BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report - List of Effective Pages

Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report - List of Effective Pages

ML15266A272
Person / Time
Site:	Vermont Yankee
Issue date:	09/01/2015
From:	Entergy Nuclear Operations
To:	Office of Nuclear Reactor Regulation
Shared Package
ML15266A267	List: BVY 15-046, Rev. 0 to Defueled Safety Analysis Report, Drawing G-200347, Circulating Water System Aerating Structure-MAS & Reinf. BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report - List of Effective Pages BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, DSAR Changes BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, DSAR Drawing List BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing 5920-00526, Process Radiation Monitoring System BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing 5920-06245, Plan Showing Property Lines & Plan Site. BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191142, Site Plot Plan. BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191144, General Arrangement Turbine Building Ground Floor Plan. BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191148, Rev. 23, General Arrangement Reactor Building Plans Sheet 1. BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191149, Rev. 27, General Arrangement Reactor Building Plans Sheet 2. BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191150, Rev. 20, General Arrangement Reactor Building Section. BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191151, Rev 19, General Arrangement Radwaste Building Plans. BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191152, Rev. 14, General Arrangement Radwaste Building Section. BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191159, Rev. 100, Flow Diagram Service Water System. Sheet 2 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191159, Rev. 92, Flow Diagram Service Water System. Sheet 1 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191160, Rev. 24, Flow Diagram Instrument Air System. Sheet 2 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191160, Rev. 25, Flow Diagram Instrument Air System. Sheet 4 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191160, Rev. 25, Flow Diagram Instrument Air System. Sheet 5 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191160, Rev. 25, Flow Diagram Instrument Air System. Sheet 6 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191160, Rev. 25, Flow Diagram Instrument Air System. Sheet 7 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191160, Rev. 25, Flow Diagram Instrument Air System. Sheet 8 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191160, Rev. 31, Flow Diagram Instrument Air System. Sheet 3 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191160, Rev. 37, Flow Diagram Instrument Air System. Sheet 1 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191161, Rev. 22, Flow Diagram, Maxe-up Water Treatment System. BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191163, Rev. 51, Flow Diagram Fire Protection System Inner Loop. Sheet 1 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191163, Rev. 51, Flow Diagram Fire Protection System Inner Loop. Sheet 2 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191164, Rev. 29, Flow Diagram Sampling System. Sheet 1 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191165, Rev. 46, Flow Diagram Sampling System. Sheet 2 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191173, Rev. 12, Flow Diagram Fuel Pool Cooling & Clean Up System. Sheet 2 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191173, Rev. 42, Flow Diagram Fuel Pool Cooling & Cleanup System. Sheet 1 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191177, Rev. 23, Flow Diagram Radwaste Systems. Sheet 3 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191177, Rev. 44, Flow Diagram Radwaste Systems. Sheet 1 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191177G, Rev. 28, Flow Diagram Radwaste Systems. Sheet 2 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191238, Rev. 38, HVAC - Flow Diagram Reactor Building. BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191254, Rev. 43, HVAC Heating Flow Diagram & Boiler Room Layout. BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191298, Rev. 5, Main One Line Wiring Diagram. Sheet 3 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191299, Rev. 37, 4KV Auxiliary One Line Diagram. BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191300, Rev. 27, 480V Auxiliary One Line Wiring Diagram Swgr BUS-8 MCC-8A, 8C. Sheet 1 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191300, Rev. 38, 480V, Auxiliary One Line Diagram MCC-8B, MCC-8E, MCC-89B. Sheet 2 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191301, Rev. 30, 480V Aux. One Line Diagram Swgr BUS-9, MCC-9A, 9C. Sheet 1 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191301, Rev. 39, 480V Aux. One Line Diagram MCC-9B, 9D, 89A. Sheet 2 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191372, 120/240V Vital AC and Instrument AC One Line Diagram, Sheet 4 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191372, 125V DC One Line Diagram, Sheet 2 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191372, 125V DC One Line Diagram, Sheet 3 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191372, 24V DC Neutron Monitoring & 120V AC RPS One Line Diagram, Sheet 5 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191372, Rev. 72, 125VDC One Line Wiring Diagram. Sheet 1 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191451, Circulating Water System Intake Structure-M, Sheet 1 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191452, Circulating Water System Intake Structure-V, Sheet 2 BVY 15-046, Vermont Yankee Nuclear Power Station, Rev. 0 to Defueled Safety Analysis Report, Drawing G-191453, Circulating Water System Intake Structure-M, Sheet 3 ... further results
References
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{{#Wiki_filter:Vermont Yankee Defueled Safety Analysis Report-List of Effective PagesRevision0 VYNPS DSARRevision0List of Effective Pages1of1Section# PagesRevisionTitle Page 10TOC20Section 1170Section 21200Section 3980Section 4540Section 550Section 6370Section 7160Appendix A2160Appendix G.2400 VYNPS DSAR Revision 0 DEFUELED SAFETY ANALYSIS REPORT VERMONT YANKEE NUCLEAR POWER STATION DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS VYNPS DSAR Revision 0

TOC-1 of 2 SECTION 1 UFSAR, REV 17, TOC

1.0 INTRODUCTION

AND

1.1 INTRODUCTION

1.2 DESIGN CRITERIA 1.3 FACILITY DESCRIPTION 1.4

SUMMARY

OF RADIATION EFFECTS 1.5 GENERAL CONCLUSIONS SECTION 2 2.0 STATION SITE AND ENVIRONS 2.1

SUMMARY

DESCRIPTION 2.2 SITE DESCRIPTION 2.3 METEOROLