -1. SHEET NO. I CONTAINMENT STRUCTURE lSOLATlON VALVE ARRANGEMENT I F 5 -10 PENE. NO. 18 lC SERVICE RC DRAIN TANK N0.11 VENT HEADER TO WASTE GAS SURGE TANK REACTOR COOLANT PUMP SEALS CONTROLLED BLEED OFF ORIG. FSAR PENE. NO. 36 45 PENE. VALVE TYPE ARRGT. I I I 12 6 PENETRATION PENE. LINE LOCATION srzE WEST PIPING 2" PENETRATION RM. WEST PIPING 3/4" PENETRATION RM. DWG. NUMBER OM-078 SH0001 OM-077 SH0001 OM-073 SH0002 ISOLATION VALVE CS l CV-2180 CV-2181 CV-505 CV-506 POST-I NC IDE NT POSITION CLOSED CLOSED CLOSED CLOSED INSIDE CTMT. I -RCW*337 I -RCW-340 1 *RCW*339 SIAS CLOSES CTMT. TRENCH f (if L_. L __ ;/-1 .... I .. t I I "{. ' I I 1 1 *CVC*370 L.!.J TELL TALE AIS DIAGRAM STRUCTURE SIAS Cl.OSES I AIS ll I o-WGS-668 ' G }r O-WGs-669 LI L.:.J TEST CONN. CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT FIGURE 5-10 SHEET NO .. 2 \ OUTSIDE CTMT. STRUCTURE f*CVC*358 L.:.J TELL TALE TO WASTE GAS SURGE TANK NO.I I REVISION: 21 PENE. NO. 10 2A SERVICE POST-ACCIDENT SAMPLING LI au ID RETURN TO RC DRAIN TANK LETDOWN LINE TO PURlFICATION DEMIN NOTE 1: EITHER CVC-103 OR CVC-105 IS NORMALLY OPEN. THE SECOND VAL VE IS NORMALLY CLOSED. ORIG. PENE. FSAR PENE. TYPE NO. IV 2 VALVE PENE. PENETRATION DWG. ARRGT. LOCATION LINE NUMBER SIZE 37 WEST PIPING 114* OM-066 PENETRA Tl ON RM. SH0003 7 WEST PIPING 2* OM-073 PENETRATION RM. SH0002 SH0003 POST-ISOLATION VALVE CS l INCIDENT POSITION SV-6529 LOCKED SHUT CV-515 CLOSED CV-516 CLOSED CVC-103 SEE CVC-105 NOTE 1 I NS IDE CHAT. TO RC DRAIN TK. NO. 11 FC eves SIG CLOSE SIAS CLOSES 1 *CVC*193 I *CVC*319 STRUCTURE TEST CONN., r.1 I *PS 1012 DIAGRAM l*PS 1013 J I *CVC* 364 DRAIN TEST CONN. I *CVC* I -eve. 299 298 CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT FIGURE 5-10 SHEET NO. 3 OUTS IDE CTMT. STRUCTURE FC ICIOI r 9 I t*PS t*PS Sii 1020 L:.J TEST CONN. t*CVC* I *PS 1019 L:.J TELL TALE. 1 *CVC* 300 1 *CVC* 302 I *CVC* 106 366 1 *CVC* TO t *CVC* LJ TELL TALE 367 LJ TEI.I. TALE 306 LETDOWN HEAT EXCH NO.II REVISION: 21 PENE. NO. 28 ORIG. PENE. PENETRATION PENE. SERVICE FSAR VALVE DWG. TYPE ARRGT. LOCATION LINE NUMBER PENE. SIZE NO. REACTOR COOLANT 3 9 WEST PIPING 2" OM-73 CHARGING L rNE PENETRATION RM. SH. 2 OF 3 POST-ISOLATION VALVE<Sl INCIDENT POSITION CVC-184 -----CVC-435 -----CV-519 OPEN CV-518 OPEN CV-517 CLOSED TO RC LOOP N0.126 INSIDE CTMT. STRUCTURE AUXILIARY SPRAY TO PRESSURIZER DIAGRAM LhELL TALE. LJ TELL TALE. 51'3 1co1 NDE AIS I LJ TEST CONN. CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT FIGURE 5-10 SHEET NOo 4 OUTSIDE CTMT. STRUCTURE REGENERATIVE EXCHANGER NO.I I TEST CONN. TELL TALE REVIS I ON : 1 7

PENE. NO. 7A 78 8 SERVCCE ILRT PENETRATCON ILRT PENETRATION CONTAINMENT NORMAL SUMP TO MISC. WASTE DRAIN TK. OR[G. FSAR PENE. NO. ----14 PENE. TYPE IV IV [ [ VALVE ARRGT. 45 45 13 PENETRATION LOCATCON WEST PIPING PENETRATION ROOM WEST PCPCNG PENETRATION ROOM CONTAINMENT RECCRC. PCPE TUNNEL PENE. L!NE SIZE 314* 314 .. 4* DWG. ISOLATION POST-[NCCDENT NUMBER VALVE POSITION OM-065 !LRT-1 CLOSED SH0002 BLIND FLANGE CLOSED OM-065 CLRT-2 CLOSED SH0002 BLIND FLANGE CLOSED OM-076 MOV-5463 CLOSED SH0001 MOV-5462 CLOSED SH0004 I NS CDE CTMT. _ n ====::::::1::J I I i-1, ... I I NORMAL SUMP 1 _, DIAGRAM STRUCTURE n-ICIO r'l SUS (JC , . CL(ISES ye OUTSIDE CTMT. STRUCTURE TEST CONN. f ** -ILrT* ... t><l-I* I* ILRT* ILRT* I 3 TEST CONN, r* _":'" ... C>4-I* 1* ILRT* ILRT* 2 "' ICIO {iii!l y I SIAS <r-CLOSES 1 .Y 1c10{0co r -_IMJ-I -@Lc::>DL yc I sm M TO 12 ECCS -.. '-..:* PUMP RN. NO. tt8 "-r "" -..-.... -, , I *EA0*125 H TEST CONN. SUMP N0.13 CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT I FIGURE 5-10 SHEET NOo 7 I RE v Is I ON : 21 PENE. NO. 9 10 SERVICE CONTAINMENT SPRAY WATER N0.12 CS HEADER CONTAINMENT SPRAY WATER C NO.II CS HEADER ORIG. rSAR PENE. NO. 4 5 PENE. VALVE TYPE ARR GT. II 17 II 17 PENETRATION LOCATION WEST PIPING PENETRATION ROOM EAST PIPING PENETRATION ROOM PENE. DWG. ISOLATION LINE NUMBER VALVE ISi SIZE 8" OM-74 SI-326 SH.3 OF 3 OM-52 SI-340 8" OM-74 SI-316 SH.3 OF 3 OM-52 51-330 POST-INCIDENT POSITION TO CTMT. SPRAY HllR.N0.12 TO CTMT. SPRAY HOR.NO.II INSIDE CTMT. STRUCTURE TO CHARCOAL. FILTER SPRAY SYSTEM FO TO CHARCOAL Fii. rER SPRAY SYSTEM FD DIAGRAM OUTS IDE CTMT. STRUCTURE I AIS 61CIO I* SI* 388 TEST CONN. L..!.J I Al'S 6*cos I* SI* 378 TEST CONN. I* Ji. I* Ji. SI* H6 387 L!...J TEST CONN. VENT r.l L!.J TEST CONN. L!..I SI* 325 328 TELL*TALE DRAIN L!...J TELL-TALE DRAIN CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT FIGURE 5-10 SHEET N 0. 8 REV* 15 5-10 CONTAINMENT STRUC1URE-ISOLATION VALVE ARRANGEMENT (Sheet 09) PENE-I I ORIG. SERVICE FSAR NO. PENE. NO. 11 I CONT A!NMENT I 12 STRUCTURE SUMP WEST REC!RC.HEADER 1 2 I CONT A!NMENT I 1 3 STRUCTURE SUMP EAST REC!RC.HEADER PENE. VALVE TYPE ARRGT. 11 16 t1 36 PENETRATION LOCATION CONT A!NMENT RECIRC.PIPE TUNNEL CONTAINMENT RECIRC.PIPE TUNNEL PENE. ISOLATION LINE owe. SIZE NUMBER VALVE IS I 24" OM-074 MOV-4145 SH0003 24" I OM-074 I MOV-4144 SH0003 POST-INCIDENT POSITION OPEN OPEN DIAGRAM INSIDE CTMT. STRUCTURE I OUTSIDE CTMT. STRUCTURE CGNUINIENT SUll' --r CONUINIENT -SUll' .......... !l RAS Ol'EHS't---do9! ... __ I \---+-_:WEST RECIRC. --HEADER TEST CONN. ftt] I :c14i l*SI J 4170 090 VALVE ENCAPSULA Tl ON r I-SI 4174 LJ RAS OPEH5'---, @1coa' I .4*441 Ill _.__ '--1>-t--J l-Sl-t090l VALVE ENCAPSULATION 11-51' 417S L.:.J rj1 I-SI 4177 -I-SI 149 mll L.J L.J CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT 1 FIGURE 5-10 SHEET NO. 9 I REV=32 I DATE=ll/12/02 B007..1..DGN F5M Rev.32 FIGURE 5-10 CONTAINMENT 5TRUCTURl:.-l:SULA 1 IUN VALVt:. AK11AN\>t:.Mt:.N I <;,nt1t1< nil PENE. ORIG. PENE. VALVE PENETRATION PENE. ISOLATION POST-DIAGRAM SERVICE FSAR LINE DWG. INCIDENT NO. PENE. TYPE ARRGT. LOCATION SIZE NUMBER VALVECSI POSITION INSIDE CTMT STRUCTURE OUTSIDE CTMT. STRUCTURE NO. 13 PURGE AIR 40 Dr EAST PIPING 46" OM-065 INLET PENETRATION SHOOOl BLINO FLANGE INSTALLED ROOM w .. CRS C\.OSES ' ' '\ Cl* 1:l>> ' .... ' ' ' ND< TEST ' ' L--CONN. L--\--tii Alli r, j-171 ' (* --> ... ' AO CPA-111 ... I I re 14 PURGE AIR 41 Dr EAST PIPING 48" OM-065 OUTLET PENETRATION SHOOOl BLINO FLANGE INSTALLED ROOM a r--1> CRS CLOSES I I I ---, I n I I UST N _! c-. 171 ' AO 'tT AIS * *r I L-r,s,*-** :=1 I ,0 I 1' re CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT FIGURE 5-10 !SHEET NO. 10 I REVISION: 35 Revision 35 PENE. NO. 15 -16 SERVICE CONTAINl.1E:NT ATt.405. AND PURGE AIR MONITOR COMPONENT COOL ING WATER INLET TO RE:AC TOR COOLANT PUMPS ORIG. PENE. VALVE FSAR PENE. TYPE ARR GT. NO. '12 II s 24 f II 25 PENETRAl ION LOCAl ION EAST PIPING PENETRATION ROOt.I EAST PIPING PENETRATION ROOM PENE. SIZE , .. 10" OWG. NUt.4BER OM-98 SH.I Or Z OM-51 SH.2 OF' :: ISOLATION VALVEISl CV-5292 CV-5291 CV-3832 POST* INC JOE:NT POSIT ION CLOSED CLOSED CLOSED DIAGRAM INS lDE CT MT. Sl RUC1'U,RE 1-sv SZ'31 ! I I 1 l*SV J 383Z l*SV 529Z I ICIO @-*---:3"E : 1,\,15 :I .L '" "' ** C IS /'" i CLOSES fF' CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT I OUTSIDE CH.AT. STRUCTURE: *' SlilS CL.05[5 1r10 : I I Al5 I L .. *" cv 3832 TELL Till.I! r.i AO IC10 C 3832 SHEET NO.II I FIGURE 5-10 CONTAINMENT STRUCTURE -ISOLATION VALVE ARRANGEMENT (Sheet 12) ' PENE. ORIG. PENE. PENETRATION PENE. DWG. POST-DIAGRAM FSAR V-'LVE ISOLATION NO. SERVICE PENE. TYPE ARRGT. LOCATION LINE NUMBER VALVE ISi INCIDENT NO. SIZE POSITION INSIDE CTMT. STRUCTURE OUTSIDE CTMT. STRUCTURE 17A STEN.I GEN. N0.12. 22 Ill 14 EAST PIPING 2" OM*J5 CV-4012 CLOSED SURFACE SLOWDOWN PENETRATION ROOM SH.1 OI' J .r OM-454 .... .... .. ,_____ C:SAC Q.DIU DI c ., .. *'\T .,. (\ . -JI -.. l : ... ...... -.. "' ....,.coeteur """" ao* DOWN -\.....,/ . -....... 178 STEAM GEN. N0.12. 23 Ill 14 EAST PIPING 2" CV-401J CLOSED BOTlOM BLOWDOWN PENETRATION ROOM SM.1 OI' J -:1: OM-464 -t-----1: ...... -=-:= 'tT.,.. .,. /it'll . -L q -.. ! ,......__. -.. .. '...,, .18 LAN,P PSmO ....,...,.,..,, rll!:f/.m 22571.DGH FSNl-12 CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT I FIGURE 5-10 I SHEET NO. 12 I REVISION: 42 L ...I Revision 42 PENE. NO. 18 19A OR[G. PENE. VALVE SERVlCE FSAR TYPE ARRGT. PENE. NO. COMPONENT COOLING 29 I I I 2 WATER OUTLET FROM REACTOR COOLANT PUMPS INSTRUMENT AIR 38 I I I 31 PENETRATION LOCATION EAST PIPING PENETRATION ROOM EAST PIPING PENETRATION ROOM PENE. LINE SlZE 10" DWG. NUMBER OM-051 SH0002 OM-053 SHOOOl SH0003 ISOLATION VALVE<Sl CV-3833 MOV-2080 IA-337 POST-[NCI DENT POS[TION CLOSED CLOSED DIAGRAM INSIDE CTMT. STRUCTURE OUTSIDE CTMT. STRUCTURE .... l __ y 3833 '---NE -... A/$ 31133 l*ZS t* F1 ... 833 3833 .__ _ __._ __ --! AO , r I *CC-287 r I -CC*482 1 *CC*288 I *CC*289 L:....J L:....J L:....J I* [A-339 TEST CONN. 1-u-331 I \\"7 1 , jj l*IA*:Sl8 I l*U*336 a L!.J L.:.J VENT VENT ICIO tlCIO , , 1 *[A-636 L.:J VENT TEST CONN. TELL TALE , , 1

• IA*638 L:...J VENT CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT I FIGURE 5-101 SHEET NOo13 I RE v Is I ON : 21 PENE. NO. 198 20A SERVICE PLANT A[R N2 SUPPLY TO SAFETY INJECTION TANKS 11A.11B.12A 8 128 ORIG. FSAR PENE. NO. 39 34 PENE. TYPE [V I I I VALVE PENETRATION ARRGT. LOCATION 3 EAST PIPING PENETRATION 38 EAST PIPING PENETRATION ROOM PENE. DWG. ISOLAT[ON LINE NUMBER VALVE CS) SlZE 2" OM-479 PA-1044 SH0002 PA-1040 1
• OM-068 OM-074 N2-344 SH0002 CV-612 CV-622 CV-632 CV-642 POST-[NCI DENT POSlTION LOCKED SHUT LOCKED SHUT --CLOSED CLOSED CLOSED CLOSED DIAGRAM INSCOE CTMT. STRUCTURE OUTS£DE CTMT. STRUCTURE TO 51 TANK llA TO SI TANK ttB TO SI TANK 12A TO SI TANK 12B LJ ORN. 1*5*1 612 I I C Mi----.! ;---!.-, *(. <$ @AO -I o.. _.. I t-Sl-491 l*CV 622 1C08 l -SI *492 ""i:c....,. ... *----1,./T-I *SI *494 I *CV 632 1C09 LJ TEST CONN. LJ TEST CONN. LJ TELL TALE ' r O*N2 ' .. 363 LI TELL TALE CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT I FIGURE 5-101 SHEET NOo14 REVISION: 21 PENE. NO. 208 20C SERVICE NITROGEN SUPPLY TO RC DRAIN TANK NITROGEN SUPPLY TO STEAM GENERATORS N0.11 8 12 AND TO PRESS. QUENCH TANK ORIG* FSAR PENE. PENE. TYPE NO. 34 I I I 34 [[I VALVE ARRGT. 18 39 PENETRATION LOCATION EAST PIPING PENETRATlON ROOM EAST PIPING PENETRATION ROON PENE. LINE SIZE 1. 1
• POST DIAGRAM DWG. ISOLATlON [NCIDENT NUMBER VALVE<Sl POSITION OM-068 OM-077 N2-345 SH0001 OM-068 OM-035 SHOOOI OM-072 SH0001 INSIDE CTMT. STRUCTURE I R.C. DRAIN TANK I NO.II I TO ... O*Nz*2412 TO PRESSURIZER QUENCH H.NK NO.II TEST CONN. OUTSIDE CTMT. STRUCTURE J O*N *592 l*PI -6320 --.... O*N *JOSI CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT I FIGURE 5-101 SHEET NOo15 I RE v Is I ON :: 21 PENE. NO. 21 22 SERVICE AUX FEEDWATER TO STEAM GENERATOR N0.12 AUX FEE OWAIER TO 5 TE AM GENERATOR NO.II ORIG. FSAR PENE. NO. 17 18 PENE. VALVE TYPE ARR GT. 111 15 If I 15 PENETRAl ION LOCAl ION EAST PIP PENETRATION ROOM EAST PIP JNG PENETRATION ROOM F'E NE. LINE SIZE .... OWG. ISOLATION NlJ,._.6ER VALVE IS I OM-800 AFW-200 CV*4'51Z AfW-165 OM-BOO AFW-199 CV-4511 AFW-163 POST-INC !DENT POSIT ION OPEN LOCKED SHUT OPEN LOCKE:D SHLIT DIAGRAM i '----INS IDE CHAT, STRUCTiJRE OUTSIDE CTtJT. STRUCTURE -t I I I i \ !ITEAM N0.*1 __ l_S_O_L_A_T_I_o __ N_V_A_L_V_E __ AR __ R_A_N_GE __ M_E_N_T--'-___ __ 5_-_l_O __ -L--__ s_H_E_E_T __

ORIG. PENE. POST-DIAGRAM PENE. SERVICE FSAR PENE. VALVE PENETRATION LINE DWG. ISOLATION INCIDENT NO. PENE. TYPE ARRGT. LOCATION SIZE NUMBER VALVE POSITION INSIDE CTMT. STRUCTURE CTMT.STRUCTURE NO. 23 DRAINS FROM REACTOR 35 111 24 WEST PIPING 2* OM-077 CV-4260 CLOSED COOLANT SYSTEM PENETRATION ROOM SHOOOl DRAIN TANK 1C:U r ---1*$V I 1-PT --@ 060 42S9 42!19 I 1C10 1 *RCW-3036 ... ... -... DRAIN 1CIO A"' .. 'A 1 *RCW-3035 I! 1-cv 42&0 426 4260 AO 1 *RCW-128 J TEST -i p .... CONN. I -RCW-319 I ----1'01 ..... ..._. JJ "'FC 1*RCW-314f L!.J DRAIN 24 OXYGEN SAMPLE LINE -11 I 37 WEST PIPING 114* OM-463 SV-6531 LOCKED <FROM PRZR. QUENCH PENETRATION ROOM SH0001 SHUT TANK N0.11 l r _I_ --1-sv r --TEST $ li531 CONN. I ICIOI r.l -I@ , , B NOE I .. ---li531 l I *PS 101s, l 1 *PS-S15 1-sx ... 400 -.... 1-PS u ,, 1016 1-PS-610 1 *PS-611 H H L..:J L'...J TEST TELL CONN. TALE 22577.0GN CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT I FIGURE 5-10 I SHEET N0.17 I REVISION: 26 W,H,RICE 10/19/99 PENE. ORIG. SERVICE* FSAR NO. PENE. NO. 25 SERVICE WATER INLET 25 TO CONTAINMENT COOLING UNIT N0.11 26 SERVICE WATER INLET 26 TO CONTAINMENT COOLING UNIT N0.13 PENE. TYPE I I I I I I PENE. ISOLATION VALVE PENETRATION LINE DWG. ARRGT. LOCATION SIZE NUMBER VALVECSl 26 EAST PIPING 8" OM-46 SRW-317 PENETRATION ROOM SH.2 OF 2 26 EAST PIPING g* OM-46 SRW-319 PENETRATION ROOM SH.2 OF 2 POST-I NC IDE NT POSITION --I NS I DE CTMT. STRUCTURE } -L'..J DR t -SR'tl-1151 r: ...... DRAIN L.: .,..-...., !-SRW-1152 [ ORA!N .,..--., I -SRW-317 I t -SR'tl-1148 DRAIN. [ : 1-SRW-1147 ORA IN [ :...-...; I -SRW-319 I I -FT 1589 !'--" DIAGRAM OUTS IDE CTMT. STRUCTURE SIAS __ RAS OPENS I 1 -sv J_ 1589 l-cv E 1589 AIS I-& 0 y; )(' AC CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT I FIGURE 5-10j SHEET NOa18 I RE v Is I ON 1 7 PENE. SERVICE NO. 27 SERVICE WATER INLEI TO CONTAINMENT

• COOLING UNIT N0.14 28 SERVICE WATER INLET TO CONTAINMENT COOLING UNIT NQ.12 ORIG. FSAR PENE. PENE. TYPE NO. 27 I I I 28 I I I PENE. ISOLATION VALVE PENETRATION LINE DWG. ARRGT. LOCATION SIZE NUMBER VALVECS) 26 WEST PIPING 8" OM-46 SRW-320 PENETRATION ROOM SH.2 OF 2 26 WEST PIPING 8" OM-46 SRW-318 PENETRATION ROOM SH-2 OF 2 POST-I NC !DENT POSITION --DIAGRAM I NS IDE CTMT. STRUCTURE OUTS IDE CTMT. STRUCTURE J J I *SRll-1126 DRAIN [ """" ........ 1 -SRll-1125 DRAIN [ 1 *SRW-320 1 t -SRW-lf 31 DRAIN [ ""-... -... 1*SRW*1132 1 *FT 1592 .__., 2 I -FT 1584 ,___.... DRAIN [ .._ " ... -... 7
• j 1-SRW-1 se I I SIAS THROTTLEsfs-I RAS OPENS 1c13 I 1-cv 1584 AIS -* :,-1--. .c._ ./,;..-'AC 11 yF 1-SRW-144 1-SRW-318 I CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT j FIGURE 5-101 SHEET NOo19 ] REVIS I ON 1 7 PENE. NO. 29 30 SERVICE SERVICE WATER RETURN CONTAINMENT COOLING UN1T#l3 SERVICE WATER RETURN CONTAINMENT COOLING UN IHll ORIG. f SAR PENE. TYPE PENE. NO. 30 III 31 III VALVE ARR GT. 22 22 PENETRATION LOCATION EAST PIPING PENETRATION RM. EAST PIPING PENETRATION RM. PENE. LINE SIZE 8" 8" DWG. ISOLATION NUMBER VALVE ISi OM-46 CV-1590 SH.2 OF 2 CV-1591 SRW-154 OM-46 CV-1582 SH.2 OF 2 CV-1583 SRW-140 POST-INC IDE NT POSITION OPEN OPEN LOCKED SHUT OPEN OPEN LOCKED SHUT INSIDE CTMT. STRUCTURE l*SRW* 473 LJ DIAGRAM 1*5RW* 474 OUTSIDE CTMT. STRUCTURE l*SRW-154 H '-... J..--Lr F.D. l*SRW-153 .,--$_ I c I _J l*SRW-138 1-------11 / -, I I o 1------11 / I-----F.O. l*SRW-139 1*5RW*140 CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT FIGURE 5-10 SHEET NO. 20 REV 15 PENE. NO. 31 32 ORIG. SERVICE FSAR PENE. NO. SERVICE WATER 32 RETURN CONTAINMENT COOLING UN IT#l4 SERVICE WATER 33 RETURN CONTAINMENT COOLING UNIHl2 PENE. VALVE TYPE ARR GT. III 22 III 22 PENETRATION LOCATION WEST PIPING PENETRATION RM. WEST PIPING PENETRATION RM. PENE. LINE DWG. ISOLATION SIZE NUMBER VALVE ISi 8" OM-46 CV-1594 SH.2 OF 2 CV-1593 SRW-161 8" OM-46 CV-1586 SH.2 OF 2 CV-1585 SRW-147 POST-INCIDENT POSITION OPEN OPEN LOCKED SHUT OPEN OPEN LOCKED SHUT DIAGRAM INSIDE CTMT. STRUCTURE OUTSIDE CTMT. STRUCTURE l-SRW-162 f l-SRW-161 F.D. l-SRW-160 1-..f' I c I -' 1-SRW-"'5 F.O. l-SRW-146 l*SRW-147
• CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT FIGURE 5-10 SHEET NO. 21 REV 15 I PE NE. I SERVICE NO. : i i I 1 MAit-; rEEOWA1E:R TO 33 I l*1I STEAM GENERATOR i ! I I I i I i i i i I ! ! i ! : I ' l ' i I l ! I i ' I ! ; ; ! ORIG. f"SAR PENE. N Cl. 15 I i I I I i ' PENE. TYPE Ill ! ,, , ,, I I i I ' I I I I I I I VALVE ARR GT. 32 32 I PE NE lRAl 10"1 LOCAl ION MAIN STE.AM PENETRATION RM. I ! I I I I I ! t.!A!"I STEAM I PENETRATION F.M. I ! ! ! i I j i I I I I I I I I PENE. L It-IE: 5 IZE: 16" i6" I OWG. OM*39 SH.4 OF 4 I I I I i I I I I i OM-39 I SH.4 OF I I ISOLATION VALVE IS I MOV*4516 .
• I l.<OV-45'7 I I I I I I POST-INC IDE NT POSIT ION INS IOE -CLOSED I I ! l CLOSED CONTAINMENT STRUCTURE lSOLATION VALVE ARRANGEMENT C rt.n. STRUCTURE l*l"W* 130 zzz ,.,.,., . . ZZJ :___ii"'"' .. DIAGRAM OUTSIDE STHUCTURE . ' !C'OJ I O*E'l'll!r.: i ,,,., tS3!5/CSASI l*HS Ct.. VAL \o.( I iCOJ I I @--1 __ I I I IC03 I I VC -I l I I I I I @ c I J;-, t . ,7 .. -l *A FI GU RE 5 -10 SHEET NO. 22 PENE. NO. 35 SERVICE MAIN STEAM FROM *II STEAM GENERATOR ORIG. FSAR -PENE. NO. 19 PENE. VALVE TYPE ARR GT. III 23 PENETRATION PENE. DWG. LINE LOCATION SIZE NUMBER MAIN STEAM 34" OM-35 PENETRATION SH.I OF 3 ROOM OM-114 SH.I OF 2 OM-800 ISOLATION POST-INC IDE NT VALVEISJ POSITION CV-3938 CLOSED MOV-6621 CLOSED CV-4043 CLOSED MOV-4045 CLOSED MOV-6612 CLOSED CV-4070 OPEN/ CLOSED MS-102 CLOSED MOV-6615 CLOSED CV-4070A OPEN/ CLOSED DIAGRAM INSIDE CHAT. STRUCTURE OUTSIDE CTMT. STRUCTURE STEAM GENERATOR I-MS 250 I-MS 249 l-MS-102 NOTES: I. FOR CONTROL DETAILS, SEE DWG. OM -35, SH. I OF 3. 2. FOR CONTROL DETAILS, SEE DWG. OM-114, SH.I OF 2. 3. FOR CONTROL DETAILS, SEE DWG. OM-480. r--, I I I I CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT FI GU RE 5-10 SHEET N 0. 23 REV* 15 PENE. NO. 36 SERVICE MAIN STEAM FROM *12 STEAM GENERATOR ORCG. FSAR PENE. NO. 20 PENE. TYPE III VALVE PENETRATION PENE. LINE ARR GT. LOCATION SIZE 23 MAIN STEAM 34" PENETRATION ROOM DWG. ISOLATION NUMBER VALVEISl OM-35 CV-3939 SH.I OF 3 OM-114 MOV-6620 SH.I OF 2 CV-4048 OM-800 MOV-4052 MOV-6611 CV-4071 MS-222 MS-105 MOV-6613 CV-4071A POST-INCIDENT POSITION CLOSED CLOSED CLOSED CLOSED CLOSED OPEN/ CLOSED CLOSED CLOSED CLOSED OPEN/ CLOSED DIAGRAM INSIDE CTMT. STRUCTURE OUTSIDE CTMT. STRUCTURE STEAM GENERATOR *12 l*MS-104 DLC)---, AFAS SSIAIJ \--i ti-SV\ I I J:o---..---ND IC04 l*MS-105 8 RELIEF VALVES 1 I -Ye J NOTES: I. FOR CONTROL DETAILS, SEE DWG. OM-35, SH.I OF 3. 2. FOR CONTROL DETAILS, SEE DWG. OM -114, SH.I OF 2. 3. FOR CONTROL DETAILS, SEE DWG. OM-480. TO AUX FEED PUMPS CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT FI GU RE 5-10 SHEET NO. 24 REV 15 ORIG. PENE. POST-DIAGRAM PENE. SERVICE FSAR PENE. VALVE PENETRATION LINE DWG. ISOLATION INCIDENT NO. PENE. TYPE ARR GT. LOCATION SIZE NUMBER VALVEISl POSITION INSIDE CTMT. STRUCTURE OUTSIDE cnn. STRUCTURE NO. 37 PLANT SERVICE WATER --II 43 WEST PIPING 3'" OM-479 PSW-1008 CLOSED PENETRATION ROOM SH.I OF 2 PSW-1019 CLOSED 2 OF 2 TEST TELL TO U-1 TEST CONN. TALE RWT STA. CONN. r.i r.i r.t O*PSW IOI '°-PSW '°-PSW ..... $-1010 co ,tzy .......... ...,,,,. ...-I ...-.... O*PSW 0-PSW 0-PSW 201 1001 1019 0-PSW 1008 38 DEM INER AL IZED WATER I III I EAST PIPING 2" OM-72 CV-5460 CLOSED TO QUENCH TANK PENETRATION ROOM SH.I OF 2 ICIO l*SV I 5460 I NDE* -0 /(-HS\ I ' .. A IVS E :f 0 , I AO I I-CV ..... &460 I O-DW-251 l-DW-1200 ... .JI. , ...-,..i: .... v""" 'O-DW-281 'O-DW-282 .a ,j L.:.J L.'..J TELL TALE TEST CONN I I CONTAiNMENT STRUCTURE iSOLATION VALVE ARRANGEMENT I FI GU RE 5-10 SHEET N 0. 25 I REV . 15 .

PENE. NO. 39 ORIG. PENE. PENETRATION PENE. SERVlCE FSAR VALVE DWG. PENE. TYPE ARRGT. LOCATlON UNE NUMBER NO. SIZE SAFETY INJECTION 11 IV 20 WEST PlPING 2" OM-074 TANK TEST LINE PENETRATION ROOM SH0002 POST-I SOL A TI ON VALVE CS l I NC IDE NT POSITION SI -463 LOCKED SHUT SI -455 LOCKED SHUT I NS rDE CTMT. VENT VENT r:"1 r:"1 l*SI-I -SI-503 502 SIAS CLOSES5--. I I DIAGRAM STRUCTURE VENT n I *SI -501 CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT FIGURE 5-10 SHEET NOo 26 OUTSIDE CTMT. STRUCTURE LI DRAIN l*SI* TO RWT 455 NO.JI I *SI-I *SI* 495 461 LI TEST CONN. l"ROM S I TANK N0.128 w TELL TALE REVIS I ON : 21 22588.DGN W.H.FllCE 10119199 PENE. NO. 41 42 SERVICE REACTOR COOLANT SHUTDOWN COOL [NG FUEL TRANSFER TUBE ORIG. F'SAR N . 10 43 PENE. VALVE TYPE ARRGT. [V 11 [V 8 PENETRATION LOCATION EAST P[P!NG PENETRATION SPENT FUEL POOL PENE. POST-DWG. ISOLATION LINE NUMBER VALVE INCIDENT SIZE POSITION 14. OM*074 MOV-651 LOCKED SH0002 SHUT MOV-652 LOCKED SHUT 36" OM-058 SFP* 1 CLOSED DIAGRAM INSIDE CTMT. STRUCTURE OUTSIDE CTMT. STRUCTURE I NTERl.OCk FROM t *PIC* t Ol*t PRZR PRE55.0N*721C*7 l 1C09 I ;.e __ _g__:_, I KEY OPER.UEO I r-------.J i I I M .: 1 *RV 4&9 \/£NT 1J1 *SI* j. 509 TEST CONN. n 1*MO /\ 1C09 151 'f" 1J1*51* C I 510 CTNT 1 651A TRENCH 1 ?? TO SOUTM J M I. t*SI* u t*SI* : Hri ,,, INTERLOCI< FROM l Bl. I ND FLANGE O*SFP*l [ A 1 *PIC*10l I &75 1, -........ , ... ..... ,. 1 *SI. 507 -........ 1 *51* ..,5°.!, TELL .,. ..., J TALE I INTERl.OCK FRON 1 *PIC*10l S: PRZR.PRESS. ON*721A*9l Ct! F'UEL TRANSFER TUBE x SFP*l IS NOT SUBJECT TO TYPE "C" TESTING. CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT I FIGURE 5-10 I SHEET NO. 27 REVISION: 26 FIGURE 5-10 CONTAINMENT STRUCTURE-ISOLATION VALVE ARRANGEMENT (Sheet 28) r PENE. ORIG. PENE. VALVE PENETRATION PENE. DWG. ISOLATION POST-DIAGRAM SERVICE FSAR LINE INCIDENT NO. PENE. TYPE ARRGT. LOCATION SIZE NUMBER VALVE POSITION INSIDE CTMT. STRUCTURE I OUTSIDE CTMT. STRUCTURE NO. 43A STEAM GENERATOR *11 47 Ill 14 EAST PPING 2" OM-35 CV-4010 CLOSED __J SURF ACE BLDWOOWN PENETRATION ROOM Stt.10f 3 tlftACIATICN ,.... OM-.464 -J isii' I CE --.1 --, ,----..., 'D *)t .,. $ r. -J) .. rr;--f ...... ..____ .... *-n ..____ llz*2'1 '---' ...... GDCRA10R .. 438 STEAM GENERATOR *11 Ill 14 EAST PfPIHG 2" 00>-35 cv-*on CLOSED BOTTOM BLOWDOWN PENETRATION ROOM SH.1 OF 3 OM-454 r -.. 1 I cs.'9 a.O&Ea #.t& CLCSE:a I -¢: .. tg #Ct ,tT-,,,5 m --A -IJ @ "' f v _., * ..,w PIAW mca.t10l *n .. __.MZ ._OWD01'N '--.,.,.. Rl'CKAlUR .,, 22589.DGN fSNl-28 CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT I FIGURE 5-10 I SHEET NO. 28 I REVISION: 42 L _J Revision 42 FIGURE 5-10 CONTAINMENT STRUCTURE -ISOLATION VALVE ARRANGEMENT !Sheet 29 PENE PENE. ,VALVE I PENETRATION I PENE. I DWG. IISOLATIONI POST-T DIAGRAM NO.* I SERVICE I PENE. TYPE ARRGT. LOCATION NUMBER VALVE 1----I-N_s_ID_E_C_T_M_T ___ s_TR_u_c_T_UR-E--,..----0-u_T_S_ID_E_c_TM_T ___ s_T_Ru_c_r_u_R_E__, NO. 44 47A 479 FIRE PROTECTION HYDROGEN SAMPLING N. OF PRI. SHIELDING HYDROGEN SAMPLING W. ELEV. AT 135 FT. 44 Ill 21 I EAST PIPING PENETRATION ROOM 34 WEST PIPING PENETRATION ROOM 6" 1/4" OM-56 I FP-1419 SH.2 Of 6 FP-141A Ol.l-463 SH.1 OF 2 SV-6507A SV*6540A LOCKED SHUT 34 WEST PIPING I 114" I OM-463 I SV-6507E I LOCKED PENETRATION ROOM SH.1 OF 2 SV-6540E SHUT mJ an T L:.J nST CON< CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT FIGURE 5-10 @ "' ,.,,.. CCH< LI L.:J ln..L* ORN TALC DR. SHEET NO. 29 "' T[1.L Ul.E "' ... ... I I HY ORO GEN ANALYZER CABINET 1J222 ... ... ... .... HYDROGEN ANAL VZER CABINET 1J222 REVISION: 36 Revision 36 PENL NO. 47C 47;} SERVICE t-!YOROGE N SAMPLING DOME: EL[ V ,AT 189 FT. HY JR OGEN "1.C'l"URl\J TO CTlv'T. ORIG. PENE. tSAR PENE. TYPE NO. II r I VAL VE PENElRA110N ARR GT. LOCAl ION 34 Wt ST P IP IN G PENETRATION ROOM _ 1 I . I i WEST PIPlt.G IFENEfRATfON ROOM I I PENE<< LINE SIZE I /4" 1/4" DWG. ISOLATION VALVE ISi OM-463 SV-6540F' SH,1 OF' t.' SV-6507F' OM -4s2i I SV-6540G SH.2 OF 2' SV-6507GI POST-INC !DENT POSIT ION LOCKED SHUT LOCKED SHUT LOCKED SHUT LOCKE:D SH!JT c ICIDI DIAGRAM INSIDE ClMT. S'TRUC1UR( OUTSIDE C Tt.AT, STRUC TUR[ TE-I. T,l:_I;: II *;. ,,, ':f -..e--------L'-.1....-t><J .. .. ICIDI r ....... ...,

• I* PS* £07 w TEL.L 'rA..t 1r.101 --@j I* I* FS. PS 505 .. l!YDtiO;r.:'1 CA6{ .. <;:T JJ2Z2 CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT F 5 -10 SHEET NO. 30 PENE. NO. 48A 488 SERVICE CONTAINMENT HYDROGEN PURGE OUTLET CONTAINMENT HYDROGEN PURGE INLET ORIG. FSAR PENE. NO. --PENE. VALVE TYPE ARRGT. II 35 I 1 41 PENETRATION LOCATION EAST PIPING PENETRATION ROOM EAST PIPING PENETRATION ROOM PENE. DWG. ISOLATION POST-LINE INCIDENT SIZE NUMBER VALVE ISi POSITION 4 .. OM-65 MOV-6900 CLOSED SH.2 OF 4 MOV-6901 CLOSED OM -65 HP-104 SH.2 OF 4 MOV-6903 CLOSED INSIDE CTMT. STRUCTURE SfAS CRS CTMT HI RAD CLOSES r---, I I ICIO I CRS AND CTMT -_j HI RAD OVERRICE I ICIO I ICIO .-..1..-, -II M 6900 ... :,,,,. ,... .... f'AI 1: -I -"' I I-HP-,_ ,. ,.,. HP-I 105 .. II.. L!J TEST CONN. I J DIAGRAM OUTS !DE cn.n. STRUCTURE r ---\ ICIO I I ICID CRS AND CTMT L HI RAD OVERRIDE ---I 6901 I .-..1..-, 'II J' I-HP-IOI L!J TEST COllN. ) r ,_ HP-11,. 103 L!.J TEST CONN. M ... 1,,,. ....... F'AI M ... 1 .... .. ., F'AI 1-----ll-llOV 6901 'II J' I* HP-,,j Iii.. 102 L!J TELL TALE 1-----ll*llOV 6903 --CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT I FIGURE 5-10 I SHEET NO. 31 RE v : 13 I DATE : I /9 2 FIGURE 5-10 CONTAINMENT STRUCTURE-ISOLATION VALVE ARRANGEMENT !Sheet 321 9: ORIG. I I I I E E I I I POST-DIAGRAM PENE FSAR PENE. VALVE PENETRATION p N
• DWG. ISOLATION NO. -1 SERVICE I TYPE ARRGT. LOCATION NUMBER VALVE INSIDE CTMT. STRUCTURE OUTSIDE en.ff. STRUCTURE 49A 498 49C 490 HYDROGEN SAMPLE SOUTH OF PRIMARY SHIELD HYDROGEN SAMPLE PRESSURIZER COMPARTMENT HYDROGEN SAMPLE EAST ELEVATION AT 135 FT. SPARE 11 II 11 ll 34 I WEST PIPING 34 34 PENETRATION RM. WEST PIPING PENETRATION RM. WEST PIPING PENETRATION RM. 4 7 I WEST PIPING PENETRATION RM. 114" I DM-463 I SV-65078 I LOCKED SH.1 OF 2 SV-65408 SHUT 114" 1/4 OM-463 SH.1 OF 2 OM-463 SH.1 OF 2 114" I OM-463 SH.1 OF 2 SV-6507C SV-6540C SV-6507D SV-6540D LOCKED SHUT LOCKED SHUT ,. -.L -., "><}
• o/. .. PS* ... .. ,.._ ll07 L!J T[5' CONN. r-L-, (t°SV .... .r.&b ,.,. c CA ..... c .. llD2 iT1 .... .. ,.._ .... L!J TCST CCH<. I .. ,.._ ""' L!J TCST CONN. ....,, ,.._ ... @ .. :&; L!J T[5' c:oNH. (:f TELL r: 'TAI.[ ... "' HO 1-JA. 1-74>. L!J TC5' CONN. .. ... .. ID2 "' TCST co**<. . I I I -, ... .. ... .. , .. PS* " i KTDftOGCN I NUol..Y2ER c..cr ...... .. PS* .. PS..,. .. PS* CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT I FIGURE 5-10 I SHEET NO. 32 I REVISION: 36 Revision 36 PENE. NO. 50 59 60 61 SERVICE ILRT PENETRATION REFUELlNG POOL RECYCLE INLET STEAM TO REACTOR HEAD WASHDOWN AREA REFUEL[NG POOL OUTLET TO FUEL POOL COOL[NG PIPE ORIG. FSAR PENE. NO. ---------46 PENE. VALVE TYPE ARRGT. IV 46 IV 30 IV 42 IV 27 PENETRATION LOCAT[ON EAST PIPING PENETRATION RM. WEST PIP[NG PENETRATION RM. WEST PIPING PENETRAT[ON RM. WEST PIPING PENETRATION RM. PENE. DWG. LINE SIZE NUMBER 6" OM-065 SH0002 8" OM-058 I* OM-077 SH0003 8" OM-058 ISOLATION VALVE CS) BLIND FLANGE BLIND FLANGE SFP-171 SFP-170 ES-142 ES-144 SFP-189 SFP-172 SFP-174 SFP-176 POST-I NC WENT POSITION CLOSED CLOSED LOCKED SHUT LOCKED SHUT LOCKED SHUT LOCKED SHUT LOCKED SHUT LOCKED SHUT LOCKED SHUT LOCKED SHUT DIAGRAM INSIDE CTMT. STRUCTURE 11 O*SFP*l 70 TO REFUELING POOL -N0.11 ..... 0*$FP*23' , ORN. 1 *ES*1'3 1 *ES*I UNIT N0.1 J
• 4 .._ -CNTMT VSL HD .... If ASHDOWN AREA r ORN. r.i O*SFP-233 FROM SOUTH CAVITY O*SFP*l 72 O*SFP*l 7.C ... ..... ........ ,NORMALLY OPEN ORN. ' , , O*SFP*l 77 j. ___ O*SFP-176 OUTSIDE )) II "'"'"' L..:..I TEST CONN* TEST CONN. n * ... _... }) ..... .... O*SFP*t71 II 1 *ES*1'2 H 1 *ES*i4'5 n L.:.J TEST CONN. TEST CONN. r.1 --, O*SFP*219 -.... CTMT. STRUCTURE TELL TALE n H O*SFP*220 H ......... ..... "" O*SFP*227 1 *ES-267 ... --.... , ' 1 *ES*200S . L.J TEST CONN. TEI.I.

TALE I.I A O*SFP-212 \\ <*> ......... )} CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT I FIGURE 5-10j SHEET NOo33 I RE v Is I ON : 21 NO. 72 77 78 I I i SERVICE CONTAINMENT PRESSURE \t ON ITOR INSTR. CONTAINYE:l'H ?RES SUR'='. ITCR PRESSURE: \ION lTOR INSTR. ORIG. FSAR PENE. PENE. TYPE NO, INSTRU. INS TRU. INS TAU. VALVE ARR GT, 33 "'31 33 PE NE TR AT ION LOCATION EAST ELECTRICAL PE NE TRA TlON R>.4. 11 EAST ELECTRICAL PENETRATION R'A. I I we: 5 r E:LEC TR IC AL ION RM. PENE. LINS:: 5 llE: 3/4 .. 3/4" 3/4 .. OWG. !SOL AT ION NUtJBER VALVEISI OM*65 SH.2 OF' 4 OM*65 !sl-1.2 OF 4 I I I I I l OM-65 ISH.2 OF" 41 I I i i I SV-5313A SV-53130 POST* INC ICENT POSITION OPEN OPEN INS IDE CHAT. STRUCTURE ' ICIO **y ___ J -DIAGRAM l OUTSIDE CTMr. STRIJC Tl.JAE -) 83 i"RESSURE \tON ITOR INSTR. INS TRU. 40 WEST ELECTRICAL FENE TRAT ION R\4, J/4" OM-53 SV-5313C OPEN *SH.Z OF ft*SY ,9a .SJ*JC y < I '-tff!J-* ZS /\ tJ SJIJ T -.'I/. CA I c Uo----J I 1-------------'--I .J f ca ! I I C 0 N TA I N ME N T S TR UC TUR E l S 0 L AT l 0 N VALVE ARR AN GE ME N T I F IGLJ'.RE 5 -10 l*Pf I-Pr SJ14C :lJ13C I:*CP1*7 LI SHEE f NL) .35J l -- PENE. ORIG. PENE. VALVE PENETRATION PENE. DWG. [SOLA Tl ON POST-DIAGRAM SERVlCE FSAR LINE [NCI DENT NO. PENE. TYPE ARRGT. LOCATION SlZE NUMBER VALVECSl POSI TlON [NS IDE CTMT. STRUCTURE OUTSIDE CTMT. STRUCTURE NO. ILRT 12" VENT 84 ---IV EAST ELECTR[CAL 12"' OM-065 BLIND FLG. CLOSED PENETRATION PENETRAT[ON ROOM SH0002 §J>f, 'NB-68 BLIND FLG. CLOSED / ) l-...1: / '\. 12**HS*7l 1: : l llJ Ill 114-::> I.I. I.I. 0 a: TEST < CONN.L:,I """' NOTE i VALVE ANO AIR DIFFUSER 81..0WOOWN TO TO BE INSTALLED BEFORE AUXILIARY BLOG. PERFORMING ILRT. . -----.---. '"-----*-------**-*-_, FSAR-36 CONTAINMENT STRUCTURE ISOLATION VALVE ARRANGEMENT I FIGURE 5-1 o I SHEET NOo36 IREV 0 31IDATE:10101 74853.DGN 0

• * ,.1 _____ ' ____ 'l __ l ' ' r-----------PRIMARY PIPE Tc+V STEAM 6ENERATOR ,---,' \ +--! i) ---------1* -J--PRIMARY PIPc FROM PVMP ( ', -----r'\ !: i ! -x ! : ij____ . : ; " . -------' \ --------------I ; i I ( VERTICAL SVPPOkT CWOVTffT ""'""! I \ J // FIG. 5-JI -REACTOR VESSEL SJ.ic:>e'f I 2 YERTICAl St/PPOPT 01/ c/NLET N(.;ZZL£.S
• B-8 * \ -\ \ t Rt="4CTOR I " " NOZZLE ; I \ ' I . ._ ""* I . ' / : I =+-/ / \ -------. ---/ ; "' . ___ _,,r SC//:PORT-------____ __, : .* / i 1;/;t .. 3"/i .. ,L -* :_, _-.. -UJ **--* I . "*-;jlt .. . er i'-o-9" 1'-3" -"':'"*-... -.]"...-* :,'-' 1'-a-g* 3'-8" FIG.5-11 2. cf 2. !!J'-8* A-A REACTOR VESSEL SUPPORTS
• I ' . A, . ..:_ L __ l l'J ' \<, .. c ..... cc*c*<JJ ' l\I .;, Aj
• .... ;;, ::t " . * (\, \ll
• FIG. 5-12 -STEAM GENERATOR She.,,{ lcfJ

I 4'-10"' I

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• 5.5 ** WAS llUMPED THROUGH ALUM. PIPE. SECTION A-A 4 I l I GAS & ELECTRIC CO. BASE SLAB. CONTAINMENT STRUCTURE UNIT NO. 1 FIGURE 5-19 Calvert Cliffs Nuclear Power Pl APPENDIX 5A STRUCTURAL DESIGN BASIS TABLE OF CONTENTS PAGE5A.0 STRUCTURAL DESIGN BASIS 5A.1 GENERAL 5A.2 CLASSES OF STRUCTURES, SYSTEMS, AND EQUIPMENT 5A.3 DESIGN BASES APPENDIX 5A STRUCTURAL DESIGN BASIS TABLE OF CONTENTS PAGE 5A.4 LOADINGS COMMON TO ALL STRUCTURES 5A.5 FLOODING 5A.6 MASONRY WALL DESIGN APPENDIX 5A STRUCTURAL DESIGN BASIS LIST OF TABLES TITLEPAGE APPENDIX 5A STRUCTURAL DESIGN BASIS LIST OF FIGURES FIGURE APPENDIX 5A STRUCTURAL DESIGN BASIS LIST OF ACRONYMS CALVERT CLIFFS UFSAR 5A.1-1 Rev. 47 APPENDIX 5A STRUCTURAL DESIGN BASIS 5A.05A.1 GENERAL The design basis for structures for normal operating conditions are governed by the applicable building design codes. The design bases for specific systems and equipment are stated in the appropriate Updated Final Safety Analysis Report (UFSAR) section. The design basis for the maximum loss-of-coolant accident (LOCA) and seismic conditions is that there be no loss of function if that function is related to public safety. The method used to determine the seismic response resulting from the Operating Basis (OBE) and Safe Shutdown Earthquake (SSE) is described in Sections 2.6.5 and 5.1.3.2.b. 5A.1.1 RESPONSIBLE DESIGN ORGANIZATIONS Calvert Cliffs Nuclear Power Plant (CCNPP), as the applicant, has the ultimate responsibility for the design and construction of Calvert Cliffs Nuclear Power Plant, Units 1 and 2. Calvert Cliffs Nuclear Power Plant utilizes its experienced staff to perform project management, engineering review, construction coordination and quality assurance functions.

Combustion Engineering, Inc. (CE), as the Nuclear Steam Supply System (NSSS) supplier, has supplied the components of the Reactor Coolant System, the Chemical and Volume Control System and the Safety Injection System. Babcock & Wilcox, Canada is the supplier of the replacement steam generators. As the suppliers, they are responsible for the seismic design of their components in a manner consistent with the design criteria for the project. Bechtel Associates (Bechtel), as the Architect-Engineer for CCNPP, was responsible for developing the seismic criteria and the design of Category I structures and for the approval of all other Category I equipment. Once the seismic criteria for the project were established by Bechtel with the assistance of Dames and Moore, soil consultants, Bechtel ensured that these criteria were implemented in the design of Category I structures. Bechtel Associates also developed response spectra curves and all other requirements necessary for the design of all Category I equipment including the NSSS. This information is given to all suppliers of Category I equipment including CE who has implemented these seismic criteria in their design. All interdisciplinary exchanges, between Bechtel, CE, CCNPP and any vendor supplying Category I equipment, were documented with memoranda or conference notes. Exchanges of letters, specifications and drawings in accordance with defined procedures are used to maintain uniform design throughout the plant. In 2001 Stevenson and Associates developed in-structure acceleration time-histories and seismic response spectra to be utilized for the design and evaluation of Category I equipment within the Containment, including the NSSS. 5A.1.1.1 Design Control Design control is effected by successive levels of review of seismic criteria, calculations, and seismic sections of specifications and drawings. At Bechtel, these levels are the Responsible Engineer, Group Supervisor, and Chief Engineer with final approval by the Project Engineer. These reviews also cover seismic requirements placed on suppliers, such as CE, where reviews are performed by the Specialty Group and Project Manager. CALVERT CLIFFS UFSAR 5A.1-2 Rev. 47 5A.1.2 SPECIFIC REQUIREMENTS FOR SAFETY-RELATED PURCHASES Both the NSSS supplier and the Architect Engineer included requirements for seismic design in specifications for Category I equipment. Combustion Engineering, Inc. is required to design the NSSS to withstand the load imposed by the maximum hypothetical accident, and by the maximum seismic disturbance without loss of functions required for reactor shutdown and emergency core cooling.

Definitions in typical seismic specifications for Category I equipment: a. The OBE: has a maximum horizontal ground acceleration of 0.08 g and a maximum vertical ground acceleration of 0.053 g, acting simultaneously. b. The SSE: has a maximum horizontal ground acceleration of 0.15 g and a maximum vertical ground acceleration of 0.10 g, acting simultaneously. These seismic acceleration levels were established to provide an appropriate margin of safety for withstanding stresses greater than those recorded and reflect uncertainties about the historical data and their suitability for design basis. c. All Category I systems, equipment and components shall be designed to withstand the appropriate seismic load combined with other applicable loads without loss of function. The analysis of the dynamic loads on Category I systems is accomplished by using the Response-spectrum Method as outlined in Bechtel's seismic specification. All vendors supplying Category I equipment or systems, are required to submit copies of their dynamic analyses or dynamic test results, based on seismic criteria, for approval.

CALVERT CLIFFS UFSAR 5A.4-1 Rev. 47 5A.4 LOADINGS COMMON TO ALL STRUCTURES Ice or Snow Loading - a uniformly distributed live load of 30 psf on all roofs provides for any anticipated snow and/or ice loading. Temperature - The plant is designed for a temperature range of 20°F to 90°F. CALVERT CLIFFS UFSAR 5A.5-1 Rev. 47 5A.5 FLOODING Finished grade of the plant site is at Elevation 45', therefore no loads due to floods or inundation are considered. The saltwater cooling system pump motors, located in the Intake Structure, are protected against the maximum hypothetical hurricane tide and storm surges including wave action. Maximum design wave runup is 28.5' above mean sea level.

The roof and roof hatches of the intake structure are designed for live load (250 psf), dead load (150 psf), tornado uplift (100 psf), Probably Maximum Hurricane waves (250 psf) and seismic load of 10% of the dead load acting downwards.

For all major structures below finish grades, a heavy waterproofing membrane of 40 mils thickness is provided at the exposed face of the exterior walls and below the base slab. Rubber waterstops are also provided at all construction joints up to grade elevation. Subsurface drains are provided to lower the elevation of ground water around the plant. All of these provisions are made to eliminate any possibility of flooding, by ground water infiltration, of equipment located below the elevation of highest flood water level.

CALVERT CLIFFS UFSAR 5A.6-1 Rev. 47 5A.6 MASONRY WALL DESIGN NRC Bulletin 80-11 required licensees to identify plant masonry walls and their intended functions. Licensees were also required to present reevaluation criteria for the masonry walls with analyses to justify those criteria. If modifications were proposed, licensees were to state the methods and schedules for the modifications.

In response to the bulletin, BGE provided the NRC with a description of the status of masonry walls at Calvert Cliffs Nuclear Power Plant. A total of 147 safety-related walls were initially identified. All walls subject to reevaluation were in the Auxiliary Building. The masonry construction at Calvert Cliffs Units 1 and 2 consists of single- and double-wythe walls of the running bond type whose functions include partition, shielding, blockout, bearing, and filler. Both reinforced and unreinforced walls were built in the plant. Vertical reinforcement was provided by grouting reinforcing bars in vertical cells and "Dur-o-Wall" was used for horizontal reinforcement. The masonry walls were reevaluated using the following criteria: a. The design allowables are based on ACI 531-79. b. The working stress design method and the energy balance technique were used in the analysis. Out of 147 safety-related walls, 22 were qualified by the energy balance technique. Based on a subsequent review, four of the 22 walls were reclassified as non-safety-related walls because failure of these walls would not have any impact on safety-related equipment. c. Loads and load combinations were consistent with the other parts of this appendix. d. Critical damping values of 4% for OBE and 7% for SSE were used for vertically reinforced walls which were assumed to crack under seismic conditions. A damping value of 2% was used for walls that were assumed not to crack. e. The typical analytical procedure is summarized below: 1. Determine wall boundary conditions 2. Using a one-way beam model and the floor response spectrum, determine the responses of the first three modes and combine them by the square root of the sum of the squares method. 3. Compare computed stresses with the allowable values in ACI 531-79. Because arching action had been used to qualify one of the original 147 walls, it was modified to bring it within elastic requirements. All other walls satisfied the reevaluation criteria and no other modifications were proposed.

All but two of the walls were qualified by the elastic criteria (consistent with NRC acceptance criteria) when the existing conservatism in the masonry wall analysis was accounted for. The remaining two walls were also qualified by the elastic criteria using a "plate" analysis approach rather than "beam" analysis approach. The NRC concluded there is reasonable assurance that all safety-related masonry walls will withstand the specified design load conditions without impairment of either wall integrity or the performance of the required function.

Rev.05A-1 RESPONSE SPECTRA OPERATING BASIS EARTHQUAKE Rev.05A-2 RESPONSE SPECTRA OPERATING BASIS EARTHQUAKE Rev.05A-3 RESPONSE SPECTRA OPERATING BASIS EARTHQUAKE Rev.05A-4 RESPONSE SPECTRA OPERATING BASIS EARTHQUAKE CALVERT CLIFFS UFSAR 5B-1 Rev. 47 APPENDIX 5B QUALITY CONTROLS 5BAppendix 5B, Quality Controls, is historical information about the construction phase of the plant. Appendix 5B has been removed from the Safety Analysis Report and has been sent to Plant History. An image of the contents of the appendix can be accessed in the NORMs Records System under Document ID (DOC ID) "Appendix-5B."

CALVERT CLIFFS UFSAR 5C-1 Rev. 47 APPENDIX 5C 5C STUDY OF THE EFFECTS OF MISLOCATED VERTICAL TENDONS Appendix 5C, Mislocated Tendon Study, is historical information about the construction phase of the plant. Appendix 5C has been removed from the Safety Analysis Report and has been sent to Plant History. An image of the contents of the appendix can be accessed in the NORMs Records System under Document ID (DOC ID) "Appendix-5C."

CALVERT CLIFFS UFSAR LEP-5D-1 Rev. 47 APPENDIX 5D 5D STUDY OF UPPER VERTICAL TENDON BEARING PLATES Appendix 5D, Study of Upper Vertical Tendon Bearing Plates, is historical information about the construction phase of the plant. Appendix 5D has been removed from the Safety Analysis Report and has been sent to Plant History. An image of the contents of the appendix can be accessed in the NORMs Records System under Document ID (DOC ID) "Appendix-5D."

CALVERT CLIFFS UFSAR 5E.1-1 Rev. 47 APPENDIX 5E REDUCTION IN CONTAINMENT PRESTRESS AND LONG-TERM CORRECTIVE 5E.0ACTIONS FOR VERTICAL TENDON CORROSION 5E.1 REDUCTION IN CONTAINMENT PRESTRESS 5E.1.1 DESIGN BASIS The design basis for the Containment Structure is described in Section 5.1.1. 5E.1.2 DESIGN CRITERIA FOR PRESTRESS In the concept of a post-tensioned Containment Structure, the internal pressure load is balanced by the application of an opposing external force on the structure. Sufficient post-tensioning was applied to the containment cylinder and dome to more than balance the internal pressure. Therefore, a margin of external pressure exists beyond that required to resist the design basis loss-of-coolant accident pressure. Nominal, bonded reinforcing steel was also provided to distribute strains due to shrinkage and temperature. Additional bonded reinforcing steel was used at penetrations and discontinuities to resist local moments and shears.

The internal pressure loads on the foundation slab are resisted by both the external bearing pressure due to dead load and the strength of the reinforced concrete slab. Thus, post-tensioning was not required to exert an external pressure for this portion of the structure. The post-tensioning system is described in Sections 5.1.2.1, 5.1.4.2, and 5.5.1. Design load combinations are provided in Sections 5.1.2.2 and 5A.3.1.8.

5E.1.3 DESIGN REANALYSIS As a result of some hoop tendon lift-off values being lower than expected during the third-year tendon surveillance, as required per Reference 1 at the time, a reanalysis of the Containment was performed between 1977 and 1979 to reduce the minimum required prestress. The third year lift-off force values for hoop tendons indicated that there was sufficient post-tensioning to meet design requirements, but that the losses appeared to be more accelerated than originally expected. The results of the reanalysis were used to develop minimum tendon force requirements for continual tendon surveillance.

The basic approach in the reanalysis was to take advantage of conservatism existing in the initial prestressing system and the conventional reinforcing with respect to the allowable stresses provided in Table 5-1. The reanalysis was done to more accurately reflect the expected results of future surveillances without reducing the original intended margins of the design. The strict requirements of the reanalysis were to assure that all the original design criteria was maintained. The basic approach of the reanalysis was to reduce the prestress on all three major tendon groups by a uniform percentage of 9%. The original containment vertical prestress level of 1.32P would be reduced to 1.2P. Specific load cases, as identified in Sections 5.1.2.2 and 5A.3.1.8 were reanalyzed and those components of the Containment Structure potentially affected by the reduced prestress were reevaluated. The approach was satisfactory for both the hoop and dome tendon groups with all the design criteria satisfactorily met or exceeded. However, the reinforcing steel in the shell/base slab interface was slightly overstressed for the structural integrity test (Containment) working stress design load case. To overcome this localized overstress, the vertical tendon group prestressing level was only reduced to 1.29P. Therefore, all the original design criteria, as outlined in Sections 5.1.2 and 5A.3.1, were completely satisfied, assuring sufficient prestress to be available at the end of the nominal 60-year design life. Those components CALVERT CLIFFS UFSAR 5E.1-2 Rev. 47 of the Containment Structure not affected by a reduced prestress level were not reevaluated. Since the only change made in the 1977-1979 reanalysis was to reduce the prestress level, the reanalysis concentrates on the portion of the analysis/design that was affected by prestress. Otherwise, the original analysis calculations and results remain valid.

The 1977-1979 reanalysis was performed in two parts. One part addresses the base/shell haunch issue and used separate models for the base slab, and the haunch that were analyzed manually using classical plate and shell theory and finite difference methods. Compatibility relationships were used to establish continuity at the slab/shell boundary.

The second part used a finite element analysis (FINEL CE-316) to model the Containment and obtain force and moments as output. Seismic loads were recomputed using a finite element model axisymmetric shells and solids (CE-771), which provided a more exact distribution of seismic loads, compared to those provided in the original seismic analysis. Results of the analysis were post-processed to convert the force and moments on the concrete and reinforcing steel into stresses. 5E.1.3.1 Summary of Calvert Cliffs Containment Reanalysis The Calvert Cliffs Containment was reanalyzed to check the effect of reduced prestressing forces. The following table illustrates the original design and reanalysis prestressing forces. Containment Original Prestress ForcesContainment Reanalysis Prestress Forces Reduction HOOP 630 K/Ft 573.16 K/Ft 9% VERTICAL 300 K/Ft 294.2 K/Ft 2% DOME 360 K/Ft 327.52 K/Ft 9% The reanalysis consisted of an elastic finite element analysis of the upper containment shell and dome. The lower part of the shell and base slabs were analyzed using "cracked" section properties in order to incorporate the proper redistribution of stresses. The reanalysis considered applicable dead, thermal, and pressure loadings in addition to the revised prestress forces. The results of the original seismic analysis were used in the reanalysis, as described above. The stresses derived from the reanalysis were checked against the Table 5-1 set of allowables. There is no significant overstressing in either the concrete or the reinforcing steel. The stresses checked include those in the meridional, hoop, and radial directions. The containment stresses from the reanalysis are tabulated in Table 5E-1. The location key, allowable stresses, and general notes are provided in Table 5-1. 5E.

1.4 REFERENCES

1. Nuclear Regulatory Commission Regulatory Guide 1.35, Revision 2, Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containment Structures CALVERT CLIFFS UFSAR 5E.1-3 Rev. 47 TABLE 5E-1 STRESS ANALYSIS RESULTS CONTAINMENT STRUCTURE - SUMMARY OF CONCRETE AND REINFORCING STEEL STRESSES REINFORCING STEEL COMPUTED (psi) COMPUTED vs. ALLOWABLE SECTION LOAD CASE m h II D+F+L+1.15P C C --- --- III D+F+L+TO+E C 7900 --- 0.26 A-B IV D+F+L+TA+P 10200 9900 0.34 0.33 V 1.05D+F+1.5P+TA 3300 7000 0.06 0.13 VI 1.05D+F+1.25P+TA+1.25E 2900 6900 0.05 0.13 VII D+F+P+TA+E' 2200 7500 0.04 0.14 II D+F+L+1.15P C C --- --- III D+F+L+TO+E 8800 15900 0.29 0.53 C-D IV D+F+L+TA+P 13900 20800 0.46 0.69 V 1.05D+F+1.5P+TA C 21500 --- 0.40 VI 1.05D+F+1.25P+TA+1.25E 8100 21200 0.15 0.39 VII D+F+P+TA+E' 17800 20800 0.33 0.39 II D+F+L+1.15P C C --- --- III D+F+L+TO+E 11200 12000 0.37 0.40 E-F IV D+F+L+TA+P C 5600 --- 0.19 V 1.05D+F+1.5P+TA C 2000 --- 0.04 VI 1.05D+F+1.25P+TA+1.25E C 3100 --- 0.06 VII D+F+P+TA+E' C 5200 --- 0.10 II D+F+L+1.15P C C --- --- III D+F+L+TO+E 14900 12100 0.50 0.40 G-H IV D+F+L+TA+P 27800 22400 0.93 0.75 V 1.05D+F+1.5P+TA 35900 30100 0.66 0.56 VI 1.05D+F+1.25P+TA+1.25E 29200 27800 0.54 0.51 VII D+F+P+TA+E' 24700 24500 0.46 0.45 II D+F+L+1.15P C C --- --- III D+F+L+TO+E 8500 10300 0.28 0.34 J-K IV D+F+L+TA+P 24200 26800 0.81 0.89 V 1.05D+F+1.5P+TA 36400 55600 0.67 1.03 VI 1.05D+F+1.25P+TA+1.25E 31400 36600 0.58 0.68 VII D+F+P+TA+E' 25900 27500 0.48 0.51 CALVERT CLIFFS UFSAR 5E.1-4 Rev. 47 TABLE 5E-1 STRESS ANALYSIS RESULTS CONTAINMENT STRUCTURE - SUMMARY OF CONCRETE AND REINFORCING STEEL STRESSES REINFORCING STEEL COMPUTED (psi) COMPUTED vs. ALLOWABLE SECTION LOAD CASE m h II D+F+L+1.15P 9300 C 0.31 --- III D+F+L+TO+E 15700 27900 0.52 0.93 L-M IV D+F+L+TA+P C C --- --- V 1.05D+F+1.5P+TA 16000 C 0.29 --- VI 1.05D+F+1.25P+TA+1.25E 400 C 0.01 --- VII D+F+P+TA+E' C C --- --- II D+F+L+1.15P 20800 25500 0.69 0.85 III D+F+L+TO+E 9500 C 0.32 --- N-O IV D+F+L+TA+P 18400 24200 0.61 0.81 V 1.05D+F+1.5P+TA 39300 44300 0.73 0.82 VI 1.05D+F+1.25P+TA+1.25E 25900 40500 0.48 0.75 VII D+F+P+TA+E' 20800 32500 0.39 0.60 II D+F+L+1.15P 19800 25600 0.66 0.85 III D+F+L+TO+E 8100 17200 0.27 0.57 P-Q IV D+F+L+TA+P 24400 24300 0.81 0.81 V 1.05D+F+1.5P+TA 37600 41600 0.70 0.77 VI 1.05D+F+1.25P+TA+1.25E 39900 46300 0.74 0.86 VII D+F+P+TA+E' 32900 39200 0.61 0.73 CALVERT CLIFFS UFSAR 5E.1-5 Rev. 47 TABLE 5E-1 STRESS ANALYSIS RESULTS CONTAINMENT STRUCTURE - SUMMARY OF CONCRETE AND REINFORCING STEEL STRESSES CONCRETE COMPUTED (psi) COMPUTED vs. ALLOWABLE SECTION LOAD CASE em eh am ah II D+F+L+1.15P -890-790-840 -790120.300.560.03 III D+F+L+TO+E -2630-2810-1520 -1460240.941.010.07A-B IV D+F+L+TA+P -2770-2670-930 -880270.920.620.07 V 1.05D+F+1.5P+TA -2830-2590-640 -540240.630.150.06 VI 1.05D+F+1.25P+TA+1.25E -3140-3140-780 -730210.700.180.06 VII D+F+P+TA+E' -3360-3470-930 -880180.770.220.05 II D+F+L+1.15P -480-300-430 -2401670.160.290.39 III D+F+L+TO+E -2660-1480-840 -3101310.890.560.31C-D IV D+F+L+TA+P -2190-2410-480 -2501410.800.320.33 V 1.05D+F+1.5P+TA -2010-2040-310 -210990.450.070.23 VI 1.05D+F+1.25P+TA+1.25E -2670-2200-400 -230940.590.090.22 VII D+F+P+TA+E' -3270-2410-480 -250940.730.110.22 II D+F+L+1.15P -590-290-310 -2701160.200.210.39 III D+F+L+TO+E -1130-1270-480 -3101340.420.320.44E-F IV D+F+L+TA+P -2110-2890-330 -2701280.960.220.42 V 1.05D+F+1.5P+TA -1870-2790-260 -260880.620.060.30 VI 1.05D+F+1.25P+TA+1.25E -1810-2820-300 -260970.630.070.32 VII D+F+P+TA+E' -2110-2870-340 -270980.640.080.32 II D+F+L+1.15P -160-260-130 -240300.090.160.18 III D+F+L+TO+E -2170-1870-640 -6101230.720.420.74G-H IV D+F+L+TA+P -1570-1940-240 -300400.650.200.24 V 1.05D+F+1.5P+TA T-530-40 -120130.120.030.08 VI 1.05D+F+1.25P+TA+1.25E -630-930-140 -170190.210.040.11 VII D+F+P+TA+E' -1430-1540-240 -250350.340.060.21 II D+F+L+1.15P -320-110-220 -90210.110.150.15 III D+F+L+TO+E -1860-2390-710 -1060400.800.710.29J-K IV D+F+L+TA+P -2240-1140-330 -210970.750.220.70 V 1.05D+F+1.5P+TA -740T-140 T1020.160.030.73 VI 1.05D+F+1.25P+TA+1.25E -1300T-210 T1080.290.050.78 VII D+F+P+TA+E' -1990-1040-300 -200970.440.070.70 CALVERT CLIFFS UFSAR 5E.1-6 Rev. 47 TABLE 5E-1 STRESS ANALYSIS RESULTS CONTAINMENT STRUCTURE - SUMMARY OF CONCRETE AND REINFORCING STEEL STRESSES CONCRETE COMPUTED (psi) COMPUTED vs. ALLOWABLE SECTION LOAD CASE em eh am ah II D+F+L+1.15P -1100-900-250 -7301290.370.490.47 III D+F+L+TO+E -2790-260-680 T800.930.450.44L-M IV D+F+L+TA+P -800-2410-310 -6501440.800.430.45 V 1.05D+F+1.5P+TA -1000-3350-130 -17001610.740.400.34 VI 1.05D+F+1.25P+TA+1.25E 3050-220 -9602580.680.230.54 VII D+F+P+TA+E' -800-2600-310 -8102120.580.190.44 II D+F+L+1.15P T-540T T720.30(a) 0.46 III D+F+L+TO+E -480-20-180 -150270.27(a)0.17N-O IV D+F+L+TA+P T-640-50 -301130.36(a)0.72 V 1.05D+F+1.5P+TA -180-1440-50 -201670.40(a)0.70 VI 1.05D+F+1.25P+TA+1.25E T-430-10 T1700.12(a)0.72 VII D+F+P+TA+E' T-780-40 -201500.22(a)0.63 II D+F+L+1.15P -540-660T T40.37(a) 0.07 III D+F+L+TO+E -590-990-270 -100470.55(a)0.83P-Q IV D+F+L+TA+P -950-980-170 T170.54(a)0.30 V 1.05D+F+1.5P+TA -1850-1810-170 T110.51(a)0.13 VI 1.05D+F+1.25P+TA+1.25E -850-1540-110 T480.43(a)0.56 VII D+F+P+TA+E' -980-1840-130 T580.51(a)0.68

_______________________ (a) fa = 0.3

CALVERT CLIFFS UFSAR 5E.2-1 Rev. 47 5E.2 LONG-TERM CORRECTIVE ACTIONS FOR VERTICAL TENDON CORROSION 5E.2.1 DISCOVERY OF CORROSION During performance of the 20-year (1997) Technical Specification and American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Reference 1) tendon surveillance on Unit 1, conditions that did not meet the acceptance standards were found on the Containment Structure. Conditions that did not meet the acceptance standards were found in all three containment tendon populations, i.e., hoop, dome, and vertical tendons. The abnormal conditions found on the hoop and dome tendons were considered minor enough that the acceptability of the concrete containment was not affected. The unacceptable conditions were found in the vertical tendon population. Several of the vertical tendons selected for surveillance were found to contain broken and corroded wires at their top ends, below the stressing washer. The discovery of broken wires in these tendons initiated an expansion of the Unit 1 vertical tendon inspection scope to perform visual inspections and lift-off testing on all Unit 1 vertical tendons. Subsequently, broken and corroded wires were found throughout the Unit 1 vertical tendon population at the top ends of the tendons. Following completion of the Unit 1 surveillance, the 20-year surveillance of the Unit 2 tendons was conducted. Although Unit 2 was only required to perform visual inspections, it was decided to also perform lift-off testing of all the vertical tendons in order to facilitate inspection of the tendon wires in the region of concern below the upper (top) stressing washer. Abnormal conditions very similar to Unit 1 were found on the Unit 2 vertical tendons. Precision Surveillance Corporation (PSC), performed all the inspection work. Precision Surveillance Corporation wrote a non-conformance report for every abnormally degraded condition that did not meet the acceptance standards of IWL-3220 and Reference 2. The original tendon post-tensioning system is described in Sections 5.1.2 and 5.1.4. 5E.2.2 CAUSE OF VERTICAL TENDON CORROSION During the 20-year tendon surveillance on the Unit 1 and Unit 2 Containments, corrosion and broken wires were discovered on some vertical tendons. Reference 3 and Reference 4 evaluated the findings and determined the cause of the corrosion and wire breaks. The evaluation concluded that the tendon wire failures and corrosion problems resulted from a combination of water and moist air intrusion into the vertical tendon end caps (grease cans), and inadequate initial grease coverage of wires in the area just under the top stressing washer. To address issues identified in the evaluation, short-term corrective actions were taken. These actions included spraying hot grease under the top stressing washer, reorienting the stressing shims to leave a gap between the shims to allow a vent path to help eliminate voids, re-greasing non-corroded vertical tendons, and resealing around the original tendon can all-thread penetrations with caulking. Additional inspections were performed in 1999 and 2000 to verify the assumptions that were considered in evaluation and to provide additional data to help develop a long-term corrective action plan. 5E.2.3 LONG-TERM CORRECTIVE ACTIONS The goal of the long-term corrective action plan is to ensure that the Containments meet their design basis requirements until plant end-of-life. As one part of the long-term corrective action plan, all the vertical tendons have been re-greased with new corrosion inhibiting grease (Visconorust 2090-P4). The non-corroded vertical tendons were re-greased in 2000, and the tendons with less severe corrosion were re-greased during 2001. The remaining vertical tendon population was replaced in 2001 and 2002, and had new grease put in place at that time. In addition, all of the vertical tendons had a redesigned pressure-tight, grease-filled cap installed at the upper-bearing plate to prevent water intrusion. The bottom grease cap for every vertical tendon was also replaced with a new redesigned pressure-tight grease cap. The redesigned grease cap has a flange that CALVERT CLIFFS UFSAR 5E.2-2 Rev. 47 is attached by studs and nuts to the tendon bearing plates utilizing existing taps in the plates. As mentioned in the above paragraph, another part of the long-term corrective action plan involved the replacement of a portion of the corroded vertical tendon population. Preliminary evaluations using a wire breakage predicting model had shown that without vertical tendon replacement, neither Containment would meet its design basis at plant end-of-life due to prestress loss from predicted future wire breakage. To determine which vertical tendons were to be replaced, a final vertical tendon future wire breakage prediction model and selection criteria were developed. The wire breakage prediction model is described below first, followed by the vertical tendon replacement selection criteria. 5E.2.3.1 Future Wire Breakage Prediction Model (a.k.a. Weibull Model) As discussed previously, corrosion (abnormal degradation) at some of the top ends of the Units 1 and 2 containment vertical tendons was found during the 20-year 1997 tendon surveillance. To determine the acceptability of the Containments without repairing the vertical tendon wires, a model was developed to predict how the degradation of the tendon wires would affect the wires over the plant life. Wire degradation is predicted to lead to additional future wire breakage, as was found during the 1997 inspections.

Reference 5 was used to determine the acceptability of the Containments without repairing or replacing the degraded vertical tendons. The objective of the report was to develop models for the future failures of tendon wires. The models could then be used to assess how long the vertical tendons would continue to meet structural integrity requirements, with the assumption that the short-term corrective actions and long-term corrective actions were not fully effective in stopping further corrosion degradation. The models were developed using a Unit 2 timeline. The Unit 1 tendons have been tensioned longer than those of Unit 2. However, the inspections of 1997, 1999, and 2000 indicated that the conditions of the tendons for the two units were comparable at the time. Therefore, the models are applicable and conservative for either unit on a calendar time basis, without adjustment for the longer service time under tension of the Unit 1 tendons. The models described in the report use the additional data derived from the 1999 and 2000 inspections to extend the postulated period of validity to the plant end-of-life. The model for ductile/general corrosion wire failures also uses the observed wire failures during lift-off tests to develop an additional data point for failures from that degradation mode. The report presents separate models for the two mechanisms of degradation: (1) hydrogen-induced cracking; and (2) ductile/general corrosion failures. The predictions for the two degradation mechanisms are then combined to obtain the total predicted wire failures. Because of the conservative assumptions used to develop the combined model, it is anticipated that the model predictions will be bounding for observed behavior for the remaining extended operating license of the Units. Although the model developed in the report is expected to be a conservative upper-bound estimate of what will actually occur, it is based on minimal data and plausible assumptions. The conservatism of the model will be validated by future enhanced tendon inspections. The report shows that, conservatively, 2,714 wires could break on each Unit due to abnormal wire degradation by the end of the plant operating licenses. The report shows the spread of the number of future wire breaks in individual tendons throughout the tendon population. This range is from a predicted maximum of 86 wire breaks in one tendon, to a minimum of 1 wire break in 48 different tendons. CALVERT CLIFFS UFSAR 5E.2-3 Rev. 47 In order to apply the statistical model predictions to all of the original vertical tendons for each Unit, all of the vertical tendons were first grouped by as-found corrosion level in 1997. The various corrosion levels were determined during the visual examination performed on the tendon wire surfaces behind the shim stacks at the top of each vertical tendon in 1997 and 1998. This is the area defined by References 3 and 4 as the area susceptible to wire breakage. Once the individual vertical tendons were ranked by corrosion level, a list of tendons by corrosion level group was generated ranging from "extreme corrosion" to "no corrosion." An average predicted number of wire failures was then calculated for each corrosion level group. This average number was then assigned to each tendon in a particular corrosion level group to represent future wire breaks. This approach of assigning an average number of future wire breaks to a corrosion level group was conservative in that, as vertical tendons were selected for replacement, less future wire breaks would be removed from the predicted total. By taking this approach, more corroded tendons would be required to be replaced to achieve acceptability of the Containments.

5E.2.3.2 Selection Criteria for Corroded Tendons to be Replaced To determine which vertical tendons would be replaced, a selection criterion was developed, as described below: 1. Replace all tendons that had two or more broken wires. Most of the additional broken wires discovered in 1999 were in tendons with two or more previous broken wires from 1997. Therefore, these tendons appeared to be the most likely to have future broken wires. Note: There was one exception to this criterion. Four buttonheads were found missing on the bottom of Unit 2 tendon 61V27 in 2001, and were not in the scope of Reference 4. Therefore, this tendon was not replaced. 2. Replace corroded tendons demonstrating lower lift-off forces. This applies to all tendons that were classified as having extreme or heavy corrosion and had a lift-off force of less than 649 kips in 1997. The small additional strain imparted by lift-off testing has the potential to cause additional wire breaks, as occurred in 1997. Corroded tendons with lower than predicted lift-off forces will be replaced to eliminate the possibility of premature wire breakage during future lift-off testing. Furthermore, replacing severely corroded tendons with low lift-off forces would prevent potential prestress losses associated with wire breakage from the restressing of these tendons. Restressing tendons increases the strain more than lift-off testing, and could potentially result in an even greater number of wire breaks. 3. Replace corroded tendons to ensure uniform distribution of prestress. The third criteria was specific to Unit 1 since it has two tendons that were not originally installed and, therefore, has two areas with low prestress force distribution. Calvert Cliffs replaced all the tendons that had extreme or heavy corrosion near the two empty tendon sheaths. 4. Replace corroded tendons to ensure uniform prestress force distribution after accounting for prestress losses from statistical Weibull model. This criteria ensures the loss of prestress that would result from the conservative prediction of wire breakage, would not violate design criteria described in Sections 5E.1.2 and 5E.1.3 at plant end-of-life. Calvert Cliffs applied the statistical model wire breaking predictions to all of the remaining original tendons that were not replaced under the first three criteria. This last criteria identified the areas around the Containments that, if the predicted wire breaks occurred, had the potential of driving the CALVERT CLIFFS UFSAR 5E.2-4 Rev. 47 distribution of vertical prestress force below the minimum design requirements. Once those areas were identified, appropriate corroded tendons were selected for replacement until the distribution of vertical prestress force exceeded the minimum design requirements at plant end-of-life. Once a vertical tendon was replaced, the future number of broken wires in the new tendon is assumed to be zero. The number of future wire breaks in non-replaced tendons is the average wire breaks for that corrosion level group. After tendon replacement on each Unit, the conservative predicted number of wire breaks at plant end-of-life drops to 1,195 for Unit 1 and 1,228 for Unit 2.

It was determined that 47 tendons on Unit 1 and 46 tendons on Unit 2 were the most cost-effective number of tendons to replace on each Unit that would provide the most uniform circumferential vertical prestress at plant end-of-life. It was also determined to restress 20 original vertical tendons on Unit 1, and 30 original vertical tendons on Unit 2, that had exhibited low lift-offs in 1997.

5E.2.3.3 Stressing Sequence of Tendons Replaced and Restressed The process of replacing and restressing vertical tendons on each Containment was done at full power. Therefore, to avoid operability issues during the work process, the number of tendons destressed at any one time, and the sequence in which tendons were destressed for removal, was critical to keeping the Containments within their design basis. Figures 5E-1 and 5E-2 show the final stressing sequence and individual vertical tendons replaced or restressed in 2001 and 2002 for Unit 1 and 2, respectively. 5E.2.3.4 Tendon Bearing Plate Concrete Void Repairs While performing lift-off testing on all the vertical tendons, two Unit 1 bearing plates depressed during the testing. The concrete under these bearing plates had been previously repaired as part of the tendon bearing plate study discussed in Appendix 5D. However, the repairs made to these bearing plates were not adequate to prevent bearing plate flexure. It was decided to remove these bearing plates and perform additional concrete void repairs with grout. The vertical bearing plates that received grout repairs in 2001, type of grout used, and repair method are shown on Figure 5E-1 for Unit 1. 5E.2.4 ACCEPTABILITY OF CONCRETE CONTAINMENTS Table 5E-2 provides a summary of vertical prestress conditions for both Units in 2002 for the original design and with corrective actions. Table 5E-2 also provides the predicted vertical prestress conditions in 2034 for Unit 1, and 2036 for Unit 2. The table is intended to provide a comparison of required and predicted vertical gross prestress, and a comparison of required and predicted mean average force per tendon sheath distribution.

Since not all the vertical tendons on each Unit exhibiting corrosion have been replaced, it should be noted that any corrosion on the original .25-inch diameter tendon wires could potentially reduce the effective cross-sectional area of the wire. A reduced effective cross-sectional area at the point of corrosion will cause the unit stress in the wire to increase. During initial stressing of the original tendons, the wires were left at a seating stress between 0.7 (168 ksi) and 0.73 (175.2 ksi) (Section 5.1.4.2). Over time, the original tendon wires have relaxed, reducing the stress in the wires as a percentage of . Therefore, the wire stress in the corroded areas should still be below the wire material minimum yield point of CALVERT CLIFFS UFSAR 5E.2-5 Rev. 47 0.8 (192 ksi). For the wires with severe corrosion that do become stressed beyond the wire material yield point and ultimately break, the total number has already been enveloped in the Containment acceptability evaluation. 5E.2.5 ENHANCED VERTICAL TENDON INSPECTIONS The future inspection of the vertical tendons is a two-tiered approach. First, ASME Section XI code inspections will be performed as required by the NRC-mandated ASME Boiler and Pressure Vessel Code (Reference 1). Lift-off testing will be conducted on the replacement tendons as required by the ASME Code. Second, enhanced inspections will be performed to examine the tendons for potential wire breaks. To monitor future changes in the conditions of all tendons, a database has been created to catalog the complete scope of all tendon inspection and repair activities.

The goal of enhanced inspections is to ensure that the Weibull Model bounds existing field conditions. To accomplish the enhanced inspections, the anchorhead/buttonhead region is required to be examined to determine if any wire breaks have occurred in the area under the vertical tendon top-stressing washers.

By the end of 2005 and 2007, Calvert Cliffs Nuclear Power Plant (CCNPP) will perform an inspection for wire breakage on 100% of the original vertical tendons. Unit 1 has 155 remaining original vertical tendons. Unit 2 has 158 remaining original vertical tendons. The purpose of these inspections will be to determine the number of failed (i.e., protruding) buttonheads at the top end of the vertical tendons. The resulting total number of protruding buttonheads will be compared to the number of predicted wire breaks for future specific years. If the total number of actual wire breaks is less than the number predicted for that unit in that year, and no tendon has more than the average number of failed wires in their corrosion group, then the actual condition of the containment vertical tendons are within the bounding conditions predicted by the statistical Weibull Model (Reference 5). If the number of failed wires in a tendon exceeds the average for that corrosion level group, the number will be compared to the average for that group at plant end-of-life. If a tendon's actual number of wire breaks exceeds the plant end-of-life predicted failure numbers, then an engineering evaluation will be performed.

In 2007, following the results of the ASME Code and enhanced inspections, CCNPP will assess the need to continue with enhanced inspections. This assessment will determine if the ASME Code inspections alone would provide adequate information to validate the statistical Weibull Model. If the model continues to bound field conditions, but more of a sample is required than that provided by the ASME Code surveillance population, the enhanced inspection frequency will be changed to a five-year span and then completed concurrently with the ASME Code inspections.

5E.2.6 CONTAINMENT PRESSURE TEST Reference 1, Article IWL-5000, provides requirements for pressure-testing concrete containments following repair or replacement activities. The concrete repairs to the Unit 1 Containment associated with the discovery of corrosion on the containment vertical tendons in 1997 only involved the removal of vertical tendon top bearing plates and the filling of voids with grout. These repairs were outside the outermost layer of structural reinforcing steel in the ring girder. The repairs and replacements to the containment vertical tendon system involved the exchange of post-tensioning tendons, tendon anchorage hardware, shims, and corrosion protection medium for both Containments. In accordance with the ASME Code, by performing these types of repairs and replacements only, no additional containment pressure tests were required to demonstrate containment structural integrity upon completion. The conclusions of the containment structural tests CALVERT CLIFFS UFSAR 5E.2-6 Rev. 47 to 115% of design pressure following original Units 1 and 2 construction remain valid as discussed in Section 5.5.1.2.

5E.

2.7 REFERENCES

1. American Society of Mechanical Engineers Boiler and Pressure Vessel Code, 1992 Edition through the 1992 Addenda, Section XI, Subsection IWL, "Requirements for Class CC Concrete Components of Light-Water Cooled Power Plants" 2. NRC Regulatory Guide 1.35, Revision 2, "Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containment Structures" 3. Calvert Cliffs Unit 1 20-Year Containment Tendon Surveillance Engineering Evaluation, October 28, 1997 4. CCNPP Root Cause Analysis Report, RCAR-9808, Root Cause Investigation of Containment Tendon Wire Corrosion and Failure, March 26, 1998 5. Dominion Engineering, Inc. Report, R-3648-00-01, Updated Model for Containment Structure Vertical Tendon Degradation for Calvert Cliffs 1 and 2, December 14, 2000

CALVERT CLIFFS UFSAR 5E.2-7 Rev. 47 TABLE 5E-2 PRESTRESS COMPARISON IN VERTICAL TENDON SYSTEM (Tendon relaxation losses for new tendons installed in 2001 and 2002 based on actual wire test Reports) (Values in kips) Unit 1(1) Unit 2 Gross PrestressMean Average Force per Tendon Sheath Gross Margin Mean Average Force per Tendon Sheath Margin Gross Prestress Mean Average Force per Tendon Sheath Gross Margin Mean Average Force per Tendon Sheath Margin 1. Design Basis 123,620 606 N/A N/A 123,620 606 N/A N/A 2. Predicted Prestress in 2002. No replacements, no restressing, no wire breaks. Original tendon design.(2) 130,694 641 7,074 35 133,416 654 9,796 48 3. Predicted Prestress after 2002 Corrective Actions. Includes replacements, includes restressing, no wire breaks.(3) 143,880 705 20,260 99 143,373 702 19,753 96 4. Predicted Prestress at 2034 and 2036. No replacements, no restressing, no wire breaks. Original tendon design.(4) 129,280 634 5,660 28 132,192 648 8,572 42 5. Predicted Prestress at 2034 and 2036. Includes replacements, includes restressing, includes wire breaks.(5) 132,334 648 8,714 42 131,353 643 7,733 37

CALVERT CLIFFS UFSAR 5E.2-8 Rev. 47 TABLE 5E-2 PRESTRESS COMPARISON IN VERTICAL TENDON SYSTEM _______________________ Note: All values in Table 5E-2 have been conservatively rounded down except the line 2 Unit 1 Mean Average Force per Tendon Sheath value, which has been rounded up. (1) Unit 1 has only 202 vertical tendons, but both Units have 204 sheath locations. (2) The respective data considers the prestress losses associated with concrete creep/shrinkage and wire relaxation only, and does not consider potential wire breaks. (3) Comparing Lines 2 and 3 show the predicted margin after the planned corrective actions. (4) These values can be considered the "original design" margins expected at the end of the operating licenses in 2034 and 2036, if tendon corrosion and wire breakage had never occurred. (5) Denotes the predicted prestress values following the 2001 and 2002 corrective actions of restressing and replacing tendons, while accounting for prestress losses associated with concrete creep/shrinkage and wire relaxation and a conservative amount of predicted wire breaks.

CALVERT CLIFFS UFSAR 5E.3-1 Rev. 47 5E.3 EVALUATION OF PLANT OPERATING LICENSE EXTENSION ON TENDON SYSTEM 5E.3.1 TENDON POPULATIONS Following original construction of the Units 1 and 2 Containments, there were only three distinct tendon populations: vertical, hoop, and dome. Inservice inspection (ISI) surveillance criteria for tendons during the early years of plant operation were based on Reference 1. Since the two Containments were identical in design, Reference 1 only required that lift-off testing surveillance be performed on one Unit for each tendon population. Tendon prestress losses in each population were generically calculated based on a nominal plant operating life of 40-years. In 2000, Calvert Cliffs received an extension of 20-years to its original operating license. With license extension, the prestress losses in each tendon population would require extrapolation from a 40-year nominal period to a 60-year nominal period. However, with the introduction of different NRC-mandated ISI criteria in 1996, and the discovery of corrosion in the vertical tendon population in 1997, simple extrapolation to 60 years of the existing prestress losses was not possible. Prestress losses for each tendon population on each Containment would have to be developed.

Part of the long-term corrective action plan was the replacement and restressing of vertical tendons on both Containments. The replacement vertical tendons have a different relaxation percentage than the original tendons, and a different effective prestress loss time period. The restressed original vertical tendons also have different relaxation characteristics from the original vertical tendons that were not restressed, and a different effective prestress loss time period.

These changes resulted in tendon prestress losses being developed for three vertical tendon sub-populations, the horizontal tendon population, and the dome tendon population for each Containment unit. Tables 5E-3 and 5E-4 show the tendon populations for each Unit and their effective life. 5E.3.2 ORIGINAL TENDON PRESTRESS LOSSES The Unit 1 original vertical tendons, hoop tendons, and dome tendons already had 40-year average prestress losses calculated, which had been converted into average tendon force versus time curves to be used in evaluating ISI surveillance lift-off force data. For these three original tendon population curves, the expected tendon force curves only were required to be extended to cover the new nominal 60-year plant end-of-life time period due to license extension. To develop prestress losses for the Unit 2 original vertical, hoop, and dome tendons, data from the original Unit 2 prestressing report was used. Since the prestress losses and conversion into a tendon force versus time curve represents the average tendon in a population, the average seating stress for the original three populations was taken from the original stressing report. The expected prestress losses at the end of 40 years were then determined based on the original loss values for elastic shortening of concrete, creep and shrinkage of concrete, and relaxation of tendon wire steel as provided in Section 5.1.4.2. Once a loss equation was developed to fit the 40-year values, the equation was used to extend the expected tendon force curves to cover the tendon initial stressing to plant end-of-life time period (60 nominal years). Special consideration is required for the original Units 1 and 2 tendons that were restressed. These tendons were initially stressed in the early 1970s, held under sustained tension, and then retensioned in 2002. As described in Reference 2, tendons that are restressed will reinitiate relaxation losses as if being stressed for the first time, although not to the same extent. Reference 2 documents an acceptable restress factor of 0.65. To CALVERT CLIFFS UFSAR 5E.3-2 Rev. 47 apply the factor of 0.65 to restressed tendons, the tendon wire relaxation for a given time period is determined for virgin, unstressed wire, and then multiplied by the 0.65 factor. The average date for restressing the Unit 1 vertical tendons was August 2002. The average date for initial stressing of the Unit 1 vertical tendons was March 1973. Therefore, the 2002 restress corresponds to year 29.5 of the tendon life. The end-of-plant operating life corresponds to year 61.5 of the tendon life. The new tendon force versus time curves based on prestress losses for restressed tendons have an initial stressing date of August 2002. The amount of concrete creep/shrinkage losses that are expected to occur between years 29.5 and 61.5 of the original tendon life were able to be determined since the creep/shrinkage loss is a function of time from the point that the Containment was originally stressed. No new concrete elastic shortening losses were considered. For tendons that are restressed or newly installed, there will not be an appreciable concrete elastic loss since only a few tendons were permitted to be detensioned at any given time. Since the vast majority of tendons remained in tension at all times, the concrete Containments remained under compressive stress and retained the associated elastic deformations from the original stressing in the early 1970s. Equations were developed to determine the amount of average prestress losses due to containment concrete creep and shrinkage between years 29.5 and 61.5, and the amount of prestress losses due to wire relaxation for 32 years after applying a 65% factor. Once the 32-year prestress losses were determined, another equation was fitted to the beginning average restressing force and final average calculated tendon force at the end-of-life. From this equation, tendon force versus time curves were developed to evaluate future tendon surveillance lift-off force test results. The same approach was used for the Unit 2 restressed vertical tendons in developing final force versus time curves. The only differences were the average tendon remaining life and the average initial restressing force used in the approach. 5E.3.3 NEW TENDON PRESTRESS LOSSES A total of 47 (Unit 1) and 46 (Unit 2) new vertical tendons were installed during 2001 and 2002. Each tendon consisted of 90-1/4" diameter wires with buttonheaded anchorages, just like the original tendon design. The only significant fabrication change for all the new tendons was that the upper stressing washers all utilized the Prescon 93 hole design. The three empty holes were to provide a vent path during the tendon greasing process to prevent formation of a grease void below the stressing washer. The new tendon wire is per American Society for Testing and Materials A421 Type BA. The relaxation of the prestressing wire steel was specified to be less than 4% for a design life of 40 years. Three 1000-hour relaxation tests were performed on sample wires. Test data was plotted and, using regression analysis, a best fit line was plotted. Each line was projected out to 40 years with the expected relaxation losses at 40 years ranging from 1.5% to 2.2%. An average relaxation value of 2.0% at 40 years was used in calculating the average tendon prestress loss and developing the tendon force versus time curves for the new tendons.

The new tendons will have a total life of 32 years. The tendon force versus time curves based on prestress losses for new tendons have an initial stressing date of August 2002. The amount of concrete creep/shrinkage losses that are expected to occur between years 29.5 and 61.5 of the original tendon life were determined since the creep/shrinkage loss is a function of time from the point that the Containment was originally stressed. Steel relaxation, however, is initiated at the time of stressing of the new tendons. No new concrete elastic shortening losses were considered. For tendons that are newly installed, there will not be an appreciable concrete elastic loss since only a few tendons were permitted to be detensioned at any given time. Since the vast majority of tendons remained in tension at any given time, the concrete Containments remained under compressive stress and retained the associated elastic deformations from stressing in the CALVERT CLIFFS UFSAR 5E.3-3 Rev. 47 early 1970s. Equations were developed to determine the amount of average prestress loss due to containment concrete creep and shrinkage between years 29.5 and 61.5, and the amount of prestress loss due to 2% (over 40 years) wire relaxation for 32 years only. Once the 32-year prestress losses were determined, another equation was fitted to the average new tendon initial stressing force and final average calculated tendon force at end-of-life. From this equation, tendon force versus time curves were developed to evaluate future tendon surveillance lift-off force test results. Since the tendon force versus time curves are used to evaluate tendon ISI surveillance lift-off data, the actual average wire relaxation test result of 2% was used in developing the curves instead of the maximum specified wire 40-year relaxation value of 4%. The same approach was used for the Unit 2 new vertical tendons in developing final tendon force versus time curves. The only differences were the average tendon remaining life and the average initial new tendon stressing force. 5E.

3.4 REFERENCES

1. NRC Regulatory Guide 1.35, Revision 2, "Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containment Structures" 2. Report From Ginna Nuclear Power Station, GAI Report No. 2499, Containment Vessel Tendons - Stress Relaxation Properties of Retensioned Wires, December 1, 1983

CALVERT CLIFFS UFSAR 5E.3-4 Rev. 47 TABLE 5E-3 UNIT 1 TENDON POPULATIONS AND EFFECTIVE LIVES UNIT 1 Vertical Original Vertical Restressed Vertical New Hoop Dome A 60-Year Plant License Expiration 7/31/2034 7/31/2034 7/31/2034 7/31/2034 7/31/2034B Average Stressing Date (Take as middle of month) 3/1973a 8/18/2002 8/20/2002 11/1971b 11/1971b C Years from Stressing Until Plant End-of-Life: A-B (Rounded to 1/4 year) 61.5 32 32 62.75 62.75 _______________________ a Per the original Unit 1 prestressing report, nearly the entire population of Unit 1 vertical tendons was detensioned and repairs were made to the anchor bearing plates after initial stressing. The repairs and subsequent restressing of the tendons occurred between January and April 1973. Average date taken was March 1973. b Per the original Unit 1 prestressing report, the Unit 1 tendons were initially stressed over a period of nine months between June 1971 and March 1972. Average date taken was November 1971.

CALVERT CLIFFS UFSAR 5E.3-5 Rev. 47 TABLE 5E-4 UNIT 2 TENDON POPULATIONS AND EFFECTIVE LIVES UNIT 2 Vertical Original Vertical Restressed Vertical New Hoop Dome A 60-Year Plant License Expiration 8/13/2036 8/13/2036 8/13/2036 8/13/2036 8/13/2036B Average Stressing Date (Take as middle of month) 6/1973a 5/2002 6/2002 10/1972b 10/1972b C Years from Stressing Until Plant End-of-Life: A-B (Rounded to 1/4 year) 63.25 34.25 34.25 63.75 63.75 _______________________ a Per the original Unit 2 prestressing report, nearly the entire population of Unit 2 vertical tendons was detensioned and repairs were made to the anchor bearing plates after initial stressing. The repairs and subsequent restressing of the tendons occurred between April and September 1973. Average date taken was June 1973. b Per the original Unit 2 prestressing report, the Unit 2 tendons were initially stressed over a period of eleven months between May 1972 and April 1973. Average date taken was October 1972.

CALVERT CLIFFS UFSAR 5E.4-1 Rev. 47 5E.4 VERTICAL TENDON PRESTRESS LOSSES DUE TO FUTURE WIRE BREAKAGE In accordance with American Concrete Institute Code 318-63, the prestress system has been designed for prestress losses from the following effects: a. Seating of anchorage; b. Elastic shortening of concrete; c. Creep of concrete; d. Shrinkage of concrete; e. Relaxation of prestressing steel stress; and f. Frictional loss due to intended or unintended curvature in the tendons Due to the discovery of corrosion at the top end of vertical tendons on both Units, the prestress system has also been designed for possible vertical tendon individual wire breaks. It is anticipated that over the remaining operating life of both Units, wire breakage will occur in the remaining original vertical tendon population. Unit 1 will meet its design basis with up to 1,195 additional broken vertical tendon wires, and Unit 2 will meet its design basis with up to 1,228 additional broken vertical tendon wires. The tendon force versus time curves discussed in Section 5E.3 for the three vertical sub-populations, hoop population, and dome population on each Unit are based on 90-1/4" diameter wire tendons. Should individual wire breakage be discovered in tendons during future inspections, the tendon force versus time curves will be proportioned-down by the ratio of unbroken wires to 90. The difference between the proportioned down tendon force versus time curve, and the 90-wire tendon curve, will represent the prestress loss due to wire breakage at that period of tendon life.

The prestress losses caused by the effects listed in a through f, above, occur for the most part uniformly throughout the Containment Structures. The distribution of possible future vertical tendon wire breakage is predicted to be skewed heavily toward the more severely corroded remaining vertical tendons. This was a critical factor in the evaluation to support 1,195 (Unit 1) and 1,228 (Unit 2) wire breaks.

Therefore, in addition to monitoring the loss of prestress due to the gross number of future wire breaks, the localized area effects of prestress losses will also be monitored as part of the enhanced inspections described in Section 5E.2.5.

CALVERT CLIFFS UFSAR 5E.5-1 Rev. 47 5E.5 CONCLUSION An engineering evaluation has demonstrated the acceptability of the CCNPP Units 1 and 2 concrete Containments with degraded conditions found during the 20-year (1997) containment tendon surveillance. The most severe conditions found were in the vertical tendon population. Some Units 1 and 2 vertical tendons, which have corrosion on individual wires below their upper (top) stressing washer, have not been replaced or repaired. The majority of the tendons exhibiting the greatest corrosion levels were replaced. In addition, vertical tendons that had exhibited lower than expected lift-off forces in 1997, and were not replaced, were restressed. The Containments were also demonstrated acceptable after considering the prediction that some corroded wires in tendons not replaced will break over the operating life of the two Containments. The Units 1 and 2 Containments will have sufficient vertical tendon prestress at the end of their operating licenses to meet the minimum design basis requirement of 123,620 kips gross vertical prestress, and 606 kips mean average force per tendon sheath. When its operating license expires July 31, 2034, the Unit 1 vertical tendon prestress will have appreciable margin above the two minimum design values. The 47 new tendons and 20 restressed tendons were seated to at least 742 and 725 kips, respectively. In addition, the design basis vertical prestress can be achieved with up to 1,195 additional wires breaking in the original non-replaced vertical tendon population between 2002 and July 31, 2034. When its operating license expires August 13, 2036, the Unit 2 vertical tendon prestress will have appreciable margin above the two minimum design values. The 46 new tendons and 30 restressed tendons were seated to at least 742 and 725 kips, respectively. In addition, the design basis vertical prestress can be achieved with up to 1,228 additional wires breaking in the original non-replaced vertical tendon population between 2002 and August 13, 2036. Table 5E-2 of Section 5E.2.4 shows that the Unit 1 Containment will have greater vertical tendon prestress at the end of its operating license after tendon replacement, tendon restressing, and predicted wire breaks, than it would have under the original tendon system design. However, the table also shows that the Unit 2 Containment will have slightly less vertical tendon prestress at the end of its operating license after tendon replacement, tendon restressing, and predicted wire breaks, than it would have under the original tendon system design. Although there is potentially less design margin above the minimum requirements, there is still sufficient margin. There was conservatism used in the development of the new calculated end-of-life vertical prestress. In addition, CCNPP expects that future tendon inspection data will verify that there will be far less actual wire breaks than currently assumed, allowing the predicted margin at the end of the extended operating licenses to approach (in Unit 2's case) or far exceed (in Unit 1's case) the original design margin.

Revision 41 FIGURE 5E-1 RING GIRDER BEARING PLANT PLAN - BEARING PLATE REPAIRS AND TENDON STRESSING SEQUENCE - UNIT 1 Rev. 335E-2, Ring Girder Bearing Plate Plan Unit 2

CALVERT CLIFFS UFSAR 6.5-1 Rev. 47 6.5 CONTAINMENT AIR RECIRCULATION AND COOLING SYSTEM 6.5.1 DESIGN BASIS The function of the Containment Air Recirculation and Cooling System is to remove heat from the containment atmosphere during normal plant operation. In the event of the occurrence of a LOCA, the system functions to limit the containment pressure rise to a level below the design value. In such an instance, the system also functions to reduce the leakage of airborne and gaseous radioactivity by providing a means of cooling the containment atmosphere. The containment air recirculation and cooling system is independent of the SI and Containment Spray Systems. It is sized such that, following a LOCA, three of the four cooling units will limit the containment pressure to less than the containment design pressure even if the Containment Spray System does not operate. The original sizing of the containment air coolers for design and procurement was based on the heat removal capability, using three coolers, (240x106 Btu/hr) required to maintain the post-LOCA containment atmospheric pressure within the containment design pressure. Likewise, the original sizing of the containment spray system for design and procurement was based on the heat removal capability (240x106 Btu/hr) required to maintain the LOCA containment atmospheric pressure within the containment design pressure. The analysis of these systems operating together post-LOCA in accordance with the Technical Specification requirements is presented in Chapter 14.

All system components are designed to withstand Seismic Category I loadings.

6.5.2 SYSTEM DESCRIPTION The containment air recirculation and cooling system (Figure 9-20A) includes four two speed cooling units located entirely within the containment. Service water is circulated through the air cooling coils. Each coil house is equipped with 12 individual coils piped to supply and return manifolds which connect to the SRW System.

The SRW supply line for each cooler has an air-operated valve which is normally open and de-energized. Each of these valves is designed such that it would fail in the open position. A redundant line to the coolers is normally valved off by a local, manually-operated valve.

The SRW return line for each cooler has two air-operated stop valves and one local, manually-operated valve, all in parallel. The air-operated control valves can be operated from the Control Room. One control valve is used for normal cooling requirements and the other control valve opens automatically upon receipt of SIAS. The third, parallel, local, manually-operated valve is provided to permit passage of sufficient SRW in case the full flow SIAS-actuated valve should malfunction.

Air is drawn through the coils by a vane-axial fan driven by a direct coupled two-speed motor. Normal containment recirculation requirements are satisfied at high speed operation, whereas, after a LOCA, the low speed setting is used. All fan motors may be manually started or stopped from the Control Room. Upon occurrence of LOCA, all four fan motors start automatically upon receipt of SIAS.

Upon evacuation of the Control Room due to fire, control of all fan motors may be transferred to local control stations in the electrical penetration rooms. Upon selection of local operation, the ESF signals (SIAS and under-voltage) are overridden. CALVERT CLIFFS UFSAR 6.5-2 Rev. 47 Performance data for the cooling units is given in Section 6.5.3. The materials of construction are listed in Table 6-8. The equipment is designed to withstand Seismic Category I accelerations and to operate in the LOCA environment. Service water flow is shown in Figures 9-9 (Unit 1) and 9-27 (Unit 2). 6.5.3 SYSTEM OPERATION a. Normal Operation The number of operating coolers is temperature dependant. Three cooling units are normally in operation during the warmer months and two cooling units are normally in operation during the cooler months. Each unit is sized to remove in excess of one-third of the total normal cooling load. The maximum average temperature inside the Containment is limited to 120°F by operation of the three cooling units. The maximum expected SRW inlet temperature to the coolers is 95°F. During normal operation, the full-size SRW outlet valves, which are used following a LOCA, may be closed, while the smaller (4" diameter) valves are open. Occasionally, during extended periods of high outside temperature, all four coolers are used to limit the average containment temperature to 120°F. Service water flow to the containment air coolers may be supplemented by using the 8" full size SRW outlet valves. Performance Data for Normal Operation Total heat removal capacity 2.27x106 Btu/hr(a) Motor horsepower 125 hp high speed (normal) Air flow, each 110,000 cfm(a) Fan horsepower, each 100 bhp Fan speed 1,200 rpm Cooling water flow, each 550 gpm(a) Air temperature, inlet/outlet 120/99°F(a) Water temperature, inlet/outlet 95/102.8°F(a) Fan static pressure 3.2" H20 Fan total pressure 4.2" H20 (a) Cooler heat removal capacity is a function of SRW flow and temperature, fouling, air flow, and containment temperature and pressure. b. Plant shutdown operation During plant shutdown, i.e., Modes 4, 5, 6 and defueled, the cooling units operate as necessary, based on availability and containment conditions. The availability of the cooling units is directed by administrative controls. c. Emergency Operation Upon receipt of a SIAS, any idle cooling unit is automatically started on the low speed setting and, simultaneously, any running fan is switched from their normally operating high speed setting to low speed operation. The full flow (8" diameter) SRW outlet valves for each cooler are opened upon receipt of a SIAS. The SRW inlet valves move to a throttled position upon receipt of a SIAS, and return to the full open position upon receipt of a RAS.

With off-site power available under this mode of operation, the operating cooling unit fans are switched to low speed and the idle fan(s) are started on low speed as described above. If off-site power is not available, the associated emergency CALVERT CLIFFS UFSAR 6.5-3 Rev. 47 diesel generators are started. Each emergency diesel generator supplies power to an independent safety-related bus. Each bus carries the load of two cooling units. The evaluation of post-incident containment pressure/temperature response is provided in Section 14.20. These evaluations consider the actual heat removal capacity of the containment air coolers which is a function of SRW flow and temperature, fouling, air flow, and containment pressure and temperature.

With respect to long-term cooling after a LOCA, the cooling units are designed to operate for at least one year under air-steam mixture conditions of 5 psig and 160°F. Performance Data for Emergency Post-LOCA Conditions Motor horsepower 63 hp (low speed) Fan horsepower (max.), each 33 bhp Fan speed 600 rpm Mixture temperature, inlet/outlet 275/270°F(a) Water temperature, inlet/outlet 105/204°F(a) Cooler, capacity at 273°F and 47 psig, each 90.45x106 Btu/hr(a) Cooler mixture flow, each 55,000 cfm(a) Maximum fin side pressure drop 0.5" H20 Maximum tube side pressure drop 15.6 psi Fan static pressure 1.2" H20 Fan total pressure 1.445 in H20 Water flow, each 1,900 gpm(a) (a) Cooler heat duty will vary with flow, temperature, and humidity. 6.5.4 DESIGN EVALUATION a. The coil capacity is based upon 95°F SRW inlet during normal operations and 105°F SRW inlet during LOCA. These values represent the maximum expected temperatures. The water velocity through the coils is 7.4 fps at 1,900 gpm flow. b. Total effective face area in each cooler is 144 ft2. With a normal operating air flow of 110,000 cfm, the velocity entering the coils is 765 fpm. With the emergency mixture flow of 55,000 cfm, the entering velocity is 382 fpm. c. The fin spacing is 6 fins per inch of coiled length. With this pitch, water clogging of the coil fins is avoided. d. A fouling factor of 0.0005 for the water side is included in the coil ratings. The water side of the cooling coil tubes is equipped with removable plugs on the return bends of the coils to permit cleaning in the field. e. The cooler housing is designed to ensure no loss of function when subjected to a pressure differential of 2 psi. f. Components are designed to be compatible for operation in an environment of borated water spray. The three- to four-hour short-time temperature exposure rating during a LOCA is about 280°F. The three- to four-hour short time humidity exposure rating is 100% relative humidity in a slightly acid atmosphere. Additionally, components which are considered susceptible to radiation damage, such as the gasket and motor, are designed to withstand a dose of 108 rad of gamma radiation. It has been calculated that these components will receive a dose of less than 5x107 rad during the year following a LOCA. g. With respect to normal containment recirculation and cooling, the cooler assemblies are designed for a life of 40 years. CALVERT CLIFFS UFSAR 6.5-4 Rev. 47 h. The condensate leaving the coils is conveyed over individual stainless steel drip pans to the sides of the coils out of the mixture flow stream. These pans cascade the liquid into the main sump of the housing from which it is drained via the Containment sump to one of the Auxiliary Building sumps, from which it is pumped to the Waste Processing System. 6.5.5 AVAILABILITY AND RELIABILITY a. The cooling units are located outside the secondary shield. In this location they are protected from being flooded at post-accident conditions and they also are protected against credible missiles. b. The original design heat removal capability of three of the four cooling units was to provide the same heat removal capability as the containment spray system. The analysis of these systems operating together post-LOCA in accordance with the Technical Specification requirements is presented in Section 14.20. The single failure characteristics for the cooling units are listed in Table 6-9. c. Each fan discharge duct is provided with a fusible link door. These doors open at an abnormally high containment temperature such as would occur under a LOCA. This assures the free flow of the cooled air-stream mixture to the containment environment even if the ducts collapse during or following the LOCA. d. The cooling units are designed to operate for the life of the plant. There are no belts or flexible couplings; the motor is directly connected to the fan wheel. e. Upon loss of off-site power during a LOCA, the containment cooling fans are automatically sequenced onto the emergency diesel generator buses. f. All associated system equipment, such as piping, valves, and instrumentation are also located outside of the secondary shielding to minimize the possibility of missile damage. 6.5.6 TESTS AND INSPECTIONS a. The manufacturer has developed a computer program to size cooling coil units for saturated steam-air mixtures. This program was used to size the Calvert Cliffs cooling units. Tests performed on the coils manufactured for Palisades, Fort Calhoun, Three Mile Island No. 1, Kewaunee, and Oconee have confirmed the validity of the program. These tests were conducted with coils of material and configuration identical, except for shorter length, to those used in each specified large-scale containment system. Three of the coil tests were made with the air flow horizontal and the condensate drainage perpendicular to the air flow. Water-logging problems did not arise. The coil section drainage characteristics were identical to those which were predicted for the full-size units, and therefore provide assurance that the full-size coils will adequately drain condensate from the coil surfaces. The coil was tested at a pressure loading equal to a free velocity pressure of 500 fps to demonstrate the structural integrity of the coil. This loading test was performed to simulate a pressure wave which may occur during the initial phase of containment pressure buildup in the event of a LOCA. Upon examination, the coil showed only very minor deformation of some intermediate stiffeners; hence, the structural design of the coil was proven to be adequate. b. A fan and motor have been tested to prove their ability to operate satisfactorily under conditions existing within the containment after a LOCA. Fans and motors also were tested by the fan manufacturer to assure the same characteristic performance curve for all fans. CALVERT CLIFFS UFSAR 6.5-5 Rev. 47 c. Cooling unit performance can be tested with thermometers, manometers and a Pitot tube in the field at any time the containment is accessible. d. The valves in the normal cooling water outlet lines (4") will be open during normal operation and the valves in the parallel emergency outlet lines (8") can be opened from the Control Room and the flow rate can be monitored at any time. e. All equipment and associated components are arranged so that they can be inspected at any time the containment is accessible. f. The containment air cooler blowdown door fusible links will be replaced every refueling outage to ensure that the links perform their design function.

CALVERT CLIFFS UFSAR 6.5-6 Rev. 47 TABLE 6-8 COOLING UNIT MATERIALS OF CONSTRUCTION Tubes (seamless) 90/10 copper-nickel, ASTM B111-69, Alloy 706 Fins ASTM B152 Headers ASTM B466 Coil Frame ASTM A525 Structure ASTM A501, A36, and M1020 Motor NEMA Class B, TEAO CALVERT CLIFFS UFSAR 6.5-7 Rev. 47 TABLE 6-9 SINGLE FAILURE CHARACTERISTICS FOR COOLING UNITS COMPONENT MALFUNCTION COMMENTS AND CONSEQUENCES 1. Unit Circulating Fan Fails to operate Three (or four) coolers and fans are normally operating. Fans may be tested for emergency mode of operation at any time. 2. Cooler Failure to tubes Tube failure is considered unlikely during emergency operation since the tube water to air side P is less than during normal service. If failure does occur, SRW will be spilled into the cooler since SRW pressure is above containment pressure. Tube leakage can be detected by indication of increased SRW flow to the cooler, decreased SRW head tank level, and the failed cooler can be isolated. Note: Passive failures are only considered post-RAS. In the event of tube failure associated with any one cooler after the LOCA, it is assumed, as an upper limit, that one subsystem of Service Water leaks into containment. The leak volume from one subsystem is approximately 16,000 gallons. Boron dilution, therefore, would be negligible, because the total volume of borated water in the containment structure is in excess of 400,000 gallons. 3. SRW Emergency Outlet Valve Fails to open For those coolers in operation, the valve in the normal cooling water outlet line will be open. The normal operation valve will be open. If the emergency valve fails to open, the unit will operate at reduced heat removal capability. The Containment Spray System will supplement the heat removal capability of the cooling units. 4. SRW Inlet Valve Inadvertently left throttled Valve status will be apparent from reduced flow, and the valve may be opened by operator action. If the valve fails to respond, the Containment Spray System will supplement the cooling requirements. Each cooler is provided with flow indication on the main control board. Fails to throttle Does not adversely affect containment heat removal. May result in emergency diesel generator service water inlet temperatures in excess of 105°F under certain conditions with elevated bay water temperatures. Containment spray system remains available to supplement heat removal capability of the cooling units. CALVERT CLIFFS UFSAR 6.5-8 Rev. 47 TABLE 6-9 SINGLE FAILURE CHARACTERISTICS FOR COOLING UNITS COMPONENT MALFUNCTION COMMENTS AND CONSEQUENCES Inadvertently closed Valve status will be apparent from lack of flow and position indication, and the valve may be opened by operator action. If the valve fails to respond, the Containment Spray System will supplement the cooling requirements. Each cooler is provided with flow indication in the Control Room. Note: Mechanical stops are installed on valves to limit stroke. The stops should prevent inadvertent full closure. CALVERT CLIFFS UFSAR 6.6-1 Rev. 47 6.6 CONTAINMENT PENETRATION ROOM VENTILATION SYSTEM 6.6.1 DESIGN BASIS The Containment Penetration Room Ventilation System is designed to collect and process containment penetration leakage, so as to reduce to a minimum the environmental radioactivity levels from post-accident containment leaks. See Section 14.24.3 for a discussion of the assumed operation of the system post-LOCA. 6.6.2 SYSTEM DESCRIPTION The Containment Penetration Room Ventilation System is shown schematically in Figure 9-20A. Since experience has shown that containment leakage is more likely at penetrations such as electrical cables and air purging valves, rather than through the liner plates or weld joints (Reference 1), penetration rooms are built adjacent to the outside surface of each containment and enclose the areas around the majority of the penetrations. The only penetrations which do not pass through these areas are: a. Two main steam lines; b. Two main feedwater lines;

c. Equipment hatch;

d. Normal personnel access lock;

e. Emergency personnel access lock; and, f. Refueling tube. The main steam and feedwater lines are welded to the liner plate and, therefore, are not considered as a source of leakage. The equipment hatch and access lock openings can be tested during normal operation and are not considered sources of significant leakage.

There are double seals at each of these three access openings. The refueling tube is valved on one end and blind flanged on the other end. The principal function of this system is to control and minimize the release of radioactive materials from the containment to the environment during a post-accident period. Following a LOCA, a Containment Isolation Signal (CIS) will start both of the two full-size blowers. The penetration room exhaust ventilation system design basis is the Maximum Hypothetical Accident. The system is credited in the dose calculations with filtering the radioactive material released through the 4" containment vent line at the onset of a Maximum Hypothetical Accident as well as leakage through the penetrations. A gravity damper, which opens when the blower starts, is provided at the discharge of each blower to prevent recirculation through a failed or idle unit. The entire system is designed to operate under negative pressure up to the fan discharge. To minimize the release of radioactive material to the environment, penetration room ventilation is continuously routed through a prefilter, an absolute high efficiency particulate air (HEPA) filter, and an activated charcoal filter, positioned in series. The use of these filters post-accident is described in Section 14.24.3.

In all cases, the flow rate from the penetration room will exceed the total maximum containment leakage rate. The containment purge equipment, if running, will be shut down by a containment radiation signal (CRS), and the valve in each purge line penetration will be closed. Refer to Section 14.24.3 for a discussion of relevant accident analysis assumptions. 6.6.3 SYSTEM COMPONENTS The Containment Penetration Room Ventilation System is provided with two blowers and two filter assemblies. Both blowers, each of which is aligned with a filter assembly, CALVERT CLIFFS UFSAR 6.6-2 Rev. 47 discharge through a single line to the unit vent. (Table 6-10) Power-operated dampers are provided for isolating each filter assembly from the penetration rooms. The filter assembly consists of a prefilter, a HEPA filter and an activated charcoal filter in series. The prefilter removes coarse airborne material and water droplets using pads of glass fiber, placed between perforated metal grids, as the filtering media.

The HEPA filter, which removes small airborne particles that pass through the prefilter, consists of two cells of fiber glass media mounted in a metal frame. The activated charcoal filter removes methyl as well as elemental iodine contaminants resulting from a LOCA. It consists of six cells of activated charcoal having approximately 5 wt% impregnation of iodine compounds, and an ignition temperature of 680°F, held in place by stainless steel channel clamps and galvanized bolts. As a means of checking the condition of the charcoal in each filter bank, one or more charcoal test trays, filled from the same principal batch of charcoal as the other trays, may be installed in lieu of regular trays in each filter bank. Since the test tray is a substitute for a regular tray, it experiences air flows at the same rate and angle as the other trays. This ensures that the samples taken from the test trays are representative of the charcoal in the entire bank. Trays of this type may be installed in any charcoal filter bank that is part of the iodine removal system in this power plant. For operator information, temperature and pressure monitoring is provided for all penetration rooms and an area radiation monitor is provided for the West Penetration Room. Differential pressure indicators are provided across the filters. 6.6.4 SYSTEM OPERATION During normal operation, the system is held on standby with both blowers aligned with their respective filter assemblies. A CIS will start both of the blowers. The containment purge equipment, if running, will be shut down by a CRS, and the valve in each purge line will be closed. All of the system components can be controlled from the Control Room. 6.6.5 DESIGN EVALUATION The blower capacity of 2000 +/- 200 cfm exceeds, in all cases, the total maximum Containment Building leakage rate. The blowers and the respective filters are aligned in a redundant manner to assure operation of one blower and its respective filter assembly, independent of a failure or malfunction of any of the active system components. When the system is in operation, a negative pressure may be maintained in the penetration rooms and in the ducting up to the discharge of the blower. All components are designed to Seismic Category I requirements. 6.6.6 AVAILABILITY AND RELIABILITY Redundancy of components, monitoring operation of the system from the Control Room and provision of proper instrumentation, assure proper response of the system when a LOCA does occur. Upon failure of the normal electrical power supply to the blowers, power is supplied from the emergency power source.

The system components and equipment are fully accessible during normal plant operation.

A single failure analysis for the main components of the system is given in Table 6-11.

CALVERT CLIFFS UFSAR 6.6-3 Rev. 47 6.6.7 TESTS AND INSPECTIONS All equipment and associated components are arranged so that they can be inspected at any time the containment is accessible. Testing of charcoal/HEPA filter units is established based on Technical Specification requirements. The Technical Specification specifies testing conditions based on the application of the particular filter unit. 6.6.8 REFERENCE 1. W.B. Cottrell and A.W. Savolainen, Editors, U. S. Reactor Containment Technology, ORNL-NSIC-5, Volume II, August 1965 CALVERT CLIFFS UFSAR 6.6-4 Rev. 47 TABLE 6-10 PENETRATION ROOM BLOWER DESCRIPTION Type Centrifugal Quantity 2 in each unit Capacity (cfm) 2000 at 8 1/2" w.g. (inch H20) Motor 5 hp, 460 Volt, 3 phase, 60 cycle Codes NEMA CALVERT CLIFFS UFSAR 6.6-5 Rev. 47 TABLE 6-11 SINGLE FAILURE CHARACTERISTICS FOR CONTAINMENT PENETRATION ROOM VENTILATION SYSTEM COMPONENT MALFUNCTION COMMENTS AND CONSEQUENCES 1. Blower Fails to start Spare blower is already operating. 2. Blower Fails during service Alarm in Control Room will indicate loss of negative pressure, and spare blower is already in service. 3. Blower valve Fails to open Spare blower is already operating. 4. Filter valve Fails to open Failure not considered credible since one filter will always be lined up to operate when needed. 5. Ductwork Leakage Leakage of unfiltered air may be inward since ductwork may be maintained at negative pressure. CALVERT CLIFFS UFSAR 6.7-1 Rev. 47 6.7 CONTAINMENT IODINE REMOVAL SYSTEM 6.7.1 DESIGN BASIS The containment iodine removal system is designed to collect, within the containment, the iodine released following a LOCA. 6.7.2 SYSTEM DESCRIPTION Following a LOCA, SIAS automatically starts three 20,000 cfm recirculation filter units. Each unit has the capacity of 50% of the design air flow. These units consist of activated charcoal filters preceded by HEPA filters. A moisture separator is provided upstream of the particulate air filters to remove water droplets. An electric-driven induced-draft fan located at the end of the banks of filters pulls the containment atmosphere through these components and discharges vertically back into the containment (Figure 6-6).

The three containment charcoal filter units contain a total of approximately 7300 lbs. of Barnebey-Cheney #727 coconut shell charcoal (or equivalent) impregnated with 5 wt% iodine compounds. The flow velocity through each bed is 40 fpm and the corresponding residence time is .25 seconds. Filter testing is explained in Sections 6.6.7 and 6.7.7. Testing is performed to demonstrate that the installed charcoal adsorbers will perform satisfactorily in removing both elemental and organic iodides for design conditions or flow, temperature, and relative humidity. Each of the recirculation filter units is provided with an emergency dousing system for the charcoal beds to dissipate the decay heat load in the event there is a significant rise in the charcoal bed temperature. During Modes 1, 2, 3, and 4, the dousing system is isolated by manual valves. An analysis (Reference 1) shows that maximum post-LOCA charcoal bed temperature will not cause iodine desorption or charcoal bed ignition. During Modes 5 and 6, the manual valves may be opened to allow the dousing system to be functional during iodine removal unit maintenance to provide fire protection if required.

This system is shown in diagram form on Figures 6-7 (Unit 1) and 6-11 (Unit 2). 6.7.3 SYSTEM COMPONENTS a. Unit Housing The unit housing is made of carbon steel and is capable of operating under the LOCA conditions of pressure and temperature. It is provided with service access doors, explosion-proof lights, a charcoal filter emergency dousing system, a monorail suitable for supporting an electric hoist to handle charcoal cells, test connections, connections for pressure gauges and floor drains. The housing is designed to be light-tight.

b. Moisture Separators The moisture separators consist of a steel casing, two-and-one-half-pass vertical louvers, and filter element frames. The bottom of the unit is a sump equipped with drain connections. Elements are constructed of stainless steel wire mesh. In addition to acting as moisture separators, the filter elements also act as a prefilter for the HEPA filters and remove large particles and any fibrous material present in the air steam. c. High Efficiency Particulate Air (HEPA) Filter Each element of the HEPA filter measures 24"x24"x11 1/2" and is constructed by pleating a continuous sheet of waterproof fire retardant fiberglass mat into closely spaced pleats separated by aluminum inserts. The filter medium is treated with a CALVERT CLIFFS UFSAR 6.7-2 Rev. 47 silicone-base water repellent to achieve wet strength characteristics. The filter medium and separators are encased in cadmium-plated carbon steel. The filter bank frame is galvanized steel. d. Charcoal Filters The material used in these filters is activated charcoal containing 5 wt% impregnation of iodine compounds and having an ignition temperature of 680°F. It is encased in perforated stainless steel beds. Each element is a standard manufactured size, has a frontal face size of approximately 8"x24", and contains two horizontal charcoal beds, each approximately 24"x24"x2" deep. The filter bank frame is galvanized steel.

As shown on Figures 6-7 (Unit 1) and 6-11 (Unit 2), each charcoal filter bank is equipped with an emergency cooling water dousing system. Provisions are made to collect the cooling water after it flows through the charcoal beds and onto the floor of the filter unit, and to drain it through a check valve into the Containment Structure. Each filter bank may contain a test tray which will be used to verify the efficiency of the charcoal bed periodically (Section 6.6.3). e. Fan-Motor Unit The fan is an internally-direct-driven, vane-axial type, equipped with an inlet and discharge adapter. The motor is a totally enclosed, air-over type and is provided with an insulation system capable of operating under LOCA conditions. 6.7.4 SYSTEM OPERATION The containment iodine removal filter units are not in operation during normal reactor operation. However, following a LOCA, receipt of SIAS will automatically start all three filter unit fans. These units may also be started manually by the operator from the Control Room at any time. During Modes 5 and 6, the charcoal bed emergency dousing system in each unit may be initiated manually, as needed. 6.7.5 DESIGN EVALUATION a. The system consists of three recirculating filter units with a total capacity of 150% of design flow requirements. Each unit is protected from all expected missile sources by concrete structures. b. The design life of the filter units is 40 years, the same as the design life of the plant. The units are not in operation during normal reactor operation. c. The filter units are designed to operate under the maximum temperature and pressure conditions resulting from a LOCA (original calculations show an atmosphere composed of steam-air mixture at a pressure of 47.45 psig and a temperature of 274°F). The filter units have been evaluated for the revised maximum pressure and vapor temperature contained in Section 14.20. In addition to this, the units are designed to withstand conditions of ambient temperature and 57.5 psig for 24 hours during testing of the containment. d. The units are designed to assure no loss of function when subjected to a pressure differential of 2 psi. e. The filter units are classified as Seismic Category I equipment and are designed accordingly. CALVERT CLIFFS UFSAR 6.7-3 Rev. 47 f. Filter unit components are designed to withstand a radiation level of 1 r/hr of principally gamma radiation with a 40 year cumulative dosage of 3.5x105 r under normal environmental conditions. g. The components are also designed to be capable of operation in an atmosphere of borated water spray and a maximum cumulative radiation dose of 108 r, occurring as a result of a LOCA. h. Each element of the moisture separators is designed to have a nominal flow of 1500 cfm and a clean pressure drop of less than 1" of water. i. Each element of the HEPA filters is designed to have a nominal flow rate of 1000 cfm and a clean pressure drop of less than 1" of water. j. Each element of the charcoal filter units is designed for a nominal flow rate of 333.3 cfm and a clean pressure drop of less than 1" of water. 6.7.6 AVAILABILITY AND RELIABILITY The Containment Iodine Removal System incorporates three filter units, each with a capability to handle 50% of the required air flow. The units are designed to operate without loss of function under the maximum temperature and pressure conditions resulting from a LOCA, as well as in the accompanying containment atmosphere environment, i.e., a steam-air mixture with borated water spray and radiation. Because the system is not in use during normal reactor operation, special routine tests and inspections are incorporated into plant operating procedures. These periodic tests will be performed during plant operation to assure the availability of power to each of the electrical components (both visual and audible indicators are provided in the control panel for this test). During plant shutdown the condition of each filter bank will be determined by visual inspection and pressure differential gauge indication while the fans are in operation. In addition, the water line solenoid valves will be visually checked for proper operation during periods of plant shutdown.

Failure of the normal electrical power supply will automatically place the three fans on emergency electric power. 6.7.7 TESTS AND INSPECTIONS See Section 6.6.7 for a current description of the tests and inspections performed for both systems.

6.

7.8 REFERENCES

1. Nuclear Consulting Services, Inc. Report NUCON-6BG021/01, A Computer Analysis of the Iodine Decay Heat Generated in a Carbon Bed Following a Loss of Coolant Accident, January 19, 1990, and Supplement 1 July 25, 1990 CALVERT CLIFFS UFSAR 6.8-1 Rev. 47 6.8 HYDROGEN CONTROL SYSTEMS 6.8.1 GENERAL The Calvert Cliffs containment has been designed to promote circulation of the contained air. Natural circulation is enhanced by large vent areas between compartments and by the use of gratings between elevations. A thorough review has been conducted to ascertain that no areas exist within which gases could be trapped. As a result of this review, design features such as the venting of the pressurizer compartment have been incorporated. Mechanical mixing of the air is achieved using the containment air recirculating and cooling system and the iodine removal system. Under LOCA conditions, hydrogen and radioisotopes will be rapidly distributed throughout the containment due to turbulent currents introduced by the escaping steam and water and natural convection due to the temperature differential between the sump water and the containment atmosphere. The air recirculation and Containment Spray Systems also promote mixing and minimize nonuniformities in the containment atmosphere. On March 2, 2004, the Nuclear Regulatory Commission approved a license amendment which changed the definition of DBEs to exclude hydrogen generation in Containment as a consequence of the event. The amendment was based, in part, on the size of the Containment and free circulation of air within the Containment. Hydrogen recombiners are no longer needed and were removed from the licensing basis. However, hydrogen analyzers are required to be retained as non-safety-related equipment to evaluate events beyond the design basis.

6.8.2 DELETED 6.8.3 CONTAINMENT VENT SYSTEM The Containment Vent System (Figure 9-20A) is used during power operations to vent the Containment to maintain containment pressure and airborne radioactivity within Technical Specification limits. The vented air will be introduced to the penetration room exhaust system and will be passed through the penetration room exhaust system's HEPA and charcoal filters before being discharged to the environs. The system is equipped with a flow monitor and motor-operated valves.

The containment vent may be used to purge hydrogen from Containment if desired. Upon receipt of a SIAS, CRS, or a high radiation signal, the MOVs close automatically. During post-accident conditions with the CRS or containment high-radiation signal present, the Containment Vent System can be operated by overriding the CRS or containment high-radiation closing signal using key-operated override handswitches on control panels 1C10 and 2C10.

CALVERT CLIFFS UFSAR 6.10-1 Rev. 47 6.10 ELECTRICAL HEAT TRACING SYSTEM Electrical heat tracing is installed on all piping, valves, pumps and other line-mounted components that contain concentrated boric acid. Heat tracing is needed to maintain a 12% weight boric acid solution above the 135°F saturation temperature. The heat tracing system is designed to maintain 160°F; however, the operating temperature may be set lower. The thermal insulation is designed to limit the insulation surface temperature to 140°F based on an ambient air temperature of 80°F and component temperature of 160°F. The electrical heat tracing system is designed such that a single failure will not cause loss of function and includes redundant heater elements, controls, and alarm functions. Each subsystem is supplied by separate emergency power sources.

Each subsystem is equipped with an independent alarm system. Functions which are alarmed locally and in the Control Room include high and low temperature and loss of power. Power for heat tracing is considered as part of the load for the emergency diesel generator during the first eight hours of operation.

The pre-operational test of the heat tracing system verified that the design basis, as stated above, is adequate. This test program included verification of the following: a. Proper functioning of all heating elements; b. Alarm functions for loss of power to heating elements;

c. The temperature of the boric acid solution;

d. Manual transfer to the redundant subsystem from the local control panel; and,

e. All heat tracing controls function properly.

CALVERT CLIFFS UFSAR 6.11-1 Rev. 47 6.11 ECCS LONG-TERM COOLING FLUSH During post-accident long-term cooling conditions for a large cold leg break, it is possible to have a boric acid reconcentration occur in the core area, due to the small core flow in effect. This condition may result in the crystallization of boric acid in the core and restriction of cooling flow. In order to prevent such an occurrence, two procedures have been developed. The operators will decide which method to use based on plant conditions.

Hot leg injection promotes flow through the core by establishing a flow path from the containment sump, through the LPSI system via the warm-up line, to the shutdown cooling suction line, and into the hot leg. Pressurizer injection promotes flow through the core by diverting flow from one HPSI pump through the pressurizer auxiliary spray line and into the hot leg. A minimum of 150 gpm injection flow (flow to the reactor coolant system hot side) is necessary to overcome the coolant boil-off rate and cause a net flushing flow downward through the core. The required injection flow must be provided within at least 11 hours after the accident. The HPSI pumps are the preferred method of core flush. The LPSI pumps could be used for core flush only as a backup to the HPSI pumps when pressurizer injection is not possible. The containment spray pumps can be aligned to provide core flush into the hot leg by realigning valves on the -10 foot level of the Auxiliary Building. This realignment can be made after the containment spray pumps have completed the performance of their required post-LOCA function for spray service to reduce containment pressure. The containment spray pumps would only be used as a backup to the LPSI pumps.

FIGURE 6-1 SAFETY INJECTION AND CONTAINMENT SPRAY - UNIT 1

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6-11 CONTAINMENT CHARCOAL FILTER SPRAY -UNIT 2 \: *,;: ([:1! -"' "' -G\u . -" .. :.!i -l.ll "' -"" .:: " " ".. n ** ro ' ---1 .. , "' .. ' .. l STOP, THfN11:;. ACT REVIE\f """=:!:" ***---***--.......... " Rev. 32 CHAPTER 7 INSTRUMENTATION AND CONTROL TABLE OF CONTENTS PAGE7.0 INSTRUMENTATION AND CONTROL

7.1 INTRODUCTION

7.2 REACTOR PROTECTIVE SYSTEM 7.3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEMS CHAPTER 7 INSTRUMENTATION AND CONTROL TABLE OF CONTENTS PAGE 7.4 REGULATING SYSTEMS CHAPTER 7 INSTRUMENTATION AND CONTROL TABLE OF CONTENTS PAGE7.5 INSTRUMENTATION SYSTEMS CHAPTER 7 INSTRUMENTATION AND CONTROL TABLE OF CONTENTS PAGE7.6 OPERATING CONTROL STATIONS 7.7 CONTROL ROOM ANNUNCIATION 7.8 COMMUNICATIONS 7.9 DELETED 7.10 AUXILIARY FEEDWATER ACTUATION SYSTEM 7.11 ANTICIPATED TRANSIENT WITHOUT SCRAM CHAPTER 7 INSTRUMENTATION AND CONTROL TABLE OF CONTENTS PAGE 7.12 ENVIRONMENTAL QUALIFICATION OF ELECTRIC EQUIPMENT IMPORTANT TO SAFETY CHAPTER 7 INSTRUMENTATION AND CONTROL LIST OF TABLES TITLEPAGE CHAPTER 7 INSTRUMENTATION AND CONTROL LIST OF FIGURES FIGURE CHAPTER 7 INSTRUMENTATION AND CONTROL LIST OF ACRONYMS CHAPTER 7 INSTRUMENTATION AND CONTROL LIST OF ACRONYMS CALVERT CLIFFS UFSAR 7.1-1 Rev. 47 INSTRUMENTATION AND CONTROL 7.

07.1 INTRODUCTION

The plant systems are instrumented to provide information on plant conditions at selected locations, to protect equipment and personnel from undesirable conditions and to control the plant during startup, operation, and shutdown. The principal control station for the plant is in the Control Room.

The plant is started up and shut down under remote manual control. Annunciators, indicators, and recording devices will alert the operator and provide data on plant conditions. Instrumentation and controls essential to plant safety are located in the Control Room. The instrumentation is arranged in groups on the control boards so that when corrective action is required, all pertinent indicators, recorders, and controllers are within easy reach of the operator. The control board is a duplex benchboard. Visible and audible alarms located on the superstructure over the main control board annunciate and identify abnormal operation conditions. Telephone systems provide both in-plant and external communication. The Control Room is a controlled temperature environment, kept well within the design ambient temperature requirements of the instruments. The computer room temperature and humidity are kept closely controlled. To ensure reliability, components of established quality are selected and used in the instrumentation and control equipment. All protection systems that actuate reactor trip engineered safety features (ESFs), and auxiliary feedwater (AFW) components are designed to conform to the criteria of Institute of Electrical and Electronic Engineers (IEEE) 279 and those sections that are relevant from the Commission's proposed General Design Criteria, as published February 20, 1971. The Diverse Scram System (DSS) does not meet IEEE 279. The requirements for this system are established by 10 CFR 50.62 (Section 7.11). The protection instrumentation consists of four independent multiple channels to permit system testing without reducing the degree of protection provided. Reliable sources of electrical power are provided to ensure safe and reliable plant operation (Chapter 8). The operation of the reactor within established limits is achieved by its inherent characteristics, instrumentation and control systems, and by operational procedures and administrative controls. Potential departures from these limits are audibly and visibly annunciated in the Control Room. A Reactor Protective System (RPS) is designed to protect the core and the Reactor Coolant System (RCS) pressure boundary and to initiate reactor trips. The ESF instrumentation provides the equipment necessary to initiate the required safety features functions. This system also monitors the power sources acting to assure the availability of emergency power for operation of at least the minimum ESFs (Chapters 6 and 8). This system is provided with the necessary redundant circuitry and physical isolation so that a single failure within the system would not prevent the proper system action when required. This system is provided with test facilities and alarms to alert the operator when certain components trip, malfunction or are inoperable. The controls are designed to automatically provide the sequence of operations required to initiate ESF operation with or without off-site power available. There are no RPS and ESFs instrumentation transmitters for which the trip setpoints are within 5% of the high or low end of the calibrated range, or within 5% of the overall instrument design range.

7.2 REACTOR PROTECTIVE SYSTEM 7.2.1 GENERAL 7.2.2 DESIGN BASIS

7.2.3 SYSTEM DESCRIPTION

7.2.4 SIGNAL GENERATION 7.2.5 LOGIC OPERATION 7.2.6 TESTING 7.2.7 SYSTEM EVALUATION

7.2.8 POWER SUPPLY

TABLE 7-1 REACTOR TRIP ALLOWABLE LIMITS AND PRETRIP LIMITS NO. REACTOR TRIP PRETRIP ALARM LIMIT PRETRIP ALARM LIMIT TABLE 7-2 REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES (BOTH UNITS)FUNCTIONAL UNIT RESPONSE TIME

CALVERT CLIFFS UFSAR 7.6-1 Rev. 47 7.6 OPERATING CONTROL STATIONS 7.6.1 GENERAL LAYOUT The operating control stations consist of: (a) the Control Room, used for plant control during startup, normal operation, shutdown, and emergency operation; (b) an auxiliary control station used for the waste processing systems (common to both Units 1 and 2); and, (c) various local control stations for miscellaneous noncritical systems. 7.6.2 CONTROL ROOM The Control Room, which is accessible from both the Auxiliary and Turbine Buildings, houses benchboard control boards, the combined benchboard-vertical control boards and miscellaneous vertical boards for both units. All control boards are designed to Seismic Category I requirements. For further information concerning seismic loading and design, refer to Figures 2.6-4 and 2.6-5, Response Spectra - Operating Basis (OBE) and Design Basis Earthquake, respectively. The Control Room can be occupied under all credible incident conditions. It has two separate air conditioning units, two particulate, absolute, and charcoal filter unit assemblies with dampers and fans, and an airborne radioactivity detector in the return line. Filter unit dampers and fans, which act to automatically shunt a portion of the Control Room air through the filter unit assemblies, actuate upon sensing a high airborne radioactivity level or in response to a SIAS initiation from either unit. The filter unit dampers and fans can also be remotely actuated from the Control Room.

None of the materials used in the construction of the Control Room will support combustion, and electrical wiring is flame resistant. Also, portable CO2 fire extinguishers are placed in readily-accessible stations in the Control Room, and respiratory protective equipment is available to the operators at all times. 7.6.3 RADIOACTIVE WASTE DISPOSAL SYSTEM CONTROL PANELS The waste-processing local control panel, located in the Auxiliary Building, provides the controls, instrumentation, and alarms required to initiate, operate, and monitor the waste-processing systems. Critical indications are duplicated in the Control Room. All alarms are annunciated at their local panel with a master alarm provided in the Control Room. 7.6.4 MISCELLANEOUS LOCAL CONTROL PANELS Local control panels for noncritical systems are located throughout the plant. Each panel contains the indications, controls, and alarms required for safe operation of the system. The various systems are provided with local alarms and a common alarm on the control board to alert the operator of any abnormal conditions within each of the systems. 7.6.5 FEATURES WHICH ENHANCE SAFE OPERATION In order to maintain channel separation, the control boards contain fire barriers that separate the ESF and control channels, thus preventing loss of all protection due to a single fault.

The plant annunciator system is located across the top of the main control boards, providing visual and audible indication of abnormal conditions which require operator action. 7.6.6 REMOTE SHUTDOWN CAPABILITY Numerous design features are provided to maintain Control Room accessibility. However, in the event the operator is forced to abandon the Control Room, emergency procedures CALVERT CLIFFS UFSAR 7.6-2 Rev. 47 require that the reactor shall first be tripped. During this condition, local controls and the Auxiliary Shutdown Panel, located in the 45' Elevation switchgear rooms (immediately adjacent to the Control Room), provide the instrumentation and controls necessary to safely bring the plant to the hot shutdown condition.

The Auxiliary Shutdown Panel, as supplemented by local control panels, provides the remote shutdown capability required by 10 CFR 50.48 and 10 CFR 50, Appendix R. No automatic safety features are actuated from remote shutdown monitoring instrumentation. Electrical isolation devices are installed such that a fire at the Auxiliary Shutdown Panel will not prevent shutdown of the plant from the Control Room, and vice versa. Locally available instrumentation and the instrumentation available at the Auxiliary Shutdown Panel (1C43) ensure it will be possible to perform the following functions, and monitor their effectiveness, from areas external to the Control Room. a. Insert the CEAs and trip the turbine generator; b. Borate the RCS; and,

c. Remove reactor decay heat following a reactor trip. These instruments include: INSTRUMENT READOUT LOCATION MEASUREMENT RANGE Wide Range Neutron Flux 1C43 0.1 cps-200% power Reactor Trip Breaker Indication Cable Spreading Room OPEN-CLOSE Reactor Coolant Cold Leg Temperature 1C43 212-705°F Pressurizer Pressure 1C43 0-4000 psia Pressurizer Level 1C43 0-360 inches Steam Generator Pressure 1C43 0-1200 psig Steam Generator Level 1C43 -401 to +63.5 inches Additional communications capability is provided at the control stations for use in accomplishing these functions.

It is possible to maintain the plant in a hot shutdown condition from these alternate locations until access is permitted back into the Control Room. If required, the plant can then be placed in a cold shutdown condition from the Control Room.

Instrumentation required by the reactor operator to monitor key safety parameters from outside the Control Room is specified in the Technical Specifications.

CALVERT CLIFFS UFSAR 7.7-1 Rev. 47 7.7 CONTROL ROOM ANNUNCIATION Visual and audible alarm units are incorporated into the control boards to warn the operator if limiting conditions are approached or off-normal conditions exist for any system. These visual and audible units are supplied in two stages: (1) annunciator enclosures supplied as part of the original installation; and (2) status alarm panels.

Status alarm panels are supplied in the same area as the existing annunciator panels to provide for additional alarm points. The panels furnish annunciation for one or more systems. A "summary alarm" window is included in the existing annunciator for each one of the systems contained in the status panels. The "summary alarm" has "reflash" capability to warn of additional incoming alarms on the status panel and does not clear until all status alarms are cleared. These status panels are mounted on the associated system panel or immediately adjacent to it. A list of all annunciator windows, legends, and alarm initiating devices is maintained in controlled BGE drawings.

CALVERT CLIFFS UFSAR 7.8-1 Rev. 47 7.8 COMMUNICATIONS 7.8.1 DESIGN BASIS A communication system with multiple redundancy has been provided to ensure availability and ease of operation. 7.8.2 COMMUNICATION SYSTEM DESCRIPTION The communication system consists of six subsystems: a. Plant Public Address (PA); b. Commercial Telephone; c. Sound-powered phones for plant use; d. Sound-powered phones for emergency use;

e. Microwave system; and,

f. Radio telephone system. 7.8.2.1 Public Address System The primary plant PA system utilizes the site-installed Northern Telecom administrative telephone system. The site is divided into five zones. Each zone can be accessed individually from any telephone on the site. An "ALL CALL" is available on certain site telephones that allows all five zones to be accessed simultaneously. Priority paging is also available on certain telephones that allows any individual zone or "ALL CALL" to be accessed while also overriding any non-priority page in progress. The Control Room has telephones that are able to access all five zones simultaneously and override any non-priority or priority page in progress. The plant emergency alarms are also generated by this equipment.

The primary plant PA system is powered by diesel-backed instrument bus feeder 2Y1081 via 1X61 and 1P61.

7.8.2.2 Commercial Telephones The site administrative telephone system is located in the North Service Building. This is a stand-alone system with alternative call routing that allows the site to be independent of the C&P Prince Frederick public exchange. This telephone system allows an individual to place a call within the plant, throughout BGE, or outside the company. The system is capable of routing calls throughout the Prince Frederick, Baltimore, or Annapolis public exchanges on C&P-provided facilities. The telephone system can also route calls to the public exchanges via the BGE microwave facilities into Baltimore. The Control Room has the capability of activating all remote links simultaneously in order to complete notification requirements for all agencies identified in the Emergency Response Plan.

The administrative telephone system receives its power from Motor Control Center MCC-101AT. Power is, therefore, available from either the main generator or the emergency diesel generator. The telephone system also has an emergency battery backup located in the North Service Building to ensure a communication system independent of the plant status.

7.8.2.3 Sound-Powered Phone (Plant Use) The sound-powered phone system is set up in portions of the plant where unbroken communications are needed for certain operations or maintenance. The system consists of a hard wired network with covered jacks at various stations. Phones, headsets, and handsets with extension cords are taken to these stations for remote operations and control communications. CALVERT CLIFFS UFSAR 7.8-2 Rev. 47 7.8.2.4 Sound-Powered Phone (Emergency Use) A backup system, completely redundant and maintaining physical separation from the first provides communications capability between the Control Room and areas of the plant, including the interior of containment in the event of loss of normal communications during a fire.

7.8.2.5 Microwave Communication System Automatic ringdown phones are located in the Control Room and the 500 kV switchyard and are connected to the microwave system. The signals are sent to the antenna in the 500 kV switchyard and are transmitted via microwave radio propagation to the electric system load dispatcher at the Electric Operations Building in Baltimore. The microwave system also relays radio telephone communications to radio-based stations and microwave receivers off site. Some of the normal telephone traffic between Calvert Cliffs and Baltimore is handled by the microwave system. There is one microwave channel coming into Calvert Cliffs from the Electric Operations Building, which is used by the load dispatcher to contact all BGE generating stations simultaneously. This "ALL CALL" channel does not carry voice communications away from the plant.

7.8.2.6 Radio Telephone System (Plant Use) The radio telephone system is a system of base stations and repeaters at the plant linked to the Emergency Operations Facility and base stations remote from the plant. The microwave communication system links the onsite and remote equipment. A primary radio system capable of plant wide radio communications, including communications within each containment, is installed. Through use of repeaters, outside antennas and an indoor continuous antenna, communications among control consoles, hand-held portables and nearby mobile units is possible. The Emergency Operations Facility and, state and local emergency response agencies are included in communications capabilities.

A single channel 150 MHz system is capable of direct radio contact with the Electric Operations Building. This capability is provided to ensure offsite communications in the unlikely event of a simultaneous commercial telephone line and microwave system failure. 7.8.3 RELIABILITY AND TESTING Systems of the types described above are conventional and have a history of successful operation at existing BGE Plants. Most of these systems are in routine use and this will assure their availability. Those systems not frequently used are to be tested at periodic intervals to assure operability when required. CALVERT CLIFFS UFSAR 7.9-1 Rev. 47 7.9 DELETED

CALVERT CLIFFS UFSAR 7.11-1 Rev. 47 7.11 ANTICIPATED TRANSIENT WITHOUT SCRAM Anticipated Transient Without Scram is an anticipated operational occurrence followed by the failure of the reactor trip portion of the protection system. This protection system automatically initiates the operation of systems, including the reactivity control systems, which assure that specified fuel design limits are not exceeded as a result of anticipated operational occurrences. Some examples of these occurrences are: loss of power to all reactor coolant pumps in a unit, loss of load, and loss of offsite power (Section 14.1.1.1).

Protection against ATWS events is comprised of three elements: DSS, DTT, and a diverse AFAS. These are all requirements of 10 CFR 50.62. The first two, DSS and DTT, are discussed in this section; AFAS is discussed in Section 7.10. 7.11.1 DIVERSE SCRAM SYSTEM The purpose of the DSS is to provide reactor trip capability that senses high pressurizer pressure and will function separately from the primary reactor trip system (Section 7.3). 7.11.1.1 Design Basis The DSS provides diversity from the existing RPS, electrical independence from sensor output to the final actuation device, isolation of non-safety-related from safety-related circuits, testability at power, environmental qualification for anticipated operational occurrences, and a design to prevent against inadvertent actuation and challenges to other safety systems. a. Facility electrical separation is maintained through the use of existing circuitry in the ESFAS sensor, logic, and relay cabinets, which provide physical separation for four sensor channels and two logic and relay channels. b. Seismic concerns are met by utilizing existing equipment already in place within the ESFAS sensor, logic, and relay cabinets. c. At-power testing is accomplished with the use of a bypass contactor in parallel to the existing CEDM motor-generator load contactor. d. Redundancy is not required by 10 CFR 50.62. Both channels of the DSS must trip, opening both of the CEDM motor-generator load contactors in order to cause a reactor trip. e. Diversity is provided between the RPS and the DSS through the use of equipment supplied by different manufacturers or designed differently to provide the same function. The DSS interrupts power to the CEDM power supplies by opening the motor-generator load contactors, while the RPS uses the reactor trip switchgear to interrupt CEDM power. The same sensors are used by the DSS and the RPS for pressurizer pressure which is acceptable by 10 CFR 50.62. Sensor output for DSS is made diverse from RPS, as specified by 10 CFR 50.62, through the use of an electronic isolator. f. The DSS is powered by inverter feed which are AC power sources that also supply the RPS. The use of common power supplies is acceptable for the DSS and the RPS sensors as they are not within the scope of the ATWS rule, 10 CFR 50.62. The DSS power supplies for each of the four DSS protection channels are independently breakered and fused from a different vital bus. This isolates the DSS power supplies from each of the vital buses in order to prevent common mode failures. g. Environmental qualification to accident conditions is provided for DSS equipment installed in the ESFAS cabinets. CALVERT CLIFFS UFSAR 7.11-2 Rev. 47 7.11.1.2 System Description The DSS is a four channel sensor system which through two-out-of-four logic inputs to two actuation channels. Each actuation channel opens one of the two load contactors on each CEDM motor-generator. Both load contactors on each CEDM motor-generator must open to cause a reactor trip.

The four sensor channels consist of pressurizer pressure sensors (PT-102A, B, C, D) and associated circuits. The output of the sensors, through isolators, provides pressure signals to four high-trip bistables in the ESFAS sensor cabinets. Each bistable provides channel trip annunciation, input to a two-out-of-four logic module in channel "A" of the ESFAS cabinet and input to a two-out-of-four logic module in channel "B" of the ESFAS cabinet. The logic modules energize a relay in each of the ESFAS relay cabinets to open the CEDM motor-generator load contactors. Both channels must actuate to initiate a reactor trip. At-power testing is provided through the use of a bypass contactor for the channel in test. The bypass contactor is in parallel with the load contactor and prevents the loss of output when the load contactor opens during testing. Due to the fact that the DSS is not available while the system is in bypass, administrative control will limit the time that the system may remain in bypass.

Annunciation is provided on 1(2)C05 for both "DIVERSE SCRAM SYSTEM TRIP" and "DSS LOAD CONTACTOR BYPASSED." 7.11.2 DIVERSE TURBINE TRIP Main turbine trip circuitry consists of four safety-related instrument control channels which sense CEDM power bus undervoltage. The DSS provides a diverse means of deenergizing the CEDM power bus. This satisfies the ATWS requirement for diverse means of main turbine trip. CALVERT CLIFFS UFSAR 7.12-1 Rev. 47 7.12 ENVIRONMENTAL QUALIFICATION OF ELECTRIC EQUIPMENT IMPORTANT TO SAFETY Equipment used to perform a necessary safety function must be demonstrated to be capable of maintaining functional operability for the time it is required to operate, under all service conditions postulated to occur during its installed life. This requirement, which is embodied in General Design Criteria 1 and 4 of Appendix A and Sections III, XI, and XVII of Appendix B to 10 CFR Part 50, is applicable to equipment located inside as well as outside containment. More detailed requirements and guidance relating to the methods and procedures for demonstrating this capability for electrical equipment are contained in 10 CFR 50.49, "Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants," NUREG-0588, "Interim Staff Position of Environmental Qualification of Safety-Related Electrical Equipment" and "Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors" (DOR Guidelines). On January 14, 1980, Nuclear Regulatory Commission (NRC) issued Inspection and Enforcement Bulletin (IEB) 79-01B which included the DOR Guidelines and NUREG-0588 (for comment version). Subsequently, on May 23, 1980, Commission Memorandum and Order CLI-80-21 was issued and stated that the DOR Guidelines and portions of NUREG-0588 (for comment version) form the requirements that Calvert Cliffs must meet regarding environmental qualification of safety-related electrical equipment. Supplements to IEB 79-01B, issued on February 29, September 30, and October 24, 1980, NUREG-0588, Revision 1, dated July 1981 and NRC Generic Letter 82-09, dated April 20, 1982 provide further clarification and definition of the NRC's requirements.

A final rule on environmental qualification of electric equipment important to safety for nuclear power plants became effective on February 22, 1983. This rule, Section 50.49 of 10 CFR Part 50, specifies the requirements to be met for demonstrating the environmental qualification of electrical equipment important to safety located in a harsh environment. In accordance with this rule, equipment for Calvert Cliffs may be qualified to the criteria specified in either the DOR Guidelines or NUREG-0588, (for comment version) except for replacement equipment. Replacement equipment installed subsequent to February 22, 1983 must be qualified in accordance with the provisions of 10 CFR 50.49, using the guidance of Regulatory Guide 1.89, Revision 1, unless there are sound reasons to the contrary.

The Calvert Cliffs Environmental Qualification Program complies with these requirements.

Rev.07-1 REACTOR PROTECTIVE SYSTEM - BLOCK DIAGRAM* ::::0 en Q) (") -0 ..., ""tJ a -CD (") -<" er v. -CD 3 ...... I C.C ..... c ..., C't> INPUTS FROM NSSS MEASUREMENT CHANNELS TRIP UNITS 1234567891011 -LOGIC MATRICES LOGIC MATRIX RELAYS INPUTS FROM NSSS MEASUREMENT CHANN.fil_ I -POWER LEVEL 2-RATE OF CHANGE OF POWER 3-REACTOR COOLANT FLOW 4-STE:AM GENERATOR WATER LEV£LS

• 5-STEAM GENERATOR PRESSURES 6-PRESSURIZER PRESSURE 7-THERMAL MARGIN 8-WSS OF LOAD 9-CONT'AINMENT PRESSURE 10-Axial Flux Offset 1-Asymetric Steam Generator Load
• Same Transmitter TO 120Voc VITAL INSTRUMENT BUS 01 TRIP PATHS MANUAL TRIP "*? CEOM POWER SUPPLIES CONTROL ELEMENT l'lRIVE MECHANISMS '
• 1 2 3 4 5 6 7 89 10 11 tlLJ.jj._U,L/ CHANNELS
• TO 120Vac VITAL INSTRUMENT BUS *'4 Revision 43 Rev.27-2A REACTOR PROTECTIVE SYSTEM - INTERFACE LOGIC DIAGRAM** '"""""':
• Rev.157-3 PRESSURIZER PRESSURE MEASUREMENT CHANNEL - FUNCTIONAL DIAGRAM* ""O ... CD (I) (I) c ... N. CD ... "Tl ""O c ... :J CD (') VI ..+VI a* c :J ... Ill (I> s::: 0 (I> iii' Ill (Q (I) ... c Ill ... ,3 (I> :J ... 0 ::J" Ill :J :J (I>
• f'h£&SURIUR H£AM !!!!!! lllST flUMl!}i. fflO TD .1n eu 120 \' o* VllAL
• Rev.207-4 EX-CORE NUCLEAR INSTRUMENTATION(') 0 ro z c Q. (I) Q) ..., 5' 2 3 (I) ::J -6' ::J
• Wide Range Logarithmic Channels 8 % to z.oo:Zrower ----,..:.. ---. -------AMP AMP AMP HV HV HV Power Power Power Supply Supply Supply Log Amp Log Amp Log Amp Rate Rate Rate of of of Change of Change of Change or Power f Power Power. I I Outputs to Reactor Protective System AMP HV HV Power
• Power Supp.ly Supply Log Amp Lin Amps Rate of Change or Power level Power t
• Linear Range Nuclear Instrumentation Channels 0.1 % to Power ----HV HV Power Power Supply Supply. Lin Lin Amps Amps Power Power Level Level *
• 1* Outputs to Reactor Protective System HV Power Supply Lin Amps Power Level + HV Power Supply Lin Amos Power Level + Neutron Flux Detectors Containment Wall HV Power Suppl Signal Lin Ams Processing Instr Power level '---v---/ Outputs to Reac1br Regulating System '---v / *Safety Channels Control Channels
• Calvert CliffsNuclear Power PlantLOW FLOW PROTECTIVE SYSTEM -FUNCTIONAL DIAGRAMFigure 7-5Rev. 31 Calvert CliffsNuclear Power PlantTHERMAL MARGIN/LOW PRESSURE TRIPCHANNEL - BLOCK DIAGRAMFigure 7-6Rev. 31 Calvert CliffsNuclear Power PlantTHERMAL MARGIN/LOW PRESSURE TRIPFUNCTION DIAGRAMFigure 7-6ARev. 31 Rev.07-6B ASYMETRIC STEAM GENERATOR TRIP (ASGT) - FUNCTIONAL DIAGRAM*
• If I ilPSG I > ASGT THIP scr. ASGT TBtP"" 2!.ioo PSI. ELSE ASGTTIHP = o Ps1 1F I ti.PSG I >ASGT l'BETHIP* ASGT PTRP -2500 Psi. ELSE ASG rPHETBIP -o l'Sl STEAM GENEHATOH PHESSUllES LWSG r TBIP SET MAX S[L tWSG r11 ET nu* SET ) N0.1 PSG1 1 r-t:I NO. 2 )f-PSG2 > :J TE ST <.q.-11 SET ) I I ASGT I llST I TESTl'H l,-.-1-:. _I -I 4-NEXTi* TEST PS ._L_ NlXT _ FIG. 7-6F Asymetric Steam renerator ( ASGT*) -Functional Diagrarr
• COMP 25000J ASGTTRIP ( b. -coMP 250:s ASGTprnP ( ;:i ;: . **-6A)

Rev.117-7 NEUTRON FLUX MONITORING SYSTEM STARTUP AND LOGARITHMIC RANGE CHANNELS* *

• CHANNEL 11811 DETECTOR ASSEMBLY (FISSION CHAMBER a PROPORTIONAL COUNTER) PENETRA_r_io_N_s _____ r= __ WALL -,c:::::r-----1 ---........ --PowER-I 120VAC ISOLATOR AMP AMPLIFIER OPTICAL ISOLATOR SIGNAL PROCESSOR PLANT COMPUTER CPS/POWER INDICATOR POWER INDICATOR BALTIMORE GAS a ELECTRIC CO. CALVERT CLIFFS NUCLEAR POWER PLANT I INDICATORS .---S::-:l-:-GN'."':'"AL:-:-6"""____,1---12 0 VAC I PROCESSOR --1 HAND SWITCH L FROM CH.D . _J AUXILARY SHUTDOWN PANEL (CH.Aas ONLY) RPS AUX LOGIC RPS SU RATE RPS RATE OF POWER CHANGE NI ALARM NI CH. INOP ALARM SU R.41"E . '------ti* TRIP ALARM TYPICAL FOR CH. A,B,C 8 D EXCEPT AS NOTED NEUTRON FLUX MONITORING SYSTEM STARTUP AND LOGARITHMIC RANGE CHANNELS FIGURE 7-7 REV. I I 1/91 Rev.07-9 SAFETY INJECTION ACTUATION SIGNAL* *
• low-L.oW Pressurizer Pressure or SIAS High Containment Pressure SIAS S.IAS Can Be Blocked Manually to Provide for System Maintenance and Shutdown Oepressurization. Block Is Automatically Removed Above 1 Preset Pressurizer Pressure
• GAS &. ELECTmc co. Safety Injection Actuation Signal Nnrkar Powt.'r Plant Figure 7-9 Revision 49 Revision 49 Rev. 327-11 REACTOR REGULATING SYSTEM-BLOCK DIAGRAMA 8 c D E I LOOP 1 LOOP 1 LOOP 2 T,. LOOP 2 Tc EXTERNAL. S[GNAL. INPUT TURBINE FIRST ST A.GE PRESSURE POWER RANGE NEUTRON FLUR !IS 114 II 1 TO MAIN CONTROL BOARD LIGHT w 0 v. t--..-----, TO MAIN CONTROL BOARD LIGHT 2 S4 rl CAL CHECK INPUT EXT 10PERATE I GL.! I EXT kl1 TEST RRS TEST 2 ol I 01 o--.J.-L_ TO MAIN CONTROL BOARD 3 .... ...._ K-1 !2 K* -2 K*O 519.5 TURBINE FIRST ST AG£. IDOX '12$X PRESSURE CTURSINE POWER> STEAM DUMP AREA DEMAND LSP TA.VG -TREF B 4 TAVG-TREF H[GH-1..0W ALARM STOP. THINK. ACT AND REVIEW A STEAM DUMP REFERENCE DRAWING
• AREA DEMAND D-8067-41.3-021 STEAM DUMP REFERENCE DRAWING QUICK OPEN 0-8067-413--021 PRESSURIZE REFERENCE DRAWING
• LEVEL SET PQ[NT 12132-0025 O> .., 0 CONT ACT OPENlNG ON HIGH AL.ARM * ? TAVG-TREF HIGH REFERENCE DRAWING N ALARM & AWP 61087SH0004 CONT ACT OPENING ON LOW AL.ARM * <!> TAVG-TREF LO 8 z REFERENCE DRAWING ; ALARM < 610B7SH0004 a:: Q :::E ..... ..... I-(,} NOTE .3 <n z >-LL.I <n Ct:: LL.I SPEED REFERENCE DRAWING (.!) u. ....., CONTACT Ct:: D-8067-413-450 I-< :I: ..J ::::> <.!) ; u.J a:: (/) a:: a::: 0 0 ,_ I-0 ..... w ...J STATUS NOTE 3 a:: ..... STATUS REFERENCE DRAWING c ell SIGNAL 0-8067-413-450 NOTES: 1. TIME IS ZERO WITH S1 8t S2 SWITCHES OPEN 2. s DUPLICATE INPUTS FROM SYSTEM NOT SHOWN 3. AUTO MOOE FOR CEA DRIVE IS INOPERATIVE. RELAYS WERE REMOVED D THIS DRAWING WAS REDRAWN FROM COMBUSTION ENG. INC., D-8067-413-001. CCNPP F.P. NO. 12017-0101, REV. 3. DATED 08-12-92. REV OATE 3 4 3 DESCRIPTION AS BUILT INFO ONLY ROC 1R9100258 AS BUILT DCN 12017--0101-2001SH0001 ES199901317-000 4 DWN DSGN DE IR APPROVED t.AWE GPK KAK SA BLOCK DIAGRAM REACTOR REGULATING SYSTEM SCALE: NONE INTER NO. 7"1887.0GN 0-806 7-41 3-001 OWG. NO. 12017-0101 E 4 6 LAST USER (£.A.McCOY I LAST DA.TE WORKED 110/10101 I Rev.267-12 CEA POSITION SETPOINTSfully Wittidrawn 137 T CD UC: .... ID =co ..... j_ "C s=. -;:: -0 ...., Q,) s=. ._, c *--c: c ::::::: ..,, 0 0.. < UJ u ...J -c :z -:::E 0 :z Fully 0 Inserted BALTIMORE GAS & ELECTRIC CO. Colvert Cliffs Nuclear Power Plant Shutdown CEA's Regulating CEA's . Upper Elec Upper Elec 136 Limit Limit CEA 136 Sop i36 CE Wi drowa 128 Upper Interlock Sequential -80 LOWER Permissive EXERCISE LIMIT* Lower Sequential Permissive Pre-Pa.ver Dependent Insertion -so Alarm (variable) Power Dependent I Insertion (variable) Alarm ShutdCJNn CEA Insertion Permissive 10 Lower CEA Lower Elec Group Stop l limit l 1 Dropped CEA 0 ---Lower 0 Elec Limit Dropped CEA CEA POSITION SETPOINTS REV. 26 Figure 7-12

NOTE 1: Technical Specification 3.3-4, Table 3.3.4-1

NOTE 2: UFSAR Table 7-1

Calvert Cliffs Nuclear Power Plant PRESSURE CONTROL PROGRAM Figure 7-13 Revision 39 FIGURE 7-14A FEED WATER CONTROL SYSTEM -BLOCK DIAGRAM, UNIT 1 ----------------------------, ! ( ""-.!1.f '( l _____ ..... TUU( I.. rs ! i@l : : 1\-1><>"""-,-.:>r' --0i '--: ! ;---J : n&'ftit! 1f'5 CllUERTtLIFFSIUUMl'OVERl'lt.NT IESM FIWIE 7*14A llCiE l*WllG NIA Revision 49 r*--------**-**--**-' FIGURE 7-148 FEED WATER CONTRlll SYSTEM-BLOCK OIAGRIM. UNIT 2 OvtRRIDE A 0 'o/ c ' ' ' * ---------------------------------------------------. 1------------------------------------------------, I I I I I I I I I I I I I I : : l j 10 UAIH STE"' IQO£R 10 llAIH SJE.W 1£11l£R i ; L ______________ ,..J _______ ___ PC STEAM rLow : ! * *

• re ---r------------------r-.----------... : : : r l : i : : ! 1;1 i I I I -@---, : : @' : lj: t::, .,*:,: : @' : i r--lii>--sTEAM : : : rR : ! : : a : : * : r---t><l-1 GENERATOR 1--t><::J-+--. :------:---r--:-------------'?,\V : ! ! ------------.. ---::-:-----: r-+-C><l--i i ----11.RBIHE TRIP l. fll.LRNG: J. ' ' ' ' r-----.----i @---------: -, ' ' :$ ' ' ' ' ' I I I 1S 1fJ11 : : :!::g: : l ! ! !!I!!-! STEAM GENERATOR Tl..RBD£ TRIP----O& OV[RRJDE c 'o/ ' ' ' 1-----@ l----------------*------------_J ' ' '

' ' ' ' ' ' : : rRQu UTR. DRIVEN Mil, fEEOWAT(R PIJUPS ro rROW STE.AM IJUVEN AUlC. rtEDWATER PUtiPS rAJ ILOSS AJfU 4-------1----------.... : ,.c:,----:::c--i><I--\ SIGNAL 0.2$,S 1-----1 ll[Q.\Al'Ot i AIS : i : w d ' FO ' ' ' ' ' ' ------! ' rROU sn: .... rROU UTR ORIVEN AUX f((OWAT(R Pl.IFS CALVERT currs NUCLEAR POWER PLANT IJ"SAR FIGURE 7-148 fEEOWATER 8GE DRAWING N/ A REVISION 38 Rev.157-15 BLOCK DIAGRAM - STEAM DUMP AND BYPASS SYSTEM (Sheet 1)* T AVG STEAM DUMP CONTROLLER TC 0 MANUAL SET POINT (532 F) *

• MANUAL SET POINT (900 PSIA), f c TURBINE TRIP AUXILIARY RELAY ;; CONTACTS SHOWN IN TRIPPED r REACTOR REGULATING POS I TI ON SYSTEJ.f K-7 STEAi DUMP OU I CK-OPENING OVERRIDE BISTABLE POWER SUPPLY --r--1 I I I I I I STEAM DUMP AUXILIARY RELAY CONTACTS SHOWN IN .. QUICK-OPENING" POSITION D.C. POWER SUPPLY -*-ro SH.Z. lO HIC SH.2 ALARMS WHEN EITHER HIC STATION IS ON MANUAL c l/11 OUTPUT SIGNALS FROM I/P'S TO CONTROL VALVES ARE STAGGERED (SPLIT RANGED) TO PROVIDE SEQUENTIAL OPENING OF TURBINE BYPASS VALVES A/S BLOCK DIAGRAM STEAM DUMP AND. BYPASS SYSTEM FIGURE 7-15 5HEET I TURBINE BYPASS TO CONDENSER VALVES (TYP. 4 VALVES) REV. 15 Rev.37-15 BLOCK DIAGRAM - STEAM DUMP AND BYPASS SYSTEM (Sheet 2)* r" ALTERNATE SHUTDOWN PANEL SEE SHT. I H MAIN CONTROL I ROOM c A/S AO FC
• SEE SHT. I A/S AO FC ALTERNATE SHUTDOWN PANEL
• FIGURE 7-15 SHT. 2 REV. 3

DUMP VALVE POSITION SIGNALCalvert Cliffs NuclearPower PlantPROGRAM OF STEAM DUMP BYPASS VALVES & QUICKOPENING OVERRIDESheet 3Figure 7-15Revision 37 Rev.07-16 CONTROL ELEMENT DRIVE SYSTEM-BLOCK DIAGRAM*( I -, * (-r I I I I I I I I I I I I I IA/rNWPIAIN. MilDUU'& } 7" r:> .. _,.,...,.,.,o,..Aa ,...-.-.. Powt'I! SHR,11'/NG COii. PC<<P P.eo&eMtME.es ---1 f.ft;...C7i:w-I HEAD-==VE 11 ;o Sr"STINI Z) 1-----,, _J L _ _J 1_r* -r=--TO &M. .,.._,..,, ,_ .TAWFN* t;;llU,,,,, Ct:W. '°wa P-1._-+-+--} ... , ,_,,,,-. c ... .<.,..: ,.,,D4Nlr 7P CF.4 ""'4!)/:liAta 110., .. -c i I i I I I I Q INNJafT MfNIVl.£ CNM-CIGITA.L TM/LR-TMERMAL MAAGJN/i.DN PR&a.J ... SUR-START u=t R'ATe. QPeR'A'TION MOOE!!t. Ml* MANl.W.. 1.NOtV'lCUAL "-G* MANUH.. QACUP 81!.QUeN'TIAI. .. --rtc CU.-cQllTllO\, IL&MINT A,._ ,.,_.

• COi&. '°""'El(. Pe::>&£JlllMH4C B LTIMORE GAS & ELECTRIC Co. Ca vert Cliffs Nucle r Power Plant CONTROL ELEAilENT DRIVE SYSTEM BLOCK DIAGRAM I -Figure 7-16 Rev.07-17 STEAM GENERATOR PROTECTIVE SYSTEM MEASUREMENT CHANNEL Rev.07-18 SCHEMATIC LOW STEAM GENERATOR PRESSURE BYPASS Rev.07-20 VARIABLE HIGH POWER TRIP OPERATION Calvert CliffsNuclear Power PlantDT POWER CALCULATIONFigure 7-21Rev. 31 FIGURE 7-22 LOGIC DIAGRAM - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM UNIT 2 FIGURE 7-22 LOGIC DIAGRAM - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM UNIT 2 Rev.137-24A LOGIC DIAGRAM - AUXILIARY FEEDWATER ACTUATION SYSTEM (AFAS) UNIT NO. 1* *
• 5TE"M GENERATOR NO 11 HIGH_LEVEL HL) I I I r I I I I I I I I I I I I SG.1r I "' I I I I !-'ANUAL STEAM GENERATOR NO.II
• L£\IEL zo ZE ZF uENFRATCR NO 11 LOW LEVEL .LL\ I I I I I I I I 1 ___ _ _>,Ell..§. ____ _J r---------1 I I I Pl2*Pll I Al : MAHIJAL I I : ,,_ __ : f D. : '--------------_J o TEAM GENERATOR PRES>uRE if -T T [J+-T AUX FFfDWl.W R START ST&<M GENERATOR N012 LEVEL ZO ZE '"' ?TEAM GElllERATQI; r,o .!.2 LOW LFVtL!LcJ ----Rf:ET -I * .....CAL I ---, MCR I y I I I I L ______ ___ j >TEOM NQ.12 0RE55URE ZE I I I I I G.:1 1t.P1:XTC ---rll H Fl/£ I R I SC.12. t_ ___ MCR " v STEAM GENEF<ATOR DIFFERENTIAL PRESSURE S!O:IZ F.(lelOCK(Bl.-AJ) ,:_ __ i ____ .-:....J OIFFERl:.NTl"I.. PR£SSua,t: STEAM GENERAJ"OR ti ':J STE.1"1 Gi:NER ... TCA t2 ------, ' JSI I MAllUAL t I I I I I I I SU.It . I t_ _ _____ J BALTIMORE GAS & ELECIB!C COMPANY DRAWING NO. 61-060 SH. 1 (1-L0-60) REVISION 3 NO 12 HU I I I I SGl2. : A MCR v LEGE'. ')to NO TEST DURING. CPi:""""'-IEJ &!STABLE -HI 5ET FOINT f.Il 81STABLS-Sc.-rPOusr 13:) COMPARATCR-a1ST.6.BLE * '-41 SET POINT C:0 LO>< !ET l'Ol'IT F
• Fl..NCT1CN (SS-51).Rf l sP-5TOP; 0-0PEN, C*CLOSE:, 3L.*at..oc.K) T TEST (LOCAL, PVSMBUTTON) FAC F'A.1:11.ITV CODE ISOLATION OE.VICE @ .b..UO.BLE VISIBLE. OE.\'JC.E INDICATING LIGHT-R-REO (LOCAL) C£h) DAT"-\.OGG£R-::!.E.MOTE
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CALVERT CLIFFS UFSAR 8.1-1 Rev. 47 ELECTRICAL SYSTEMS 8.

08.1 INTRODUCTION

The electrical systems include the equipment and systems necessary to generate power and deliver it to the high voltage system. They also include facilities for providing power to, and controlling the operation of, electrically-driven plant auxiliary equipment and instrumentation. Essential instrumentation, including the Reactor Protective System (RPS) and the Engineered Safety Features (ESF) instrumentation, is fed from vital instrumentation busses to provide continuous monitoring and control. The plant batteries provide circuit breaker control, Control Room emergency lighting, vital instrumentation power, and operating power for certain other equipment. 8.1.1 DESIGN BASIS The plant electrical systems are designed to ensure a continuous supply of electrical power to all essential plant equipment during normal operation and under accident conditions.

All electrical systems and components vital to plant safety, including the emergency diesel generators (EDGs), are designed as Class 1E so that their integrity is not impaired by the Safe Shutdown Earthquake (SSE), high winds, or disturbances on the external electrical system. 8.1.2 DESCRIPTION AND OPERATION The plant electrical system is shown on Figure 8-1, Main Single Line Diagram. In order to achieve maximum reliability and efficiency of operation of the electrical systems, the following criteria are employed: a. The main generators, described in Section 10, feed electrical power at 25 kV and 22 kV for Units 1 and 2, respectively, through forced air cooled isolated phase busses to two main unit transformers installed per unit. b. Plant auxiliary sources of power are the two 500 kV/14 kV plant service transformers, which are fed from separate 500 kV switchyard busses and a 13 kV line from the Southern Maryland Electric Cooperative (SMECO) system. Each 500 kV/14 kV plant service transformer is capable of supplying the total (two unit) plant auxiliary load. The 13 kV SMECO line is capable of supplying the power required to maintain both units in a safe shutdown condition. It may be substituted for one of the 500 kV/13 kV circuits as one of the two required, physically independent, offsite circuits. c. The two plant service transformers feed six 13.8 kV 2MVA +/- 10% voltage regulators which feed six 13.8 kV/4.16 - 4.16 kV service transformers, three of which are capable of supplying the total plant 4.16 kV auxiliary load. d. The 13.8 kV system consists of multiple reactor coolant pump (RCP) busses, each of which can be fed from either of the two plant service transformers. Two 13.8 kV service busses (with tie breakers) are provided for distribution to the voltage regulators and unit service transformers. e. The plant is split into two independent load groups, each with its own power supply, busses, transformers, loads, and 125 Volt DC control power (Figure 8-9). f. A reserve 125 Volt DC system, capable of replacing any of the 125 Volt DC batteries, if required, is provided. The system consists of one battery, one battery charger, and associated DC switching equipment. g. The 4.16 kV system is divided into several bus sections, each of which can be supplied from either of two unit service transformers fed from different plant CALVERT CLIFFS UFSAR 8.1-2 Rev. 47 service transformers. The four 4.16 kV ESF busses can be supplied from the EDGs. h. The plant has four safety-related EDGs, two dedicated to each unit. Any combination of two of the EDGs (one from each unit) is capable of supplying sufficient power for the operation of necessary ESF loads during accident conditions on one unit and shutdown loads of the alternate unit concurrent with a loss of offsite power and for the safe and orderly shutdown of both units under loss of offsite power conditions. The diesel generators start automatically on safety injection actuation signal (SIAS) or an undervoltage condition on the busses which supply vital loads, and are ready to accept loads within 10 seconds (Figure 8-6). A Station Blackout diesel generator can also be aligned to any of the four ESF busses. i. All necessary ESF are duplicated and power supplies are so arranged that the failure to energize any one of the applicable busses, or the failure of one diesel generator to start, will not prevent the proper operation of the ESF systems. j. The ESF electrical system has been designed to satisfy the single failure criterion as defined Institute of Electrical and Electronic Engineers (IEEE) 279, Section 4.2. k. Four vital AC instrument busses per unit are provided for essential instrumentation and reactor protection circuits. Each vital bus is fed from a separate battery supply through a dual static inverter. l. The design criteria for all electrical control cable and safety-related equipment power cable are that the cable shall not fail when subjected to associated accident conditions after the long-term, normal operating conditions. m. Power cables in 13.8 kV service are HT Kerite Permashield insulated cables rated at 15 kV. Cables are single conductor shielded and provided with Kerite type FR fire resistant jackets. n. Power cables in 4.16 kV service are HT or HV Kerite insulated cables rated at 5 kV. Cables are triplexed or single conductor, nonshielded, and provided with Kerite type NS neoprene or CSPE sheath jackets. o. Control cables are of multiconductor construction with either cross-linked polyethylene, ethylene propylene rubber or silicon rubber insulation with jackets of Hypalon, neoprene or asbestos braid. Control cables are rated at 600 Volts. Low voltage instrumentation cables are of multiconductor construction with either cross-linked polyethylene, ethylene propylene rubber or silicon rubber insulation with jackets of Hypalon, neoprene or asbestos braid with voltage ratings suitable for the application. Specialty low voltage instrumentation cables are supplied by OEM with unique constructions and voltage ratings specific for their equipment. Control cables for use in underground ducts are insulated with cross-linked polyethylene, and jacketed with neoprene, hypalon or, polyvinyl chloride. Low voltage instrument cables have total coverage electrostatic shielding, or electrostatic shielding covering individual twisted pairs or triads. p. The normal current rating of all insulated conductors is limited to the continuous value which does not cause excessive insulation deterioration from heating. Selection of conductor sizes is based on "Power Cable Ampacities," published by the Insulated Power Cable Engineers Association. q. All cables, terminations and splices within the containment associated with safety-related equipment are qualified by being type tested for the loss-of-coolant accident (LOCA) environmental conditions. r. The electrical systems have been designed in accordance with "IEEE Criteria for Class 1E Electric Systems for Nuclear Power Generating Stations," IEEE No. 308 - 1974. s. Electrical penetration qualification tests were combined using the postulated worst combination of environmental conditions as described in Section 14.20. The CALVERT CLIFFS UFSAR 8.1-3 Rev. 47 electrical penetrations were tested to verify leak integrity and also electrical integrity on those penetrations carrying ESF or reactor protective circuits. All materials used in the construction of electrical penetrations were qualified for radiation exposure of 108 rads either by materials manufacturer's or the penetration manufacturer's tests. Electrical penetration assemblies were supplied by the Amphenol Corporation and the Conax Buffalo Corporation. The Amphenol Type 1, 15 kV, medium voltage power prototype penetration canister was tested by enclosing the inside containment end of the canister in a tank and subjecting it to steam made with 1720 ppm borated water. The penetration was subjected to 275°F, at 41 psig for 15 minutes. This temperature was reached within 30 seconds. The next 45 minutes the penetration was at 260°F, at 33 psig. The following 23 hours were above 250°F, at 30 psig. Throughout the entire test the leak rate was monitored using helium and found to be within the required 1x10-6 standard cubic centimeters per second of dry helium. The Amphenol Types 1, 2, and 3, low voltage power, control and instrumentation, thermocouple and coaxial penetration canisters were tested in a prototype canister containing two coaxial conductors and three or more of each other type of conductor. The test was performed in the same manner as on the Amphenol Type 1, except that during the first 15 minutes the unit was subject to 279°F, at 44 psig, the next 45 minutes 265°F, at 35 psig, the next 23 hours were above 250°F, at 30 psig. The leak rate again was within 1x10-6 standard cubic centimeters per second of dry helium. During the test, the 480 Volt power conductors and the 120 Volt control and instrumentation conductors were energized at their operating voltage and there was no excessive leakage current. The power, control, and instrumentation conductors were also terminated in the manner they will be in the field in order to qualify the termination methods to be used inside containment. All Amphenol prototype canister penetration assemblies successfully passed the environmental test as described above. Conax Type 1, 2, and 4 penetration assemblies are header plate type and were designed, fabricated and prototype-tested to withstand Design Basis Event environmental conditions described in Section 14.20.

All Conax penetration assemblies successfully passed Design Basis Event environmental testing. 8.1.3 SHARED ELECTRICAL EQUIPMENT The following electrical auxiliary system equipment is shared by Units 1 and 2. a. Service Transformer P-13000-1 and P-13000-2 b. 13 kV Service Bus 11 and 21 c. 13 kV Service Bus 12 and 22 d. 500 kV Red Bus

e. 500 kV Black Bus

f. 125 Volt DC Plant Control Batteries 01, 11, 12, 21 and 22

g. 250 Volt DC Emergency Pump Battery 13 and 23 CALVERT CLIFFS UFSAR 8.1-4 Rev. 47 h. 125 Volt DC Busses 11, 12, 21, and 22 i. 250 Volt DC Bus 13

j. 0C Diesel Generator k. 125 Volt DC Unit Control Panel 24

CALVERT CLIFFS UFSAR 8.5-1 Rev. 47 8.5 SEPARATION CRITERIA 8.5.1 DESIGN BASIS Channels that provide signals for the same plant protective function are independent and physically separated to assure that the minimum availability required during any design basis event is met. 8.5.2 CHANNEL IDENTIFICATION Each circuit (scheme) and raceway is given a unique identification and each is additionally coded as follows: a. The channel or load group designation, b. Whether the circuit or raceway is associated with safety-related equipment,

c. Whether the circuit is associated with 125 Volt DC or 120 Volt vital AC panel feeders and non-safety-related equipment. The facility code designations are classified according to their association and redundancy with respect to one another. This classification has resulted in the seven separation groups shown here with the associated facility codes. The facility codes are assigned on the basis of the accompanying descriptions:

SEPARATION GROUP 1 A - A non-safety-related scheme or raceway, Channel A. DA - A 125 Volt DC or 120 Volt vital AC control feed to a non-safety-related item associated with Battery No. 11. ZA - A safety-related instrumentation, control, or power scheme or raceway, Channel A. ZD - One channel of a four-channel safety-related instrumentation channel or raceway, Channel D. D - One channel of a four-channel non-safety-related instrumentation channel or raceway, Channel D. SEPARATION GROUP 2 B - A non-safety-related scheme or raceway, Channel B. DB - A 125 Volt DC or 120 Volt vital AC control feed to a non-safety-related item associated with Battery No. 21. ZB - A safety-related instrumentation, control, or power scheme or raceway, Channel B. ZE - One channel of a four-channel safety-related instrumentation channel or raceway, Channel E. E - One channel of a four-channel non-safety-related instrumentation channel or raceway, Channel E. SEPARATION GROUP 3 ZC - A safety-related control or power scheme or raceway associated with Battery No. 12 and shared items; e.g., all third-pump power circuits and Diesel Generator No. 1B. DC - A 125 Volt DC or 120 Volt vital AC control feed to a non-safety-related item associated with Battery No. 12. ZF - One channel of a four-channel safety-related instrumentation channel or raceway, Channel F. CALVERT CLIFFS UFSAR 8.5-2 Rev. 47 F - One channel of a four-channel non-safety-related instrumentation channel or raceway, Channel F. SEPARATION GROUP 4 ZH - A safety-related control scheme or raceway associated with Battery No. 22. DH - A 125 Volt DC or 120 Volt vital AC control feed to a non-safety-related item associated with Battery No. 22. ZG - One channel of a four-channel safety-related instrumentation channel or raceway, Channel G. G - One channel of a four-channel non-safety-related instrumentation channel or raceway, Channel G. SEPARATION GROUP 5 A - A non-safety-related scheme or raceway, Channel A. B - A non-safety-related scheme or raceway, Channel B. SEPARATION GROUP 6 ZJ - A safety-related control scheme or raceway associated with Battery No. 01, which is capable of assuming any of the first four separation groups, one at a time. Cables from another group may not be routed with Separation Group 6. SEPARATION GROUP 7 K - An Augmented Quality-Station Blackout instrumentation, control, or power scheme or raceway related to Diesel Generator 0C or its dedicated battery (Battery No. 15). A - A non-safety-related scheme or raceway, Channel A. The facility code facilitates and ensures the maintenance of channel separation in the routing of cables and the connection of control boards and panels. All cables and raceways are physically labeled with the appropriate facility code for positive identification. Cable routing is checked and confirmed visually at the time of the cable pull. 8.5.3 CABLE ROUTING The following principles apply for the routing of cables throughout the plant: a. Cables with a facility code preceded by Z of a particular separation group are routed only in safety-related raceways of the same separation group. b. Non-safety-related cables (A or B facility code) may be routed in safety-related raceways, but cannot be routed in safety-related trays of more than one safety separation group (i.e., Groups 1 or 2 respectively). c. Non-safety-related cables (A or B facility code) of different non-safety facility codes can be routed together in non-safety-related raceways. d. Control feeders (DA, DB, DC and DH facility code) from redundant 125 Volt DC unit control or 120 Volt vital AC control panels to non-safety-related equipment are routed separately to maintain independent battery and inverter emergency power sources. e. Protective system instrumentation cables (ZD, ZE, ZF, and ZG facility code) are routed solely in "instrumentation only" raceways of the same separation group. f. Control and instrumentation cables with K facility code are related to Diesel Generator 0C or its dedicated battery and are, therefore, routed separately. CALVERT CLIFFS UFSAR 8.5-3 Rev. 47 g. The non-safety-related low voltage power, control, and instrumentation circuits related to the Station Blackout Diesel Generator Building are assigned a Facility Code A. These Facility Code A circuits will be routed in Separation Group 7 with the following exceptions: In the Auxiliary Building, these control and instrumentation circuits may be routed in either Separation Group 7 or Facility Code ZA or A raceway. However, once these circuits have been moved into Facility Code ZA or A to use existing raceway to enter equipment, they may not move back to Separation Group 7. In the Safety-related Diesel Generator Building, these circuits may be routed in the dedicated Facility Code A raceway.

h. The power cables from Diesel Generator 0C to the four safety-related emergency buses may be compatible with either safety-related Separation Group 1 or 2, but not both at the same time. i. Cables are separated into four groups according to voltage classification and function as follows: 1. Medium voltage power cables

2. Low voltage load center power cables 3. Low voltage power and control cables 4. Instrumentation cables 8.5.4 CONTROL BOARDS AND OTHER PANELS Within the control boards and other panels associated with Class 1E (in reference to electrical separation, post accident monitoring category #1 [PAM 1] circuits are included as Class 1E) systems, circuits and instruments are independent and physically separated by a distance of 6". Where physical separation is impracticable, conduit, metal barriers and fire retardant barriers are used to maintain independence.

Single-control devices to which redundant circuits are connected are avoided wherever practicable. Where single devices are unavoidable, electrical isolation is provided. Devices that provide electrical isolation include relays, isolation amplifiers, solid-state optical couplers, and resistors in instrumentation current loops across which isolated voltage signals are obtained. In the case of third-pump circuit-breaker control switches, the redundant circuits are in separate conduit and connect to the switch at separate locations. The connections are separated by an empty stage of the switch. Additional protection is obtained by the automatic disconnection of DC control power from the unused circuit. Therefore, both redundant circuits at the switch are not energized simultaneously.

With reference to the facility code designations, the separation criteria within control boards and other panels associated with Class 1E systems can be tabulated as follows to indicate compatibility of differently designated cables and devices: SEPARATION GROUP 1 - ZA, ZD, DA, D, A SEPARATION GROUP 2 - ZB, ZE, DB, E, B SEPARATION GROUP 3 - ZC, ZF, DC, F SEPARATION GROUP 4 - ZH, ZG, DH, G SEPARATION GROUP 5 - A, B SEPARATION GROUP 6 - ZJ SEPARATION GROUP 7 - K, A CALVERT CLIFFS UFSAR 8.5-4 Rev. 47 In the case of non-Class 1E A or B circuits in the control boards and other panels, the above criteria are modified to permit A and B association with all of the separation groups. Four-channel non-Class 1E circuits are permitted to associate with each other. 8.5.5 RACEWAYS Separation and independence is maintained between cable trays of different separation groups throughout the plant, including the containment, the penetration rooms, cable spreading rooms, and other congested or hostile areas. The criteria for separating are as follows: a. A minimum of 3' horizontal separation is maintained or physical fire barriers are installed between redundant cable trays. Where a barrier is required, it extends to a minimum of 1' above and below the cable tray or to the ceiling or floor, or it completely encloses each cable tray of one separation group. b. Where the vertical stacking of redundant cable trays is unavoidable, a minimum spacing of 5' is maintained, or horizontal fire barriers are installed between trays, or each cable tray of one separation group is completely enclosed with a fire barrier. c. In the case of the crossover of one redundant tray to another, a minimum of 9" vertical separation is maintained and fire barriers are installed on both top and bottom of one tray to extend 2' from the crossover. In the protected cable spreading room where arrangements preclude maintaining separation as outlined above, fire barriers are installed on the top and bottom of both redundant trays. These barriers, used in conjunction with flame-retardant cables, ensure that a fire in the cable trays, (in the cable spreading room) caused by a cable fault, will not render safety-related cables in a redundant tray inoperable. d. The arrangement and/or installation of protective barriers precludes the possibility that a locally-generated force or missile will destroy redundant systems. For example, in rooms having heavy rotating machinery or high-energy piping: 1. Redundant cable trays are maintained 20' apart, or

2. One of the redundant trays must be 20' or more from the missile source or high-energy pipe, or 3. A 6"-thick reinforced concrete wall isolates one tray from its redundant tray, or

4. A steel barrier isolates the trays from the heavy rotating machinery or high-energy piping. e. Within missile-endangered areas, a minimum horizontal separation of 20' is maintained, or a protective wall, ceiling, or floor of 6"-reinforced concrete provides isolation between redundant switchgear and between other redundant electrical equipment. f. Where routing is unavoidable through areas with potential accumulation of large quantities of oil or other combustible fluids as a result of leakage or rupture of lube oil or cooling systems, a single separation group only is routed through this area and the cables are protected from dripping oil by conduit or covered tray. g. Where it is necessary that cables of redundant systems approach the same or adjacent control panels with less than 3' separation, one system is installed in steel conduit or wireway. h. Isolation between redundant circuits is considered to be adequate where physical separation is less than that indicated above, and when one of the circuits is routed in steel conduit or wireway. i. The worst credible incident is a cable tray insulation fire in the cable spreading room caused by an electrical fault. There are no 208 Volt, 480 Volt or other high-voltage or high-fault cables in the trays of the cable spreading room. The highest CALVERT CLIFFS UFSAR 8.5-5 Rev. 47 fault current in the trays is less than 1,000 Amp or less than an equivalent energy of 360,000 ampere2 x seconds.

8.5.6 PENETRATION ROOMS Two separate penetration rooms are provided for all cables that must pass through the containment wall. The East Penetration Room is divided such that there are, to one side of the division, the Separation Group 1 penetrations and to the other side of the division, the Separation Group 2 penetrations. The horizontal separation between redundant penetrations and associated cable trays is 3'. The West Penetration Room is similarly divided except for the addition of Separation Group 3 to the same side as Separation Group 2 and Separation Group 4 to the same side as Separation Group 1. Vertical and horizontal separation between redundant penetrations will be a minimum of 3'. Cable tray separation criteria, as described earlier, are applicable for the penetration rooms. Power cable penetrations of high-energy levels are located above those of low-energy level circuits.

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07, < 1 (@'\ NOTE 6 SlATulJlllRIP t f.l;I:\: Z-NU!.!BlRS IN PMl(NIHESIS 1 ' IHE Ol!ANfl TY OFOEvlCFSREOUIR£0. '* 4. ALL £0UIPMFNT ON THIS CRhlNG rs AUGMPHEO QUALITY UNLL'>'>NOTEDOTt1Ll<WISL-o. LL :.:;[r>.JC: g vOllMLILll">wll:H 1[51 @ LIGHT KIRK HY INILl<l(](K AAU(,HUHEC QUALITY/NUN ',AfLTY U]UNCMH . H!IP "'" ,-1,: ... 1 b lj ... £ y I" I I Cl*:P*MS. '.;>SILM UNIT tJU', C' CAL VERT CLIFFS NUCLEAR POWER PLANT UFSAR FIGURE 8-3SH0003 DIESEL GENERATOR PROJECT METER AND RELAY DIAGRAM 480V SYSTEM UNIT BUS 07 BGE DRAWING 61010SH0003, REV.3 REVISION

\O.C;' l (,.,*c, \f_'Y GROUNDING 01 5tONNlC:T 11(515TOR L!Nt<. r; 4ii0n°.:: 1--1 <12"0 \? IA' __ ,/ G\' ""'7: I lA +-I \iJ.J ( [*9 c ' [ "'" L .\"..! .. 1 "I I [;; l 4KV UNIT BUS 17 1A07 2000A, 250MVA IZAl I l A-LJ--l I \9 l*;cc" A C ;y;; 1200A 120GA )n

• B 0 12QQA I )--:::i "' ,,,,)" "' ""-!< ,7\, "' ,,,,) -"'"")" .l. 1:,2 l':'.l.J'1/ l :*-l l:(\ f-J BY. ... .!.f..I..]+/-*_ 120015 ' J:?G cJ) c0 l: --MRll' 150 <J) .-\ -1 I 1;1 ) --I" j 'iOl'i 1 Tl} 4K i UN IT BU') 11 FlRfAHll 1'J? 1103 SYNC
• 0 i {1' i *Q' '[,"' -: .. ' '" . '!) ' ' 'DR \..__ 1111'> "' t ' CJ l .1'. 5 ;7'."71, c;:*0 ,/ r 140\ /'152\ "°'°'°'A 71 7\ Ll ""' ); /\, 0 -!". ..... r I),; I Tfllf' " : ... \ cYY1 fliY OwfR XHIR I' '[XlJT. PPl"l,f,QHz ). \TJ.....AJ-..J ?O 6 ' "' e o l""" GlN. fl Fl fl I r' I -'--*t,::" ____L_ l_ AUTOMATIC VOLTAGE REGULATORS ANO STATIC EXCITATION PANEL '*ill SEN\ INC; L_ SfNSINC 1 1-,lN!r 1'.olNllJ 81*-l"olU 1 SOil 51 -,-,-. l OTA DIESH Gl'oEPAIOR I"'""'"' ------r.IF'>Fi CfNERhTOR I o:KOUl RE'i AY 'HANC RE SE r I STYLE OR lilODH NO. OPERATE RllH rnr,10 1A I 1C188 f>'1AJ81Al5B _::.::_ ____ :--lRIPCG l 103 OvfftFRFOiltNCY RELAY !':J1B0'l">MN CORRECI C:ESEL Gl'l[RATQR(ORRfClSlf VQl_lllf,ffl_LAY '.);fRVOLUGF RELAY ! ABKR/W (,.'lll!81hl">tl IJNGvljlJjQo CG BUI \/Oj L:f, If NORM COllNB 1703 ----[ll'>ll NLIJlf>hl '.),'fR(LJllll[NlflflAr JPFRAlf R[LAY IBblO-lA JHBA'I !Al 3 NF J'RAL 'A1 j 1}:;1xCITATlllNHMflO,EFICUFIPfNT Pf I A Y TRIPCGflKR 1703 11 Ml l)\,LR(llflflfNJ FlfL Al I Al Tfllf' A',fl',2pl1All0 -----S[R, I ([ TFIANSf"0F!Mf J>FRCLJlll'l[Nl RlLAr NOTES : .l.
• INIHCATES NON LMlHGLNlY INPlll '0 f.C, flfltMFP 1"11P LOGIC "° I NDI CA ll '> Tfllf-' oiH[N IN PARAI. fl MODE (:'1AWIN:, <,M f IY RELA'ED LJNLl ',', (,. AlAllMANCR'U INPUTS AH[ N0N\Afl1'RtlATf1. <J. POWLR 'ACTOR MFlfR OISCONNECTLD. RlllfllU IN f-'LAll AND NOTLISFO Y, Fl "If lf;o \'AP MF!tll (") I AMP (.j) INlJl(Al I NG I I ! *2 <l Y l N :lf>L 0(< /). REFERENCE DRAWINGS r--------------------JO? HlQNI . "CL*l 006 ',I NGl l L I Nl !JU'> I I CAL VERT CLIFFS NUCLEAR POWER PLANT UFSAR FIGURE 8-4SH0002 DIESEL GENERA TOR PROJECT METER AND RELAY DI AGRAM 4KV SYSTEM UNIT BUS 17 BGE DRAWING 61007SH0001, REV.6 REVISION DISCONNECT LINK @ DIESELGENERATQROC 6750KVA,3*Ph.0.8PF. 4.16KV. 60Hz.1200RPM 11251100 illffi-'" '0 4KV UNIT BUS 07 OA07 2000A. 250MVA !FAC. Kl )\ l "' ','imc*"c' ,7\c !121 SYNC .. AUTOMATIC VOLTAGE REGULATORS ANO STATIC EXCITATION PANEL AVR#l SENSING "6" AVR#2 SENSING L 10 I "c°' \ 189-0703 3-l/C750kcmll/'1l 4 I0111m>I 5 6 10*'1 y 10 STOP, TH INK, ACT AND REVIEW I 13 '" D*OC '"' o-oc 181-U ""'O=OC 181-0 ""'O=OC 181-1 ""'O=OC "' 0-0C 15911 """il="OC 15912 o-oc !51V 0-QC 15Hl/P D-OC 151N/B --o:oc 151N/S ----o:oc 0-0C I 151G 11501151 I "' O*OC 127/151 -907 "' Q."OC 127*2 B-07 127*! B-07 186*2 B-07 166-1 607 "' II "" 'fY'PE OIESELGENERATOR ABB/II 2SOB225A101 OPEtNJa.RlcLAY OJFF RELAY ----s;;! DIESEL GENERATOR ELECT.S\o 78020 EMERG.SHUT LOCKOUTRELAYIHANDRESET) OOWNaOG BKRTRJP OGUNDERfREOUENCYRELAY AB:/" 61\B287A15BI TRJPOG BKR0703 OGOVERFREOUENCYRELAY ABB/II CF! DIESEL GENERATOR CORRECT SET 291B995Al0 I FREQUENCY RELAY OGGOVERNORCONTROL 2301A LOAD SHARING NOSPEEDCONTRO DIESEL GENERATOR CORRECT SET I 1s1ss12 I CoL'b"sBtRLoi11oc3 VOLTAGE RELAY DJESELGENERATOR A88/W OVERVOLTAGERELAY OVERCURRENTRELAY " !Ac OGVOLTAGECONTROLLEO ABB/W OVERCURRENT RELAY Ci5V7 OGGROUNDOVERCURRENTRELAY ABB/W 288B717A13 OPERATE RELAY cos 186!0-DC OlESELGENERATORNEUTRAL ABB/W 2886717A13 OPERATE RELAY OVERCURRENTRELAY cos 186!0-DC OIESELGENERATORNEUTRAL ABB/W 2888717A13 OPERATE RELAY OVERCURRENTRELAY cos 1861{)*0C OGEXCJTATlONXFMROVERCURRENT " 12PJC32J45A TRIPOG-BKR RELAY PJC 0703 GROUNDTIMEOVERCURRENTRELAY " 121AC53A801A TR!PASSOCJATEO !Ac '" SERVICETRANSFORM£R " OVlRCURR(NTRELAY w SERVICE TRANSFORMER " GROUNOINST.OVERCURRENTRELAY PJC DGNEGATIVESEOUENCERELAY /186111 coo DGLOSSOFFIELORELAY 2906481 AQ9 I TRJPOG 6KR0703 DGREVERSEPOllERRELAY ABB/II 2906038A10 TRIPOG CRN-=-1 6KR0703 VOLTAGE RESTRAINT I 121rcvs1Aou I TRlPOG OVERCURRENTRELAY BKR0703 DIESEL GENERATOR ABBIW UNDERVOLTAGERELAY CVT BUS07UNOERVOLTAGERElAY " 12NGV13625A NGV UNDERVOLTAGE RHAY(SITE POWERI ABBIW 1875508 TRIP CvT BKR0704 RELAY " 12HEA* HEA 61A223 BUS 07 LOCKOUT RELAY " 12HEA* !HANORESETI HEA 61A223 {sJTE POwtR'\ruo REVERSE POWER RELAY " 121CW-TRIP 51A2A BKR0704 "6" NOTES: \. GENERATORGROUNOOVERCURRENT TWOOUTOFTHRE£LOGIC ;. 3. TRIP LOGIC .. '* 6.RTUANDALARMll>IPUTS/IRENON-SAFETYRELATED 7.POWERFACTORMETERDISCONNECTED,REl!REDlNPLACEAND LABELLED"NOTUSED". LEGEND: 0 VOLTMETER SWITCH OW DIGITAL WATT METER OVAR DJGITALVARMETER WPC llATTPULSESCOUNTER HM HOUR METER REMOTE TERMINAL UNIT @I @ SYNCHROSCOPELAMP @ 4 © AUGf.IENTEDOUALlTY/NON-SHETYRELATEDBOUNOARY "' '"' REFERENCE DRAWINGS: ELECTRICAlf.IAINSINGL°ELINEOIAGRAM METER8RELAYDIAGRAt.t.4KVSYSTEMUNITBUSES11 814 METERSRELAY01AGRAM.480VSYSTEMUNl1BUS07 SCHEMATJCOIAGRAM,PLANTSYNCHRON!Z!NG, UNITS1 82 DIESEL GENERA TOR PROJECT METER AND RELAY DIAGRAM 4KV SYSTEM UNIT BUS 07 Dit:'0:o. 1 E-007SH0003 13

N.;'< ',Hf TY 6ATTEFIYCHAFIG£FI , .. ,ir"* KI HK Y INlfRL*::K NOH I NOTES !BUSSMANN LPS-FIK100SPI 410 AWG I '°"' ,,,.r I " , J .. c,eou PANll 1 i:. M:C 12j 'UO!l lil2e -* -:,_j 'l :r,;\ 1032 I --, ©'*'"J' I'? ',HUN! :-16 : --,_, LJ y! ; )QA tJ * \: SA n JE:q )--l '>PA!l( --'°'t "' b j 1 ) 200A ' '°'l WARE 1029 y 115V DC DllTRIBUTHJ< PANEL A, 2001 CONT,10" l.C. ! l ! * ! ! ! '°'1 l i n .i ,,,-1 I I 'f '" "' "' r (64\ i fi""" ,,, 125V DC DISTRIBUTION PANEL lA, 600A CONT.20KA 1.C ** BUS 14 9'.* / 6011 ,,,6 l t* f -t *f "i SPlAf 5P>R[ 5PARf SPARl l '! 1030L L INK ,' I II '0 k-rml!>'Ol I ? I kc m, 1 1,, I f I I ?IC8AWG -I lPNI 103100/MON I Nill£ 1 1 031 T", I: i Inoa II-11'>1 JAC"K I UlSAWG I I ['] r NOT[' -14?0 -: 11*i.-. ? 100110 l 'Af.'l N.o. lf I 1 DZ§_ ' NLIES: [I. ,, °' CJ 1ffl"'>f 1L r1 1<,.J1 lltLEHC '* '.>><[ ICHES *Rt l-'OL l. '>IN(,t l T>ir.cJ;, _,NLl .)fll[fl.;ISE NOTED. LOU I PM[NT I', ',Ill l I Y Ill l h ILJ NI f">> N.; !fr Sf HAS ANl 1.:: N[ Jf tAC!I Br UILllEIO ANG 111[ Rf"',T will Fl[ ',P.;PfC. 1 I. Ill,. A Jf THI'> [;II!.* IN(, "*"' I 'JNGlf' NC. !'>1 0?4 f. SH. I. 12. AL l M[ T[R'> llNL nu A y IN ]>ff R,'1 lf py p *PF NON -If f p 1 -<l)t'.: lt.B. I l. I 1 .. 1. b II . :1 B TP 7 _EG:NC M._ SAFETY PF1t.Tfl'1N:l!i '.>Mtrr RfL'IE[ tLl'NDoRl llLfUHrl .,, ?' 4b4 FlO T '(PY P o!hl'-.* T :JTL] i'<E P'Ni l :UT. : EC:P!Nfl ,, L__ CAL VERT CLIFFS NUCLEAR POWER PLANT UFSAR FIGURE 8-5SH0002 DIESEL GENERATOR PROJECT SINGLE LINF DIAGRAM DGIA 125V DC SYSTEM BUS I 4 BGE DRAWING 61024SH0002, REV.3 REVISION afo FIGURE 8-5 DIESEL GENERATOR PROJECT SINGLE LINE DIAGRAM DGOC 124V DC SYSTEM BUS 15 (Sheet 3) I ., .......... .. . .. t3 ii r. .J " ............ ! . ! -........ *-. .!. ! .. . **.;.:,* *-* * ".;! ** J J. '* !. .. I -* I .. -,!, !, !. l .. !. _ .... \ .J. .!, _ ... -.. ,,I ----.:.i IH =* .::;*-* = .. :.: ( I * ...... t *t . c" .. .._,, ..... _ "-.?. l *-,.'!t ......... --_, .. :5. .,..... .. .... ..... *.:* ... .:---.. ..:!" ..... ::--.-:J _ .. _ .......... .. Revision 39

FIGURE 8-llA DIESEL NO. 2A -STARTING AIR, FUEL OIL, AND LUBE OIL 1 I 2 I J a j '-1--.....-----------+--b *-; ... .,."::* ... ,__ _ r.=============::::::::-. _T .. --r-*c .. I r .. -*1--Revision 38 § ' D @ . . , FUEL Oil TPAN'HRPJMP yf--18-0L0-112 *'.i) G "' (P::, NOTES 10 STOP, THINK, ACT ANO REVIEW 13 12 (j) SIMPLIFIED SYSTEM DRAWING DIESEL NO 18 STARTING FUEL & LUBE OIL SL-0698 CALVERT CLIFFS NUCLE.AR POWER PL.allT CALVERT CUFFS UNIT 1 &2 64321 13 LVB<:OlL COOLERN0.2B T rnELOIL TRANSFERP\Jlf' ENGH£ ORIVC:N LUBEOlLPUll' ENGINEORIVEN JACKETCOCUM1PU141' PRIMING EllG!h(ORIVEN AIR COOLER CO!X.AHTPUll' AIRCOOll"IGSYSTE* ENERGENC1 DIESEL GENERUORUNll N0.28 10 STOP, THINK, ACT AND REVIEW 1l !fil!lli l.FORDRAWINGSY"'8Cl.0GT.SEEBGEORAWING Sllf'llFIEO mm1 2. s PLANT. SEETHECORRESPONOlllGOll 0ROlr;G,BG£N0.60127!01H91. UFSAR FIG. 8-SC REVISION __ SIMPLIFIED SYSTEM DRAWING DIESEL NO. 28 *oo fl I \*I"" I STARTING a LUBE OIL gr.* 64322 10 11 1""" ll 8 0 G 1A1 FUEL INJECTORS 1A1-DF0-101: 1A LO FILL PUMP ORUM CONNECTION 1A1-DF0-82 1A1 ENGINE om---( DRIVEN FUEL -) Oil PUMP p ENGINE DRIVEN LUBE OIL PPS 1A1-DFO-103 1A1 HT COOLANT 1A1 PRE LUBE HEATER 1A1 AUXILIARY DESK <NOTE 1) 1A1 PRE LUBE PUMP LUBE OIL DAY TANK 1-TCV-10172 1A1 SOUTH L.O. COOLER 1A1-DL0-51 BS @-al TO UNLOADING STATION 1A2-DL0-61 1-TCV-10211 lo---@ L.O. COOLER lo---@ 1A1 AIR DRIVEN PRELUBE PP HP AIR BOTTLE 1A1 FUEL Oil RECIRC PP 1A OG FUEL OIL STORAGE TANK ENGINE DRIVEN LUBE OIL PPS 1A2 PRE LUBE PUMP 1A2 HT COOLANT 1A2 PRE LUBE HEATER 10 STOP, THINK. ACT AND REVIEW 13 L 1A2-1A2 ELECTRIC FUEL OIL PP NOTES: AUXILIARY DESKS ARE ST AND-ALONE PANELS CONTAINING TEMP, PRESS, dP INSTRUMENTS, AND SUPPORT SYSTEMS PUMPS, HEATERS, VALVING, ETC. 2. FOR DRAWING SYMBOLOGY, SEE BCE DRAWING NO. 64329 CSL-080J, SIMPLIFIED SYSTEM DRAWING LEGEND. 3. THIS DRAWING IS NOT TO BE USED FOR MAINTENANCE OR OPERATION OF THE PLANT. SEE THE CORRESPONDING OM DRAWING, BCE NO. 60727 (QM-69). y' "'-L/l""-Y ----I 1A2 FUEL --1----------------------!-* L==> INJECTORS 1A2 AUXILIARY DESK <NOTE 1) AS BUILT INFO ONLY ESPES199700070-000 1 1ikk I 10 ,.,,_1_1_1 j"" 11 UFSAR FIGURE 8-80 SIMPLIFIED SYSTEM DRAWING NO.IA EMERGENCY DIESEL GENERATOR FUEL OIL, LUBE OIL BAL 64331SH0001 13 SIZE ' I 8 I C io I i r 1A1 HT EXPANSION TANK 01A1LT EXPANSION TANK NOTES: 1. AUXILIARY DESKS ARE STAND-ALONE PANELS CONTAINING TEMP, PRESS & dP INSTRUMENTS AND SUPPORT SYSTEMS PUMPS, HEATERS, VALVING, ETC. 2. HT COOLANT SUPPLIES THE ENGINE BLOCK, TURBOCHARGER ANO GOVERNOR OIL TEMPERATURE CONTROLLER. 3. LT COOLANT SUPPLIES THE COMBUSTION AIR INTERCOOLERS ANO LUBE OIL COOLERS 4. FOR DRAWING SYMBOLOGY, SEE BGE DRAWING NO. 64329 CSL-080>, SIMPLIFIED SYSTEM DRAWING LEGEND. 5. THIS DRAWING IS NOT TO BE USED FOR MAINTENANCE OR OPERATION OF THE PLANT. SEE THE CORRESPONDING OM DRAWING, BGE NO. 60727 COM-69). 1A1 RADIATOR l T RADIATOR ELECTRIC PREHEAT ER PUMP 1A1 AUXILIARY DESK CNOTE 1J 1A1 LT ENGINE DRIVEN PUMP 1A1 HT ENGINE DRIVEN PUMP 1A1 OUT BO EXH ATMOS 1A1 EXH SILENCER TURBO CHARGERS 1A1 STARTING AIR COMPRESSOR SKID 1A1 INBO EXH 1A2 INBD EXH 1A1-DSA-54 ATMOS 1A2 EXH SILENCER TURBO CHARGERS 1A2 ST ART ING AIR COMPRESSOR SKID 1A2 OUTBD EXH 1-TCV-10152 1A2 l T ENGINE DRIVEN PUMP 1A2 HT ENGINE DRIVEN PUMP 10 1A2 RADIATOR RADIATOR FANS l T RADIATOR ELECTRIC PREHEATER 1A2 AUXILIARY DESK (NOTE 1) 0 ¥111J.i.i I 10 STOP, THINK, ACT AND REVIEW 13 1A2 LT D EXPANSION TANK 1A2 HT EXPANSION TANK UFSAR FIGURE 8-8E SIMPLIFIED SYSTEM DRAWING NQ.1 A EMERGENCY DIESEL GENERATOR STARTING AIR, COOLING, LT COOLING 8 I C I 0 I

• *
• oc LO FILL PUMP DRUM CONNECTION OC1-DF0-82 OC1 OQJ---\ OC1 ENGINE DRIVEN FUEL OIL PUMP __ .. _____ .,. _____ _ I I OC1-DF0-101: I OCt ELECTRIC : to FUEL OIL PUMP1 OC1-' DFO-: 102 I P) FILTER OC1-0FO-103 FUEL INJECTORS I ---------------p OC1 ENGINE DRIVEN LUBE OIL PPS PRE LUBE HEATER OC1 AUXILIARY DESK CNOTE 1} LUBE OIL DAY TANK OC1-DL0-61 O-DL0-10172-TCV OC1 DIESEL OC2-DL0-61 O-DL0-10211-TCV O-DL0-10171-TC"o-DL0-10212-TCV I 11 FUEL OIL STORAGE TANK ENGINE DRIVEN LUBE OIL PPS PUMP OC2 HT COOLANT , ... p O-DF0-145 NOTE: UEL 1. AUXILIARY DESKS ARE ST AND-ALONE PANELS CONTAINING TEMP, PRESS, dP INSTRUMENTS, AND SUPPORT SYSTEMS PUMPS, HEATERS, VALVING, ETC. O-DF0-129 OC FUEL OIL TRANSFER PUMP OC2 ENGINE DRIVEN FUEL OIL PP od ELECTRIC FUEL OIL PP OC2-0F0-102 '-----------OC2 FUEL INJECTORS QQJ
• I OC1 AIR DRIVEN PRELUBE PP HP AIR BOTTLE OC2 DRIVEN PRELUBE PP OC2 AUXILIARY DESK CNOTE n CAL VERT CLIFFS NUCLEAR POWER PLANT UFSAR FIGURE 8-8F NO. QC EMERGENCY DIESEL GENERATOR FUEL OIL, LUBE OIL REVISION 21
• *
• OCl HT EXPANSION TANK NOTES: OC1 RADIATOR T HT RADIATOR ATMOS OC1EXH SILENCER ATMOS OC2 EXH SILENCER OC2 RADIATOR HT RADIATOR T ;"";;-', ..... ---, , ..... I '

  noo::; 1 :"'JDC}/14Kv -r 13K, SER'v'IC:F NO 11 I [)1FSF1 GEN NO 2A UNIT-2 _e-S.M.l C 0 J SULlS I A I *ON 69/1.)KV l1.3KV SERVIO RUS NO 73 Dl[SLL Gl:_N NO. 1H LOAD GROUP 8 UNIT-1 UNIT-2 I -l I I -* i SFRVICE TRNSF U-4000 11 13.8/4.16KV /l'\ ;;;N N0.28 l_ . . ! !J '\ l\IR,irr lR"SF I . ? \ SFR\ICF fRN'.11 , --+-----1 40UU-12 \ _13 814 16K, '......._ , U.\-ICL IRNSI 1* ' II ;o00*?1 ____. v/ .. ". )' *. U.8p'* 15K. j .... " I L. -1 I L 4KV UNI 1 LJU'.::. NO 11 4K\/ UNIT BUS NO 21 l 0 11 KV UNIT BUS NO 17 1srr D<NG 51os1sH0002i 4801/ UNIT BUSlS 11A &r 11R . I 21A & 218 .... 'lJ .. I I ... HJ CHARCFR CHARGFR CHARGFR 2os112ov INS r N0.11 I NO 14 NO 23 NO 22 BUS N0.11 L!AT I LIU NO 11 1r=1 I I l MCC 21'\R 70811/0V INS I. 1:3US NO Ll R/\TTFRY NO 72 11 1 1 M .. r1-12ov DC "j I NO 22 NO 11 l I ' [::;"' Nlf<OIJ NELS 12.15 [ ,JNll I IN, I INJ 11 I "' IN\' INO 71 r11[5[L GEN NO 2A AUXILIAl-?ILS CFN NO lH AUXILIAl?IL'.:i '------?G /f REACTOR PROTECTION AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM CHANNELS UNITS 1 AND 2 INO '2 I LL I N ['j 1

• i**r. ' UNll """I -___.__ 5' "'* I 1 . NO 2'1 I N(; 21 288/1/,,, l3US NO 22 r 8ATT[RY NU 12 [ -1 "JO 21 r-1 I I "{f.::" ""' I II ' l u 'j NO 1.'__J -* I i ti NIT I CONTRJI Pt..N[LS I I I L L_1r-..,T ,,)r-,r;:,. .. ' ':.' NOTE OiRK I INf-S rJui-; rr*DFRS L::'.rll llNFS .\Rf flF[)tRS CAL VERT CLIFFS NUCLEAR POWER PLANT UFSAR FIGURE 8-9SH0001 BLOCK DIAGRAM AUXILIARY SYSTEM LOAD GROUPS UN I TS 1 8 2 BGE DRAWING 6105 7, REV.9 4KV UNIT BUS N0.17 480V UNIT BUS 17 BATTERY CHARGER N0.16 125V DC BUS N0.14 PANEL I D28 UNIT CONTROL PANEL I Cl 88 TO 4KV UNIT BUS N0.11 ISEE DWG 610571 DGI A MCC 123 BATTERY N0.14 BATTERY N0.15 LOAD GROUP SBO DIESEL GENERATOR DG-OC BATTERY CHARGER N0.17 125VDC BUS N0.15 UN IT CONTROL PANEL OCI 88 s.M.E.c.o. SUBS TAT I ON SERVI CE TRANSF. OX01 13.2/4.16KV DIESEL GEN.QC 4KV UNIT BUS N0.07 480V UNIT BUS NO. 07 DG-OC MCC 023 TO 4KV UNIT BUS N0.11,14,21 AND 24 ISEE DWG 610571 NOTES Re H Rt NC:C !;CL':. IN' CALVERT CLIFFS NUCLEAR POWER PLANT UFSAR FIGURE 8-9SH0002 BLOCK DI AGRAM AUXILIARY SYSTEM LOAD GROUPS DG-OC SBO 8 DGI A DIESEL GENERA TORS BGE DRAWING 61057SH0002, REV.5 RE VISION ;;i(o

.--... I I i 1 i i i i ! g i 4 I=' I ' I;:: I ©I ! ! >-.>.* I L.J cl5 gm l:izd;::ooO !iN r "'o"')>:"'1Z,: ' . 0 SERVICE TRANSF (U-4000-22) TO BKR 152-2311 SERVICE TRAN SF <U-440-24Al LOW PRESS SAFETY INJ PUMP 22 SALT WATER """' 22 TO BKR 1!.2*0701 DIESEL GEN OC TO DISC SW 189-2106 <THIS OWCJ HIGH PRESS SAFETY lNJ PUMP 22 SERVICE WATER PUMP 22 tlGH PRESS SAF'ETY INJ PUMP 2J <TO DISC SW 189*2110) SERVICE WATER PUMP 2J <TO DISC SW 189*211U SALT WATER PUMP 2J <TO DISC SW 189*2112> S£RVICE TRANSF <U*'40*248J SERVICE TRAN SF (U-4000-12) TO BKR 152-2301 152-2415 © 1011 100/5 1200A MJX FEEDWATER g I" E@1 MOTOA OAIVEN o o "' -6 PUMP 2J © og 0-3 " " __ "'_J i i i i ' o' " ' >o ,, << * .!" p I !( i *o 0 fil H 0. I ' i G § £ ro c v> SERVICE TRANSF <U-4000-22l TO BKR 152-2201 SERVICE TRANSF CU*440*21AJ SWITCHYARD FEEDER C500 KVAl SM.T WATER P<JljP2J no DISC SW 189-2412> SERVICE WATER PWP 23 <TO DISC SW 189*24111 HIGH PRESS SAFETY INJ PUMP 2J <TO DISC SW 189-2410) SERVICE WATER PUMP 21 lo9CH PRESS SAF'ETY INJ P\U'21 COHTAN.IENT SPRAY P\U'21 TO DISC SW 189-2406 mes owe> SM.T WATER P\U'21 LOW PRESS SAF'ETY INJ P\U'21 2-5001.ACM/0 SERVICE TRAN.SF <U*440*218J SERVICE TRAN SF (U-4000-12) TO BKR 152-2209 _, I z ""

CALVERT CLIFFS UFSAR 9.4-1 Rev. 47 9.4 SPENT FUEL POOL COOLING SYSTEM 9.4.1 DESIGN BASIS The SFPC system is common to both units. The pool contains water with the proper dissolved concentration of boron and has the capacity to store 1830 fuel assemblies. The SFPC system is designed to remove the maximum decay heat expected from 1613 fuel assemblies, not including a full core off-load. The maximum pool temperature in this case is 120°F. The system is also capable of being used in conjunction with the SDC system to remove the maximum expected decay heat load from 1830 fuel assemblies, including a full core discharge. The maximum SFP temperature in this case is 130°F. The maximum decay heat load expected from 1613 fuel assemblies, not including a full core off-load, is a function of decay time. For a limiting decay time of 3.5 days, which results in an initial core alteration time of 3.0 days after reactor shutdown, the decay heat load is 22.33x106 Btu/hr. The fuel is assumed to have undergone steady-state burnup at 2738 MWt for an average of 1562.4 days for an 100 assembly batch reload. The total SFP decay heat load as a function of decay time is compared to the heat removal capacity from the two SFP heat exchangers as a function of SRW temperature to show what time after shutdown is acceptable for each SRW temperature condition to maintain the pool at a temperature of 120°F. A maximum SRW temperature of 65°F is required to support a minimum decay time of 3.5 days. In the event that one SFP cooling loop is lost, the remaining loop can remove the heat load while maintaining the pool temperature at 155°F. The maximum decay heat rate for 1830 fuel assemblies stored in the SFP is a function of decay time. For a limiting decay time of 4.5 days, which results in an initial core alteration time of 3.0 days after reactor shutdown, the decay heat load is 45.96x106 Btu/hr based upon the following hypothetical sequence of events: 1. Eighty-four fuel assemblies are removed from Unit 1 after an average of 1860 days of reactor operation at 2738 MWt, and are replaced with fresh fuel. Unit 1 is then returned to full power. 2. Three-hundred-sixty-five days after the Unit 1 refueling, 84 fuel assemblies are removed from Unit 2 after an average of 1860 days of irradiation and are replaced with fresh fuel. Unit 2 is then returned to full power. 3. Three-hundred-sixty-five days after the Unit 2 refueling, 84 fuel assemblies are removed from Unit 1 after an average of 1860 days of irradiation and are replaced with fresh fuel. Unit 1 is then returned to full power. 4. This refueling cycle continues until the pool contains 1613 fuel assemblies at the end of a Unit 2 refueling. It has been conservatively assumed that the 67 oldest assemblies have been removed from the pool to allow for complete filling of the racks with newer fuel. 5. Unit 1 is then shutdown 60 days after the previous Unit 2 shutdown and the entire core is offloaded after a minimum of 4.5 days of decay. At this point, it is conservatively assumed that the fuel has completed its current cycle, and is therefore at maximum irradiation. CALVERT CLIFFS UFSAR 9.4-2 Rev. 47 Upon completion of the last operation, the pool will contain 1830 fuel assemblies, with each discharge subjected to different periods of irradiation and decay, in accordance with the table below assuming the minimum decay time of 4.5 days: Number of Assemblies Irradiation Period (Days) Decay Period (Days) a. 17 1860 6964.5 b. 84 1860 6599.5 c. 84 1860 6234.5 d. 84 1860 5869.5 e. 84 1860 5504.5 f. 84 1860 5139.5 g. 84 1860 4774.5 h. 84 1860 4409.5 i. 84 1860 4044.5 j. 84 1860 3679.5 k. 84 1860 3314.5 l. 84 1860 2949.5 m. 84 1860 2584.5 n. 84 1860 2219.5

o. 84 1860 1854.5 p. 84 1860 1489.5 q. 84 1860 1124.5 r. 84 1860 759.5 s. 84 1860 394.5 t. 84 1860 64.5 u. 217 1860 4.5 The total SFP decay heat load as a function of decay time is compared to the heat removal capacity from both loops of SFPC as a function of SRW temperature, supplemented with one loop of SDC to show what time after shutdown is acceptable for each SRW temperature condition to maintain the pool at a temperature at 130°F. A maximum SRW temperature of 75°F is required to support a minimum decay time of 4.5 days. 9.4.2 SYSTEM DESCRIPTION The SFPC System shown in Table 9-16 and Figure 9-7 is a closed-loop system consisting of two half-capacity pumps and two half-capacity heat exchangers in parallel, a bypass filter that removes insoluble particulates, and a bypass demineralizer that removes soluble ions. The SFPC heat exchangers are cooled by service water (SRW).

Skimmers are provided in the SFP to remove accumulated dust from the pool. The clarity and purity of the water in the SFP, refueling pool, and the RWT are further maintained by passing a portion of the flow through the bypass filter and/or demineralizer. The SFP filter and demineralizer removes fission products from the cooling water in the event of a leaking fuel assembly.

Connections are provided for tie-in to the SDC system to provide for additional heat removal in the event that 1830 fuel assemblies are contained in the pool. When the pressure in the SDC system is greater than the design pressure of the SFPC system, the SFPC system is isolated from the SDC system via two manual isolation valves. Although not required by the design code, double valve isolation is provided at this system interface to meet the original FSAR design basis (FCR 90-87).

CALVERT CLIFFS UFSAR 9.4-3 Rev. 47 The entire SFPC system is tornado-protected and is located in a Seismic Category I structure. Borated makeup water comes from the RWT. Non-borated makeup water comes from the demineralized water system. 9.4.3 COMPONENTS 9.4.3.1 Functional Description A description for the spent fuel pool cooling system is contained in Table 9-16. 9.4.3.2 Codes and Standards The following codes and standards were used in the design of the SFPC System components: Pump Standards of: ASME (III, VIII, IX, PTC8.2), ASTM, NEMA, ANSI Heat Exchanger Standards of: Tubular Exchanger Manufacturers Association (TEMA), ASME (III, VIII, IX), ASTM, ANSI Filter ASME III C and ASME VIII paragraph UW-2(a) Ion Exchanger ASME III C and ASME VIII paragraph UW-2(a) Valves, Piping, Fittings ANSI B31.7 Class III 9.4.3.3 Tests and Inspections Each component is cleaned and inspected before installation and the assembled systems flushed with demineralized water. The flow paths, flow capacity and mechanical operability are tested by operation. The head and capacity of the pumps are also tested.

Instruments are calibrated prior to tests. Alarm functions are checked for operability and limits during preoperational testing. During normal operation, periodic tests will be made to confirm design criteria. 9.4.4 SYSTEM OPERATION AND RELIABILITY In the normal case (i.e., with no full-core off load), if one SFPC loop is lost, the remaining loop can remove decay heat while maintaining the pool temperature at 155°F. In the case of total loss of SFPC with 1613 fuel assemblies in the pool, it would take more than 8 hours to raise the pool temperature from 155°F to 210°F. The case of total loss of SFP cooling is only discussed to demonstrate the time available to take appropriate action in such an event to preclude boiling, and the resulting loss in pool water level. The design of the SFPC System and pool structural components (e.g., pool liner plate, SFPC piping and pumps) for total loss of cooling is not part of the system's design basis.

The most serious failure to the system is the loss of SFP water. This is avoided by routing all SFP piping connections above the water level and providing them with siphon breakers to prevent gravity drainage.

The SFP is designed to preclude the loss of structural integrity. Section 5.6.1 describes the analysis made to verify that the structural integrity cannot be impaired. Additional design and quality control requirements for the SFP are given in Section 6.3.5.1. However, if a leak from the SFP is postulated, the capabilities for controlling the leak are as follows:

CALVERT CLIFFS UFSAR 9.4-4 Rev. 47 Makeup water can be supplied indefinitely to the SFP at a rate of at least 150 gpm. It can usually be supplied at a greater rate for a period of many days, but this depends upon plant conditions. The makeup water flow path is as follows: a. Source - Well water b. Portable Makeup Demineralizers - Typical capacity 150 gpm or more c. Demineralized Water Storage Tank - Storage capacity 350,000 gallons d. Four Reactor Coolant Makeup Pumps (Normally run one per unit) - Capacity 165 gpm each, less the amount required for reactor coolant makeup e. Two RWTs (One per unit) - Storage capacity 420,000 gallons - Required to have 400,000 gallons during operation - During refueling this water has been transferred to the refueling pool where it is also available for pumping if conditions permit f. Two Spent Fuel Cooling Pumps (One per RWT) - Capacity 1390 gpm each g. Spent Fuel Pool The two halves of the SFP can be isolated from each other and 830 fuel assemblies, as a minimum, can be stored in the non-leaking half.

The four Emergency Core Cooling System (ECCS) equipment rooms on the lowest level of the Auxiliary Building (Figure 1-5) can be prevented from flooding by shutting their watertight doors. In addition, each ECCS pump room is also drained by an 80 gpm sump pump. The remainder of this level is drained by two sump pumps at a rate of 160 gpm. The sump pumps discharge to the Miscellaneous Waste Processing System (MWPS), which has storage capacity of 8000 gallons and can process 128 gpm.

CALVERT CLIFFS UFSAR 9.4-5 Rev. 47 TABLE 9-16 SPENT FUEL POOL COOLING SYSTEM COMPONENT DESCRIPTION Pump Type Horizontal, centrifugal with mechanical seals Number 2 Capacity (each) 1390 gpm TDH 200 feet Materials Casing American Society for Testing and Materials (ASTM) A296, Gr CA-15 or ASTM A217, Gr CA-15 Stuffing Box Extension Assy. (Backhead) ASTM A296, Gr CA-15, ASTM A217, Gr CA-15, ASTM A487 Gr CA-15, or ASTM A487 Gr CA6NM Class A Motor 100 hp, 460 Volt, 60 Hz, 3 phase, 3550 RPM Heat Exchanger Type Horizontal counter flow Straight tube rolled and seal welded into tube sheets Number 2 in parallel Heat Transfer area (each) 1920 ft2 Materials Shells C.S. SA-285-C Tubes SS-304, SA-213 Tube Sheets SS-304, SA-240 Shell side relief valve setpoint 150 psig Fuel Pool Filter Type Cartridge Number 1 Design/Operating Flow 128/120 gpm Design Pressure 175 psig Design Temperature 250°F Material ASTM SA240, Type 304 Fuel Pool Demineralizer Type Mixed bed, non-regenerable Number 1 Design/Operating Flow 128/120 gpm Design Pressure 200 psig Design Temperature 250°F Resin Mixed (anion, cation) Materials ASTM SA240, Type 304 CALVERT CLIFFS UFSAR 9.4-6 Rev. 47 TABLE 9-16 SPENT FUEL POOL COOLING SYSTEM COMPONENT DESCRIPTION SFP Piping, Fittings, Valves Material Stainless Steel 304 Design Pressure 160 psig Design Temperature 150°F/155°F(a) Joints 2-1/2" and Larger Butt-welded except at flanged equipment Joints 2" and Smaller Socket weld except at flanged equipment Valves 2-1/2" and Larger Stainless steel, butt weld-ends, 150 psi Valves 2" and smaller Stainless steel, socket weld ends, 150 psi Relief valve setpoint 150 psig (on tube side of spent fuel pool cooling heat exchanger) Butterflies 3" and larger Rubber seated carbon steel lug type, 150 psi _______________________ (a) Portions of the SFP Cooling System are designed for a maximum postulated temperature of 155°F [Section 9.4.4, Doc. No. 92-769(M601)].

CALVERT CLIFFS UFSAR 9.6-1 Rev. 47 9.6 SAMPLING SYSTEMS 9.6.1 DESIGN BASIS The sampling systems are designed to permit the sampling of liquids, steam, and gases for radioactive and chemical control of the plant primary and secondary fluids. 9.6.2 SYSTEM DESCRIPTION The sampling system consists of six subsystems; reactor coolant sampling, steam generator blowdown sampling, radioactive miscellaneous waste sampling, turbine plant sampling, gas analyzing sampling, and post-accident sampling systems (PASS). Figure 9-10 shows the reactor coolant, the steam generator blowdown, PASS and the waste process sample systems. Figure 9-30 shows the turbine plant sample system. Figure 9-11 shows the gas analyzing system. 9.6.2.1 Reactor Coolant Sampling Each reactor coolant sampling system consists of one stainless steel sink enclosed inside a hood. The hood is ventilated by an individual blower through a high-efficiency filter and located inside the sample room (Auxiliary Building). Interlocking high-density concrete block shielding separates the hood from the rest of the sample room, which also contains the steam generator blowdown system. The reactor coolant hood is used to determine the chemical and radiochemical condition of the reactor coolant and related auxiliary systems. The hood contains piping, valves, coolers, instrumentation, and sample bombs necessary to take liquid and gaseous samples from various systems. Two samples from the pressurizer (liquid, vapor) and one from the reactor coolant hot leg system can be controlled by three handswitches located on the steam generator blowdown panel. Should any one of the remotely-operated sampling valves fail to close after a sample is taken, a second remotely-operated valve can be shut from the Control Room. These valves are also closed by SIAS. The remotely-operated valves are backed up by manually-operated valves at the reactor coolant sampling hood. High-pressure samples flow through metering valves in order to reduce their pressure. One high-temperature sample is cooled in a sample cooler supplied with CC. All analyses on these samples are performed in the laboratory located in the Auxiliary Building.

9.6.2.2 Post-Accident Sampling If needed (see Section 1.8.1, Item II.B.3), post-accident samples can be obtained in Unit 1 or Unit 2 Nuclear Steam Supply System Sample Room in the 45' Auxiliary Building. The Unit 1 or Unit 2 Nuclear Steam Supply System Sample Room contains piping, valves, coolers, and instrumentation necessary to sample either Unit 1 or Unit 2 RCS via either the normal RCS sampling line, or Unit 1 or Unit 2 Containment sump via the LPSI system header. A grab sample is used to obtain a liquid sample from the RCS. In the Chemistry Lab, the grab sample is depressurized, degassed, and diluted as necessary to enable handling the sample without excessive radiation exposure. This grab sample capability can be used to obtain samples from the RCS or the SI system. Sample purge waste is sent to the Reactor Coolant Drain Tank of the Unit being sampled, or alternatively to the Unit 1 or Unit 2 VCT. There is a provision to analyze the dissolved gasses in the liquid sample as well as chloride and boron. The gases from the degassed coolant are vented to atmosphere via Unit 2 Plant Vent via the Chemistry Lab hoods.

CALVERT CLIFFS UFSAR 9.6-2 Rev. 47 9.6.2.3 Steam Generator Blowdown Sampling Each steam generator blowdown sampling system consists of one conditioning rack-panel unit and one ventilating hood, and is located inside the same sample room as the reactor coolant hood. The conditioning rack section of the steam generator blowdown system contains isolation valves, primary coolers, rod-in-tube devices, an isothermal bath and chiller. High pressure samples are passed through a pressure-reducing valve (rod-in-tube type) located downstream of the primary coolers and upstream of the isothermal bath. High temperature samples first pass through a primary cooler (supplied with CC) and then through the isothermal bath. All samples pass through the isothermal bath which is capable of maintaining each sample at 77°F at the coil outlet. The chiller is supplied with cooling water from the component cooling system. Sample outlets from the conditioning rack are connected to the hood.

The panel section of the steam generator blowdown system contains conductivity and pH monitors, three hand switches for pressurizer sample selection, chiller controls, and an annunciator. The pH and conductivity samples are continuously monitored and alarmed on high conductivity. In addition, pH and conductivity are trended on the computer-based display in the chemistry laboratory. High sample temperature (downstream of the isothermal bath) actuates a common alarm point. Any point alarming on the local annunciator will actuate a master alarm in the Control Room (trouble alarm). The ventilating hood contains two stainless steel sinks and is ventilated by an individual blower through a high-efficiency filter. The ventilating hood is used to obtain samples for determining the chemical and radiochemical content of the steam generator blowdown system. The radioactive miscellaneous sample system is also located inside the steam generator blowdown hood for Unit 1. The steam generator blowdown part of the hood contains all piping, grab sample valves, instrumentation including pH cells and conductivity analyzers, and all equipment necessary for this system.

9.6.2.4 Radioactive Miscellaneous Waste Sampling The radioactive miscellaneous waste sampling is located inside the ventilating hood for the steam generator blowdown (Unit 1) and is used to obtain samples from which the chemical and radiochemical content of miscellaneous waste is determined. This system is common to both units. All samples are low pressure and are cooled, as necessary, in sample coolers (supplied with CC). This part of the hood contains isolation valves, piping, valves, and instrumentation necessary for obtaining liquid samples from both units. The analyses of these samples are performed in the laboratory located in the Auxiliary Building.

9.6.2.5 Turbine Plant Sampling System Each turbine plant sampling system is used to obtain samples for determining the chemical condition of the steam, feed, and condensate systems associated with the turbine plant. The system consists of one sampling station per unit (stainless steel sink and panel) and one mechanical chiller as a separate unit. These sampling systems are located in the Turbine Building. The sink contains the isolation valves, piping, instrumentation, coolers, and grab valves necessary to take samples from the steam, condensate, and feedwater systems. High-pressure samples pass through a pressure-reducing valve (rod-in-tube device). All samples pass through individual primary coolers supplied with SRW. Every sample then CALVERT CLIFFS UFSAR 9.6-3 Rev. 47 passes through cooling coils immersed in the isothermal bath that maintains each sample at 77°F at the coil outlet. The mechanical chiller circulates chilled water in the isothermal bath and is supplied with SRW. Each sample is provided with one grab sample valve for taking liquid samples as necessary. The steam generator feed pump headers are continuously monitored and recorded for hydrazine, oxygen and pH, any of which can cause an alarm on the annunciator. All samples are continuously monitored for conductivity and an alarm occurs when an abnormal condition is reached. In addition, samples are trended on the computer-based display in the Chemistry Laboratory. The turbine plant system contains conductivity, pH, and oxygen recorders, oxygen analyzers, handswitches (to control the hotwell sample pumps and the chiller circulating pump), and an annunciator. The annunciator alarms on high conductivity, high pH, high oxygen, high hotwell temperature, and low hotwell sample pump discharge pressures. Any annunciator alarm will activate a master alarm in the Control Room. 9.6.2.6 Gas Analyzing System Control of hydrogen in Containment during and following a Design Basis Event is no longer required. On March 2, 2004, the NRC issued a license amendment that allows removal of the hydrogen recombiners and hydrogen analyzers from the Technical Specifications. The NRC has required retention of the hydrogen analyzers as non-safety-related equipment for recording hydrogen concentrations in a beyond Design Basis Event.

The gas analyzing system is used to determine the hydrogen concentration of six points inside the containment and of four samples from the reactor coolant waste tanks (receiver and monitor tanks), as well as the oxygen concentration of several samples from the reactor coolant and miscellaneous waste systems. The gas analyzing system is installed in the sample room located in the Auxiliary Building (Elevation-10') and consists of two hydrogen analyzer cabinets and separate manifolds for the isolation valves and sample selection solenoid valves and one oxygen analyzer cabinet with a manifold for the isolation valves. Two of the analyzer cabinets are for hydrogen measurement and include a sample pump, cooler, piping, valves, and instrumentation. Each hydrogen cabinet panel contains one hydrogen analyzer, one multipoint recorder for recording each measured sample, one programmer for random selection of individual readout, and alarm contacts for activation of a master alarm in the Control Room. The third analyzer cabinet is for oxygen grab sample measurement and includes a sample pump, cooler, piping, valves and sample syringe. An exhaust system on the oxygen analyzer cabinet purges any hydrogen that may leak into this cabinet. The H2 and 02 sample points are routed to the analyzer in accordance with the following table: Sample Point H2 Analyzer 0-AE-6519 H2 Analyzer 0-AE-6527 02 Grab Sample Containment 1 - North of Primary Shield No Yes No Containment 1 - South of Primary Shield Yes No No Containment 1 - Pressurizer Compartment Yes No No Containment 1 - East at Elevation 135' Yes No No Containment 1 - West at Elevation 135' No Yes No Containment 1 - Dome at Elevation 189'(b)No Yes No Containment 2 - N Yes No No Containment 2 - S No Yes No CALVERT CLIFFS UFSAR 9.6-4 Rev. 47 Sample Point H2 Analyzer 0-AE-6519 H2 Analyzer 0-AE-6527 02 Grab Sample Containment 2 - Press No Yes No Containment 2 - E No Yes No Containment 2 - W Yes No No Containment 2 - Dome Yes No No RC Waste Rec Tank 11(a) Yes Yes No RC Waste Rec Tank 12(a) Yes Yes No RC Waste Mon Tank 11(a) Yes Yes No RC Waste Mon Tank 12(a) Yes Yes No Waste Gas Decay Tank 11 No No Yes Waste Gas Decay Tank 12 No No Yes Waste Gas Decay Tank 13 No No Yes Waste Gas Surge Tank No No Yes Degasifier Accumulator 11 No No Yes Degasifier Accumulator 21 No No Yes Evaporators Discharge Gas Cooler No No Yes Przr. Quench Tank 11 No No Yes Przr. Quench Tank 21 No No Yes Misc. Waste Evap No No Yes ______________________ (a) These samples would normally be routed to either analyzer. (b) The 189' sample line in Unit 1 is inoperable because it is no longer seismically supported. For this reason it is not credited for post-accident sampling. The six containment samples of the hydrogen analyzer cabinets and two samples of the oxygen analyzer cabinet can be controlled through remotely-operated solenoid valves. To provide a post-accident containment air sampling capability, a sample vessel was placed into the sampling lines from containment 1 and 2 west at Elevation 135' to allow a syringe sample to be taken and analyzed in the laboratory. This sample vessel is located on the 45' Elevation of the Auxiliary Building. 9.6.3 SYSTEM RELIABILITY All piping, tubing, fitting, and valves (exception listed under d and e below) in contact with fluids is 316 stainless steel and complies with the following codes: a. ASME B&PV Code, Section III, Class 3 (Nuclear Power Plant Components) for the gas analyzing system. b. ANSI B31.1 for turbine plant, steam generator blowdown, post-accident sampling, reactor coolant, and miscellaneous waste sampling systems. The exception to this is the normally-closed isolation valves located in the cabinets which constitute the boundary from ASME Section III piping to non-Class piping. These valves are listed below. (NOTE: The reactor coolant, the miscellaneous waste, and steam generator blowdown sampling systems, originally designed to meet Seismic Category I requirements, were downgraded to Category II via FCR 88-0074). c. ASTM 450-68 which requires an eddy-current test for all tubing. d. Pressure relief valves that are in contact with fluid shall be made of 304 or 316 stainless steel material. CALVERT CLIFFS UFSAR 9.6-5 Rev. 47 e. Pressure reducing valves for the turbine plant and steam generator blowdown sample systems shall be constructed of Types 303, 304, or 316 stainless steel. All applicable valves, piping, and coolers are designed to accept full steam pressure and temperature.

The gas analyzing system, the component cooling portion of the sample coolers in the reactor coolant hood, and the valves listed below are designed to meet Seismic Category I requirements. The reactor coolant, miscellaneous waste, and steam generator blowdown sampling systems (within the hoods and excluding those portions delineated above) are designed to meet Seismic Category II requirements. (NOTE: The reactor coolant, the miscellaneous waste, and the steam generator blowdown sampling systems, originally designed to meet Seismic Category I requirements, were downgraded to Category II via FCR 88-0074). The following valves must be normally closed and will retain their current ASME Section III Seismic Category I classifications. Post-Accident Sampling 1-PS-172 2-PS-172 1-PS-193 2-PS-193 Miscellaneous Waste 0-PS-226 0-PS-229 The following valves were Seismic Category I and designed in accordance with ANSI B31.1. Steam Generator Blowdown 1-PS-126 2-PS-126 1-PS-128 2-PS-128 1-PS-129 2-PS-129 1-PS-137 2-PS-137 1-PS-139 2-PS-139 1-PS-140 2-PS-140 The turbine plant (Turbine Building) is designed to meet Seismic Category II requirements. 9.6.4 TESTING AND INSPECTION Each component is inspected and cleaned prior to installation into the system. Instruments were calibrated during testing. Automatic controls were tested for actuation at the proper setpoints. Alarm functions were checked for operability and limits during preoperational testing period. The system will be operated and tested for flow, capacity, and mechanical operability.

CALVERT CLIFFS UFSAR 9.10-1 Rev. 47 9.10 COMPRESSED AIR SYSTEM 9.10.1 DESIGN BASIS The Compressed Air System consists of the instrument air and plant air subsystems. The instrument air subsystem is designed to provide a reliable supply of dry and oil-free air for the pneumatic instruments and controls and pneumatically operated containment isolation valves. The plant air subsystem is designed to meet necessary service air requirements for plant maintenance and operation. The designs of each subsystem are based on an estimated instrument air requirement of 260 scfm and an estimated plant air requirement of 600 scfm. The instrument air subsystem compressor is sized for 450 scfm. 9.10.2 SYSTEM DESCRIPTION The Compressed Air System is shown schematically on Figures 9-23 (Unit 1) and 9-28 (Unit 2). The Plant Water and Air Service System is shown in Figure 9-29. The system incorporates two full-capacity, non-lubricated compressors for instrument air, each having a separate inlet filter aftercooler and moisture separator. The instrument air compressors then discharge to a single header which is connected to two air receivers. Both air receivers discharge to a compressed air outlet header which supplies instrument air to the air dryers and filter assembly. The compressed air header then divides into branch lines supplying the pretreatment and tank storage area, the Intake Structure, the service building, the water treatment area, the Turbine Building, the containment structure, and the Auxiliary Building.

An emergency back-up tie from the plant air header has been provided to automatically supply air to the instrument air system if the pressure to the instrument filter and dryer assembly falls below a preset value. Local controls are provided to prevent plant air use when this occurs. For the transition from normal to emergency service, air storage tanks provide an approximate 20-minute supply (Table 9-21).

Particle size, dew point, and oil hydrocarbons are controlled for instrument air supply in accordance with Instrument Society of America standards. Additionally, the Calvert Cliffs approach to controlling air quality was submitted to the NRC in response to Generic Letter 88-14.

One full-capacity plant air compressor with an inlet filter, and integral air coolers and moisture separators, discharges to the plant air receiver. The receiver outlet header is connected to the prefilter assembly, which is followed by an outlet header branching into two separate air headers, one to the instrument air dryers and filter assembly, and the other to the plant air pretreatment and storage tank area, the Intake Structure, the service building, the water treatment area, the Turbine Building, the Containment Structure, and the Auxiliary Building. A system cross-tie between Unit 1 and Unit 2 has been provided for the plant air headers. Additionally, each plant air system has a permanent connection for the installation of a portable air compressor to allow for maintenance of the compressors or SRW system during Modes 3, 4, 5, 6 and defueled. This connection may also be used in Modes 1 and 2 to provide a contingency backup to an operating plant air compressor should the other installed plant air compressor be unavailable.

9.10.3 SYSTEM COMPONENTS Ratings and construction of system components are listed in Table 9-21. 9.10.4 SYSTEM OPERATION A continuous supply of instrument air is provided to hold various pneumatically-operated valve actuators in the positions necessary for operating conditions. Normally, the plant air CALVERT CLIFFS UFSAR 9.10-2 Rev. 47 compressor and one instrument air compressor will operate and the second instrument air compressor will be on automatic standby.

9.10.5 SYSTEM RELIABILITY The power supply for the normal compressors is the normal distribution system and can be backed up by the EDG. Additional emergency air compressors, known as the saltwater air compressors (SWACs), provide redundant air supply to most safety-related components when the normal air compressors are lost. The SWACs (Table 9-16B) are seismically qualified, air-cooled, and oil-free. The instrument air portion of the compressed air system is primarily used for valve actuation and is not used in any reactor indication, control, or protective circuitry. These valve actuators are designed to fail in the safe position after loss of the instrument air supply. The design of the system and installed equipment redundancy ensure that total loss of instrument air supply is highly improbable. Concurrently, attention has been given to ensure that valve failures from loss of instrument air supply are consistent with the capability to maintain the plant in a safe condition and mitigate the consequences of any simultaneous incident or accident. 9.10.6 TESTS AND INSPECTIONS Each component is inspected and cleaned prior to installation into the system. Instruments were calibrated during testing and automatic controls were tested for actuation at the proper setpoints. Alarm functions were checked for operability and limits during plant operational testing. The systems were operated and tested initially with regard to flow paths, flow capacity, and mechanical operability.

CALVERT CLIFFS UFSAR 9.10-3 Rev. 47 TABLE 9-21 COMPRESSED AIR SYSTEM COMPONENT DESCRIPTION A. INSTRUMENT AIR SYSTEM Air Compressor Type Vertical, non-lubricated reciprocating, two state Y-angle type Quantity 2 (per unit) Design capacity (scfm) 470 (each) Discharge pressure (psig) 100 Motor 100 hp, 3 phase, 60 Hz, 460 Volt Code ASME Section VIII, NEMA Intake Filter - Silencer Type dry Quantity 2 per Unit Base size 8" Aftercooler and Moisture Separator Type Shell and tube Quantity 2 (1 per compressor) Code TEMA Class C, ASME Section VIII Air Receiver Type Vertical Quantity 2 (1 per compressor) Design pressure (psig) 115 Actual volume (ft3) 96 Code ASME Section VIII Prefilters Type Cartridge Quantity 2 per Unit Capacity (scfm) 720 Filtration 99% removal of all liquids, oil, and water droplets Air Dryer Type Heatless Desiccant Activated alumina absorbent Quantity 2 per unit Capacity (scfm) 475 (Nos. 12 and 22), 700 (Nos. 11 and 21) Outlet moisture content with saturated air inlet -40°F dew point at 100 psig Afterfilters Type Cartridge Quantity 2 per Unit Capacity (scfm) 600 Filtration 100% removal of all particulates over 0.9 microns CALVERT CLIFFS UFSAR 9.10-4 Rev. 47 TABLE 9-21 COMPRESSED AIR SYSTEM COMPONENT DESCRIPTION Piping and Valves Valves 150 psi ANSI for 2-1/2" and larger, 600 psi ANSI for 2" and smaller Piping Seamless ASTM A106, Grade B (2-1/2" through 24") Code ANSI B31.1 (ANSI B31.7 - penetration piping) B. PLANT AIR SYSTEM Air Compressor Type Centrifugal, two stage, with integral air coolers and moisture separators Quantity One per Unit Design capacity (scfm) 600 Discharge pressure (psig) 100 Motor 200 hp, 3 phase, 60 Hz, 460 Volt Code NEMA Intake Filter Silencer Type Dry Quantity One per Unit Air Receiver Type Vertical Quantity 1 Design pressure (psig) 115 Actual volume (ft3) 96 Code ASME Section VIII Prefilter Type Cartridge Quantity 2 per Unit Capacity (scfm) 720 Filtration 99% removal of all liquids, oil, and water droplets Piping and Valving Valves 150 psi ANSI for 2-1/2" and larger, 600 psi ANSI for 2" and smaller Piping Seamless ASTM A106, Grade B (2-1/2" through 24") Code ANSI B31.1 (ANSI B31.7 - penetration piping) C. INSTRUMENT BACKUP AIR SYSTEM Storage Tank Type Vertical Quantity 4 Capacity 300 ft3 Design pressure (psig) 225 Code ASME Section VIII Air Amplifier Ratio 2:1 IXl M t><l 1-....1 I+ I t::<I .... ll<l GATE* GATE

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CHAPTER 10 STEAM AND POWER CONVERSION SYSTEM TABLE OF CONTENTS PAGE10.0 STEAM AND POWER CONVERSION SYSTEM 10.1 MAIN STEAM SYSTEM 10.2 CONDENSATE AND FEEDWATER SYSTEM 10.3 AUXILIARY FEEDWATER SYSTEM 10.4 AUXILIARY BOILER STEAM SYSTEM 10.5 TURBINE-GENERATOR AND CONDENSER SYSTEM APPENDIX 10A HIGH ENERGY PIPE RUPTURE OUTSIDE CONTAINMENT CHAPTER 10 STEAM AND POWER CONVERSION SYSTEM LIST OF TABLES TITLEPAGE CHAPTER 10 STEAM AND POWER CONVERSION SYSTEM LIST OF FIGURES FIGURE CHAPTER 10 STEAM AND POWER CONVERSION SYSTEM LIST OF ACRONYMS

CALVERT CLIFFS UFSAR 10.4-1 Rev. 47 10.4 AUXILIARY BOILER STEAM SYSTEM 10.4.1 DESIGN BASIS The Auxiliary Boiler Steam System is shown on Figure 10-6. Two auxiliary boilers are provided, each having the following principal characteristics: Steam generating capacity: 125,000 lbs/hr Steam pressure: 180 psig (operating) 250 psig (design) Steam temperature: 380°F Feedwater temperature: 180°F (normal) 40°F (at startup) Each auxiliary boiler was sized to satisfy the plant heating plus the condensate startup deaeration steam requirements. If required, the auxiliary boilers can be used to supply the necessary steam supply for the steam generator auxiliary feed pumps and the SGFPs. Component design data is contained in Table 10-1. 10.4.2 SYSTEM DESCRIPTION Three fuel oil pumps are provided for the two boilers. Each pump has a capacity of 10,000 lbs/hr at a discharge pressure of 150 psig. Three feedwater pumps are provided for the auxiliary boilers and take suction from a common deaerator. The design flow of each pump is 265 gpm at 580' total discharge head. The plant fuel oil system consists of two outdoor bulk storage tanks for No. 2 fuel oil serving the auxiliary boilers, the emergency diesel generators, the Station Blackout Emergency Diesel Generator, and the diesel-driven fire pump. There is an additional fuel oil system for the Societe Alsacienne De Constructions Mecaniques De Mulhouse, safety-related diesel generator, No. 1A. This fuel oil has a viscosity range as specified in American Society for Testing and Materials (ASTM) D975-81, Table 1. Two of the storage tanks are tornado-protected. Truck unloading pumps and tank transfer pumps are also provided.

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• i ; *.; ... :: ; i .. I!" *Pl s li i *** e ! !' i ; lw 4 .. ; 34 x 32 x 36 FIG.612 GJMMPOTY LIST OF MATERIAL FOR VALVE ONLY CAI.VOIT Q.lfTS MJCl.EAR POlt:R Pl.NIT EHC&l*tRPG suv1as DCPMlWDft CAl.vPI Q.ffS 11111 "'2 I i I i . 085*JJ7SJ-OISHOOOS 1*<v. r:_c-15382-0030 I 3 Revision 48 r ."""' '!'o'* .. , *--*---*-*-.. *-*---____ _ LIST OF MATERIALS QUANTITIES ARE FOR ONE VALVE STOP, THINK, ACT ORDER DATE BM ND 00720997-33753-01 , ! A WHERE SPECIFICATIONS ARE JNOICATEO, THE LATEST REVISION APPLIES ! PC ROCKWELL ! A
• NO. DESCRIPTION QTY, RMC NO. ! DESCRIPTION i l ; i ! B c D 51 52 *SJ 54 55 55.1 55.2 *56 57 58 59 60 61 62 63 *64 *65 66 67 68 69 70 71 72 73 74 75 76 77 78 79 BO 81 82 83 FLANGE, GLAND RING, SPACER BONNET IPTI STUD, SPECIAL GLAND ASSEMBLY GLAND BUSHING, GLAND DISK !PT> GASKET, P.S. RING, JUNK PACKING COVER, SPRING GUIDE SPRING, CONICAL WASHER, BELLEVILLE RETAINER, BONNET RETAINER, GASKET BODY ASSEMBLY RING, YOKE LOCK NUT, HEX. STUD F.T. CONNECTOR, 45" CONNECTOR 90" CONN./CABLE ASSY. 10" LG. AOAPATOR L LOCK RING ASSY. BOX, JUNCTION HUB, MEYERS HUB, MEYERS CONNECTOR , WIRE TERMINAL BLOCK TERMINAL MARKER, VINAL PANEL, JUNCTION BOX TUBING, MET AL FLEXIBLE CONNECTOR, STRAIGHT SCREW, RO. HO. MACHINE SCREW, RD. HD. MACHINE 1 1 1 2 2 1 72 I 1 I I 4 2 2 6 6 6 2 6 2 54 6 I 2 8' LG. 4 8 20 24074/06080 01111 01112 20003/06040 02360/060BO 02360 35091 01112 05101106150 04200 55555 01200 04980 01430 01150 02841 01023 01021 01290 02862 55555 55555 55555 55555 55555 55555 55555 55555 55555 55555 55555 55555 55555 55555/06040 55555106040 Al.LOY STL.ISOLID LUBRICANT CTG. CARBON STEEL CARBON STL. Al.LOY STL./CAO. PL. Al.LOY STL./SOLID LUBRICANT CTG. Al.LOY STL. ALUMINUM BRONZE CARBON STL. CARBON STL.ISILVER NICKEL BASE Al.LOY PACKING CARBON STL. INCONEL X750 HIGH CARBON STEEL MILO CARBON STEEL STAINLESS STEEL CAST CARBON STEEL CAST CARBON STEEL Al.LOY STEEL Al.LOY STEEL ST ANOARD PART ST ANOARD PART ST ANOARD PART STANDARD PART STANDARD PART ST ANDARO PART STANDARD PART STANDARD PART ST ANOARD PART STANDARD PART STANDARD PART STANDARD PART ST ANOARD PART ST ANDARO PART /CAO. PL. STANDARD PART /CAO. PL. NONDESTRUCTIVE EXAMINATION CODES* IPTI
• LIQUID PENETRANT TEST !RT>
• RADJOGRAPH TEST *PRESSURE RETAINING PARTS SPECIFICATION 4140/MOL YBOENUM OJSULnDE ASTM A-105 !SEE NOTE 2l ASME SA-105 COMMENTS USE EXISTING PART ;:.4 ;l;UlfIDE "8 11 ASTM A-331 GR.4140 HT (<SEE NOTE Jl).... J ASTM B-148 GR. C95200, OR GR. C9S400 (1SEE NOTE 4j ASME SA-105 AISJ 1005-1010/BRIGHT SILVER WAUKESHA-88 CHESTERTON STYLE 1 ASTM A-108 GR. 1018-103D AMS 5699 C104574QLT ASTM A-515, GR. 70 ASTM A-461, GR. 660 ISEE NOTE 11 ASTM A-216, GR. wee ASTM A-216, GR. wee ASTM A-194, GR. 7 ASTM A-540, GR. B23, CL.4 COMMERCIAL COl.tMERCIAL COMMERCIAL COMMERCIAL COMMERCIAL COMMERCIAL COMMERCIAL COMMERCIAL COMMERCIAL COMMERCIAL COMMERCIAL COMMERCIAL COMMERCIAL COMMERCIAL/ ASTM A-165 TYPE TS COMMERCIAL/ ASTM A-165 TYPE rs SUPPLIED BY CUSTOMER USE EXISTING PART USE EXISTING PART USE EXISTING PART USE EXISTING PART USE EXISTING PART USE EXISTING PART ASME SECTON III 1971 EDITION SUMMER 71 ADDENDUM NUCLEAR CLASS 2 VALVE "N" STAMP e! I l -! ! i ! cl ! I I j j .,__, ! ! ! i ! ! 01 j 1. REPLACEMENT MATERIAL TO BE SA638 GR.660 TYPE 2. THIS DRAWING WAS REPRODUCED FROM ROCKWELL INTERNATIONAL MEASUREMENT L FLOW CONTROL DIVISION DWG. NO. DB5-33753-01, REV. C, DATED 1D/06/BB. BCE F.P. DOC. NO. 153B2-0031SHD005A. ! ! IREF. t:.51997017!>7-<JOO REV.OJ "8" 2. REPLACEMENT MATERIAL TO BE ASTM A-1;68, GR.4140 CL.L, -"" nv <;:fs:'s:'1 ,,..,.. ..,, ..,,,,. r..in """'"""-3. ASTM A-322 GR. 4140 HT IS AN ACCEPTABLE ALTERNATE MATERIAi.i i IECP-13-0005681. I 4* IS AN ACCEPTABLE ALTERNATE MATERIAL I! [ --CUST.* BALTIMORE GAS & ELECTRIC CO. P.O. NO.' 81008-{;X r TAG NO'S, 1-CV-4048/LINE NO. 34EBl-1002 11(¥ DAT[ AS BUU.T, Jf#"O ONL'f ESP £5199800048-000 JA ---KWJ 34 x 32 x 36 FIG. 612GJMMPOTY LIST OF MATERIAL FOR VAL VE ONLY i -i ; ! j I I $. I C.11. VERT CLJrrS lllCl.ENI POWER Pl.NIT [ l DClfrfiRING SDtYICES D£PARfMENT I!
• 2-CV-4048/LINE NO. 34EBJ-2002 Ll= __ , _______ J_ ______ -1c.11.VERr a.m !Hr 2 le I '1 i oes-33153.01 .. 15382-<lOJlSHOOOSA 8 Revision 48

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CALVERT CLIFFS UFSAR 10A.1-1 Rev. 47 10A.1 MAIN STEAM Each steam generator is connected with a single 34" pipe line to the steam header near the turbine. Each steam line will carry approximately 5.6 million pounds of steam per hour during rated power operation. These lines penetrate the Containment at Elevation 38'0" and pass through the Auxiliary Building to the Turbine Building. The MS System, shown in Figure 10A.1-1, will vary normally between 900 and 850 psia for no-load and full-load operation, respectively. A flow-limiting nozzle, located in the Containment, will protect the primary system against an excessive cooldown rate in the event of a main steam line break (MSLB). Pressure in the MS system is maintained primarily by the reactor coolant temperature. A turbine by-pass system, with a capacity of 40% of the rated steam flow, and an atmospheric dump system, with a capacity of 5% of the rated steam flow, provide additional control of the MS pressure during load changes. In addition, 16 relief valves protect the MS system from abnormal pressure above 1050 psia. 10A.1.1 PIPE WHIP The MS system normally operates at a pressure above 275 psig and 200°F and, therefore, protection is provided for pipe whip following a longitudinal or circumferential break. 10A.1.2 CRITERIA FOR PIPE BREAK LOCATION Pipe breaks are postulated to occur at the following locations: a. Terminal ends; b. Any intermediate locations between terminal ends where either the circumferential or longitudinal stresses derived on an elastically-calculated basis under the loadings associated with seismic events and operational plant conditions exceed 0.8 (Sh + SA)* or the expansion stresses exceed 0.8 SA; and, c. Two additional intermediate locations are selected on the following reasonable bases: 1. The points of highest stress, Figure 10A.1-2 identifies the location of the high stress points. Table 10A-1 lists the stress values for these points; and, 2. No break in short-run pipes up to five pipe diameters. d. A critical crack defined as one-half the pipe diameter in length and one-half the pipe wall thickness in width is postulated to occur at any location. 10A.1.3 CRITERIA FOR PIPE BREAK ORIENTATION A longitudinal pipe break is considered for lines 4" and larger. The break is assumed to be parallel to the pipe axis and oriented at any point around the pipe circumference. A circumferential break is considered for lines exceeding a nominal pipe size of 1". The break is assumed to be oriented perpendicular to the pipe axis. A critical crack is assumed to be oriented at any point around the pipe circumference. 10A.1.4 SUMMARY OF PIPE WHIP DYNAMIC ANALYSIS 10A.1.4.1 Location of Number of Breaks The locations and number of design basis breaks are chosen in accordance with the criteria in Section 10A.1.2. Two types of breaks, longitudinal break and circumferential break, are considered in accordance with the criteria in Section 10A.1.3.

• Pressure greater than 275 psig and/or temperature greater than 200°F.

CALVERT CLIFFS UFSAR 10A.1-2 Rev. 47 The critical crack is considered to occur anywhere on the line. Figure 10A.1-2 shows the postulated pipe break locations for the MS line. 10A.1.4.2 The Postulated Rupture Orientation The longitudinal break is parallel to the pipe axis and oriented at any point around the pipe circumference. The longitudinal break area is equal to the effective cross-sectional flow area upstream of the break location. The circumferential break is perpendicular to the pipe axis, and the break area is equivalent to the internal cross-sectional area of the ruptured pipe. Dynamic forces resulting from a circumferential break are assumed to separate the piping axially, and cause whipping in any direction normal to the pipe axis.

The critical crack is oriented at any point around the pipe circumference. 10A.1.4.3 Description of Forcing Function Design parameters to estimate steam-water blowdown thrust and jet impingement forces expressed in term of FT/POAB as a function of the friction parameter. Flow resistance coefficient (fL/D), and upstream area restriction parameter, AB/AR, are respectively presented in Figures 10A.1-9 and 10A.1-10. In addition, graphical solutions to predict the impingement force experienced by the target object as a function of fL/D are also plotted in Figures 10A.1-11 and 10A.1-12.

DISCUSSION AND APPLICATION 1. Blowdown Thrust Loads Thrust and jet impingement forces are produced during a rapid blowdown of a high pressure vessel. Thrust reaction force is a summation of the momentum expulsion rate and the exit plant pressure force. Momentum flow rate is the product of velocity and mass flow rate. Furthermore, the blowdown mass flow rate, velocity, and exit plane pressure are determined by vessel pressure, fluid properties, and the escape geometry. It follows that, for a given pressure vessel blowdown, thrust reaction force is totally determined. The total steady thrust reaction force may be written as follows: (1) or (2) For definition of terms, refer to the list of notations at the end of this section. Figure 10A.1-8 shows a blowdown of steam or saturated water from a vessel through an arbitrary pipe. The impingement target is located sufficiently far away that full jet expansion to environ pressure, P has occurred. It follows from conservation of momentum equations that the steady thrust reaction force and total jet impingement force per unit break area AB can be expressed by: (2) (3) CALVERT CLIFFS UFSAR 10A.1-3 Rev. 47 Furthermore, a simple force balance on the steady jet which impinges normally on the flat wall shows that the total jet force and the total thrust are equal but opposite in direction giving (4) If ideal gas is assumed as the fluid flowing and blowdown through an isentropic nozzle (zero friction), it follows that for k=Cp/Cv = 1.3 the thrust may reach the maximum value (5) However, pipe friction and upstream area restrictions significantly affect the steady thrust loads. Pipe friction effects on steam or saturated water blowdown steady thrust can be incorporated in the steady thrust loads from Figure 10A.1-9. Figure 10A.1-9 can also be used to estimate the thrust load for pipe break of any water line that is directly connected to the pressure vessel provided that subcooling of the blowdown zone in the pressure vessel is not greater than 22 Btu/lb. If the postulated rupture pipe has an upstream area restriction such as flow-limiting venturi or feedwater orifice, the steady thrust loads can be seriously affected. Figure 10A.1-10 should be used to determine steady thrust loads for various break-to-restriction-area ratios in circumferentially ruptured pipes that initially contained steam or water.

2. Jet Impingement Loads Blowdown flow will form a jet which can produce impact forces on pipes or other mechanical target objects in its path. Total steady-state jet impingement force per unit break area is given in Equation (3). It follows from Equations (4) and (5) that the maximum value of total steady-state jet load for saturated steam or steam/water mixture blowdown through an isentropic nozzle where entire jet intercepted by target is (6) If the blowdown pipe friction is significant, Figure 10A.1-9 would be used to determine Fj/A. If there is an area restriction in the line, use Figure 10A.1-10 with Fj = FT.

Total force on target objects, which are submerged in a jet (i.e., target area AT is less the fully expanded free jet area A) and do not fully intercept the jet, can be estimated from the product of "jet pressure" Fj/A (Figure 10A.1-11) and projected target area, AT. If AT is greater than A (target intercepts the jet), the full jet load Fj, which is equal to total thrust in Figure 10A.1-9 should be used. If the target is very close to a break where jet originates, full expansion will not occur so that Figure 10A.1-11 is invalid. Data of Faletti (Reference 1) indicates that full jet expansion probably occurs about five pipe diameters of axial travel after leaving the break. Therefore, whenever L/D5, jet pressure of Figure 10A.1-11 is valid. However, if L/D<5, a jet pressure equal to FT/A, and jet area A would be more appropriate. CALVERT CLIFFS UFSAR 10A.1-4 Rev. 47 Whether or not the target is fully submerged in a jet can be determined from the jet expanded area as follows: A) L/D5. Figure 10A.1-12 gives the expanded area, A. If AT<A target is fully submerged and impingement load = jet pressure x AT. If ATA target intercepts entire jet and impingement load = Fj. B) L/D<5. If AT<A target is fully submerged and impingement load = FjAT/A. If AT>A target intercepts entire jet and impingement load = Fj. Notations A = Jet area, ft2 AB = Pipe flow area or break flow area, ft2 AR = Restriction flow area, ft2 A = Fully expanded free jet area, ft2 AT = Projected target area, ft2 D = Pipe hydraulic diameter, ft FT = Total thrust, lbf Fj = Jet impingement force, lbf f = Moody friction factor L = Equivalent pipe length for pressure loss from vessel L = Distance from pipe break to target, ft Po = Vessel pressure, lbf/ft2 Psat = Saturation pressure, lbf/ft2 P2 = Exit plane pressure, lbf/ft2 P = Atmospheric pressure, lbf/ft2 = Fluid density, lbm/ft3 = Fluid specific volume, ft3/lb G = Mass flow rate per unit area V = Velocity CP = Constant pressure specific heat CV = Constant volume specific heat gc = Newton's constant, 32.2 lbm-ft/lbf-sec2 10A.1.4.4 Mathematical Model and Dynamic Analysis A large pipe break is assumed to be a one-time event, requiring a plant shutdown and necessary repairs. Permanent deformation of the pipe and restraint are allowed. An energy balance method was used for the pipe whip restraint design. This method is similar to the maximum deflection of a system subjected to a long duration loading relative to the natural period as presented on Page 222 of Reference 6. The mathematical model is shown in Figure 10A.1-5. When required to accommodate the thermal movement of the pipe, gaps were provided CALVERT CLIFFS UFSAR 10A.1-5 Rev. 47 between the pipe and the restraint, and the effects of these gaps were considered in the dynamic analysis. Thus, the formula shown on Page 222 of "Introduction to Structural Dynamics" is modified as follows: where: F = Jet force acting upon the pipe Yg = Gap between the pipe and the restraint m = Ductility ratio, i.e., Ymax/Yel Rm = Restraint resistance force Yel = Deflection of the pipe and the restraint at the yield stress As suggested by the AEC, a comparison was made between a time-history analysis method and the energy balance method of the restraint design. The results of this comparison are shown in Table 10A-7. A comparison was made on four different restraints. As shown in this table, a stepped forcing function was used in the time-history analysis, compared to a straight line forcing function (Figure 10A.1-15) used in the energy balance method of analysis. Two different sets of analyses were performed in the comparison. In the first set of calculations, the elastic deflections given by the time-history method, were used in the energy-balance method. In the second set of calculations, properties of a given restraint were used in the energy-balance method.

The results of this comparison indicated that the resistance force on the restraint (yield capacity of the restraint) will be similar in both analyses. 10A.1.4.5 Unrestrained Motion of the Ruptured Line The MS line is restrained at the postulated break locations and additional restraints are provided to preclude axial movement within the encapsulation sleeve. No damage, therefore, can occur to structures, systems and components important to the plant safety due to a MSLB. 10A.1.5 PROTECTION AGAINST PIPE WHIP, JET IMPINGEMENT, AND REACTIVE FORCE 10A.1.5.1 Pipe Whip Restraints and Encapsulation 10A.1.5.2 The MS line is encapsulated and restrained to prevent pipe whip, jet impingement, or reactive forces from damaging other plant components and structures required for safety following a longitudinal or circumferential break. The MS line encapsulation sleeve is designed in accordance with the following criteria: a. The encapsulation sleeve is designed and supported in a manner which will not introduce significant strain concentrations on the encapsulated section of piping. b. The piping beyond the encapsulation sleeve is provided with pipe whip restraints (or anchors) which restrict its axial displacement and motion within the sleeve following a postulated circumferential pipe break. c. The encapsulation sleeve is designed (a) to withstand the dynamic forces of internal pressurization resulting from the escape of high energy fluid at the postulated pipe break location assuming complete pipe severance and axial separation to the extent permitted by the pipe restraints, and (b) to CALVERT CLIFFS UFSAR 10A.1-6 Rev. 47 restrict the flow at the open ends of the sleeve to a level required to preclude compartment pressurization beyond the allowable structure design limits. d. The stresses imposed on the encapsulation sleeve during dynamic pressurization are limited to the design limits associated with "emergency condition" as permitted by the rules of American Society of Mechanical Engineers (ASME) Section III - Nuclear Power Plant Components Code, for Class 2 components. e. All material for use in the encapsulation sleeves was procured to the requirements of Article NC-2000 of ASME Code, Section III, 1971. f. Fabrication of the encapsulation sleeves is in accordance with the requirements of Article NC-4000 of ASME Code, Section III, 1971. g. Full-penetration shop welds were radiographed in accordance with ASME Code, Section III, Class 2, and Code Case 1554. h. Full-penetration field welds of the encapsulation sleeve were magnetic particle or liquid penetrant examined in accordance with the procedures described in Appendix IX-3500 or IX-3600 of ASME Code, Section III, 1971, with the acceptance standards of paragraph NB-5320 of the Code. Examinations were performed at the one-third level, two-thirds level and of the final welded surface. i. The design of the encapsulation sleeve permits either its removal by machinery or flame-cutting techniques, or the replacement of encapsulated pipe section in the event leaks develop which require repair or replacement of the pipe. j. Pipe weld joints located within the encapsulation sleeve and not accessible for subsequent ISI were non-destructively examined prior to the assembly of the encapsulation sleeve. The results satisfy the acceptance standards of ASME Section XI, Inservice Inspection Code. k. The encapsulation sleeve is provided with open vent and drain pipe nipples as a means of monitoring the encapsulated pipe section for any leaks which might develop in service. These nipples extend beyond the pipe insulation. l. The piping welds not encapsulated within the piping runs traversing safety-related areas, or within compartments adjoining safety-related areas were subjected to periodic inservice examinations in accordance with ASME Section XI Code Class 2 component requirements except that 100% of such welds were examined during each inspection interval. Alternatively, a risk-informed process for piping outlined in Reference 8 may be used for the weld selections and the determination of additional examinations when defects are discovered. This applies to the MFW and MS systems within the Auxiliary Building. Figure 10A.1-3 shows an encapsulation detail for the MS System. The fluid head at the containment penetration is designed for pressure build-up or movements due to a pipe break in the Auxiliary Building.

The jet forces from a critical crack of less than 10 kips are not significant enough to create a pipe whip affecting Category I structures, systems, and components. The jet forces will produce low bending stresses well within the elastic range of the pipe with an expected pipe movement of less than 1/4" in the worst case. The jet impingement force resulting from a single critical crack will be shielded as required to prevent damage to the safety-related components, systems, and equipment. For further discussion see Section 10A.1.13C. CALVERT CLIFFS UFSAR 10A.1-7 Rev. 47 For those locations where the postulated break area would exceed 28.9 in2, (the area of the largest branch line) the pipe is encapsulated to limit the blowdown to less than 291 lbm/sec (analysis in Section 10A.1.20). This is accomplished by limiting the release area between the ends of the encapsulation and the pipe to a net area less than 28.9 in2. The encapsulation will also dissipate the jet impingement forces.

Following a steam line break, the pressure will instantaneously build up inside the encapsulation because of the restriction of blowdown through the gap. Supports are located between the encapsulation and the pipe to prevent displacement of the pipe normal to its axis. Whipping restraints are located so that the encapsulations are rigidly held in place and the MS lines are prevented from pulling out of the encapsulations. A vent stack has been provided to vent the MS line Penetration Room to a compartment pressure below the acceptable level, which would affect the integrity of the Category I structures, system or components important to plant safety (Section 10A.1.20). 10A.1.5.3 Separation Provisions 10A.1.5.4 The MS lines are run parallel to each other approximately 5'10" apart. Separation of redundant features of the MS lines is accomplished by a combination of encapsulation and properly placed restraints. The safety relief valves are arranged such that jet forces from the safety relief valves on one line will not affect the valves on the adjacent line.

The existing exhaust stack support steel (12" structural members) between each relief valve inlet will provide protection and separation of adjacent MS relief valves from the jet impingement force resulting from a circumferential or slot break at the 6" MS nozzle to the relief valves.

Additional steel was provided in the area of the relief valves where required to ensure complete protection against jet impingement from a 6" MSLB. A jet impingement barrier, which consists of a steel plate, is provided between the MS line and Lb wall to protect the wall from the jet force resulting from a 6" MSLB. 10A.1.5.5 Description of a Typical Pipe Whip Restraint The pipe whipping restraints are provided at the postulated break locations. Additional restraints are provided near elbows and other critical locations to control the pipe whip impact and axial movement due to a full break at the postulated break locations. Figure 10A.1-2 shows the location of restraints for the MS line.

The design and detail of a pipe whip restraint depends upon many variables, such as physical location, amount of force to be sustained and thermal movement of the pipe. A typical pipe whip restraint is a rigid structure of heavy structural steel members and/or steel plates. It is supported from the existing structural components, such as floors, walls and columns. When the restraint loads cannot be sustained by the existing structure of structural components, these loads are transferred to the foundation level using additional supports. Figure 10A.1-4 shows details of a pipe whip restraint for the MS line.

CALVERT CLIFFS UFSAR 10A.1-8 Rev. 47 10A.1.6 EVALUATION OF SEISMIC CATEGORY I STRUCTURES 10A.1.6.1 Method of Evaluating Stresses Category I existing and added structures were evaluated for structural adequacy following a postulated rupture using the design bases shown in Appendix 5A. Ultimate strength design method for concrete was used as given in the above reference.

10A.1.6.2 Allowable Design Stress Design stresses are proportioned such that the combined stresses are within the limits established in Appendix 5A. 10A.1.6.3 Load Factors and Load Combinations Load factors and load combinations are discussed in Section 10A.1.7. A further discussion of load factors and load combination is provided in Section 5. 10A.1.6.4 Stresses in Category I Structure Main steam line is encapsulated at the postulated break locations. The jet impingement forces resulting from a postulated pipe break are retained in the encapsulation pipe and are not taken by the structure or structural components. Any jet forces escaping from the encapsulation pipe are distributed such that they will not affect the structure.

The magnitude of a jet impingement force, due to a critical crack in the MS line, will be less than 10 kips. A simplified approach to impingement forces assumes the jet to disperse uniformly at the half angle of incidence between jet axis and the target surface. The half angle, , is taken as 10°. Thus, the pressure at distance X is: Pj = Fj/Aj where, Pj = Effective jet pressure on the target Fj = Jet impingement forces in kips Aj = The cross-sectional area, in square inches, normal to the jet Table 10A-2 shows the concrete and steel stresses due to jet impingement forces resulting from a critical crack plus 1 psi compartment pressurization on various structural components in the vicinity of the MS line. Table 10A-3 shows the concrete and reinforcing steel stresses due to the pressurization of 2.6 psi resulting from a postulated pipe rupture. The calculated stresses shown in the above tables are combined stresses, including the effects of pipe rupture, plus the effects of live load, dead load, equipment load, and Safe Shutdown Earthquake (SSE) loads. Allowable stress for the concrete is taken at 85% of the ultimate strength. Concrete, having an ultimate strength of 4,000 psi, is used. The allowable stress for the reinforcing steel is taken at 90% of the yield strength. Reinforcing steel, having a minimum yield of 40,000 psi, is used. The structures are also evaluated for the effects of pipe breaks which are transmitted through the restraints.

10A.1.6.5 Erosion of Concrete from Jet Impingement Forces Since encapsulation pipes are used to prevent full area pipe rupture jet forces from effecting the structure, the only jet impingement force that must be considered is CALVERT CLIFFS UFSAR 10A.1-9 Rev. 47 from the critical crack. The most severe jet force condition occurs where the steam line is 1' away from the concrete. The exit velocity is expected to be approximately 1500 fps with a total force of 10,000 lbs distributed over an area of 84 in2. Most of the work done relating to blast erosion of concrete has been with reference to blast from jet engines of aircraft. Some of the effects of jet blast have been discussed in Reference 7. The work has been done at 1250°F and velocities at 3500 fps. The results of these tests and actual service showed that concrete pavement suffered light damage. Since our velocities and temperatures are considerable less than those obtained from the jet engine, excessive erosion of our structure concrete will not be a problem. The jet impingement forces which are expected will be on a local area for a relatively short duration and should not damage the structure adequacy. However, if the effects of these jet forces are determined to significantly erode the concrete, steel shielding plates will be provided as required. 10A.1.7 STRUCTURAL DESIGN LOADS The following design loads are used to evaluate the adequacy of Category I structures following a postulated rupture: Dead Load - Actual weight of structural elements supported. Live Load - Maximum expected live load in the area under consideration. Equipment Load - Actual static load of equipment. Pipe Load - Maximum calculated forces expected under normal operating and upset conditions. The forces include dead load, seismic forces, and thermal forces. Pressurization - The maximum expected compartment pressure build up that would result from a postulated rupture. Jet Impingement - Jet impingement forces resulting from full pipe area breaks are retained in the encapsulation pipe and are not taken by the structure. The forces resulting from critical cracks were considered. Temperature - The effects of temperature increase from a pipe rupture are considered to be short term increases and will not affect the structure adequacy. Seismic Forces - Seismic forces as shown in Appendix 5A. These loads are combined using the following load combination equations to evaluate the structural integrity of a Category I structure following a postulated high energy pipe line rupture. Y = 1/ (1.25D +1.00R + 1.25E) Y = 1/ (1.25D +1.25H + 1.25E) Y = 1/ (1.00D +1.00R + 1.00E') Y = 1/ (1.00D +1.00H + 1.00E') Y = required yield strength of the structure D = dead load of structure, actual static weight of equipment, expected live load in the area under consideration. In addition, any other permanent loads contributing stress, such as soil or hydrostatic loads. R = reactions from the pipe whip restraints, the maximum expected compartment pressure build-up that would result from a postulated rupture and jet impingement forces resulting from the critical crack (jet impingement forces CALVERT CLIFFS UFSAR 10A.1-10 Rev. 47 resulting from a postulated pipe break are retained in the encapsulation pipe and are not taken by the structure). H = maximum calculated forces expected under normal operating and upset conditions. The forces include dead loads, seismic loads and thermal expansion of restrained pipes under normal operating conditions. E = Operating Basis Earthquake (OBE) load. E' = SSE load. = yield capacity reduction factor as defined in Appendix 5A. 10A.1.8 REVERSAL OF LOADS ON THE STRUCTURE The forces which could cause reversal of loadings due to the postulated accident, on the Seismic Category I structures or structural components are: a. Jet Impingement Force b. Compartment Pressurization c. Reaction from Pipe Whip Restraint Since the MS line is encapsulated at the postulated full break locations, the existing Category I structures or structural components will not be affected by the jet impingement forces.

A vent stack is provided to vent the MS line compartment at Elevation 27'0" (Figure 10A.1-7). The pressure in the MS line compartment, due to a postulated full break, will be limited to an acceptable level by providing the vent stack and the encapsulation pipe (Section 10A.1.20). The maximum pressure, in the MS line compartment, will not affect the integrity of the Category I structures or structural components.

Pipe whip restraints are supported by the existing structural components. When the restraint loads cannot be sustained by the existing structure of structural components, these loads are transferred to the foundation level using additional supports. The effects of jet impingement forces and the pressurization due to the postulated single critical crack were insignificant except in the existing pipe tunnel. The roof of this pipe tunnel was adequately strengthened in order to make the tunnel safe against the reversal of loads due to the postulated single critical crack. 10A.1.9 STRUCTURAL EFFECTS OF OPENINGS The openings are designed and located such that no adverse structural effects are incurred. Venting from the MS compartment was accomplished by the use of the existing pipe tunnel and the addition of a vent to the roof. The vent to the roof was made through existing tendon access openings, which required no additional reinforcing. 10A.1.10 EFFECT OF STRUCTURAL FAILURE There will not be a failure of any structure, including Category II (non-seismic Category I) structures, due to the accident, that could cause failure of any other structure in a manner to adversely affect: a. Mitigation of the consequences of the accident; and b. Capability to bring the unit(s) to a cold shutdown condition. CALVERT CLIFFS UFSAR 10A.1-11 Rev. 47 10A.1.11 VERIFICATION THAT PIPE RUPTURE WILL NOT AFFECT SAFETY In the event of a MSLB in the Auxiliary Building the only region affected is the MS Penetration Room (Figure 10A.1-2). The structures are designed to contain the escaping high energy fluid and to vent and/or drain this fluid safely.

The only passages through which steam can pass are the vents to the outside atmosphere and the tunnel to the Turbine Building (Section 10A.1.20.4). All doors, piping penetrations, and electrical penetrations are leak tight and capable of withstanding the pressure in the Penetration Room. The plant ventilation system does not communicate with this Penetration Room. A separate duct to the atmosphere is provided for normal ventilation in this room. Both the vent stack (Figure 10A.1-7) and the normal ventilation duct are designed to withstand a Penetration Room pressure of 5.0 psig so that steam cannot break through to any other region in the Auxiliary Building. As this paragraph illustrates, steam is prevented from propagating into the other areas of the Auxiliary Building.

Any steam escaping to the Turbine Building will not reach any vital equipment, instruments, electrical supplies, or cables, and will not flow back into the Auxiliary Building.

The only safety-related equipment located in the MS Penetration Room are: a. Main steam isolation valves (MSIVs) b. Main feedwater isolation valves (MFIVs) c. Control valves (CVs) on the MS line to the AFWP turbines d. Cable trays Section 10A.1.13 describes the qualification of items a, b, and c above to properly function in the steam environment. Section 10A.1.20 describes the methods used to determine the standards to be met for environmental qualification and lists the resulting values. Section 10A.1.5 describes the protection provided against pipe whip, jet impingement, and reactive forces. The cable trays are enclosed in metal shielding to prevent damage from jet forces or a steam environment. The steel conduits will withstand a pressure of 5 psig and 300°F of steam environment. The junction boxes are designed to withstand the above conditions (5 psig and 300°F). In the event of a MS line rupture in the Turbine Building, the only regions affected are in the Turbine Building. The pressure levels that would result are insufficient to cause damage to the adjoining Auxiliary Building structures or security doors (Section 10A.1.20). The steam will not propagate into the Auxiliary Building.

Safety-related portions of main steam drains 5 and 6 penetrate the K-line wall and are located adjacent to it on the 12' and 27' Elevations of the Turbine Building. While some of this piping downstream of the included level switches exceeds 1", it is supplied by 1" piping. Therefore, no pipe breaks are required to postulated. Even if a break were postulated, there are no other safety-related components around this piping on the Unit 1 side and only the 36" saltwater ram's head on the Unit 2 side. Clearly, this small piping poses no threat to the ram's head's integrity. However, the main steam headers pass over these drains on the 27' Elevations of both Units 1 and 2. A postulated break in these main steam headers could potentially rupture either or both of these drains. This condition was evaluated, and the results showed that this event would not impair the ability to achieve shut down and would not increase the consequences beyond that of the CALVERT CLIFFS UFSAR 10A.1-12 Rev. 47 ruptured team line alone. Therefore, no barriers or restrains are required to protect these drains from a break in the main steam headers. The location of the instrumentation associated with the Reactor Protection System and the ESF Systems in relation to the high energy piping is such that the instruments will not be affected by pipe whip or jet impingement. The instruments have been qualified for a high temperature and pressure environment.

The safety relief valve operation will not be affected by the steam environment in the MS valve room after a pipe rupture. The turbine bypass line and the turbine bypass valve are located in the Turbine Building and their operation will not be affected by the steam environment.

The atmospheric steam dump valves are discussed in Section 10.1.2.1. These valves are located in the Auxiliary Building immediately above the MS Penetration Room and are in an enclosure that communicates with the MS Penetration Room (Section 10A.1.20.5). The atmospheric dump valves are not safety-related and are not required to operate during this accident. Should one of these valves inadvertently open, the operator has sufficient time to feed the unaffected steam generator with the AFW System. 10A.1.12 EFFECT ON CONTROL ROOM The results of the analysis presented in Section 10A.1.20 show that the Control Room will not be affected by a break in the MS line. 10A.1.13 ENVIRONMENTAL QUALIFICATION OF AFFECTED REQUIRED EQUIPMENT A. Identification The following electrical equipment and valves must be qualified to meet the requirements of Section 10A.1.20: 1. Both MSIVs and both gas/hydraulic actuators 2. Main Steam to AFWPs CVs, 1/2-CV-4070/4071 and 1/2-CV-4070A/4071A 3. Both MFIVs, Motor-Operated Valves (MOVs)-4516 and 4517 B. Testing The Limitorque valve operators on the MFIVs are similar to Limitorque valve operators which have been tested in simulated reactor containment post-accident steam environment conditions. These tests were performed by Franklin Institute Research Laboratories and are summarized in their report Number F-C3441. The valve operators were exposed to a steam environment for 30 days, including two temperature cycles going to 340°F during the first day. The resulting pressure/temperature profile closely followed that recommended by a cognizant IEEE committee. (Reference 3)

The AFWP steam isolation valves are Fisher Controls Valves [1/2-CV-4070/4071] and Rockwell bypass valves adapted to accept Valtek actuators [1/2-CV-4070A/4071A]. These valves and associated appurtenances have been analyzed to be qualified for the anticipated environments. The worst-case environment in the MS piping Penetration Room following a MS line critical crack will be a wet steam and air mixture at 2.23 psig and 331°F. The MSIV actuator has been tested and demonstrated its ability to perform its design function under the above environmental conditions.

CALVERT CLIFFS UFSAR 10A.1-13 Rev. 47 C. Criteria for Protecting Category I Systems, Components, or Equipment The MS line is encapsulated at the postulated break locations. The jet impingement forces due to a postulated break are retained in the encapsulation pipe and the Category I systems, components, or equipment will not be affected by the jet impingement forces.

The magnitude of a jet impingement force resulting from a critical crack in the MS line will be less than 10 kips. A simplified approach to impingement forces assumes the jet to disperse uniformly at the half angle of incidence between the jet axis and the target surface. The half angle, , is taken as 10°. Thus, the jet pressure at distance X is: Pj = Fj/Aj where, Fj = Jet impingement force Aj = Cross-sectional area normal to the jet (expanded jet area) When the target (Category I equipment or systems such as cable, cable trays, instruments) is fully submerged in a jet, jet impingement force = Pj x AT where, Pj = Jet pressure AT = Target area When the target intercepts the jet, that is, target area is larger than the expanded jet area. Jet impingement force = Fj When necessary, barriers are provided to protect Category I systems, components, and equipment against the jet impingement forces. A detail of a typical barrier for protecting cable trays is shown in Figure 10A.1-13. The design criteria for barrier design are similar to the Category I structure design criteria and are discussed in Section 10A.1.7.

To prevent steam from escaping into areas affecting vital equipment and instrumentation, pressure seals designed to withstand the necessary pressure and temperature have been provided where required. The doors in the compartment walls are also designed as pressure retaining doors and are sealed accordingly.

All conduits at the junction boxes in the MS valve room are sealed against the steam environment such that no steam can pass through conduits to any other areas. D. Control Room A break at any of the postulated locations has no effect on the Control Room environment. E. Onsite Power The steam is prevented by walls from propagating into the Switchgear and the Emergency Diesel Generator (EDG) Rooms and, therefore, the onsite power sources and distribution systems will remain operable. CALVERT CLIFFS UFSAR 10A.1-14 Rev. 47 10A.1.14 DESIGN DIAGRAMS AND DRAWINGS Figure 10A.1-1 is the MS System diagram. The routing of the MS line through the Auxiliary Building is shown on Figure 10A.1-2. This drawing shows the location of safety-related equipment located near the MS lines. It also shows the pressure retaining walls that will prevent the propagation of steam, and the vents that limit the pressure rise in the MS Penetration Room.

Figure 10A.1-3 shows a pipe encapsulation detail and Figure 10A.1-4 shows a whipping restraint detail. Figure 10A.1-5 shows a mathematical model for pipe whip restraint. Figure 10A.1-14 shows the AFWP Room and Service Water (SRW) Pump Room ventilation system. 10A.1.15 FLOODING The postulated break of the MS line in the Auxiliary and Turbine Buildings will release high quality steam, most of which is vented so as not to damage vital equipment or structures. Any moisture separated from the escaping steam or formed by condensation on cold surfaces can be adequately handled by two 6"-diameter drain lines penetrating the tunnel wall at floor level and gravity draining to the turbine room floor drain system at floor Elevation 12'0". Watertight doors are provided in the Auxiliary Building to prevent flooding the Penetration Room to other parts of the building. There are administrative controls (including Technical Specifications) on the open/closed status of the doors. 10A.1.16 QUALITY CONTROL The quality control and quality assurance for the safety-related piping is in accordance with Appendix 1B. The level of quality control coverage for the remainder of the piping runs was selected on the basis of importance to plant operating reliability and it is intended that the same degree of quality controls will be maintained. 10A.1.17 LEAK DETECTION Temperature switches are located within the MS line Penetration Room, which will alarm and will alert the operator to the abnormally high temperatures that could result from a small crack. No credit is taken for these switches. In the event there is a large rupture, the instrumentation associated with the steam generators will alert the operator to an MSLB (Section 10A.1.18).

10A.1.18 EMERGENCY PROCEDURES Following a steam line rupture in the Auxiliary Building or the Turbine Building, the applicable emergency operating procedure would be implemented.

Depending on the size of the leak, indications may or may not be received. If indications are not received after a small leak has occurred, the operator would note the leak on his rounds in the Auxiliary Building and Turbine Building (four hours maximum). The operator would then evaluate the need to shut down the plant. 10A.1.19 SEISMIC AND QUALITY CLASSIFICATION The MS lines are designed and constructed to meet American National Standards Institute (ANSI) B31.1 requirements with 100% radiograph of butt-welds in piping greater than 2" NPS, except for the portion that penetrates the Containment out through the MSIV. This portion is designed and constructed to meet the requirements of ASME Section III, Class 2. The non-destructive examination requirements of butt-welds in those

CALVERT CLIFFS UFSAR 10A.1-15 Rev. 47 portions of the MS piping built to ANSI B31.1 and outside of the ISI boundary have been revised to the following requirements: NPS > 8" 100% Radiographic Examination NPS 8" The weld root pass is to be fabricated by GTAW method. A surface examination will be performed on the weld root and the final weld. Examination method is to be magnetic particle when practical, otherwise the liquid penetrant method shall be used. Also, a radiographic examination may be performed as an alternative to the above requirements. For 2" NPS and under piping, weld inspection shall be per code requirements. The MS line from the steam generator outlet to the Turbine Building wall is designed as a Category I (seismic) system. The design data for the MS line are given in Table 10-1. The seismic requirements for Category I (seismic) systems are described in Appendix 5A.

10A.1.20 DESCRIPTION OF ASSUMPTIONS, METHODS AND RESULTS OF ANALYSIS FOR PRESSURE AND TEMPERATURE TRANSIENTS IN COMPARTMENTS The opening at the end of the tunnel between the MS Penetration Room and the Turbine Building serves as a vent to relieve the pressure buildup. A wall obstructs the end of the tunnel; however, this wall is of lighter construction than any other structural component of the tunnel. With the wall in place, the pressure will build up until the wall collapses (at less than 1.0 psig). The pressure will immediately decay in the tunnel. The results of this analysis indicate, therefore, that the maximum pressure in that area will not exceed 1.0 psig. The wall will be designed to fail at 0.5 psi or a hydrostatic pressure of 3' of water. The construction will consist of gypsum wallboard with 24 gauge metal frame work. The retainer clips used in the construction of the wall will be designed to fail on the application on either of the above design loads. No credit is taken in the following analyses for the tunnel or the blow-out wall. 10A.1.20.1 Circumferential Break of 34" Line - 1608 in2 In accordance with the break location criteria, presented in Section 10A.1.2, circumferential breaks are postulated in the MS line Penetration Room in the Auxiliary Building and in the Turbine Building (Figure 10A.1-2). The effect of a full, double-ended break and its associated blowdown was not considered in the Auxiliary Building due to the encapsulation design which limits the steam release rates (Section 10A.1.20.3). The postulated break in the Turbine Building, however, is not encapsulated and the effects of a full, double-ended break were studied. The break is located at the turbine nozzle (terminal end). The mass and energy release data was computed by the Nuclear Steam Supply System supplier for the accident postulated in Section 14.14. The assumptions used to generate system blowdown data for this accident are applicable to a break located in the Turbine Building. This blowdown data, which is given in Table 10A-4, was used to evaluate the pressure transient in the Turbine Building.

The Bechtel computer code Compartment Pressure Analysis (COPRA) was used to analyze the pressure transient for this problem. Compartment Pressure Analysis is a computer code for predicting the pressure differential across the walls of two adjoining compartments following rapid steam and water blowdown within CALVERT CLIFFS UFSAR 10A.1-16 Rev. 47 one of the compartments. The code is intended as a design tool. It is written in FORTRAN IV for the Honeywell 635 computer. The equations and corresponding solutions are divided into two phases: an initialization or steady-state phase, and a calculational or transient phase. The initialization phase sets up quantities such as pressure, masses and temperatures in the compartment atmospheres for the steady state just prior to the blowdown accident. The transient phase described the transient behavior of these quantities during and after blowdown. The following assumptions are incorporated in the analysis: A. The steam, water, and air throughout each compartment are in thermal equilibrium at all times. B. Water, steam, and air entering a compartment are mixed homogeneously and instantaneously; no accumulation of water occurs on the walls or in the sump. C. There is no heat transfer to the compartment walls or floor. D. The blowdown expands into Compartment 1 by the following thermodynamic process: first the mass expands isenthalpically to the total compartment pressure. The water present at that time could form more steam only by relatively slow evaporation. The water is assumed to undergo no further change of phase, maintaining thermal equilibrium with steam and air. The steam then completes its isenthalpic expansion to the partial pressure of the steam already in the compartment. E. If equilibrium calculations result in a superheated atmosphere, then a sufficient quantity of the water suspended in the atmosphere is flashed into steam such that the atmosphere is just saturated. The energy to flash to water is taken from the atmosphere. If all the suspended water were ever used up, the containment atmosphere would remain superheated. F. For masses passing between compartments, the thermodynamics differ slightly. The steam-air-water mixture entering from the other compartment will be brought to thermodynamic equilibrium without the intermediate step of flashing at the total pressure. G. The mass flowing between the compartments is a homogeneous, two-phase, water-steam-air mixture. The flow equations, in addition, assume a frictionless, compressible, adiabatic, no-slip model. H. Two completely separate flow equations are used: one for sharp-edged orifices, and one for all other apertures. I. The first law of thermodynamics for an open system with no heat transfer, as applied to each compartment, is: where: E = Total system energy dMi = Mass transfer into compartments hi = Enthalpy of mass being transferred t = Time The break was postulated at the turbine nozzles which are below the operating deck. The pressure buildup in this region is relieved by venting through the floor grating areas to the upper and lower elevations. These openings were treated as sharp-edged orifice type passages. CALVERT CLIFFS UFSAR 10A.1-17 Rev. 47 The following assumptions are made for the Turbine Building: Room Volume below operating deck = 869,000 ft3 Room Volume above operating deck = 9.0x106 ft3 Vent Area = 1492 ft2 Vent Flow coefficient = Orifice type coefficient (supplied by COPRA) Additional vent areas will occur when the Turbine Building siding breaks off. The siding is released when the differential pressure across the siding exceeds 0.45 psi. The value of 0.45 psi is based on the failure of the retaining clips which hold the siding in place. The failure load of the retaining clips was determined by the siding manufacturer's laboratory test of the ultimate strength of the clips. On this basis there is no reason to expect any pressure greater than 0.45 psig in the region above the operating deck.

The COPRA analysis indicated that the differential pressure across the operating deck reached a maximum value of 0.39 psi and was decreasing before the pressure above the operating deck reached 0.45 psig. This decrease in differential pressure is caused by more mass flowing through the grating area than is coming out of the break; hence, the entire Turbine Building is tending toward an equilibrium pressure. The differential pressure is below 0.30 psi across the operating deck when 0.45 psig is reached above the operating deck. Since the pressure will not exceed 0.45 psig above the operating deck after the siding is released, the pressure below the operating deck will not exceed 0.75 psig after this occurrence. An examination of the COPRA analysis also shows that the pressure below the operating deck never exceeds 0.71 psig prior to the release of the siding. The conclusion of this analysis is that the pressure will not exceed 0.45 psig above the operating deck, and will not exceed 0.75 psig below the operating deck, following a double-ended MSLB in the Turbine Building. The wall separating the Auxiliary Building and the Turbine Building is a 3'-thick reinforced concrete wall capable of withstanding an external pressure of 15 psi.

All ventilation openings between the Turbine Building and the Auxiliary Building are protected with quick closing dampers which are actuated by a gauge pressure of 0.125" water column in the Turbine Building. These dampers are designed and tested to withstand a 1.0 psi differential pressure. The roll-up doors which are located below the operating deck are reinforced by adding removable vertical columns at the third points. This strengthens the doors to withstand pressures in excess of 1 psi. The remaining doors to the Auxiliary Building are personnel doors which swing open into the Turbine Building. Personnel doors located above the operating deck have been tested to withstand 90 psf (0.65 psi) differential pressure. Personnel doors located below the operating deck will be designed to withstand 1.0 psi differential pressure (Table 10A-8). 10A.1.20.2 Circumferential Break (Single Ended) of 6" Line - 28.9 in2 The MS branch lines to the relief valves (6" lines), to the AFWP turbines (6" lines) and to the atmospheric dump valves (4" lines) join the 34" MS line in the MS Penetration Room (Figures 10A.1-2 and 10A.2-2). These connections are CALVERT CLIFFS UFSAR 10A.1-18 Rev. 47 considered terminal ends which are circumferential break locations as defined in Section 10A.1.2. The maximum mass and energy release rates for a 6" circumferential break were calculated using the two-phase, single component, annular flow model developed by Moody (References 4 and 5). To account for the stored energy in the MS lines, an "infinite" reservoir was conservatively assumed just up-stream of the break. To account for the break entrance and exit effects, one velocity head loss was assumed. Normal maximum system pressure of 900 psia and temperature of 532°F were assumed. With this approach, the maximum blowdown is 1450 lb/sec-ft2. The maximum exit conditions at the break were found to be as follows: Mass discharge rate, lb/sec - 291 Enthalpy (average), Btu/lbm - 1150 (November 1981 analysis used 1200) The Bechtel computer code COPATTA was used in November 1981, to analyze the pressure transient in this compartment following the postulated break. The COPATTA program is discussed in Section 14.20. The discharge mass rate and enthalpy were assumed constant at the maximum values given above.

The following assumptions are made for the Penetration Room: Room Volume = 24,000 ft3 Vent Area = 44.1 ft2 Vent flow coefficient = 0.71 Initial Room Temperature = 160°F Relative Humidity = 70% Total Heat Sink Area = 6765 ft2 (Concrete Walls, Floor, Ceiling) The following additional assumptions were used in the analysis. a. No credit is taken for equipment such as heat sinks. b. Room volume does not include the adjacent pipe tunnel nor is credit taken for steam flow into that area. c. No credit is taken for heating, ventilation and air conditioning operation and the capability to remove heat. d. Process heat loads are 130,285 Btu/hr until isolation and 65,142 Btu/hr for 3-1/2 hours thereafter. Heat load becomes zero 3-1/2 hours after isolation. Results of the analysis are presented in Table 10A-1A. This table indicates that the maximum sustained temperature continues until the blowdown is isolated (also Figures 10A.1-15 and 10A.1-16).

10A.1.20.3 Circumferential Break Inside Encapsulation - 53.8 in2 Clearance The MS line encapsulations are located inside the MS valve compartment of the Auxiliary Building as shown in Figure 10A.1-2. The construction tolerances imposed on the design limit the gap between the MS line and the closure plates on the ends of the encapsulation and between the MS branch lines and the encapsulation. The total escape area of all openings in any encapsulation section will not exceed 53.8 in2 for a guillotine break inside the encapsulation. This area does not include the 1" drain and 3/4" vent or take credit for the restraint bolt CALVERT CLIFFS UFSAR 10A.1-19 Rev. 47 obstructions; however, these are insignificant in the pressure calculations. This will correspond to a break flow of approximately 545 lbm/sec using the same mass flux employed in the preceding section. The November 1981, COPATTA analysis of this break assumed the same room characteristics as listed above. The flow rate of 545 lbm/sec and the enthalpy of 1200 Btu/lbm are assumed constant.

Results of this analysis are presented in Table 10A-1A. This table indicates that the maximum sustained temperature continues until the blowdown is isolated (also Figures 10A.1-17 and 10A.1-18.)

10A.1.20.4 Critical Crack in 34" Line - 8.5 in2 A major MS line rupture in the tunnel between the MS valve compartment in the Auxiliary Building and the Turbine Building is not a credible event because it does not meet the criteria presented in Section 10A.1.2. Specifically, the steam lines in the tunnel are long, straight sections that have no high stress points or terminal ends. A critical crack, however, is postulated anywhere in the MS line. The maximum mass and energy released from a critical crack were calculated in the same manner as discussed in Section 10A.1.20.2. The maximum exit conditions were found to be as follows: Mass discharge rate, lbm/sec - 86 Enthalpy (average), Btu/lbm - 1150 (November 1981 analysis used 1200) The Bechtel computer code COPATTA was used in the November 1981 analysis of this problem. The discharge mass rate and enthalpy were assumed constant at the maximum values given above. Assumptions are listed in 10A.1.20.2. Results of this analysis are presented in Table 10A-1A. This table indicates that the maximum sustained temperature continues until the blowdown is isolated (also Figures 10A.1-19 and 10A.1-20.) 10A.1.20.5 Critical Crack in 4" Atmospheric Dump Line-0.32 in2 The 4" lines between the MS header in the MS Penetration Room and the atmospheric dump valves are high energy lines. The atmospheric dump valves are normally closed; hence, the lines downstream of these valves are not in the high energy class. All of the postulated break locations in the high energy portion of this line are in the MS Penetration Room. These breaks in the 4" lines are smaller than the encapsulated guillotine break discussed in Section 10A.1.20.3, and result in smaller peak compartment pressure. To protect against critical cracks in those portions of the high energy atmospheric dump valve lines in the compartment above the MS Penetration Room, the lines are completely enclosed from the floor penetrations to a point beyond the valves.

These enclosures are sealed around the lines downstream of the valves and attached to the floor to prevent propagation of steam into the compartment. Steam leakage will be vented from the enclosure into the MS Penetration Room through the floor penetrations.

CALVERT CLIFFS UFSAR 10A.1-20 Rev. 47 Therefore, the postulated break discussed in Section 10A.1.20.4 is controlling for the design of the enclosure. That break resulted in a maximum pressure in the MS Penetration Room of 2.2 psig. The enclosure is designed to contain an internal pressure of 5 psig. This affords protection against a critical crack in the atmospheric dump line and also against the postulated line breaks in the MS Penetration Room. 10A.1.21 INTEGRITY OF THE CONTAINMENT STRUCTURE AND A PIPE RUPTURE OUTSIDE THE CONTAINMENT The Containment Structure is designed using load combinations as discussed in Appendix 5A. The method used in the analysis of the Containment Structure is discussed in Chapter 5.

Since the MS line is encapsulated from the containment boundary to the first high stress point, there will be no jet impingement forces due to a full postulated break to impair the structural integrity of the prestressed concrete Containment Structure. 10A.1.22 EFFECT OF THE BABCOCK & WILCOX, CANADA REPLACEMENT STEAM GENERATORS The replacement of the original steam generators with Babcock & Wilcox, Canada replacement steam generators does not involve changes to the design of the main steam line piping or its supports and restraints located outside Containment. Babcock & Wilcox, Canada has evaluated the effect of the replacement steam generators on those hydraulic parameters that are significant in determining the magnitude of the forces caused by a pipe break. It is concluded that the loads from secondary side pipe breaks are either unchanged or are reduced with the replacement steam generators in place. 10A.1.23 REFERENCES 1. D. W. Faletti, "Two-Phase, Critical Flow of Steam-Water Mixture," AICHE Journal 9, No. 2, 1963 2. General Electric Company Document No. 26A2625, "Systems Criteria and Application for Protection Against the Dynamic Effects of Pipe Break," February 1972 3. Proposed Guide for Type Test of Class I Electrical Valve Operators for Nuclear Power Generating Stations, Draft 13, IEEE Project Number 382, JCNPS/SC2.3, June 1972 4. Maximum Flow Rate of a Single Component, Two Phase Mixture," by F.J. Moody; ASME Transactions, Series C, Volume 87, 1965 5. Maximum Two Phase Vessel Blowdown from Pipes," by F.J. Moody, APED-4827; General Electric Company, April 20, 1965 6. Introduction to Structural Dynamics by Professor John M. Biggs, 1964 Edition, published by McGraw-Hill Book Company 7. Perry H. Peterson "Resistance to Fire and Radiation," American Society for Testing and Materials (ASTM) Special Technical Publication No. 169, page 201, dated 1956 8. EPRI Topical Report No. TR-1006937, Extension of the EPRI Risk-Informed ISI Methodology to Break Exclusion Region Programs, Rev. 0-A, August 2002 CALVERT CLIFFS UFSAR 10A.1-21 Rev. 47 TABLE 10A-1A MAIN STEAM PENETRATION ROOM ANALYSIS RESULTS INITIAL ROOM TEMP. = 160°F 6" 34" (Encapsulation) CRITICAL CRACK Maximum Sustained Temperature(d) 316°F for 13.5 min(a) 316°F for 7.7 min(b) 308-320°F for 4 hrs(c) Peak Temperature 327°F < 20 sec 318°F < 20 sec 331°F < 20 sec Peak Pressure 0.52 psig at 0.35 sec 1.49 psig at 0.50 sec 2.23 psig at 101.5 sec Time to Return to 160°F 16.9 hrs 9.7 hrs 40.4 hrs _______________________ (a) Based on time to isolate feedwater plus time to empty one steam generator. (b) Based on time to empty one steam generator (c) Based on leak being located and isolated during required four-hour tours. (d) Times listed are those taken to isolate blowdown; i.e., maximum temperature persists until blowdown isolation in each analyzed case.

CALVERT CLIFFS UFSAR 10A.1-22 Rev. 47 TABLE 10A-1 MAIN STEAM STRESS VALUES Following is a stress summary of the intermediate points considered between the terminal ends. The postulated full break locations are shown on Figure 10A.1-2. All values are in psi. PRIMARY STRESS POINT NUMBER SECONDARY STRESS(a) (0.8SA) LONGITUDINAL PRESSURE LONGITUDINAL WEIGHT SEISMIC OBE OTHER(c) TOTAL SPR(b,d) (<0.8 Sh) TOTAL STRESS(e) [<0.8 (SA + Sh)] 2 14,513 4,910 415 50 569 5,944 20,457 3 18,121 4,910 1,968 210 323 7,411 25,532 4 22,131 4,910 4,154 343 304 9,711 31,842 6 10,943 4,910 315 22 539 5,786 16,729 7 18,845 4,910 1,726 88 375 7,099 25,944 8 20,069 4,910 2,908 326 107 8,148 28,217 _______________________ (a) SA = The larger of f [1.25 Sc + 0.25 Sh + (Sh - SPR)] or f(1.25 Sc = 0.25 Sh) as per paragraph 102.3.2(c) and (d) of the USAS Code for Pressure Piping, USAS B31.1.0-1967, and as per NC-3600 of Section III (Nuclear Power Plant Components), ASME Boiler and Pressure Vessel (B&PV) Code. 0.85 SA taken as 21,000 psi. (b) Sc and Sh are the allowable stresses at cold and hot conditions, respectively, for Class 2 and Class 3 components as per ASME B&PV Code, Section III (Nuclear Power Plant Components). (c) Other stresses are: 1) Due to steam or water hammer 2) Due to relief valve discharge. (d) SPR is the total of columns 3 through 6. (e) 0.8 (SA + Sh) taken as 35,000 psi.

CALVERT CLIFFS UFSAR 10A.1-23 Rev. 47 TABLE 10A-2 STRESSES ON STRUCTURAL COMPONENTS DUE TO JET IMPINGEMENT FORCES RESULTING FROM A CRITICAL CRACK OF THE MAIN STEAM LINE CALCULATED STRESSES(a) DUE TO JET FORCE PLUS 1 psi CALCULATED STRESSES VS. ALLOWABLE STRESSES STRUCTURAL COMPONENT THICKNESS DIST FROM RUPTUREJET FORCE 1.26 PA CROSS SECTION TARGET AREA ft2 COMPRESSIVE CONC (psi) TENSILE REINF (psi) CONCRETEREINFORCINGNorth Tunnel Wall 1'0" 2'0" 9,640 1.40 710 26,000 0.209 0.722 South Tunnel Wall 2'0" 3'0" 9,640 2.40 539 18,400 0.158 0.511 Tunnel Floor 1'3" 1'0" 9,640 0.60 1,012 20,950 0.297 0.582 Tunnel Ceiling EL 33'6" 1'4" 2'6" 9,640 2.00 650 14,039 0.191 0.390 New Wall @ Col. Line 6.1 1'3" 4'0" 9,640 3.90 1,065 25,885 0.313 0.719 Wall Lb 1'9" 3'6" 9,640 3.15 518 16,791 0.152 0.466 Wall 17 2'0" 3'0" 9,640 2.50 539 18,400 0.158 0.511 New Wall @ Col. Mc 1'3" 10'0" 9,640 17.30 376 9,129 0.110 0.253 Floor EL 27'0" 2'6" 1'0" 9,640 0.60 655 20,600 0.193 0.572 Ceiling EL 45'0" 1'3" 11'5" 9,640 21.80 480 11,640 0.141 0.323 Ceiling EL 45'0" 2'6" 3'0" 9,640 2.50 50 1,580 0.01 0.044 _______________________ (a) Stresses shown are maximum stresses and include live load, dead load, equipment load & SSE seismic stresses. CALVERT CLIFFS UFSAR 10A.1-24 Rev. 47 TABLE 10A-3 STRESSES ON STRUCTURAL COMPONENTS IN THE MAIN STEAM COMPARTMENT DUE TO POSTULATED MAIN STEAM PIPE RUPTURE CALCULATED STRESSES(a) DUE TO PRESSURIZATION OF 2.6 psi(b) CALCULATED STRESSES VS. ALLOWABLE STRESSES STRUCTURAL COMPONENT THICKNESS COMPRESSIVE IN CONCRETE (psi) TENSILE IN REINFORCING (psi) CONCRETE REINFORCING North Tunnel Wall 1'0" 230 8,420 0.068 0.234 South Tunnel Wall 2'0" 492 16,312 0.145 0.453 Tunnel Floor 1'3" 555 11,480 0.163 0.319 Tunnel Ceiling EL 33'6" 1'4" 245 5,292 0.072 0.147 New Wall @ Col. Line 6.1 1'3" 592 14,388 0.174 0.40 Wall Lb 1'9" 438 13,454 0.129 0.374 Wall 17 2'0" 492 16,312 0.145 0.453 New Wall @ Col. Mc 1'3" 592 14,388 0.174 0.40 Floor EL 27'0" 2'6" 615 19,330 0.181 0.537 Ceiling EL 45'0" 1'3" 525 12,710 0.154 0.353 Ceiling EL 45'0" 2'6" 38 1,184 0.01 0.033 _______________________ (a) Stresses shown are maximum stresses which include live load, dead load, equipment load, and SSE seismic forces. (b) This pressure is based on original design analyses, later reviews have shown this value to be conservative - Section 10A.1.

CALVERT CLIFFS UFSAR 10A.1-25 Rev. 47 TABLE 10A-4 STEAM LINE RUPTURE INCIDENT NO LOAD, 1-LOOP OPERATION BREAK INSIDE TURBINE BUILDING TIME (sec) TOTAL BLOWDOWN (lb/sec) TOTAL MASS (lbs) 0.00 6655.01 0.00 1.00 6215.92 6.409x103 2.00 5836.57 1.241x104 3.00 5509.71 1.806x104 4.00 5219.13 2.338x104 5.00 4959.82 2.838x104 6.00 4731.27 3.315x104 7.00 4523.20 3.767x104 8.00 4339.18 4.201x104 9.00 4177.27 4.619x104 10.00 4033.25 5.022x104 20.00 1860.79 8.048x104 30.00 1725.64 9.850x104 40.00 1485.61 1.146x105 50.00 1300.55 1.283x105 60.00 1175.84 1.407x105 70.00 1067.60 1.518x105 80.00 980.12 1.620x105 90.00 908.10 1.714x105 100.00 848.02 1.802x105 150.00 658.80 2.171x105 180.00 585.99 2.358x105 200.00 546.93 2.470x105 CALVERT CLIFFS UFSAR 10A.1-26 Rev. 47 TABLE 10A-7 COMPARISON OF TIME-HISTORY PIPE WHIP DESIGN METHOD WITH ENERGY BALANCE DESIGN METHOD TIME-HISTORY DESIGN METHOD CALVERT CLIFFS ENERGY BALANCE USING GIVEN RESTRAINT DEFLECTION USING PROPERTIES OF GIVEN RESTRAINT LOAD. COND. FORCING FUNCTION TIME SECONDS MAX.LOAD ON RESTRAINT RESTRAINTCAPACITY AT YIELD FORCING FUNCTION RESTRAINT CAPACITY AT YIELD DUCTILITY RATIO FORCING FUNCTION RESTRAINTCAPACITY AT YIELD DUCTILITY RATIO (kips) (kips) (kips) (kips) (kips) 1 258 0.0027 615 930 324 1080 5 324 994 5 180 0.11 103 0.30 2 258 0.0018 646 1050 324 1080 5 324 994 5 180 0.0625 281 0.30 3 310 0.0063 793 930 324 1080 5 324 994 5 216 0.031 303 0.09 257 0.30 4 310 0.0063 850 1050 324 1080 5 324 994 5 216 0.031 303 0.09 257 0.30 CALVERT CLIFFS UFSAR 10A.1-27 Rev. 47 TABLE 10A-8 LOCATION OF DOORS BETWEEN TURBINE BUILDING AND AUXILIARY BUILDING UFSAR FIGURE NO. ELEVATION LOCATION(a) CAPABILITY 1-6 5'0" Between columns 105 & 106 10 psig 1-6 5'0" Between columns 107 & 108 10 psig 1-10 5'0" Between columns 206 & 207 10 psig 1-10 5'0" Between columns 208 & 209 10 psig 1-7 27'0" Between columns 105 & 106 1 psig 1-7 27'0" Between columns 107 & 108 (roll-up door) 1 psig 1-7 27'0" Between columns 110 & 111 1 psig 1-7 27'0" Between columns 112 & 113 1 psig 1-11 27'0" Between columns 203 & 204 1 psig 1-11 27'0" Between columns 206 & 207 1 psig 1-11 27'0" Between columns 208 & 209 1 psig 1-8 45'0" Between columns 105 & 106 0.65 psig 1-8 45'0" Between columns 107 & 108 (roll-up door) 1 psig 1-8 45'0" Between columns 110 & 111 0.65 psig 1-8 45'0" Between columns 112 & 113 (double door) 0.65 psig 1-12 45'0" Between columns 201 & 202 (double door) 0.65 psig 1-12 45'0" Between columns 203 & 204 0.65 psig 1-12 45'0" Between columns 206 & 207 (roll-up door) 1 psig 1-12 45'0" Between columns 208 & 209 0.65 psig _______________________ (a) All doors are through the Turbine Building/Auxiliary Building wall.

Rev. 2010A.1-1 MAIN STEAM SYSTEM Rev.1810A.1-2 HIGH ENERGY MAIN STEAM SYSTEM ROUTING PLAN - FLOOR EL. 27'0" Rev.010A.1-3 ENCAPSULATION DETAILS MAIN STEAM SYSTEM Rev.010A.1-4 HIGH ENERGY PIPE WHIP RESTRAINT Rev.010A.1-5 MATHEMATICAL MODEL FOR PIPE WHIP RESTRAINT Rev.010A.1-7 VENT STACK FROM MAIN STEAM COMPARTMENT Rev.010A.1-8 BLOWDOWN MODEL Rev.010A.1-9 EFFECT OF PIPE FRICTION OF STEAM-WATER Rev.010A.1-10 EFFECT OF AREA RESTRICTION ON STEAM WATER Rev.010A.1-11 STAGNATION PRESSURE - ASYMPTOTIC JET PRESSURE Rev.010A.1-12 STAGNATION PRESSURE - JET ASYMPTOTIC AREA Rev.010A.1-13 TYP. JET IMPINGEMENT BARRIER FOR CABLE TRAYS Rev.010A.1-14 HIGH ENERGY AUX. FEED PUMP ROOM AND SERVICE WTR PUMP ROOM VENTILATION SYSTEMS Rev.010A.1-15 6" STEAM LINE BREAK - MAIN STEAM PENETRATION ROOM - INITIAL TEMPERATURE = 160F Rev.010A.1-16 6" STEAM LINE BREAK - MAIN STEAM PENETRATION ROOM - INITIAL TEMPERATURE = 160F Rev.010A.1-17 BREAK OF ENCAPSULATED STEAM LINE - MAIN STEAM PENETRATION ROOM - INITIALTEMPERATURE = 160F Rev.010A.1-18 BREAK OF ENCAPSULATED STEAM LINE - MAIN STEAM PENETRATION ROOM - INITIALTEMPERATURE = 160F Rev.010A.1-19 CRITICAL CRACK IN MAIN STEAM LINE - MAIN STEAM PENETRATION ROOM - INITIAL TEMPERATURE= 160F Rev.010A.1-20 CRITICAL CRACK IN MAIN STEAM LINE - MAIN STEAM PENETRATION ROOM - INITIAL TEMPERATURE= 160F Rev.2010A.2-1 MAIN STEAM TO AUX. STEAM GEN. FEED PUMP TURBINE Rev. 1510A.2-2 HIGH ENERGY MAIN STEAM TO AUX. FEED PUMP TURBINE ROUTING PLAN - FLOOR EL. 27'0" Rev.1510A.2-2A MAIN STEAM TO AFW TURBINE - BREAK LOCATIONS Rev.1510A.2-2B MAIN STEAM TO AFW TURBINE - SMALL PIPE LOCATIONS Rev.610A.2-3 HIGH ENERGY MAIN STEAM TO AUX. FEED PUMP TURBINE ROUTING PLAN - FLOOR EL. 5'0" Rev.2410A.3-1 STEAM GENERATOR BLOWDOWN Rev.2010A.3-2 HIGH ENERGY STEAM GENERATOR BLOWDOWN SYSTEM ROUTING PLAN - FLOOR EL. 27'0" Revision 46 FIGURE 10A.3-3 HIGH ENERGY STEAM GENERATOR BLOWDOWN SYSTEM ROUTING PLAN - FLOOR EL. 450 Rev.010A.3-4 STEAM GENERATOR BLOWDOWN LINE BREAK - EAST PIPING PENETRATION ROOM - INITIALTEMPERATURE = 160F Rev.010A.3-5 STEAM GENERATOR BLOWDOWN LINE BREAK - ELEVATION 45' - INITIAL TEMPERATURE = 160F Rev.1810A.4-1 MAIN FEEDWATER SYSTEM Rev.010A.4-2 HIGH ENERGY MAIN FEEDWATER ROUTING PLAN - FLOOR EL. 27'0" Rev.610A.4-3 HIGH ENERGY MAIN FEEDWATER ROUTING PLAN - FLOOR EL. 5'0" Rev.1810A.4-4 FEEDWATER TO HP HEATERS Rev.1810A.4-5 FEEDWATER HEATER DRAIN TO CONDENSER Rev.010A.4-6 HIGH ENERGY MAIN FEEDWATER AND HEATER DRAIN SYSTEM ROUTING PLAN - TURB BLDG. FLR EL.12'0" Rev.510A.4-7 PIPE SLEEVE DETAILS MAIN FEEDWATER SYSTEM UNIT NO. 1 Rev.2010A.5-1 AUX. FEEDWATER SYSTEM

Rev.1510A.5-5 POSTULATED PIPE CRACK TARGETS - AUX. FEEDWATER SYSTEM UNITS 1 AND 2 Rev.2010A.7-1 CVC SYSTEM REACTOR COOL. CHARGING LINE Rev.010A.7-2 HIGH ENERGY R.C. CHARGING LINE - ROUTING PARTIAL PLANS - @ FLOOR EL. 5'0" AND (-)10'0"

Rev.2010A.7-4 CVC SYSTEM REACTOR COOLANT LETDOWN LINE

Rev.2010A.8-1 STEAM GEN. AND REACTOR COOL. SAMPLING SYSTEMS

Rev.1810A.9-1 AUX. STEAM TO WASTE EVAPORATORS

Rev.010A.9-3 HIGH ENERGY AUXILIARY STEAM ROUTING PLAN - FLOOR EL. 27'0"

CALVERT CLIFFS UFSAR 11.3-1 Rev. 47 11.3 RADIATION SAFETY 11.3.1 GENERAL The Plant General Manager-Calvert Cliffs Nuclear Power Plant Department is responsible for the Radiation Safety Program at Calvert Cliffs. This responsibility is shared by all supervisors and plant personnel. Personnel assigned to the plant and all visitors are required to follow all rules and procedures established for protection against radiation, contamination and airborne activity.

Administration of the Radiation Safety Program at Calvert Cliffs is the responsibility of the General Supervisor-Radiation Protection. The implementation of this program is the responsibility of the Health Physics Operations Unit and Health Physics Support Unit, with their primary purpose being to administrate, control, and eliminate, if possible, any and all radiological hazards within the plant. Additional responsibilities are in the areas of assisting the various plant training, operations, and maintenance sections in assuring the plant is operated and maintained in a safe condition, and that compliance with Company, State, and Federal regulations in regard to radiation safety are adhered to. 11.3.2 ACCESS CONTROL A radiologically controlled area (RCA) is defined as an area within the plant site in which radioactive materials and/or radiation are present, or where there is a potential for their release in sufficient quantities (as designated by 10 CFR Part 20) to require protective measures. Entry into radiologically controlled areas is limited to those persons authorized to accomplish a specific task.

Radiologically controlled areas are designated by appropriate signs and barriers. Prior to entering these areas, personnel will meet the requirements for dosimetry, protective clothing and procedures as detailed in Calvert Cliffs Radiation Safety Procedures. 11.3.3 PERSONNEL PROTECTIVE AND MONITORING EQUIPMENT 11.3.3.1 Protective Equipment All personnel entering a RCA may be required to wear anti-contamination clothing. This is directly dependent on the locations, plant condition and the task to be performed.

Generally the standard dress will be coveralls (over personal undergarments or medical "scrubs"), shoe covers, cotton glove liners, rubber gloves, rubber boots, and a hood. The requirements may be increased or decreased depending upon contamination and airborne radioactivity concentrations and the task to be performed.

Respiratory protection devices may be required when high or the potential for high airborne radioactive material exists. In such situations, the air will be monitored by the Health Physics Technicians and the required protective devices will be issued as appropriate for the type and concentrations of airborne radioactive material present. Monitoring and evaluating airborne radioactive material and the use of respiratory protection is performed according to the Calvert Cliffs Radiation Safety Procedures and 10 CFR Part 20. 11.3.3.2 Personnel Monitoring Program The personnel monitoring program at Calvert Cliffs is based on the use of Dosimeter of Legal Record (DLR), Electronic Dosimeter (ED), and direct-reading dosimeters (DRD) for determining personnel exposures due to external Beta, Gamma, and neutron sources. When neutron exposures may be expected, a CALVERT CLIFFS UFSAR 11.3-2 Rev. 47 combination of DLR and/or a direct reading portable neutron survey instrument will be used for exposure determination. All personnel working in an RCA will be issued DLRs. Dosimeters of legal record will be processed routinely in accordance with Radiation Safety Procedures. Additional processing will be in accordance with applicable Radiation Safety Procedures. Additional DLRs will be issued for critical organs and/or extremity monitoring where prescribed by applicable Radiation Safety Procedures. Under conditions specified in 10 CFR 20.1502 and the Calvert Cliffs Radiation Safety Manual or upon entering radiologically controlled areas of the plant, all personnel are required to wear a DLR and/or ED or DRD. In the case of visitors, they shall wear an DRD and be accompanied by an escort wearing a DLR and/or ED for the group. Direct-reading dosimeters shall be read daily or upon exit from the RCA to provide estimates of personnel exposure between DLR processing periods. The DRD also provides data as needed for evaluation of a lost or damaged DLR or evaluations of equipment, jobs, techniques, shielding, etc. Electronic Dosimeters may be used in place of DRDs. The bioassay program requires all employees designated for assignment to Calvert Cliffs to pass a physical examination prescribed by the Calvert Cliffs Nuclear Power Plant, Inc. (CCNPP) Environmental, Safety & Health Department. For individuals required to work in RCAs of the plant, this examination will include a whole body count or passive screening. Periodic physical reexaminations, including in vivo counting or passive screening, will be given to plant employees who do significant work in an RCA. Employees terminating employment with CCNPP who have worked in an RCA at Calvert Cliffs will receive an in vivo analysis or passive screening.

Special health examinations may be required for any individual whose records show they are exceeding of yearly whole body or tissue/organ dose limits, or for any individual suspected of assimilating radioactive material. This examination may include in vivo analysis and/or in vitro analysis whenever uptake of significant radioactive materials is suspected.

Records showing the occupational dose accumulated at Calvert Cliffs of all individuals provided with personnel monitoring shall be maintained in accordance with 10 CFR 20.2106. Lists of the current status of personnel dose are available to plant supervision to aid in job planning. In addition, an alert list system will be used to emphasize those individuals who are approaching the administrative annual individual ALARA dose goal. An individual radiation history record folder and/or electronic media will be maintained for each occupational worker. The folder contains records of: external and internal occupational dose received at Calvert Cliffs; prior occupational exposure history, to the extent required by revised 10 CFR Part 20; radiation orientation and/or training received; and special measurement results (in vivo, in vitro, respirator tests). Additional records and reports to employees, other individuals, and the US NRC, shall be in accordance with 10 CFR 20.2202 through 20.2206.

CALVERT CLIFFS UFSAR 11.3-3 Rev. 47 11.3.4 RADIATION SAFETY FACILITIES 11.3.4.1 Change Room and Decontamination Facilities Change room facilities are provided where personnel change into the protective clothing required for RCA work. Showers, sinks and appropriate monitoring equipment are provided to aid in personnel decontamination.

Equipment decontamination facilities are also provided at the plant for large and small equipment and components. 11.3.4.2 Health Physics Laboratory Facilities The radiation safety laboratory contains facilities and equipment for detecting, analyzing, and measuring all types of ionizing radiation. In addition, a small source "calibrator" is available to perform operational checks of portable gamma survey instruments. The chemistry laboratory includes a counting room for analyzing environmental survey samples, including identification of specific radionuclides.

11.3.4.3 Health Physics Instrumentation (Excluding Process and Area Monitoring Systems) Portable radiation survey instruments are provided for use to Health Physics Technicians as well as for operating and maintenance personnel. A variety of instruments are selected to cover the spectrum of radiation measurement requirements anticipated, i.e., instruments for detecting and measuring alpha, beta, gamma, and neutron radiation. Sufficient quantities are provided to allow for routine and emergency use, and allowing for unavailability of instruments due to maintenance and calibration. In addition to the portable instruments, appropriate monitoring instruments are located at the exits from an RCA and at various locations within an RCA. These instruments are intended to detect contamination on personnel, materials, or equipment, so as to prevent contamination from being spread within or beyond controlled areas. Portal monitors will also be utilized, as appropriate, to monitor for radioactive material at plant ingress and egress.

Details on the Health Physics instrumentation used are contained in the Calvert Cliffs Radiation Safety Procedures and Instrument Test Equipment and Calibration procedures.

CALVERT CLIFFS UFSAR 11.4-1 Rev. 47 11.4 RADIOACTIVE MATERIALS SAFETY 11.4.1 MATERIALS SAFETY PROGRAM The overall radiation safety program includes provisions to assure the safe storage, handling, and use of sealed and unsealed special nuclear, source, and byproduct materials (Section 11.3).

In addition to indoctrination and training, personnel dosimetry, radiation and contamination surveys, and contamination control methods, the Radiation Safety Program includes specific provisions for radioactive material control. These provisions include procedures for the proper receipt of radioactive materials, the storage and movement of material within the plant, radioactive source control and inventory, and the packaging and labeling of radioactive materials for shipment. 11.4.2 SEALED SOURCE CONTAMINATION Sealed sources containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material are periodically verified to be free of 0.005 microcuries of removable contamination. The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. This limitation ensures that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values. The test method has a detection sensitivity of at least 0.005 microcuries per test sample.

Each sealed source is tested for leakage and/or contamination by CCNPP personnel or other persons specifically authorized by the NRC or an Agreement State. Each category of sealed sources (excluding startup sources and fission detectors previously subjected to core flux) will be tested at the frequencies described below: a. Sources in use - At least once per six months for all sealed sources containing radioactive material: 1. With a half-life greater than 30 days (excluding Hydrogen 3), and 2. In any form other than gas. b. Stored sources not in use - Each sealed source and fission detector is tested prior to use or transfer to another licensee unless tested within the previous six months. Sealed sources transferred without a certificate indicating the last test date are tested prior to being placed into use. c. Startup sources and fission detectors - Each sealed startup source and fission detector is tested within 31 days prior to being subjected to core flux or installed in the core and following repair or maintenance to the source or detector. Sealed sources with removable contamination in excess of 0.005 microcuries are immediately withdrawn from use and either decontaminated and repaired, or disposed of in accordance with NRC requirements. 11.4.3 FACILITIES AND EQUIPMENT Plant laboratory facilities and equipment, survey and measuring instruments, and monitoring devices are described in Sections 11.2.3, 11.3.3, and 11.3.4.

11.4.4 PERSONNEL AND PROCEDURES The experience and qualifications of key Radiation Safety personnel responsible for handling and monitoring radioactive materials are described in Sections 12.1 and 12.2.

CALVERT CLIFFS UFSAR 11.4-2 Rev. 47 11.4.5 REQUIRED MATERIALS MATERIAL FORM AND USE POSSESSION LIMIT 1. Any byproduct, source, and special nuclear material As reactor fuel; as sealed neutron sources for reactor start-up; as sealed sources for reactor instrument and radiation monitoring equipment calibration; and as fission detectors. As required by Unit 1 or Unit 2 License. 2. Any byproduct, material Any form for sample analysis or counting equipment calibration.As required by Unit 1 or Unit 2 License. 3. Any source or special nuclear material Any form for sample analysis or instrument calibration. As required by Unit 1 or Unit 2 License. 4. Sodium-24 Liquid form for tracer measurements for steam. As required by Unit 1 or Unit 2 License. 5. Byproduct and special nuclear materials Possess but not separate such materials in such form as may be produced by operation of the facility. As required by Unit 1 or Unit 2 License. I NO 11WASTE EVAPORATOR I I I 11 10 11 10 STOP, THINK, ACT ANO REVIEW 13 12 SIMPLIFIED SYSTEM DRAWING RC WASTE PROCESSING SL -077 CALVERT CLIFFS NUCLE.AR POWER PL.allT ENGINEER!NGSERVICESOEPARTl.IENT CALVERT CUFFS UNIT 1&2 SL-077 64323 13

11-5 REACTOR CAVITY SEAL Hatches at 90* and 270" locations are as shown for Unit 1

• For Unit 2 these hatches are curved. '270" I I I I I I I er BALTIMORE GAS & ELECTRIC CO. Calvert Cliffs Nuclear Power Plant I I 90* REACTOR CAVITY SEAL Figure 11-5 Rev. 19 Rev. 19 11-6 NEUTRON SHIELDING SECTION, UNITS 1 AND 2 BALTIMORE GAS & ELECTRIC co_ Calvert Cliffs Nuclear Power Plant .. .. A .. .. .... A .. ... .... .... ..:. .. .. ... .. < * .. .. A .... A .. ,;. ,. . < " .. ,;,
• A .. . A < .. .. * ... . .. NEUTRON SHIELDING SECTION UNITS 1AND2 Figure 11-6 Rev.18 Rev. 18

CALVERT CLIFFS UFSAR 12.3-1 Rev. 47 12.3 OPERATIONAL PROCEDURES AND CONTROLS The plant is operated and maintained in accordance with approved procedures. (Technical Specifications) These procedures are reviewed and approved in accordance with administrative procedures. Fuel Handling Procedures These procedures prescribe the general preplanning for the fuel handling program and its associated safety measures and identify those aspects of the program for which procedures are to be prepared for each refueling outage. Locking Devices Locking Devices have been installed on selected valves or their controls to prevent their inadvertent operation. Administrative controls have been established to ensure that these devices remain in place as intended, and that they are re-installed after periods of approved removal.

12.4 RECORDS

CALVERT CLIFFS UFSAR 12.6-1 Rev. 47 12.6 EMERGENCY RESPONSE PLAN (FORMERLY SITE EMERGENCY PLAN) 12.6.1 OVERALL CONCEPT OF OPERATION The CCNPP Emergency Response Plan (ERP) has been developed to protect the general public and site personnel from possible consequences of an emergency condition. This plan, combined with its implementation procedures and the Radiological Emergency Plans of the State and local agencies, allows for (a) early recognition and classification of a possible emergency condition; (b) prompt notification, via reliable communication channels, of agencies and personnel to augment the normal operating personnel; (c) prompt preplanned actions to be taken to protect the population-at-risk. The CCNPP staff is trained to cope with emergencies. Written agreements with Federal agencies, private contractors, and coordinated State and local agency emergency plans (required by law) provide assistance to ensure resources can be readily available in as short a time as possible to cope with emergencies and protect the population-at-risk. The agencies and the resources they will provide are described in the ERP and the "Maryland Emergency Operations Plan, Annex Q, Radiological Emergency Plan." Both plans describe the roles of the various State and local agencies and their interfaces for carrying out protective and parallel actions in both a 10-mile-radius plume zone and a 50-mile-radius ingestion zone. The ERP describes (1) the emergency classification system used at the plant; (2) the organizational control of emergencies including onsite, offsite, and augmentation organizations; (3) the emergency measures to be taken; and (4) available emergency facilities and equipment. Procedures for implementation of the CCNPP ERP are contained in the Emergency Response Plan Implementation Procedures (ERPIPs). The procedures are distributed to those individuals, facilities, and organizations where immediate availability of such procedures would be required during an emergency. The ERPIPs provide the following information: 1. Means of classifying emergencies, lists of available equipment, and emergency information; 2. Directions for meeting notification requirements; 3. Directions for seeking emergency assistance; and,

4. Detailed instructions to individuals responsible for (a) assessing emergency conditions and (b) providing steps to be taken to mitigate the consequences of the accident. The ERPIPs are used in conjunction with applicable plant operating, radiological control, and security procedures to correct the emergency condition and to mitigate the consequences of the accident. 12.6.2 ESSENTIAL ELEMENTS FOR ADVANCE PLANNING The CCNPP Emergency Response Program, is defined by two separate but totally coordinated documents. The first document, the ERP, provides the basis for performing advance planning and for defining specific requirements and commitments to be implemented by other documents and procedures. The second document, the ERPIPs, provides the detailed information and procedures that will be required to implement the ERP in the event of an emergency at CCNPP. These two documents are briefly described below.

CALVERT CLIFFS UFSAR 12.6-2 Rev. 47 12.6.2.1 Emergency Response Plan The CCNPP ERP ensures that all emergency situations, including those that involve radiation or radioactive material, are handled properly and efficiently. It covers the entire spectrum of emergencies from minor, localized emergencies to major emergencies involving action by offsite emergency response agencies and organizations. The CCNPP ERP includes a system for classifying emergencies. Thus, the CCNPP ERP provides the overall advance planning required for the development of methods of implementation that are included in the ERPIPs. In summary, the CCNPP ERP provides the following: 1. A means for classifying emergency conditions in a manner compatible with a system utilized by Federal, State and County emergency response agencies and organizations; 2. A means of reclassifying such emergency conditions should the severity increase or decrease; 3. Identification of normal and emergency operating organizations; 4. General guidelines as well as specific details as to which County, State, and Federal authorities and agencies and other outside organizations are available for assistance; 5. Information pertaining to the Emergency Operations Facility, Technical Support Center, Operational Support Center, Joint Information Center, and equipment available both onsite and offsite; 6. Requirements for training, drills, reviews, and audits to ensure a high degree of emergency preparedness and operational readiness; 7. Figures and tables that display such information and data as organization charts, maps and population distributions; 8. Specific plans and agreements pertaining to participating offsite organization and agencies. 12.6.2.2 The Emergency Response Plan Implementation Procedures The purpose of the ERPIPs is to provide (1) a single source of pertinent information and data and (2) the procedures that would be required by or useful to various emergency response agencies and organizations in the event of an emergency at CCNPP. The ERPIPs consolidate and integrate specific material required for personnel to implement or support the CCNPP ERP and the State Radiological ERP. The ERPIP document is organized to provide the following: 1. Means of classifying emergencies, lists of available equipment, and emergency information. 2. Procedures that: a. Provide the CCNPP staff and supporting agencies with specific instructions for the implementation of the CCNPP ERP; b. Assign specific responsibility and authority to emergency response personnel; c. Provide a single source of pertinent information, forms, data and step-by-step instructions to ensure prompt actions and proper notifications and communications are carried out; d. Provide a record of the completed actions; and, e. Provide the mechanism by which emergency preparedness will be maintained at all times. CALVERT CLIFFS UFSAR 12.6-3 Rev. 47 12.6.3 ORGANIZATIONAL CONTROL OF EMERGENCIES 12.6.3.1 Emergency Response Organization The first line of control of any emergency at CCNPP lies with the normal shift personnel on duty at such time as an emergency situation should occur. Assistance is available within about one hour from other plant staff and operating personnel. Additional assistance is available from Corporate, local, State, Federal Agencies, and contractor personnel. The Emergency Director/Recovery Manager is in charge of onsite emergency activities, described in this plan and further delineated in the ERPIPs, as the main contact at the site. He/she also directs communications to and from the site. In addition to directing the staff and operating personnel, he/she can call on additional Company and outside agency assistance as needed. 12.6.3.2 Recovery Organization The Recovery Organization is activated at the direction of the Emergency Director/Recovery Manager. The Recovery Organization is responsible for providing additional personnel and technical assistance from offsite sources.

12.6.3.3 Offsite Emergency Organization Calvert Cliffs Nuclear Power Plant is equipped and staffed to cope with many types of emergency situations. However, if a fire or other type of incident occurs that requires outside assistance, such assistance is available from state agencies, local services, and contractors including: A. State of Maryland 1. Governor 2. Maryland Emergency Management Agency 3. Department of Health and Mental Hygiene 4. Department of Agriculture

5. Department of Natural Resources

6. Maryland State Police

7. Department of Human Resources

8. Department of Transportation 9. Department of Education 10. Department of Housing and Community Development

11. Maryland Military Department/National Guard

12. Maryland Institute for Emergency Medical Service Systems

13. Office of the Comptroller of the Treasury 14. State Fire Marshall 15. Maryland Department of Environment B. Calvert Memorial Hospital C. Local Fire and Rescue Service D. Naval Oceanographic Command Detachment E. Emergency Medical Assistance Program Contractor CALVERT CLIFFS UFSAR 12.6-4 Rev. 47 The primary functions of the above-listed agencies and services can be found in the ERP.

12.6.4 EMERGENCY PLANNING ZONES The ERP identifies the interface between CCNPP and the governmental agencies having action responsibilities to ensure the protection of the population-at-risk with Emergency Planning Zones of CCNPP. The Plume Exposure Pathway extends to 10 miles from CCNPP and the Ingestion Exposure Pathway extends 50 miles from the site. CHAPTER 13 INITIAL TESTS AND OPERATION TABLE OF CONTENTS PAGE13.0 INITIAL TESTS AND OPERATION 13.4 POST-REFUELING STARTUP TESTING CHAPTER 13 INITIAL TESTS AND OPERATION LIST OF ACRONYMS CALVERT CLIFFS UFSAR 13.0-1 Rev. 47 INITIAL TESTS AND OPERATION 13.0Sections 13.1 through 13.3 are historical information about tests that were conducted when the units were first started up. They can also be accessed through the NORMS Records System. The Document IDs (Doc ID) are "Section-13.1, Section-13.2, and Section-13.3."

CALVERT CLIFFS UFSAR 13.4-1 Rev. 47 13.4 POST-REFUELING STARTUP TESTING The following discussions represent the major startup tests conducted for Calvert Cliffs subsequent to each refueling. Sufficient data is obtained to verify that the plant operates in a safe condition within the bounds of the applicable acceptance criteria and, therefore, the Safety Analysis. 13.4.1 HOT FUNCTIONAL TESTING 13.4.1.1 CEDM Performance Testing During this testing, the proper functioning of the control element assemblies (CEAs), control element drive mechanisms (CEDMs), and CEA position indication are verified through the insertion and withdrawal of the CEAs. Rod drop times are measured and evaluated. Proper latching of the single CEA to the respective CEA extension shaft is confirmed. Any irregularities are analyzed. 13.4.1.2 RCS Flow Verification Reactor Coolant System (RCS) flow rates are verified based on differential pressure measurements obtained across the reactor coolant pumps and the reactor vessel. These values are compared for consistency to those obtained during previous testing. 13.4.2 INITIAL CRITICALITY Approach to criticality commences with the withdrawal of the Shutdown CEA Groups, followed by the withdrawal, in sequence, of the Regulating CEA Groups, concluding with Group 5 at mid-core. Criticality is established through boron dilution. The plant is allowed to stabilize, then proceed to the low power physics tests to verify physics design parameters. 13.4.3 LOW POWER PHYSICS TESTING 13.4.3.1 Deleted 13.4.3.2 Critical Boron Concentration Critical Boron Concentrations are determined for all CEAs withdrawn and with CEA Groups 1 through 5 inserted.

13.4.3.3 Isothermal Temperature Coefficient The Isothermal Temperature Coefficient is determined by varying the RCS temperature. Control element assembly Regulating Group 5 is used to maintain flux and reactivity within a defined operating band.

13.4.3.4 CEA Group Worth Measurements The RCS is diluted/borated while the CEAs are inserted/withdrawn in order to compensate for a change in reactivity. These changes are monitored via the reactivity computer. 13.4.4 POWER ASCENSION TEST Power ascension testing consists of frequent monitoring of core power distribution with specific acceptance and review criteria established for 30, 60 and 85% power levels. Equilibrium conditions are established near full power. The Isothermal Temperature Coefficient is measured and compared with predictions. CALVERT CLIFFS UFSAR 13.4-2 Rev. 47 13.4.5 ACCEPTANCE CRITERIA Acceptance and review criteria for the above startup tests are listed below: Parameters Acceptance Criteria Review Criteria CEA Groups Worth Greater of +/- 15% Or +/- 0.1% Greater of +/- 15% Or +/- 0.1% Total Regulating CEA Group Worth +/- 10% of predicted +/- 10% of predicted Critical Boron Concentration +/- 0.75% of predicted +/- 0.5% of predicted Isothermal Temperature Coefficient Technical Specification limits for Moderator Temperature Coefficient +/- 0.2x10-4 /°F Power Distribution Technical Specification limits on and Tq Measured radial assembly power distributions (the greater of) 30% +/- 15% or +/- 0.15 RPD, or as limited by the applicable core misload detection analysis 60% +/- 10% or +/- 0.10 RPD 85% +/- 10% or +/- 0.10 RPD Full Power +/- 10% or +/- 0.10 RPD Where RPD is the relative power density Core Symmetry Evaluation

a. Tilt 30% Power None +/- 3% 60,85,97% Power +/- 3% +/- 2% b. Symmetric ICI None +/- 10%

Box Powers 13.4.6 ACTION AND REVIEW PLAN The Engineering Supervisor-Fleet Nuclear Fuels, will review the comparison of measurements with Review/Acceptance Criteria. If any Review Criteria are exceeded, an evaluation will be made to determine first, the applicability of the prediction to the precise plant conditions under which the measurement was performed and, second, the accuracy of the measurement. As a result of this review, the measurement may be repeated and/or the prediction may be updated, if required, to reflect actual plant conditions at the time of measurement.

If any measurement from the lower power physics tests exceeds its Review Criteria, the Plant Operations and Safety Review Committee will review results of the low power physics tests and ensure that Acceptance Criteria are met prior to recommending operation above 5% of Rated Thermal Power. If, as a result of this review, it is determined that a Technical Specification limit has been exceeded, then appropriate action as required by Technical Specifications will be taken. A similar action plan for power ascension testing will be followed prior to increasing power beyond the 60 and 85% power plateaus. If any Acceptance Criteria, except for bank worth, are exceeded, the validity of the physics data input to the Safety Analysis for the entire cycle will be determined. If it can be demonstrated that the measured value of the particular parameter in question, when CALVERT CLIFFS UFSAR 13.4-3 Rev. 47 combined with the values of the other safety-related parameters, does not increase the severity or consequences of accidents or anticipated operational occurrences, the test results will be deemed acceptable. Additional measurements of safety-related parameters may be performed in order to support this demonstration.

If any regulating bank worth measurement falls outside of its acceptance criterion or if the total worth of the regulating banks falls outside of its acceptance criterion, shutdown Bank C shall be measured and compared with its acceptance criterion. If shutdown Bank C worth falls outside of its acceptance criterion or if the accumulated total worth of all the banks measured falls below their total worth acceptance criterion (after appropriate corrections and adjustments), then an evaluation shall be made of the validity of the safety analyses for the entire cycle, similar to the procedure discussed above for other measurement data. If the combination of safety parameters determined above falls outside of the range of safety parameters used to support the proposed operation of the plant, the plant operating limits will be adjusted to prevent conditions that could result in exceeding the Specified Acceptable Fuel Design Limits.

CALVERT CLIFFS UFSAR 14.2-1 Rev. 47 14.2 CONTROL ELEMENT ASSEMBLY WITHDRAWAL EVENT 14.2.1 IDENTIFICATION OF EVENT AND CAUSE The CEAs, in a pre-programmed sequence (according to the PDIL), are used to control (dampen) xenon oscillations and rapidly control core power. The Regulating and Shutdown CEA Groups also provide the required negative reactivity for shutdown during DBEs.

The action of the control element assembly withdrawal (CEAW) Prohibit will stop the CEAs from withdrawing under the following conditions and, thus, prevent the CEAs from aggravating the situation.

a. High neutron flux power level pre-trip; b. High rate of change pre-trip; or,

c. Thermal Margin/Low Pressure pre-trip. The action of the CEA motion inhibit will prevent any CEA from being raised or lowered under the following conditions.

a. PDIL alarm; b. CEA Regulating group out of sequence alarm; or, c. CEA Regulating or Shutdown group deviation alarm. Consequently, the CEA motion inhibit prevents the groups from being moved outside the pre-programmed sequence and prevents a single CEA from being misaligned. Therefore, only a sequential CEA group withdrawal needs to be addressed. A CEAW event is defined as any event caused by a single malfunction in the Reactor Regulating System (RRS) or Control Element Drive Mechanism (CEDM) control system that results in a continuous sequential CEA group withdrawal. The CEA position indication systems are programmed to produce no more than a 40% overlap between CEA Regulating groups with Group 1 being withdrawn first and Group 5 last.

The CEAW event has been re-analyzed to support the use of AREVA fuel at Calvert Cliffs. The results of the analysis are presented in Section 14.2.4 and support the transition fuel cycle, as well as, full core implementation of AREVA fuel at Calvert Cliffs. 14.2.2 SEQUENCE OF EVENTS A CEAW event can approach the DNBR and linear heat generation rate (LHGR) SAFDLs and the RCS Pressure Upset Limit. Initial margins maintained by the LCOs in conjunction with the RPS (VHPT, TM/LP trip, or LPD trip) ensure that these design limits will not be exceeded. Since no pin failures are postulated to occur, the site boundary dose criteria in 10 CFR 50.67 guidelines will not be approached. AREVA analyzed HFP and HZP conditions, based upon the assertion that the range of initial reactor power levels is bounded by analyzing only full-power cases due to the VHPT setpoint automatically resetting to track the current operating power level, resulting in a proportionately lower setpoint for a part-power case than for a full power case. The NRC asserts that the possible transient variations in the core power distribution may lead to a more limiting DNBR at lower power levels. The response to this concern incorporated an explicit analysis of part-power transients using an operating envelope that bounded Calvert Cliffs. However, the analysis did not provide a Calvert Cliffs specific basis to conclude that changes in the core operating limits are acceptable with respect to the part-power transient. Therefore, a licensing condition is imposed in the Technical CALVERT CLIFFS UFSAR 14.2-2 Rev. 47 Specifications to restrict certain Core Operating Limits Report limits from being changed without prior NRC review and approval until an NRC-accepted, or Calvert Cliffs specific, basis is developed for analyzing the CEAW event at full power conditions only (Reference 3). 14.2.2.1 Zero Power Case The zero power case is assumed to initiate at a HZP, critical condition. For events with high reactivity insertion rates, the positive reactivity insertion caused by CEA withdrawal will cause power to increase at an exponential rate. Since the event initiates at zero power with no heat being produced in the fuel, the heat flux will also be zero. If the reactor power goes above 10-4% of rated power and the rate of change of neutron flux is greater than 1.5 decades per minute, a CEAW Prohibit will be initiated. If the rate exceeds 2.6 decades per minute, with the power between 10-4 and 15% of rated power, a reactor trip will be initiated. For conservatism, no credit for the High Rate-of-Change of Power Trip is taken in the analysis presented, i.e., for an event that initiates from a critical condition. However, the high rate-of-change of power trip is credited as justification for not analyzing subcritical CEAW events. By the time core power reaches 1% of rated power, the neutron flux will be increasing exponentially at an extremely high rate. Although the reactor trip occurs at the minimum setting on the VHPT (30%; analysis assumes 36.4%), the core power could peak above 100% power depending on the worth of the CEAs. The core power peaks after trip due to the RPS electronic and the CEA holding coil delays, and the time necessary to insert enough SCRAM reactivity to offset the positive reactivity insertion. As core power increases, the fuel temperature will increase and result in Doppler feedback, which will reduce the peak core power. Due to the fuel time constant, the core heat flux will lag the core power and consequently result in a lower peak. The core power will rapidly decrease as the CEAs are inserted, thus terminating the power excursion. The core average temperature will slowly increase as it follows the increase in core heat flux. Due to the loop cycle time, the core inlet temperature will lag the average temperature. With the exit temperature increasing faster than the inlet temperature, more moderator feedback will occur in the top portion of the core. Since the MTC is generally negative, more negative reactivity will be inserted in the top portion of the core resulting in the power peak going to the bottom of the core. For conservatism in the analyses, a positive MTC is used thereby adding positive reactivity.

During the event, the RCS pressure will increase and follow the core average temperature rise. Depending on the CEA worth, the temperature rise could increase the pressurizer pressure above the power-operated relief valve (PORV) setting. The action of the pressurizer pressure and level control systems will moderate the pressure peak. For peak pressure consideration, no credit is allowed for these systems. The PORVs act to decrease primary pressure, resulting in more adverse DNBR consequences. The peak primary system pressure is not explicitly calculated. It is expected to be benign due to the MSSVs being available in Mode 3 and higher, and in Modes 1 and 2, the secondary system is available for heat removal until after reactor trip. The SG temperature and pressure will increase as the core temperature increases. Upon the reactor trip, the atmospheric dump and steam bypass valves CALVERT CLIFFS UFSAR 14.2-3 Rev. 47 will normally modulate the core average temperature and SG pressure to 532°F and below 900 psia, respectively. With the quick opening of the valves, the core inlet temperature will initially decrease and then follow core average temperature. 14.2.2.2 Full Power Case The full power case is initiated at 100% of rated power and at the LCOs. As the CEAs are withdrawn at the preprogrammed rate, the core power will steadily increase at a rate dependent on the worth of the CEAs. If the CEAs are being withdrawn from a high worth region (i.e., a region in which the CEAs are suppressing the power), the core power will increase at a fast rate. Conversely, if the CEAs were initially in a low worth region, the power will increase at a slow rate.

The withdrawal of CEAs will cause the axial power distribution to shift to the top of the core. The associated increase in the axial peak is compensated by a decrease in the integrated radial peaking factor. The magnitude of the 3-D peak change depends primarily on the initial CEA configuration and the initial axial power distribution. The withdrawal of CEAs will also cause the neutron flux power measured by the excore detectors to be decalibrated due to rod shadowing. This decalibration of excore detectors, however, is partially compensated for by neutron attenuation due to moderator density changes (temperature shadowing).

As the core power increases, the fuel temperature will increase and result in negative Doppler reactivity feedback. The core average heat flux will slowly increase and lag the core power at an increment dependent on the clad-fuel gap conductance. With the heat flux increasing, the core average temperature will increase. With the core average temperature increasing, the moderator feedback will increase or decrease the rate of reactivity addition depending on whether the MTC is positive or negative. As a result of the increase in core average temperature, the RCS pressure will increase. If the CEAs are fully withdrawn before any trip is reached, a new steady-state at a higher core power and core average temperature will result. With the turbine still demanding 100% of rated power, the atmospheric dump and bypass systems will pick up the additional power (load). Assuming a large enough withdrawn CEA worth, a trip will occur on either the Variable High Power, Axial Flux Offset, TM/LP, or High Pressurizer Pressure Trips. The amount of withdrawn CEA worth to cause a trip depends on the MTC, Doppler coefficient, and the position of the CEAs. During a CEAW with the fuel and the RCS heating up, the MTC (usually negative during power operation) and the Doppler coefficient (always negative during power operation) will offset part of the withdrawn CEA worth. With the RCS temperature increasing, the pressurizer pressure and the level will increase. Although no credit is taken in the analysis, the pressurizer sprays will partially suppress the pressure increase and the level control system will maintain the programmed level. For cases where the pressurizer pressure exceeds 2400 psia, a reactor trip will be initiated and the PORVs will open, thereby reducing the number of times the PSVs are actuated. For peak pressure consideration, the PORVs are assumed to be inoperable as they are non-safety grade. In addition, no credit is allowed for the action of the atmospheric dump and turbine bypass valves which would normally maintain the SG below 900 psia and regulate the average RCS temperature at 532°F. CALVERT CLIFFS UFSAR 14.2-4 Rev. 47 When addressing the fuel DNB SAFDLs, the pressurizer sprays and PORVs are assumed operable to minimize system pressure. These systems act to maximize the margin required to account for transient shifts, which are necessary due to lack of dynamic compensation in the TM/LP trip. Transient shifts account for changes in monitored parameters that occur between the time a TM/LP trip pressure is sensed and the time of MDNBR. 14.2.3 CORE AND SYSTEM PERFORMANCE 14.2.3.1 Mathematical Models The transient response of the RCS and steam systems to the CEAW event was simulated using the S-RELAP5 thermal-hydraulic system code, described in Section 14.1.4.1, consistent with the methodology in Reference 2. The XCOBRA-IIIC fuel assembly thermal-hydraulic code, described in Section 14.1.4.1, was used to calculate the flow and enthalpy distributions for the entire core and the DNB performance for the DNB-limiting assembly. The limiting assembly DNBR calculations were performed using an NRC-approved AREVA DNB correlation. The overall core conditions calculated by S-RELAP5 during the transient were used as the input to the XCOBRA-IIIC calculation. The limiting design axial power profile (a top peaked axial power distribution) was used for this simulation. 14.2.3.2 Input Parameters and Initial Conditions The input parameters and initial conditions used in the analysis are listed in Table 14.2-1. Those parameters that are unique to the analysis are discussed below. For DNB cases, reactivity parameters were chosen such that a new steady-state power level was reached at the VHPT setpoint to maximize DNBR degradation. For FCM cases, reactivity parameters were set to ensure the greatest power excursion and maximum peak FCM. The key parameters for the CEAW event initiated from both HFP and HZP for determining minimum DNBR are the reactivity insertion rate due to rod motion, the MTC and the FTC. The maximum CEAW rate is calculated by combining the maximum CEA differential worth (/inch) and the maximum CEAW speed of 30 in/min. The minimum transient DNBR is calculated using the most adverse DNBR initial conditions. The analysis conservatively assumed a positive MTC value since this, in combination with increasing coolant temperatures, inserts a positive reactivity and thus maximizes the power and heat flux transients. The analysis also assumed a BOC FTC with uncertainty. This, in combination with increasing fuel temperatures, inserts the least amount of negative reactivity due to Doppler feedback. This minimizes the transient minimum DNBR. For HZP conditions, the scram worth of the CEAs is set to the Technical Specification minimum shutdown margin, which is more limiting than assuming that the highest-worth CEA is stuck in the fully withdrawn position. The initial power level used for the analysis is the lowest following an extended shutdown (assumed to be 10-9 times the rated power). The analysis assumes that the event is preceded by an extended shutdown because the extremely low neutron population under such a condition delays the power increase as the CEAs are withdrawn until a significant amount of positive reactivity has been added, which maximizes the subsequent power excursion. The combination of the highest reactivity addition rate and the lowest initial power level produces the CALVERT CLIFFS UFSAR 14.2-5 Rev. 47 highest peak values of the fuel rod surface heat flux and centerline temperature, which result in limiting DNB and FCM values. The VHPT setpoint is further decalibrated by a factor that accounts for changes in the peripheral power that may occur as CEAs are withdrawn from the interior of the core.

For HFP conditions, a spectrum of positive insertion rates is analyzed from very slow to fast, limited only by bank worth and maximum drive speed. Two reactivity feedback matrices of cases are evaluated: for most-positive reactivity feedback (most-positive MTC and least-negative Doppler coefficient) and the other for most-negative feedback (most-negative MTC and most-negative Doppler coefficient). For both matrices of reactivity feedback cases, reactivity insertion rate ranges bounding the respective lowest MDNBR point and the maximum value for CEA bank withdrawal are considered. The lower bound of the reactivity insertion rate range analyzed is also considered to be bounding of a reasonable minimum reactivity insertion rate for bounding Mode 1 Boron Dilution. Protection against violation of the SAFDLs is provided by the VHPT or the TM/LP trip in the analysis.

14.2.3.3 Results Table 14.2-2 contains the sequence of events for the zero power case for the maximum withdrawal rate. Figures 14.2-1 through 14.2-4 present the transient behavior of the core power, core average heat flux, RCS temperatures, and RCS pressure as a function of time. Also, the analysis revealed that the fuel centerline temperatures are well below those corresponding to the acceptable FCM limit provided in the Technical Specifications. Table 14.2-3 contains the sequence of events for the full power case for the limiting withdrawal case with respect to DNBR SAFDL. Figures 14.2-5 to 14.2-8 present the transient behavior of the core power, core average heat flux, RCS temperatures, and RCS pressure as a function of time for this case. The limiting case is a CEAW from EOC HFP conditions with a withdrawal rate of 7.55 pcm/second. The analysis also concluded that the fuel centerline temperatures are well below those corresponding to the acceptable FCM limit. The S-RELAP5 plant simulation results from the analysis of the CEAW event were used as input into the MDNBR calculations. The S-RELAP5 plant simulation was adjusted to account for power uncertainty. The temperature, pressure, and flow measurement uncertainties are accounted for in the MDNBR calculations. The MDNBR was above the high thermal performance DNB correlation upper 95/95 limit plus a 2% mixed core penalty for both the zero and full power cases.

The FCM calculation for the HZP case results in a fuel centerline temperature that is significantly less than the melt temperature provided in the Technical Specifications. The HFP case results in LHGR less than the LHGR FCM safety limit.

The CEAW event is not limiting with respect to peak RCS pressure. With no loss of secondary load or feedwater and no loss of offsite power, a reactor trip on VHPT or high pressurizer pressure, along with primary safety valve capacity, is sufficient to maintain peak RCS pressure well below the over pressurization limit. CALVERT CLIFFS UFSAR 14.2-6 Rev. 47 Other events, such as Loss of Load, Loss of Normal Feedwater and Feedline Break, all exceed this event with respect to peak RCS pressure. The radiological consequences of opening the atmospheric dump valve during the most adverse CEAW event is less adverse than the LOAC event. 14.

2.4 CONCLUSION

S The analysis of the CEAW event demonstrates that the initial margin maintained by the LCOs in conjunction with the action of the RPS prevents exceeding the fuel SAFDLs and the RCS Pressure Upset Limit during an uncontrolled CEAW transient. The radiological consequences of opening the atmospheric dump valve upon reactor trip during the most limiting CEAW event is a site boundary dose, which is negligible compared to the 10 CFR 50.67 guidelines. Since the DNBR and centerline temperature melt (CTM) design limits are not exceeded for this event and no fuel pins are predicted to fail, it is concluded that extended burnup has no adverse impact during this event. 14.

2.5 REFERENCES

1. Deleted 2. EMF-2310(P)(A), Revision 1, "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors," May 2004 3. Letter from Mr. D. V. Pickett (NRC) to Mr. G. H. Gellrich (CCNPP), dated February 18, 2011, Calvert Cliffs Nuclear Power Plant, Units Nos. 1 and 2 - Amendment Re: Transition from Westinghouse Nuclear Fuel to AREVA Nuclear Fuel CALVERT CLIFFS UFSAR 14.2-7 Rev. 47 TABLE 14.2-1 INITIAL CONDITIONS AND INPUT PARAMETERS - CEAW EVENT HZP PARAMETER UNITS HZP VALUE Initial Core Power MWt 2.737 x 10-6 Initial Core Inlet Temperature °F 532 Initial RCS Pressure psia 2250 Initial Vessel Flow Rate gpm 370,000 Combined Bank Differential Worth pcm/inch 32.0 Bank Withdrawal Rate inches/minute 30 Scram Worth pcm 3500 VHP Trip Setpoint %RTP 36.4 VHP Trip Delay sec 0.4 MTC pcm/°F +7.0 Maximum Predicted FQ for HZP CEA Withdrawal --- 3.688 Rod Shadowing Power Decalibration --- 0.677 HFP PARAMETER UNITS HFP VALUE Initial Core Power MWt 2754 Initial Core Inlet Temperature °F 548 Initial RCS Pressure psia 2250 Initial Vessel Flow Rate gpm 370,000 Combined Bank Differential Worth pcm/sec 0.0002 to 5.0 for positive feedback 3.00 to 8.00 for negative feedback Scram Worth pcm 5277.6 VHP Trip Setpoint %RTP 110.33 VHP Trip Delay sec 0.9 MTC pcm/°F +7.0 for positive feedback -33 for negative feedback Doppler Temperature Coefficient pcm/°F -0.80 for positive feedback -1.85 for negative feedback CALVERT CLIFFS UFSAR 14.2-8 Rev. 47 TABLE 14.2-2 SEQUENCE OF EVENTS FOR ZERO POWER CEAW EVENT TIME (sec) EVENT VALUE 0.0 Bank Withdrawal Begins 16.0 pcm/sec 37.55 Core Power Reaches VHP Trip Setpoint 53.767 %RTP 37.95 Reactor Trip Signal Generated 80.3 %RTP 38.45 Control Rods Released 98.9 %RTP 38.55 Maximum Nuclear Power 99.5 %RTP 40.16 Maximum Heat Flux Power 1198.8 MWt 43.8 %RTP CALVERT CLIFFS UFSAR 14.2-9 Rev. 47 TABLE 14.2-3 SEQUENCE OF EVENTS FOR FULL POWER CEAW EVENT TIME (sec) EVENT VALUE 0.0 Bank Withdrawal Begins --- 109.06 Trip Setpoint Reached --- 109.73 Maximum Power 111.01 %RTP 109.96 Reactor Trip Signal Generated TM/LP 110.32 Maximum Heat Flux Power 110.58 %RTP 110.46 Control Rods Released TM/LP CALVERT CLIFFS UFSAR 14.3-1 Rev. 47 14.3 BORON DILUTION EVENT 14.3.1 IDENTIFICATION OF EVENT AND CAUSE The CVCS regulates both the chemistry and the quality of coolant in the RCS. Changing the boron concentration in the RCS is a part of normal plant operation, compensating for long-term reactivity effects such as fuel burnup, xenon buildup and decay, and plant startup. During refueling operations, borated water is supplied from the refueling water tank (RWT).

Boron concentration in the RCS can be decreased either by controlled addition of demineralized water or by using a purification ion exchanger with a deborating resin. During normal operation, concentrated boric acid solution and demineralized water is introduced into the volume control tank in concentrations corresponding to the required concentration for proper plant operation. When the specified amount has been injected, the makeup controller automatically shuts the demineralized water and boric acid control valves. A purification ion exchanger with a deborating resin is normally used for boron removal when the boron concentration in the RCS is low (less than 50 ppm) and the feed and bleed method becomes inefficient.

The CVCS is equipped with the following indications which could inform the operator when a change in boron concentration in the RCS is occurring: a. Volume control tank level; b. Makeup controller flow; and,

c. CVCS valve position lineup. A Boron Dilution event is defined as any event caused by a malfunction or an inadvertent operation of the CVCS that results in a dilution of the active portion of the RCS. The active portion of the RCS is defined as that volume of water that circulates through the core. For example, when in shutdown cooling (SDC), no credit is allowed for the volume of water in the SG and other stagnant portions of the RCS. A dilution of the RCS can be the result of adding borated water, which has a boron concentration that is less than the system boron concentration, or by the removal of boron using a purification ion exchanger with a deborating resin.

14.3.2 SEQUENCE OF EVENTS The analysis of the Boron Dilution event covers the six modes of operation (Table 14.3-1). The modes of operation are defined in Technical Specification Table 1.1-1. A Boron Dilution event can approach the DNBR and LHGR SAFDLs and the RCS Pressure Upset Limit. In all cases, operator action is required to prevent exceeding these limits by securing the dilution and borating, if necessary, to maintain the required shutdown boron concentration. The calculated time-to-criticality is dependent upon the critical and shutdown boron concentrations, the RCS coolant mass and the flow rate of the dilution stream. In addition, for the dilution front model, a range of SDC flow rates is required.

Assume the boron concentration is exactly the amount needed to maintain the required Technical Specification shutdown margin. Also assume that under the worst conditions the operator has 30 minutes in the refueling mode and 15 minutes in the other modes of operation from the time of initiation of the event to secure the dilution to prevent losing the minimum shutdown margin. The DNBR and LHGR SAFDLs and the RCS Pressure Upset Limit criteria will be met if the entire shutdown margin is not lost.

CALVERT CLIFFS UFSAR 14.3-2 Rev. 47 14.3.2.1 Power Operation and Startup An inadvertent Boron Dilution event at power and startup (Modes 1 and 2) can be postulated as a result of various malfunctions of, or inadvertent operation of, the CVCS. The sequence of events starts with the decrease of the boron concentration in the RCS. All three charging pumps are on and adding 150 gpm of unborated (demineralized) water into the RCS. The effect of decreasing the boron concentration is to add positive reactivity. With the reactor initially critical, the core power, heat flux, and RCS temperatures will increase the pressurizer pressure and level. Although no credit is taken for them in the analysis, the pressurizer pressure and level control systems will maintain programmed pressure and level. The combination of the pressurizer sprays and RCS letdown will accommodate these slow increases in pressure and level respectively. The SG temperature and pressure will slowly increase with the increasing average RCS temperature. This is a similar sequence to a slow reactivity addition due to a CEA withdrawal. The increasing fuel and moderator temperatures will result in a negative reactivity feedback due to the Doppler and the MTC being generally negative. The negative reactivity feedbacks will partially offset the positive reactivity insertion due to the dilution, thus further slowing down the power rise. If the dilution is not secured, the reactor will be shut down by either the TM/LP or the VHPT. The action of the pressurizer control system will prevent the pressure from exceeding the High Pressurizer Pressure Trip setpoint. Operator action is required to secure the dilution. 14.3.2.2 Hot Standby, Hot Shutdown For the Hot Standby (Mode 3) and Hot Shutdown (Mode 4) cases, the analysis assumes all three charging pumps are on (as in Section 14.3.2.1 above) and assumes a boron concentration to meet the required shutdown margin, and shutdown to critical boron concentration ratios are as in Table 14.3-1. Mode 3 assumes RCPs are operating and mix the entire RCS loops.

While in Hot Shutdown the limiting shutdown to critical boron ratio occurs while on SDC. The active volume of the RCS includes only that volume to the top of the Hot Leg plus the SDC system. Rapid mixing cannot be assumed in the reactor vessel head or the SG. 14.3.2.3 Cold Shutdown Two cases are run for Cold Shutdown (Mode 5): a three-pump case and a two-pump case. The three-pump case uses the same volume as the hot shutdown case. For the two-pump case, Cold Shutdown (Mode 5), the NSSS could be partially drained due to repairs or inspections on the RCS (e.g., RCS pump seal replacement, SG inspection, etc.). Therefore, the analysis assumes an active volume of water which is sufficient to fill the RCS to the bottom of the hot leg plus fill the SDC System. Technical Specifications require at least one train of SDC to be in operation. Since the Technical Specifications only allow 88 gpm when the pressurizer level is below 90" in Mode 5, the analysis uses 100 gpm when the level is at the bottom of the hot leg.

14.3.2.4 Refueling The analysis for Refueling (Mode 6) uses an active volume in the RCS that is the same as cold shutdown. The refueling boron concentration is defined in terms of a ratio of refueling to critical concentration. Changes in the boron concentration CALVERT CLIFFS UFSAR 14.3-3 Rev. 47 which occur each cycle, can be easily evaluated by comparing the ratio to the minimum allowable ratio presented in Table 14.3-1. 14.3.3 CORE AND SYSTEM PERFORMANCE 14.3.3.1 Mathematical Models Since the NSSS response to a Boron Dilution event is basically the same as a slow CEA Withdrawal event, only the time to criticality is calculated. The rate of change of boron concentration as a function of time can be described by the below equation. The boron in the active volume will be uniformly mixed when sufficient flow exists. Instantaneous mixing is assumed when an RCP is operating. Therefore, the time to lose the prescribed shutdown is: or where: M = active volume Co = initial boron concentration Cc = critical boron concentration W = charging volume flow rate tc = time to lose a prescribed shutdown margin RCS = density of active volume chgg = density of charging water The ratio of shutdown boron to critical boron is: where: Csdm = shutdown boron concentration CRCS = critical boron concentration tsdm = criterion for minimum time to lose prescribed shutdown margin The dilution front model is used when the RCS flow is much slower than would occur with at least one RCP running. Typical loop transient times of several minutes are associated with these low flow conditions. The assumption of instantaneous mixing is no longer valid in these low flow conditions; however, the dilution flow is assumed uniformly mixed with the RCS coolant in the vessel downcomer and the lower plenum regions prior to reaching the core inlet.

The time for the first dilution front to reach the core is calculated by dividing the RCS mass from the mixing location to the bottom of the core by the shutdown cooling + dilution flow. The time for subsequent front is calculated by dividing the mass of the RCS and shutdown cooling systems by the shutdown cooling + dilution flow. The time to criticality is determined by iteratively tracking the number of dilution fronts. 14.3.3.2 Input Parameters and Initial Conditions The input parameters and initial conditions used in the analysis are listed in Table 14.3-1.

CALVERT CLIFFS UFSAR 14.3-4 Rev. 47 The instantaneous mixing and dilution front models are dependent upon the mass of liquid in the active volume, the dilution stream mass flow rate, and the initial to critical boron ratio. The shutdown margin requirement, inverse boron worth and time in cycle are inherently included in the boron ratio. The dilution front model is also dependent upon the RCS flow rate, which is set equal to the shutdown cooling flow rate.

14.3.3.3 Results Table 14.3-2 contains the minimum time to lose the prescribed shutdown margin for Modes 3-6 of operation. The results show that the operator has sufficient time to take appropriate action to mitigate the consequences of this event for each operating mode.

For Modes 1 and 2, an inadvertent charging of unborated water at the maximum rate would result in a maximum rate of reactivity addition that is within the range evaluated for a CEA Withdrawal event and is therefore not as limiting. 14.

3.4 CONCLUSION

In Modes 1 and 2, the RPS initially mitigates the consequences of a Boron Dilution event; after which the operator has sufficient time to terminate the dilution. In all other modes, sufficient time is provided to allow operator action to mitigate the consequences before shutdown margin is lost.

The worst time in life for this event is at BOC when boron concentration is highest and MTC is least negative. Therefore, increased burnup has no adverse effect on this transient. 14.3.5 NRC ACCEPTANCE LIMIT The acceptance criteria for this event are that the times between initiation of a boron dilution event and loss of shutdown margin are not less than 15 minutes for Modes 2, 3, 4 (Reference 1), and 5, and 30 minutes for Mode 6 (Reference 2). The SER (Reference 2) also states that the analysis of boron dilution for power operation is acceptable because the operator has adequate time to terminate the boron dilution event due to the TM/LP trip and VHPT.

Standard Review Plan Section 15.4.6 requires plants licensed to these requirements to demonstrate that Control Room operations will have a positive alarm indicating the onset of a boron dilution event. Generic Letter 85-05, dated January 31, 1985 and titled "Inadvertent Boron Dilution Events," documents the NRC determination that the consequences of this event do not warrant backfitting this requirement to plants (such as Calvert Cliffs) that are not currently licensed to the Standard Review Plan for this event. 14.

3.6 REFERENCES

1. Letter from R. A. Clark (NRC) to A. E. Lundvall, Jr., dated December 12, 1980, Issuance of Amendment No. 48 to Facility Operating License No. DPR-53 2. Letter from S. A. McNeil (NRC) to J. A. Tiernan, dated May 4, 1987, Issuance of Amendment No. 108 to Facility Operating License No. DPR-69 CALVERT CLIFFS UFSAR 14.3-5 Rev. 47 TABLE 14.3-1 INITIAL CONDITIONS AND INPUT PARAMETERS FOR THE BORON DILUTION EVENT PARAMETER Minimum Ratio of Shutdown to Critical Boron Concentration Ratio SDM % Power Operation (Mode 1 and 2) --- --- Hot Standby (Mode 3) 1.05 -3.5 Hot Shutdown (Mode 4), RCP Running 1.04 -3.5 Hot Shutdown (Mode 4), SDC 1.17 -3.5 Hot Shutdown (Mode 5), 3 Charging Pumps 1.16 -3.0 Hot Shutdown (Mode 5), 2 Charging Pumps 1.11 -3.0 Refueling (Mode 6), 3 Charging Pumps 1.28 -6.263 Refueling (Mode 6), 2 Charging Pumps 1.18 -6.263 RCS Volume and Charging Flow Volume ft3 Method Power Operation (Mode 1 and 2) --- --- Hot Standby (Mode 3) 8861 Instant Mix Hot Shutdown (Mode 4), RCP Running 8861 Instant Mix Hot Shutdown (Mode 4), SDC 4513 Dilution Front Hot Shutdown (Mode 5), 3 Charging Pumps 4513 Dilution Front Hot Shutdown (Mode 5), 2 Charging Pumps 3657 Dilution Front Refueling (Mode 6), 3 Charging Pumps 3657 Dilution Front Refueling (Mode 6), 2 Charging Pumps 3657 Dilution Front CALVERT CLIFFS UFSAR 14.3-6 Rev. 47 TABLE 14.3-2 RESULTS OF BORON DILUTION EVENT MODE TIME TO LOSE PRESCRIBED SHUTDOWN MARGIN (MIN) CRITERION FOR MINIMUM TIME TO LOSE PRESCRIBED SHUTDOWN MARGIN (MIN)(a) Hot Standby 16.5 15 Hot Shutdown 16.5 15 Cold Shutdown - three pumps 21.8 15 Cold shutdown - two pumps 20.0 15 Refueling 30.9 30 _______________________ (a) Assumed time between initiation of event and termination of the dilution by the operator.

14.5 LOSS OF LOAD EVENT 14.5.1 IDENTIFICATION OF EVENT AND CAUSE 14.5.2 SEQUENCE OF EVENTS 14.5.3 CORE AND SYSTEM PERFORMANCE 14.

5.4 CONCLUSION

S TABLE 14.5-1 INITIAL CONDITIONS AND INPUT PARAMETERS FOR THE LOSS OF LOAD EVENT TO CALCULATE MAXIMUM RCS PRESSURE PARAMETER UNITS VALUE TABLE 14.5-2 SEQUENCE OF EVENTS FOR LOSS OF LOAD EVENT TO MAXIMIZE CALCULATED RCS PEAK PRESSURE TIME (sec) EVENT SETPOINT OR VALUE TABLE 14.5-3 INITIAL CONDITIONS AND INPUT PARAMETERS FOR THE LOSS OF LOAD EVENT TO CALCULATE MAXIMUM SECONDARY PRESSURE PARAMETER UNITS VALUE TABLE 14.5-4 SEQUENCE OF EVENTS FOR LOSS OF LOAD EVENT TO MAXIMIZE CALCULATED SECONDARY PEAK PRESSURE TIME(sec) EVENT SETPOINT OF VALUE 14.6 LOSS OF FEEDWATER FLOW EVENT 14.6.1 IDENTIFICATION OF EVENT AND CAUSE 14.6.2 SEQUENCE OF EVENTS 14.6.3 CORE AND SYSTEM PERFORMANCE

14.

6.4 CONCLUSION

S TABLE 14.6-1 INITIAL CONDITIONS AND INPUT PARAMETERS FOR THE LOFW EVENT TO MAXIMIZE CALCULATED PEAK PRESSURE PARAMETER UNITS PEAK RCS PRESSURE PEAK SECONDARY PRESSURE TABLE 14.6-2 INITIAL CONDITIONS AND INPUT PARAMETERS FOR THE LOFW EVENT TO MAXIMIZE SG INVENTORY DEPLETION PARAMETER UNITS SETPOINT OR VALUE TABLE 14.6-3 SEQUENCE OF EVENTS FOR LOFW EVENT TO MAXIMIZE CALCULATED PEAK RCS PRESSURE TIME (sec) EVENT SETPOINT OR VALUE TABLE 14.6-4 SEQUENCE OF EVENTS FOR THE LOFW EVENT TO MAXIMIZE CALCULATED PEAK SECONDARY PRESSURE TIME (sec) EVENT SETPOINT OR VALUE TABLE 14.6-5 SEQUENCE OF EVENTS FOR THE LOFW EVENT TO MAXIMIZE STEAM GENERATOR INVENTORY DEPLETION TIME (sec) EVENT SETPOINT OR VALUE CALVERT CLIFFS UFSAR 14.7-1 Rev. 47 14.7 EXCESS FEEDWATER HEAT REMOVAL EVENT 14.

7.1 INTRODUCTION

The condensate and feedwater system is designed to provide a means for transferring the condensate from the condenser hotwells to the SGs (while at the same time raising the temperature and pressure) and providing a means for controlling the quantity of feedwater into the SGs (Section 10.2).

Condensate from the three condenser hotwells is pumped first through the two lowest pressure feedwater heating stages (three heaters per stage), and then through two parallel sets of three low pressure feedwater heaters to the SG feed pumps. The feedwater is then pumped through two parallel feedwater heaters to the SGs. Steam generator level is modulated by the feedwater regulating valves, the feedwater regulating bypass valves, and the associated control systems. An Excess Feedwater Heat Removal event is defined as a reduction in SG feedwater temperature without a corresponding reduction in steam flow from the SGs. This could be caused by the loss of one or more of the feedwater heaters, or due to a feedwater controller malfunction at steady-state power that causes an increase in feedwater flow. Proposed General Design Criterion 6, Reactor Core Design, requires that the reactor core function without exceeding fuel damage limits under all normal operating conditions and plant transients. This transient, the Excess Feedwater Heat Removal event, was analyzed to ensure the DNB and LHR SAFDLs are not exceeded. The computer models and methods used in this analysis are those described in Section 14.1.4, specifically S-RELAP5 and XCOBRA-IIIC. As discussed in the following sections, during the core and system response to the Excess Feedwater Heat Removal event, the SAFDLs are within the required limits and the proposed General Design Criterion is met. 14.7.2 PHYSICAL DESCRIPTION OF EVENT The most limiting Excess Feedwater Heat Removal event is postulated to occur at HFP and is caused by the assumed loss of both high pressure feedwater heaters. This is modeled by a reduction in SG feedwater enthalpy. The immediate system response to this malfunction is a decrease in feedwater temperature to the SGs. The cooler water entering the SGs causes the SG temperature and pressure to slowly decrease, and more heat is extracted from the RCS. In response, the RCS temperature and pressure will decrease and cause pressurizer level to decrease. When there is a negative MTC, a positive reactivity feedback occurs in the core in response to the decreasing core average temperature. This increases core power. The core average heat flux will also increase and partially offset the RCS temperature decrease resulting from the feedwater temperature decrease, and the reactor reaches a new (higher) steady-state power. Although the VHPT is approached, no reactor trip on nuclear instrument power occurs due to the temperature shadowing of the excore detectors. The delta T portion of the VHPT and the TM/LP trip are not credited. The plant remains at the steady-state power until operators manually trip the plant. Table 14.7-2 depicts the sequence of events for the Excess Feedwater Heat Removal event. An increase in feedwater flow rate to 155% of rated full power flow has also been analyzed. However, the results of the increased feedwater flow transient were bounded by the results of the loss of feedwater heater transient. 14.7.3 METHODOLOGY The NSSS response to the Excess Feedwater Heat Removal event was simulated using S-RELAP5. The S-RELAP5 results were subsequently used as input to the XCOBRA-IIIC CALVERT CLIFFS UFSAR 14.7-2 Rev. 47 code to evaluate the DNB response. Fuel centerline melt is bounded by that calculated for an Excess Load event initiated at HFP.

14.7.4 INPUTS AND ASSUMPTIONS Initial Conditions Steam and main feedwater flow are initially assumed equal. The remaining initial plant conditions for the Excess Feedwater Heat Removal event were selected to maximize the NSSS cooldown and the core power increase to ensure the SAFDLs are maintained. Key inputs such as power, Tin, RCS pressure, core mass flow rate, MTC, and the feedwater enthalpy were selected to achieve these conditions (Table 14.7-1).

Concurrent Events/Single Failures There are no concurrent events or single failures assumed in the analysis. Automatic RPS/ESFAS Functions No RPS actuations occurred. No ESFAS equipment is actuated during this event. Other Equipment Safety Functions The pressurizer pressure and level control systems are not credited. Since this is an overcooling event and the RCS/SG pressure upset limits are not approached, the PSVs, PORVs, and MSSVs are not actuated. In addition, the AFW system is not actuated.

Operator Actions The analysis assumed that operator actions mitigate the event (i.e., manually trip the plant) at 1800 seconds in accordance with applicable plant procedures. Status of Non-safety Related Control Systems The steam dump and bypass system is not actuated. 14.7.5 RESULTS Figures 14.7-1 through 14.7-6 present, as a function of time, the transient core power, core average heat flux, RCS temperatures, RCS pressure, SG pressure, and SG temperature. These results support the determination that the DNB and FCM SAFDLs are not exceeded.

Results of all cases show that this event is bounded by the Excess Load event for all criteria. 14.

7.6 CONCLUSION

S The loss of both high pressure feedwater heaters is the most limiting HFP Excess Feedwater Heat Removal event (i.e., results in higher core power and lower RCS temperature and pressure). The analysis demonstrates that the SAFDLs (DNB and FCM) are not exceeded. Since this is an overcooling event, the RCS pressure upset limit is not approached. In addition, since there are no fuel failures, the radiological consequences of the Excess Feedwater Heat Removal event are negligible.

CALVERT CLIFFS UFSAR 14.7-3 Rev. 47 TABLE 14.7-1 INITIAL CONDITIONS AND INPUT PARAMETERS FOR THE EXCESS FEEDWATER HEAT REMOVAL EVENT PARAMETER UNITS HFP VALUE Initial Core Power MWt 2754 Initial Core Inlet Temperature °F 548 Initial RCS Pressure psia 2250 Initial Vessel Flow Rate gpm 370,000 Effective MTC pcm/°F -33 EOC Kinetics, eff --- 0.005237 ASI for MDNBR (Limiting Design Axial Profile) --- -0.3 Doppler Coefficient pcm/°F -1.11 Integrated Radial Peak Factor (Fr) --- 1.65 Maximum Feedwater Temperature Decrease °F 100.0 CALVERT CLIFFS UFSAR 14.7-4 Rev. 47 TABLE 14.7-2 SEQUENCE OF EVENTS FOR THE EXCESS FEEDWATER HEAT REMOVAL EVENT TIME (sec) EVENT SETPOINT OR VALUE0.0 Loss of Both High Pressure Feedwater Heaters --- 160.2 Secondary Pressure Reaches a Minimum Value 815.0 psia 161.4 RCS Pressure Reaches a Minimum Value 2228.5 psia 162.6 Core Power Reaches a Peak Value 3208 MW 117.2% of 2737 MWt 162.6 Minimum DNBR is Reached > MDNBR SAFDL 163.6-169.4 Core Inlet Temperature Reaches a Minimum Value 540.0°F 167.6 Core Average Heat Flux Reaches a Maximum Value 3205.53 MW 117.1% of 2737 MWt 1800 Operator Action Mitigates the Event --- CALVERT CLIFFS UFSAR 14.8-1 Rev. 47 14.8 REACTOR COOLANT SYSTEM DEPRESSURIZATION 14.8.1 IDENTIFICATION OF EVENT AND CAUSE The primary function of the PSVs is to prevent over-pressurization of the RCS. There are two valves in the system which are located on two parallel pipes off the top of the pressurizer. These valves are spring-loaded and have an opening pressure of 2485 and 2550 psig, respectively. To reduce the number of challenges to the PSVs and to prevent over-pressurization at low system temperature, two PORVs are also installed. These valves tee-off the pipes to the PSVs. Both the PSVs and the PORVs discharge to the quench tank. An RCS Depressurization event is defined as a rapid, uncontrolled decrease in RCS pressure other than a loss of coolant (Section 14.17). Inadvertent opening the PSVs or PORVs during steady-state operation would result in an RCS Depressurization event. The most limiting RCS depressurization at HFP is an inadvertent opening of both PORVs. The two PORVs have a larger relieving capacity than one PSV.

If the RCS Depressurization event is not terminated, the event would turn into a small break loss-of-coolant accident (Section 14.17). Therefore, this analysis will only follow the RCS Depressurization event until just after a reactor trip. 14.8.2 SEQUENCE OF EVENTS An RCS Depressurization event can approach the DNBR SAFDLs. The action of the TM/LP Trip will prevent exceeding the DNBR limit. The LHGR SAFDL and RCS Pressure Upset Limit will not be approached as there is essentially no power rise and no pressure increase for this event. Since no fuel pin failures are postulated to occur, the site boundary dose criteria in the 10 CFR 50.67 guidelines will not be approached.

An RCS Depressurization event is postulated to be initiated at HFP by the inadvertent opening of both PORVs. The immediate system response is a rapid depressurization of the RCS. The level in the pressurizer will initially increase as voids form in response to the decrease in pressure. As the pressure continues to decrease, the level in the pressurizer will decrease due to the steam mass leaving the pressurizer. The discharged steam goes to the quench tank where it is condensed and stored.

To compensate for the decreasing pressure, the water in the pressurizer flashes to steam and the PPCS actuates the proportional heaters in an attempt to maintain pressure. As the pressurizer level decreases, the PLCS will reduce RCS letdown flow to a minimum of 29 gpm and actuate the remaining charging pumps. With the pressure continuing to decrease, all backup heaters will be energized to assist in maintaining pressure. For conservatism in the analysis, no credit is allowed for the pressurizer pressure control system. The pressurizer level control system was not modeled. This has little impact on MDNBR and does not significantly impact the depressurization rate. For both Units at this time, RCS temperatures, core power, core average heat flux, and secondary system pressure will be essentially constant.

With a maximum depressurization rate and the pressurizer pressure and control system inoperable, the RCS pressure will rapidly approach the TM/LP Analysis Trip setpoint. Upon reactor trip, the core power and core average heat flux will rapidly decay. The RCS will approach the saturation temperature corresponding to the normal main steam bypass analysis setpoint pressure of 900 psia.

CALVERT CLIFFS UFSAR 14.8-2 Rev. 47 14.8.3 CORE AND SYSTEM PERFORMANCE 14.8.3.1 Mathematical Models The transient response of the RCS and steam systems to the RCS Depressurization event was simulated using the S-RELAP5 thermal-hydraulic system code consistent with the methodology in Reference 1. The XCOBRA-IIIC fuel assembly thermal-hydraulic code was used to calculate the flow and enthalpy distributions for the entire core and the DNB performance for the DNB-limiting assembly as part of the AREVA TM/LP setpoint verification analysis (Reference 2). The limiting assembly DNBR calculations were performed using a NRC-approved DNB correlation. Both of these computer codes are described in Section 14.1.4.1. 14.8.3.2 Input Parameters and Initial Conditions The input parameters and initial conditions used in the analysis are listed in Table 14.8-1. Those parameters which are unique to the analysis are discussed below. For this event, the TM/LP Trip is the primary reactor trip. A single calculation is performed at BOC HFP conditions, maximum Technical Specification core inlet temperature, and minimum Technical Specifications RCS flow rate. This produced the minimum margin to the DNB limit. A conservative moderator density reactivity feedback is used, which is based on the HZP Technical Specification MTC. The event is assumed to be caused by an inadvertent opening of both pressurizer PORVs while operating at RTP. This results in a rapid drop in the RCS pressure and, consequently, a rapid decrease in DNBR. The initial axial power shape and the corresponding SCRAM worth versus insertion used in the analysis is a bottom peaked shape. This power distribution maximizes the time required to terminate the decrease in DNBR following a trip. The pressurizer heaters were assumed inoperable. The charging and letdown system is not modeled. This is due to the event presenting a benign challenge to MDNBR and the event being used to determine the TM/LP pressure bias used in the setpoint verification analysis. The depressurization rate at the time of trip is not significantly impacted by letdown. 14.8.3.3 Results Table 14.8-2 contains the sequence of events for the event at HFP. Figures 14.8-1 through 14.8-4 present the transient core power, core average heat flux, RCS temperatures, and RCS pressure behavior. 14.

8.4 CONCLUSION

The analysis of the RCS Depressurization event demonstrates that the action of the RPS prevents exceeding the fuel SAFDLs. The radiological consequence of opening the atmospheric dump valves upon reactor trip during the most limiting RCS Depressurization event is a site boundary dose which is negligible compared to the 10 CFR 50.67 guidelines. The key transient parameters for this event are independent of burnup, therefore, extended burnup has no impact on this event.

CALVERT CLIFFS UFSAR 14.8-3 Rev. 47 14.

8.5 REFERENCES

1. EMF-2310(P)(A), Revision 1, SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors, May 2004 2. EMF-1961(P)(A), Statistical Setpoints for Combustion Engineering Type Reactors CALVERT CLIFFS UFSAR 14.8-4 Rev. 47 TABLE 14.8-1 INITIAL CONDITIONS AND INPUT PARAMETERS FOR RCS DEPRESSURIZATION EVENT PARAMETER UNITS VALUE Initial Core Power MWt 2754 Initial Core Inlet Temperature °F 548 Initial RCS Pressure psia 2250 Initial Vessel Flow Rate gpm 370,000 TM/LP Trip Delay sec 0.9 Scram Worth pcm 5277.6 MTC pcm/°F +7 Doppler Reactivity Coefficient pcm/°F -0.8 Effective Cross-sectional Area of the PORVs ft2 0.02008 PLCS operating condition Not Modeled PPCS operating condition Heaters Disabled Spray Available SDBC operating condition Not actuated SG Tube Plugging  % per SG 0 CALVERT CLIFFS UFSAR 14.8-5 Rev. 47 TABLE 14.8-2 SEQUENCE OF EVENTS FOR THE RCS DEPRESSURIZATION EVENT TIME (sec) EVENT VALUE 0.0 Inadvertent Opening of PORVs --- 37.45 TM/LP Analysis Trip Setpoint is Reached 1862.8 psia 38.35 TM/LP Trip Breakers Open --- 38.80 Time of Maximum Heat Flux 2819.9 MW 38.85 CEAs Begin to drop into Core --- 53.85 Transient Evaluation Terminated(a) ---

_______________________ (a) Since the event trips on TM/LP for all cases, the DNBR calculation is covered in the TM/LP setpoint verification analysis. No event specific DNBR calculation is required. CALVERT CLIFFS UFSAR 14.9-1 Rev. 47 14.9 LOSS-OF-COOLANT FLOW EVENT 14.9.1 IDENTIFICATION OF EVENT AND CAUSE The primary function of the RCPs is to provide forced coolant flow through the core. There are four RCPs in the RCS which are located in the SG cold legs. The RCS is a two-loop, two SG system with four cold legs.

Electrical power for the RCPs is provided from separate busses which are connected to a service transformer. Both Unit 1 and Unit 2 service transformers receive their power from the main switchyard which is arranged in a ring bus configuration. The power from each unit's main generator goes directly to the main switchyard and then comes back into the service transformers.

In the event that the main turbine-generator should trip, electrical power for the RCPs is provided by offsite sources through the ring bus. If a turbine trip occurs on Unit 2 during a period when offsite power sources are unavailable, the four RCPs would be initially energized by the transient output of the turbine-generator. During this period, the high rotational energy of the turbine-generator during initial stages of coast down is used to provide power to the RCPs. This turbine-generator coast down assist feature provides a slower decrease in RCS flow rate than would occur due to the influence of the RCP flywheels alone. Unit 1 does not have this assisted coast down feature and coasts down on the RCP flywheels. Should a failure occur in the service transformer, the affected pumps will coast down with their own flywheel energy. The turbine trip would result in a reactor trip. However, no credit is allowed for this trip in the analysis.

A Loss-of-Coolant Flow event is defined as a loss of forced reactor coolant through the core with offsite power available, but without a seized RCP rotor (Section 14.16). A loss-of-coolant flow with offsite power unavailable is discussed in Section 14.10.

The most limiting Loss-of-Coolant Flow event is a concurrent loss of power to all four RCPs. A loss of four RCPs is more limiting than a loss of one, two, or three RCPs as the coolant through the core will decrease faster with a loss of all four RCPs. 14.9.2 SEQUENCE OF EVENTS A Loss-of-Coolant Flow event can approach the DNBR and LHGR SAFDLs and the RCS Pressure Upset Limit. The action of the low flow trip in conjunction with the steady-state margin ensured by the LCOs will prevent exceeding these limits. Since no fuel pin failures are postulated to occur, the site boundary dose criteria in 10 CFR 50.67 guidelines will not be approached.

A Loss-of-Coolant Flow event is initiated at HFP and within the LCOs by the concurrent loss of power to all four RCPs. The immediate system response to the coast down of the pumps is a rapid decrease in coolant mass flow rate through the core. In one second, the flow decreases by approximately 11% and in ten seconds, the flow is down to approximately 50% of full flow.

With the coolant mass flow rate decreasing, the core temperatures and the enthalpy rise across the core will start increasing. In the presence of a positive MTC that is normally negative, the increasing core temperatures will result in positive reactivity feedback. The core power will subsequently increase. Due to the fuel time constant and increased enthalpy, the core average heat flux will lag the core power rise.

Depending on the initial core flow measured by the SG differential pressure transmitters, the low flow analysis trip setpoint is reached in approximately one second. After sufficient time for trip signal processing and decay of the CEA holding coils (i.e., 1 second), the CALVERT CLIFFS UFSAR 14.9-2 Rev. 47 CEAs begin to drop into the core and insert negative reactivity. Consequently, the increase in core power is terminated and starts to decrease. Since the core flow is still decreasing and the heat flux has not decreased, the DNBR will continue to decrease at this time. After the CEAs have been sufficiently inserted (depending on the axial power distribution) and taking into account the lagging heat flux, the DNBR transient will terminate within 3 to 5 seconds of the event.

Shortly after the core heat flux starts to decrease, the RCS temperatures and pressurizer pressure will begin to decrease. Since the loop cycle time is longer than the DNBR transient time (i.e., approximately 10 seconds compared to less than 5 seconds), the SG temperature and pressure remain essentially constant during the initial sequence of events. After the SDBS actuates, the system will stabilize to a HZP condition. 14.9.3 CORE AND SYSTEM PERFORMANCE 14.9.3.1 Mathematical Models The transient response of the RCS and steam system to the Loss of Coolant Flow event was simulated using the S-RELAP5 thermal-hydraulic system code consistent with the methodology in Reference 1. The S-RELAP5 code is described in Section 14.1.4.1. The time-dependent coolant flow data was developed based on measured data for a zero plugged tube condition. The data was conservatively adjusted for the effect of SG tube plugging using the COAST computer code. The transient results were subsequently used as input for the evaluation of MDNBR using the XCOBRA-IIIC computer code. The code is described in Section 14.1.4.1. 14.9.3.2 Input Parameters and Initial Conditions The input parameters and initial conditions used in the analysis are listed in Table 14.9-1. Those parameters which are unique to the analysis are discussed below. The MTC and FTC are the only key parameters which are impacted by extended burnup. Since this transient is more limiting at BOC, corresponding MTC and FTC values were assumed in the analysis. Hence, extended burnup has no adverse impact on this event.

The coolant flow coast down calculated by S-RELAP5 was benchmarked to flow coast down data reflecting the effects of 10% SG tube plugging. Reactor trip for the Loss of Coolant Flow event was initiated by low coolant flow rate as determined by a reduction in the sum of the total loop flow. The plant protection system setpoint credited in the analysis was adjusted to account for uncertainties and time delays. The initiation of scram was delayed after the setpoint was reached to account for appropriate instrumentation and other time delays in the safety system including initiation of actual rod movement. The analysis is conducted at HFP BOC conditions, using a limiting axial power distribution. The BOC most-positive MTC Technical Specification value, independent of power level, and the BOC HFP nominal FTC (biased less negative) were used. The minimum transient DNBR is calculated using the most adverse DNBR initial conditions. 14.9.3.3 Results Table 14.9-2 contains the sequence of events for the Loss-of-Coolant Flow event. Figures 14.9-2 through 14.9-5 present the transient core power, core average heat flux, RCS temperatures, and RCS pressure behavior. CALVERT CLIFFS UFSAR 14.9-3 Rev. 47 Core boundary conditions from each case are used for the evaluation of the MDNBR, via the AREVA DNB LCO setpoint verification analysis (Reference 2). The MDNBR is above the NRC-approved DNB correlation upper 95/95 limit plus a 2% mixed core penalty. The results show that the DNBR SAFDL is not exceeded. The peak RCS pressure scenario is bounded by the more limiting Feedline Break event. 14.

9.4 CONCLUSION

The analysis of the Loss-of-Coolant Flow event demonstrates that the action of the RPS in conjunction with the LCOs will prevent exceeding the fuel SAFDLs and RCS Pressure Upset Limits. The radiological consequences of opening of the atmospheric dump valves upon reactor trip is a site boundary dose which is negligible compared to the 10 CFR 50.67 guidelines. Since DNBR design limits are not exceeded and no fuel pins are predicted to fail, extended burnup has no adverse impact during this event. 14.

9.5 REFERENCES

1. EMF-2310(P)(A), Revision 1, SRP Chapter Non-LOCA Methodology for Pressurized Water Reactors, May 2004 2. EMF-1961(P)(A), Statistical Setpoints for Combustion Engineering Type Reactors CALVERT CLIFFS UFSAR 14.9-4 Rev. 47 TABLE 14.9-1 INITIAL CONDITIONS AND INPUT PARAMETERS FOR LOSS-OF-COOLANT FLOW EVENT PARAMETER UNITS VALUE Initial Core Power MWt 2754 Initial Core Inlet Temperature °F 548 Initial RCS Pressure psia 2250 Initial Vessel Flow Rate gpm 370,000 Minimum RCS Flow Trip Setpoint % of 370,000 gpm 90 RCS Flow Trip Response Time sec 0.5 Minimum CEA Worth Available at Trip pcm 5277.6 Doppler Fuel Temperature Coefficient pcm/°F -0.8 Effective MTC pcm/°F +7 4-Pump RCS Flow Coast Down (Includes effects of 10% plugged tubes) --- Figure 14.9-1 CALVERT CLIFFS UFSAR 14.9-5 Rev. 47 TABLE 14.9-2 SEQUENCE OF EVENTS FOR LOSS-OF-COOLANT FLOW EVENT TIME (sec) EVENT VALUE 0.0 Loss of Power to all Four RCPs --- 0.91 Low Flow Trip Analysis Setpoint Reached 90% of 370,000 gpm 1.40 Trip Breakers Open --- 1.89 CEAs Begin to Drop Into Core --- Various(a) Minimum DNBR Occurs > 1.164

_______________________ (a) Less than 5 seconds. Time depends on axial shape.

CALVERT CLIFFS UFSAR 14.11-1 Rev. 47 14.11 CONTROL ELEMENT ASSEMBLY DROP EVENT 14.11.1 IDENTIFICATION OF EVENT AND CAUSE The primary function of the CEA is to control the core axial power distribution and to provide instantaneous reactivity to shut down the reactor during controlled procedures and during abnormal and emergency conditions. Under normal operating conditions (i.e., controlled procedures), CEAs are inserted and withdrawn in a pre-programmed group sequence according to the PDIL curve in the Technical Specifications. There are 3 shutdown groups and 5 regulating groups for a total of 77 CEAs. Presently there are no PLCEAs (Chapter 3) in the core. The shutdown groups are dual CEAs (i.e., two CEAs with one extension shaft) and the regulating groups are single CEAs. The CEAs are withdrawn, inserted, and held by the CEDMs which are located on top of the reactor vessel. The CEDM is a magnetic jack-type drive system which operates at a constant speed. A CEA is released from the CEDM holding coil grippers by removing the power to the holding coil. After approximately half a second, the CEDM holding coil magnetic flux will decay and the weight of the CEA and the CEA extension shaft will cause the CEA to drop into the core.

A CEA Drop event is defined as an uncontrolled insertion of a CEA. A loss of power to the CEDM holding coil or a mechanical fault in the CEDM magnetic jack drive will result in a CEA Drop event. The most limiting CEA Drop event is an uncontrolled CEA insertion at HFP. Operation at HFP is most limiting as the reactor is then operating closest to the fuel SAFDLs. Appendix C of the Technical Specifications prohibits changing Core Operating Limits Report Figures 3.1.6, 3.2.3, and 3.2.5 until an NRC accepted generic, or Calvert Cliffs specific, basis is developed for analyzing this event at full power conditions only.

The CEA Drop event represents the limiting event with respect to challenging the LHGR FCM safety limit. The LHGR FCM safety limit that is imposed to compensate for the RODEX2 methodology, which does not explicitly model degraded fuel thermal conductivity, is 21 kW/ft (Reference 3). 14.11.2 SEQUENCE OF EVENTS A CEA Drop event can approach the DNBR and LHGR SAFDLs. The steady-state margin ensured by the LCOs will prevent exceeding these limits. The RCS Pressure Upset Limit is not approached as the system cools down during the event. Since no fuel pin failures are postulated to occur, the site boundary dose criteria in 10 CFR 50.67 guidelines will not be approached.

A CEA Drop event is initiated at HFP from within the LCOs by a failure of the CEDM holding coil. The immediate system response caused by the inserted negative reactivity worth is a reduction in the local core power in the vicinity of the dropped CEA. The local heat flux will be suppressed and the local moderator temperature will decrease. The azimuthal core power tilt will increase due to shutting down of part of the core. The core power and core average heat flux will correspondingly decrease. The amount of decrease is dependent on the dropped CEA worth. The reduction in core average temperature, in conjunction with a negative moderator temperature (normally negative), will result in positive moderator feedback. The resulting positive reactivity addition from the moderator feedback partially compensates for Doppler and the dropped CEA negative reactivity. CALVERT CLIFFS UFSAR 14.11-2 Rev. 47 The RCS pressure will decrease in response to the reduction in the core average temperatures. The analysis assumes the pressurizer pressure and level control systems, which would mitigate the pressure decrease, are inoperable. A decreased RCS pressure results in a lower minimum DNBR. The decrease in core outlet temperature will cause the SG temperature and pressure to decrease. In the analysis, the turbine load is assumed to remain at full power. To maintain the same load, the turbine control valve is assumed to open further to compensate for the reduction in SG pressure. The result is an additional decrease in core inlet temperature and positive moderator feedback. Normal operating procedures maintain the turbine control valve at a set valve position which would prevent any further decrease in core inlet temperature. The cooldown of the RCS continues until the power mismatch between the RCS and the power demand is eliminated. The core average temperature continues to decrease until sufficient positive moderator reactivity feedback offsets the Doppler and the dropped CEA negative reactivity and returns the core power to its pre-drop level. Consequently the power mismatch will be eliminated and no further cooldown of the RCS will occur. The RCS coolant temperatures will reach a new equilibrium value that is slightly lower than the initial values. During the return to the initial core average power level and with part of the core power suppressed, the local power peaks will increase to make up the power difference. The local peaks that will occur are dependent upon the worth and position of the dropped CEA. After detection of CEA drop/misalignment, the operator will initiate a power reduction as required by Technical Specifications. 14.11.3 CORE AND SYSTEM PERFORMANCE 14.11.3.1 Mathematical Models The transient response of the reactor coolant and steam systems to the CEA Drop event was simulated using the S-RELAP5 thermal-hydraulic system code consistent with the methodology in Reference 1. The S-RELAP5 results were subsequently used as input for the evaluation of MDNBR and FCM. S-RELAP5 is described in Section 14.1.4.1. 14.11.3.2 Input Parameters and Initial Conditions The input parameters and initial conditions used in the analysis are listed in Table 14.11-1. Those parameters that are unique to the analysis are discussed below. The analysis assumes the most negative MTC and FTC of reactivity (including uncertainties), because these coefficients produce the minimum RCS coolant temperature decrease upon return to 100% power level and lead to a minimum DNBR.

Charging pumps and pressurizer heaters are assumed to be inoperable during the transient. This maximizes the pressure drop during the event. All other systems are assumed to be in manual mode of operation and have no impact on this event. The analysis uses the maximum radial peaking distortion factors which, for conservatism, are the ratio of the post-drop to the pre-drop Fr. Calculations were performed to cover a range of dropped CEA worths from 10 pcm to 200 pcm. The CALVERT CLIFFS UFSAR 14.11-3 Rev. 47 dropped CEA event is usually terminated by a TM/LP trip, a VHPT, or may potentially reach a new equilibrium state without trip. As seen from the above discussion, the MTC and FTC are the only key parameters which are impacted by extended burnup. The analysis conservatively assumed an EOC MTC value of -3.3x10-4 /°F. In addition, the analysis assumed an EOC FTC value with an uncertainty and bias. Hence, the effects of extended burnup have been explicitly and conservatively included in the analysis. The event was initiated by dropping a full-length CEA over a period of 3.0 seconds. The maximum increases in radial peaking factors in either rodded or unrodded planes were used in all axial regions of the core once the power returns to the initial level. The axial power shape in the hot channel is assumed to remain unchanged, therefore, the increase in the three-dimensional peak is proportional to the maximum increase in radial peaking factor. Since there is no trip assumed, and the secondary side continues to demand 100% power, the peaks will stabilize at these asymptotic values after a few minutes. 14.11.3.3 Results Table 14.11-2 contains the sequence of events for the CEA Drop event at HFP. Figures 14.11-1 through 14.11-4 present the transient behavior of the core power, core average heat flux, RCS temperatures, and RCS pressure as a function of time for the 200 pcm case. Core boundary conditions from each case are used for the evaluation of the MDNBR, via the AREVA DNB LCO setpoint verification analysis (Reference 2) and FCM. The resultant peak LHGR is less than the more limiting FCM LHGR limit provided by either Reference 3, or by cycle specific analysis, and the MDNBR is above the NRC-approved DNB correlation upper 95/95 limit plus a 2% mixed core penalty. 14.

11.4 CONCLUSION

The analysis of the CEA Drop event demonstrates that operating within the LCOs will prevent exceeding the fuel SAFDLs, maintain the integrity of the RCS, and ensure negligible radiological release to the site boundary compared to 10 CFR 50.67 guidelines. 14.

11.5 REFERENCES

1. EMF-2310(P)(A), Revision 1, SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors, May 2004 2. EMF-1961(P)(A), Statistical Setpoints for Combustion Engineering Type Reactors 3. Letter from Mr. D. V. Pickett (NRC) to Mr. G. H. Gellrich (CCNPP), dated February 18, 2011, Calvert Cliffs Nuclear Power Plant, Units Nos. 1 and 2 - Amendment RE: Transition from Westinghouse Nuclear Fuel to AREVA Nuclear Fuel CALVERT CLIFFS UFSAR 14.11-4 Rev. 47 TABLE 14.11-1 INITIAL CONDITIONS AND INPUT PARAMETERS FOR CEA DROP EVENT PARAMETER UNITS VALUE Initial Core Power MWt 2754 Initial Core Inlet Temperature °F 548 Initial RCS Pressure psia 2250 Initial Vessel Flow Rate gpm 370,000 Effective MTC pcm/°F -33 Excore Detector Decalibration Factor %/°F 0.70 Axial Power Distribution ASI -0.20 to +0.20 Maximum Fz --- 1.485 Distortion Factor (Full Power) Fr post/ Fr pre 1.13 CALVERT CLIFFS UFSAR 14.11-5 Rev. 47 TABLE 14.11-2 SEQUENCE OF EVENTS FOR THE CEA DROP EVENT TIME (sec) EVENT SETPOINT OR VALUE 0.0 Rod Drop Initiated --- 0.0 Core Heat Flux Reaches Minimum --- 3.0 CEA Fully Dropped -200 pcm 3.0 Core Power Reaches Minimum 78.02% of RTP 300.0 MDNBR Boundary Conditions: Core Heat Flux Reaches Final Value 100.76% of RTP Core Inlet Temperature 542.14 RCS Pressure 2214.1 MDNBR(a) > 1.164

_______________________ (a) Results shown for the maximum dropped CEA worth. Core boundary condition from each dropped CEA case are used for the evaluation of MDNBR, via the setpoint verification analysis.

CALVERT CLIFFS UFSAR 14.12-1 Rev. 47 14.12 ASYMMETRIC STEAM GENERATOR EVENT 14.12.1 IDENTIFICATION OF EVENT AND CAUSE The primary function of the SGs is to remove heat from the RCS. Any perturbation within the SGs will affect the RCS due to the close coupling of both systems. An Asymmetric SG event is defined as any initiator that affects only one of the two SGs. A loss of load, an excess load, a LOFW, or an excess feedwater to only one SG would result in an Asymmetric SG event. Asymmetric SG events, which are the result of a malfunction of one SG, cause a non-uniform core inlet temperature distribution. The non-uniform core inlet temperature distribution in conjunction with the moderator temperature reactivity feedback produces asymmetric local power peaking in the core. The most limiting Asymmetric SG event is a loss of load to one SG. An asymmetric loss of load would produce the largest core inlet temperature differential across the core. Based on a negative MTC, the RCS temperature tilt across the core will cause an increase in local core power peaking and an approach to the fuel SAFDLs.

14.12.2 SEQUENCE OF EVENTS An Asymmetric SG event can approach the DNBR and LHGR SAFDLs. The action of the Asymmetric Steam Generator Protection Trip (ASGPT), and the Low SG Pressure, Low SG Level, TM/LP, and HPTs will prevent exceeding these limits. The primary trip for the most adverse Asymmetric SG event is the ASGPT. The RCS Pressure Upset Limits will not be approached during the event. Since no fuel pin failures are postulated to occur, the site boundary dose criteria in 10 CFR 50.67 guidelines will not be approached. 14.12.2.1 Asymmetric Excess Feedwater An Asymmetric Excess Feedwater event is initiated at HFP from within the LCOs by a malfunction in one of the feedwater controllers, which instantaneously fully opens the feedwater regulator valve to one SG. The full opening of the feedwater regulator valve causes additional subcooled feedwater to enter the SG which lowers the temperature and pressure. The result is a reduction in the steam flow from the affected SG. The excess feedwater also causes the affected SG cold leg temperature to decrease because additional heat is being extracted. The analysis assumes the turbine demand remains constant, which causes the unaffected SG to pick up part of the load by further opening the turbine control valve. Present operating practices maintain the turbine control valve flow area constant, thus avoiding the increased demand. The increased steaming rate results in lowering the temperature of the SG and therefore the cold leg temperatures. The result of the asymmetric decrease in the core inlet temperature is a temperature and power tilt across the core. Since the increased feedwater flow rate only decreases the temperature slightly, there will be a small increase in radial peaks and core power. The event will be terminated by the ASGPT. This event is less limiting than a loss of load to one SG (Section 14.12.2.4) because it produces a smaller temperature tilt across the core.

14.12.2.2 Asymmetric Loss of Feedwater An Asymmetric LOFW event is initiated at HFP from within the LCOs by a malfunction in one of the feedwater controllers which instantaneously shuts the CALVERT CLIFFS UFSAR 14.12-2 Rev. 47 feedwater regulator valve to one SG. The closure of the feedwater regulator valve causes a LOFW to the SG. The LOFW will cause the temperature and pressure to increase in response to the decreasing SG level. The temperature and pressure in the unaffected SG (i.e., with feedwater flow available) also increases in response to the increased turbine header pressure. The core inlet temperature from both SGs will increase with the decreased secondary heat transfer. A slight core inlet temperature asymmetry occurs with the higher inlet temperature resulting from the affected SG. The small core inlet temperature tilt will not cause a significant radial power tilt. The slight increase in core temperatures in conjunction with a negative MTC will result in a decrease in core average power. The event will be terminated by the Asymmetric SG Pressure Trip or a Low SG Level Trip. This event is less limiting than a loss of load to one SG (Section 14.12.2.4) because it produces a smaller temperature tilt across the core. 14.12.2.3 Asymmetric Excess Load An Asymmetric Excess Load event is initiated at HFP from within the LCOs by the inadvertent opening of a single secondary safety valve on one SG. The excess load on a single SG causes its pressure and temperature to decrease which results in a decrease in the core inlet temperature. Since the temperature from only one SG decreases, a core inlet temperature distribution tilt occurs across the core. In the presence of a negative MTC, positive moderator reactivity feedback occurs that increases the core power. A new steady-state condition is obtained once the core power increases to match the excess load demand. The event will be terminated by the Asymmetric SG Pressure Trip or Low SG Level Trip. This event is less limiting than a loss of load to one SG (Section 14.12.2.4) because it produces a smaller temperature tilt across the core.

14.12.2.4 Asymmetric Loss of Load An Asymmetric Loss of Load event is initiated at HFP from within the LCOs by an inadvertent closure of a single MSIV on one SG. The loss of load to a single SG causes the pressure and temperature on the SG to increase. With the decrease in SG heat transfer, the core inlet temperature from the isolated SG will increase. The isolated SG water level drops rapidly as the increasing pressure collapses the steam bubble in the liquid inventory. The pressure will continue to increase until the MSSVs open. The analysis assumes the turbine load demand remains constant, which causes the turbine control valves to open further. The increased load demand will decrease the other (i.e., unaffected) SG pressure and temperature. In response to the decreased temperature, the core inlet temperature from the SG will also decrease. Present operating practice maintains the turbine control valve flow area constant, which will lessen the severity of the event.

The result of the outlet temperature increase and decrease from their respective SGs is a severe core inlet temperature maldistribution. In the presence of negative MTC and FTC (normally negative at power), the coolant temperature tilt will cause a radial power shift toward the cold side of the core. The power in the outermost fuel bundles, where there is almost no mixing of the inlet flow, will experience the greatest local power increase. The power on the hot side of the core will decrease due to the negative moderator reactivity feedback. CALVERT CLIFFS UFSAR 14.12-3 Rev. 47 The ASGPT will initiate a reactor trip to terminate the event when the absolute differential SG pressure (i.e., Psg1-Psg2) exceeds a preselected analysis setpoint value. 14.12.3 CORE AND SYSTEM PERFORMANCE 14.12.3.1 Mathematical Models The transient response of the RCS and steam systems to the Asymmetric Steam Generator Loss of Load event was simulated using the S-RELAP5 thermal-hydraulic system code consistent with the methodology in Reference 1. The event is analyzed with an S-RELAP5 model which captures the asymmetric core inlet temperature distribution and applies local peaking augmentation factors (Reference 2).

The XCOBRA-IIIC fuel assembly thermal-hydraulic code was used to calculate the flow and enthalpy distributions for the entire core and the DNB performance for the DNB-limiting assembly. The limiting assembly DNBR calculations were performed using an approved AREVA DNB correlation. The overall core conditions calculated by S-RELAP5 during the transient were used as the input to the XCOBRA-IIIC calculation. The limiting design axial power profile (a top peaked axial power distribution) was used for this simulation for conservatism. Both of these computer codes are described in Section 14.1.4.1.

14.12.3.2 Input Parameters and Initial Conditions The input parameters and initial conditions used in the analysis of the Asymmetric Loss of Load event are listed in Table 14.12-1 and Figure 14.12-1. Those parameters that are unique to the analyses are discussed below. The analysis used the radial peaking distortion factor as a function of core inlet temperature for calculating the minimum DNBR and peak linear heat generation rate (PLHGR). To increase the power tilt, a negative MTC was assumed. During a severe asymmetric transient, the cooler core inlet temperature and the temperature tilt decalibrate the neutron power and the DT power, respectively.

Asymmetric tube plugging is bounded by assuming no tube plugging in both generators and minimum MSSV setpoints. No tube plugging increases the initial pressures in both generators and causes earlier opening of the affected SGs MSSVs, thereby delaying actuation of the ASGPT. The MTC is the only key parameter which is adversely impacted by extended burnup. The analysis assumed an EOC MTC value. Hence, the effects of extended burnup have been explicitly and conservatively included in the analysis. 14.12.3.3 Results Table 14.12-2 contains the sequence of events for the Asymmetric Loss of Load event at HFP. Figures 14.12-2 through 14.12-6 present the transient behavior of the core power, core average heat flux, RCS temperatures, pressurizer pressure, and SG pressure. The Asymmetric Loss of Load event at HFP conditions peak power result combined with the Technical Specification peaking factor, inlet temperature augmentation and uncertainties, and accounting for control rod position, axial peak, and engineering uncertainty, results in a conservative PLHGR that is below the cycle specific limit and is below the LHGR limit imposed in Reference 2. CALVERT CLIFFS UFSAR 14.12-4 Rev. 47 The S-RELAP5 plant simulation results from the analysis of the Asymmetric SG Loss of Load event were used as input into the minimum DNBR calculations. The plant simulations were adjusted to account for power, temperature, pressure, and flow measurement uncertainties in the minimum DNBR calculations. The MDNBR was above the NRC-approved DNB correlation upper 95/95 limit plus a 2% mixed core penalty. In addition, adequate FCM margin exists for this event. 14.

12.4 CONCLUSION

The analysis of the Asymmetric SG event demonstrates that operating within the LCOs in conjunction with the LSSS will prevent exceeding the fuel SAFDLs, and maintain the integrity of the RCS. The radiological consequence of opening the atmospheric dump valve upon reactor trip is a site boundary dose that is negligible compared to the 10 CFR 50.67 guidelines.

14.

12.5 REFERENCES

1. EMF-2310(P)(A), Revision 1, May 2004, SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors 2. Letter from Mr. D. V. Pickett (NRC) to Mr. G. H. Gellrich (CCNPP), dated February 18, 2011, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 - Amendment Re: Transition from Westinghouse Nuclear Fuel to AREVA Nuclear Fuel (TAC Nos. ME2831 and ME2832)

CALVERT CLIFFS UFSAR 14.12-5 Rev. 47 TABLE 14.12-1 INITIAL CONDITIONS AND INPUT PARAMETERS FOR THE LOSS OF LOAD TO ONE STEAM GENERATOR PARAMETER UNITS VALUE Initial Core Power MWt 2754 Initial Core Inlet Temperature °F 548 Initial RCS Pressure psia 2250 Effective MTC pcm/°F -33 Fuel Doppler Temperature Feedback --- EOC Fuel Temperature Dependent ASGPT Setpoint psid 186 Minimum CEA Worth Available at Trip pcm 5740.8 SG Tube Plugging % 0 Asymmetry in SG Tube Plugging Assumed % difference between SGs 0 Initial Vessel Flow Rate gpm 370,000 CALVERT CLIFFS UFSAR 14.12-6 Rev. 47 TABLE 14.12-2 SEQUENCE OF EVENTS FOR THE LOSS OF LOAD TO ONE STEAM GENERATOR TIME (sec) EVENT VALUE 0.0 Spurious closure of MSIV on SG-1 --- 0.0 Steam flow from unaffected SG increases to maintain turbine power --- 6.04 ASGPT Analysis Setpoint reached (differential pressure) 186.0 psid 6.94 Trip breakers open --- 7.44 CEAs begin to insert --- Varies(a) Minimum DNBR occurs > 1.164 _______________________ (a) Near time of CEA insertion. Time depends on core conditions.

CALVERT CLIFFS UFSAR 14.19-1 Rev. 47 14.19 TURBINE-GENERATOR OVERSPEED INCIDENT This is an analyzed event. For a discussion of this material, see Section 5.3.1.2.

CALVERT CLIFFS UFSAR 14.21-1 Rev. 47 14.21 DELETED Hydrogen Accumulation in Containment, was deleted per License Amendment Nos. 262/239.

14.26 FEEDLINE BREAK EVENT 14.26.1 IDENTIFICATION OF EVENT AND CAUSE 14.26.2 SEQUENCE OF EVENTS 14.26.3 CORE AND SYSTEM PERFORMANCE

14.

26.4 CONCLUSION

14.

26.5 REFERENCES

TABLE 14.26-1 INITIAL CONDITIONS AND INPUT PARAMETERS ASSUMED IN THE FEEDWATER LINE BREAK EVENT PARAMETER UNITS VALUE(a) _______________________ TABLE 14.26-2 ASSUMPTIONS FOR THE RADIOLOGICAL EVALUATION FOR THE FEEDLINE BREAK EVENT PARAMETER UNITS VALUE TABLE 14.26-3 SEQUENCE OF EVENTS FOR FEEDWATER LINE BREAK WITH LOAC FOLLOWING REACTOR TRIP TIME (sec) EVENT SETPOINT OR VALUE

Calvert Cliffs Nuclear Power Plant LOSS OF LOAD EVENT CORE POWER VS TIME Figure 14.5-1 Revision 49 Calvert Cliffs Nuclear Power Plant LOSS OF LOAD EVENT CORE AVERAGE HEAT FLUX VS TIME Figure 14.5-2 Revision 49 Calvert Cliffs Nuclear Power Plant LOSS OF LOAD EVENT RCS PRESSURE VS TIME Figure 14.5-3 Revision 49 Calvert Cliffs Nuclear Power Plant LOSS OF LOAD EVENT RCS TEMPERATURES VS TIME Figure 14.5-4 Revision 49 Calvert Cliffs Nuclear Power Plant LOSS OF LOAD EVENT STEAM GENERATOR PRESSURE VS TIME Figure 14.5-5 Revision 49 Calvert Cliffs Nuclear Power Plant LOSS OF LOAD EVENT PRESSURIZER WATER VOLUME VS TIME Figure 14.5-6 Revision 49 Calvert Cliffs Nuclear Power Plant LOSS OF FEEDWATER FLOW EVENT MAXIMUM RCS PEAK PRESSURE CORE POWER VS TIME Figure 14.6-1 Revision 49 Calvert Cliffs Nuclear Power Plant LOSS OF FEEDWATER FLOW EVENT MAXIMUM RCS PEAK PRESSURE RCS TEMPERATURES VS TIME Figure 14.6-2 Revision 49 Calvert Cliffs Nuclear Power Plant LOSS OF FEEDWATER FLOW EVENT MAXIMUM RCS PEAK PRESSURE RCS PRESSURE VS TIME Figure 14.6-3 Revision 49 Calvert Cliffs Nuclear Power Plant LOSS OF FEEDWATER FLOW EVENT MAXIMUM RCS PEAK PRESSURE STEAM GENERATOR PRESSURE VS TIME Figure 14.6-4 Revision 49 Calvert Cliffs Nuclear Power Plant LOSS OF FEEDWATER FLOW EVENT MAXIMUM SECONDARY PEAK PRESSURE CORE POWER VS TIME Figure 14.6-5 Revision 49 Calvert Cliffs Nuclear Power Plant LOSS OF FEEDWATER FLOW EVENT MAXIMUM SECONDARY PEAK PRESSURE RCS TEMPERATURES VS TIME Figure 14.6-6 Revision 49 Calvert Cliffs Nuclear Power Plant LOSS OF FEEDWATER FLOW EVENT MAXIMUM SECONDARY PEAK PRESSURE PRESSURIZER PRESSURE VS TIME Figure 14.6-7 Revision 49 Calvert Cliffs Nuclear Power Plant LOSS OF FEEDWATER FLOW EVENT MAXIMUM SECONDARY PEAK PRESSURE STEAM GENERATOR PRESSURE VS TIME Figure 14.6-8 Revision 49 Calvert Cliffs Nuclear Power Plant LOSS OF FEEDWATER FLOW EVENT MAXIMUM STEAM GENERATOR INVENTORY DEPLETION CORE POWER VS TIME Figure 14.6-9 Revision 49 Calvert Cliffs Nuclear Power Plant LOSS OF FEEDWATER FLOW EVENT MAXIMUM STEAM GENERATOR INVENTORY DEPLETION RCS TEMPERATURES VS TIME Figure 14.6-10 Revision 49 Calvert Cliffs Nuclear Power Plant LOSS OF FEEDWATER FLOW EVENT MAXIMUM STEAM GENERATOR INVENTORY DEPLETION PRESSURIZER PRESSURE VS TIME Figure 14.6-11 Revision 49 Calvert Cliffs Nuclear Power Plant LOSS OF FEEDWATER FLOW EVENT MAXIMUM STEAM GENERATOR INVENTORY DEPLETION STEAM GENERATOR PRESSURE VS TIME Figure 14.6-12 Revision 49 Calvert Cliffs Nuclear Power Plant LOSS OF FEEDWATER FLOW EVENT MAXIMUM STEAM GENERATOR INVENTORY DEPLETION STEAM GENERATOR INVENTORY VS TIME Figure 14.6-13 Revision 49

Calvert Cliffs Nuclear Power Plant FEEDLINE BREAK EVENT WITH LOAC FOLLOWING REACTOR TRIP RCS PEAK PRESSURE VS BREAK SIZE Figure 14.26-1 Revision 49 Calvert Cliffs Nuclear Power Plant FEEDLINE BREAK EVENT WITH LOAC FOLLOWING REACTOR TRIP CORE POWER VS TIME Figure 14.26-2 Revision 49 Calvert Cliffs Nuclear Power Plant FEEDLINE BREAK EVENT WITH LOAC FOLLOWING REACTOR TRIP CORE AVERAGE HEAT FLUX VS TIME Figure 14.26-3 Revision 49 Calvert Cliffs Nuclear Power Plant FEEDLINE BREAK EVENT WITH LOAC FOLLOWING REACTOR TRIP RCS TEMPERATURES VS TIME Figure 14.26-4 Revision 49 Calvert Cliffs Nuclear Power Plant FEEDLINE BREAK EVENT WITH LOAC FOLLOWING REACTOR TRIP RCS PRESSURE VS TIME Figure 14.26-5 Revision 49 Calvert Cliffs Nuclear Power Plant FEEDLINE BREAK EVENT WITH NO LOAC FOLLOWING REACTOR TRIP STEAM GENERATOR PRESSURE VS TIME Figure 14.26-6 Revision 49 Calvert Cliffs Nuclear Power Plant FEEDLINE BREAK EVENT WITH NO LOAC FOLLOWING REACTOR TRIP STEAM GENERATOR INVENTORY VS TIME Figure 14.26-7 Revision 49 Calvert Cliffs Nuclear Power Plant FEEDLINE BREAK EVENT WITH NO LOAC FOLLOWING REACTOR TRIP AUXILIARY FEEDWATER FLOW VS TIME Figure 14.26-8 Revision 49 Calvert Cliffs Nuclear Power Plant FEEDLINE BREAK EVENT WITH NO LOAC FOLLOWING REACTOR TRIP INTEGRATED BREAK FLOW VS TIME Figure 14.26-9 Revision 49 020406080100120140160050010001500200025003000350040004500INTEGRATED BREAK FLOW (thousands 1bm)TIME (seconds) Calvert Cliffs Nuclear Power Plant FEEDLINE BREAK EVENT WITH NO LOAC FOLLOWING REACTOR TRIP BREAK FLOW VS TIME Figure 14.26-10 Revision 49 CHAPTER 15 TECHNICAL REQUIREMENTS MANUAL TABLE OF CONTENTS PAGE15.0 TECHNICAL REQUIREMENTS MANUAL CALVERT CLIFFS UFSAR 15.0-1 Rev. 47 TECHNICAL REQUIREMENTS MANUAL 15.0The Technical Requirements Manual consists of material that was removed from the Technical Specifications on conversion to Improved Standard Technical Specifications. The material removed was selected because it did not meet any of the four criteria the Nuclear Regulatory Commission has established for material that is in Technical Specifications. These four criteria, which set a level of safety significance for Technical Specification content, are detailed in 60 FR 36953. The four criteria are summarized in two general classes: (1) those related to the prevention of accidents, and (2) those related to mitigation of the consequences of accidents. The Technical Requirements Manual is a stand-alone, licensee-controlled document, changes to which are controlled by the 10 CFR 50.59 process. It describes the normal operating condition for the systems and components listed below, and specifies actions that would be taken when these system and components are not in their normal conditions. These actions are consistent with the guidance provided in Nuclear Regulatory Commission Generic Letter 91-18 for degraded and nonconforming conditions and will the requirements of 10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance," regarding removal of equipment from service for maintenance or testing.

The following systems, components, and process limits are addressed in the Technical Requirements Manual: 1. Boron dilution and flow paths 2. Control element assembly position indication

3. Radiation monitoring instrumentation

4. Meteorological instrumentation

5. Incore Detector System 6. Seismic monitoring instrumentation 7. Fire detection instrumentation

8. Reactor Coolant system chemistry

9. Pressurizer pressure/temperature limits

10. American Society of Mechanical Engineers Code components 11. Deleted 12. Letdown line excess flow

13. Reactor Coolant System vents

14. Containment structural integrity

15. Steam generator pressure/temperature limits

16. Snubbers 17. Sealed source contamination 18. Watertight doors

19. Fire suppression water system

20. Spray and sprinkler system

21. Halon system 22. Fire hose stations 23. Yard fire hydrants and hydrant hose houses

24. Fire barrier penetrations

25. Fuel decay time

26. Refueling communications

27. Refueling machine 28. Spent fuel pool crane travel 29. Explosive gas mixtures

30. Gas storage tanks

31. Fire detection instruments

32. Snubber visual inspection interval 33. Sprinkler locations 34. Fire hose stations CHAPTER 16 LICENSE RENEWAL TABLE OF CONTENTS PAGE16.0 LICENSE RENEWAL

16.1 INTRODUCTION

16.2 AGING MANAGEMENT PROGRAMS AND ACTIVITIES 16.2.1 ADDITIONAL BASELINE WALKDOWNS AGING MANAGEMENT PROGRAM 16.2.2 AGE-RELATED DEGRADATION INSPECTION (ARDI) AGING MANAGEMENT PROGRAM 16.2.3 ALLOY 600 AGING MANAGEMENT PROGRAM 16.2.4 ASME SECTION XI IN-SERVICE INSPECTION (ISI), SUBSECTIONS IWB, IWC, AND IWD AGING MANAGEMENT PROGRAM 16.2.5 ASME SECTION XI IN-SERVICE INSPECTION (ISI), SUBSECTIONS IWE AND IWI AGING MANAGEMENT PROGRAM 16.2.6 ASME SECTION XI IN-SERVICE INSPECTION, SUBSECTION IWF AGING MANAGEMENT PROGRAM 16.2.7 BORIC ACID CORROSION INSPECTION (BACI) PROGRAM 16.2.8 BURIED PIPING INSPECTION AGING MANAGEMENT PROGRAM 16.2.9 CABLE AGING MANAGEMENT PROGRAM 16.2.10 CAST AUSTENITIC STAINLESS STEEL (CASS) AGING MANAGEMENT PROGRAM 16.2.11 CAULKING AND SEALANTS INSPECTION AGING MANAGEMENT PROGRAM 16.2.12 CHEMISTRY (WATER) AGING MANAGEMENT PROGRAM 16.2.13 COMPREHENSIVE REACTOR VESSEL SURVEILLANCE AGING MANAGEMENT PROGRAM 16.2.14 CONTAINMENT LOCAL LEAKAGE RATE TESTING AGING MANAGEMENT PROGRAM 16.2.15 DESIGN CHANGE AND MODIFICATION IMPLEMENTATION AGING MANAGEMENT PROGRAM 16.2.16 DIESEL FUEL OIL (TANKS AND CHEMISTRY) AGING MANAGEMENT PROGRAM 16.2.17 ENVIRONMENTAL QUALIFICATION (EQ) AGING MANAGEMENT PROGRAM 16.2.18 FLOW ACCELERATED CORROSION AGING MANAGEMENT PROGRAM 16.2.19 FATIGUE MONITORING AGING MANAGEMENT PROGRAM 16.2.20 FIRE BARRIER PENETRATION SEAL INSPECTION AGING MANAGEMENT PROGRAM 16.2.21 FIRE PROTECTION AGING MANAGEMENT PROGRAM CHAPTER 16 LICENSE RENEWAL TABLE OF CONTENTS PAGE16.2.22 LOAD HANDLING AND FUEL HANDLING EQUIPMENT AGING MANAGEMENT PROGRAM 16.2.23 PREVENTIVE MAINTENANCE (PM) AGING MANAGEMENT PROGRAM 16.2.24 PROTECTIVE COATINGS AGING MANAGEMENT PROGRAM 16.2.25 REACTOR COOLANT AGING MANAGEMENT PROGRAM 16.2.26 REACTOR VESSEL INTERNALS (RVI) AGING MANAGEMENT PROGRAM 16.2.27 SPENT FUEL POOL (SFP) AGING MANAGEMENT PROGRAM 16.2.28 STEAM GENERATOR (SG) AGING MANAGEMENT PROGRAM 16.2.29 STRUCTURE AND SYSTEM WALKDOWNS AGING MANAGEMENT PROGRAM 16.2.30 SURVEILLANCE TESTING AGING MANAGEMENT PROGRAM 16.3 EVALUATION OF TIME-LIMITED AGING ANALYSES 16.3.1 CONTAINMENT TENDON PRESTRESS LOSS 16.3.2 POISON SHEETS IN SPENT FUEL POOL 16.3.3 10 CFR 50.49 ENVIRONMENTAL QUALIFICATION PROGRAM 16.3.4 REACTOR PRESSURE VESSEL TOUGHNESS REQUIREMENTS 16.3.5 REACTOR VESSEL INTERNALS 16.3.6 CONTAINMENT LINER FATIGUE 16.3.7 MAIN STEAM PIPING FATIGUE 16.3.8 NUCLEAR STEAM SUPPLY FATIGUE 16.3.9 CLASS 2 AND 3 PIPING COMPONENTS (OTHER THAN MAIN STEAM PIPING) FATIGUE 16.4 10 CFR 54.37(b) UPDATE

16.5 REFERENCES

CHAPTER 16 LICENSE RENEWAL LIST OF TABLES TITLEPAGE CHAPTER 16 LICENSE RENEWAL LIST OF ACRONYMS CHAPTER 16 LICENSE RENEWAL LIST OF ACRONYMS LICENSE RENEWAL

16.1 INTRODUCTION

16.2 AGING MANAGEMENT PROGRAMS AND ACTIVITIES 16.2.1 ADDITIONAL BASELINE WALKDOWNS AGING MANAGEMENT PROGRAM 16.2.2 AGE-RELATED DEGRADATION INSPECTION (ARDI) AGING MANAGEMENT PROGRAM Age-Related Degradation Inspection Method and Demonstration: In Behalf of Calvert Cliffs Nuclear Power Plant License Renewal Application, 16.2.3 ALLOY 600 AGING MANAGEMENT PROGRAM 16.2.4 ASME SECTION XI IN-SERVICE INSPECTION (ISI), SUBSECTIONS IWB, IWC, AND IWD AGING MANAGEMENT PROGRAM 16.2.5 ASME SECTION XI IN-SERVICE INSPECTION (ISI), SUBSECTIONS IWE AND IWI AGING MANAGEMENT PROGRAM Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containments16.2.6 ASME SECTION XI IN-SERVICE INSPECTION, SUBSECTION IWF AGING MANAGEMENT PROGRAM 16.2.7 BORIC ACID CORROSION INSPECTION (BACI) PROGRAM Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants, 16.2.8 BURIED PIPING INSPECTION AGING MANAGEMENT PROGRAM 16.2.9 CABLE AGING MANAGEMENT PROGRAM 16.2.10 CAST AUSTENITIC STAINLESS STEEL (CASS) AGING MANAGEMENT PROGRAM 16.2.11 CAULKING AND SEALANTS INSPECTION AGING MANAGEMENT PROGRAM 16.2.12 CHEMISTRY (WATER) AGING MANAGEMENT PROGRAM 16.2.13 COMPREHENSIVE REACTOR VESSEL SURVEILLANCE AGING MANAGEMENT PROGRAM 16.2.14 CONTAINMENT LOCAL LEAKAGE RATE TESTING AGING MANAGEMENT PROGRAM 16.2.15 DESIGN CHANGE AND MODIFICATION IMPLEMENTATION AGING MANAGEMENT PROGRAM 16.2.16 DIESEL FUEL OIL (TANKS AND CHEMISTRY) AGING MANAGEMENT PROGRAM 16.2.17 ENVIRONMENTAL QUALIFICATION (EQ) AGING MANAGEMENT PROGRAM 16.2.18 FLOW ACCELERATED CORROSION (FAC) AGING MANAGEMENT PROGRAM 16.2.19 FATIGUE MONITORING AGING MANAGEMENT PROGRAM 16.2.20 FIRE BARRIER PENETRATION SEAL INSPECTION AGING MANAGEMENT PROGRAM 16.2.21 FIRE PROTECTION AGING MANAGEMENT PROGRAM 16.2.22 LOAD HANDLING AND FUEL HANDLING EQUIPMENT AGING MANAGEMENT PROGRAM 16.2.23 PREVENTIVE MAINTENANCE (PM) AGING MANAGEMENT PROGRAM 16.2.24 PROTECTIVE COATINGS AGING MANAGEMENT PROGRAM 16.2.25 REACTOR COOLANT AGING MANAGEMENT PROGRAM 16.2.26 REACTOR VESSEL INTERNALS (RVI) AGING MANAGEMENT PROGRAM 16.2.27 SPENT FUEL POOL (SFP) AGING MANAGEMENT PROGRAM 16.2.28 STEAM GENERATOR (SG) AGING MANAGEMENT PROGRAM 16.2.29 STRUCTURE AND SYSTEM WALKDOWNS AGING MANAGEMENT PROGRAM 16.2.30 SURVEILLANCE TESTING AGING MANAGEMENT PROGRAM 16.3 EVALUATION OF TIME-LIMITED AGING ANALYSES 16.3.1 CONTAINMENT TENDON PRESTRESS LOSS 16.3.2 POISON SHEETS IN SPENT FUEL POOL 16.3.3 10 CFR 50.49 ENVIRONMENTAL QUALIFICATION PROGRAM 16.3.4 REACTOR PRESSURE VESSEL TOUGHNESS REQUIREMENTS 16.3.5 REACTOR VESSEL INTERNALS 16.3.6 CONTAINMENT LINER FATIGUE 16.3.7 MAIN STEAM PIPING FATIGUE 16.3.8 NUCLEAR STEAM SUPPLY FATIGUE 16.3.9 CLASS 2 AND 3 PIPING COMPONENTS (OTHER THAN MAIN STEAM PIPING) FATIGUE 16.4 10 CFR 54.37(B) UPDATE

16.5 REFERENCES

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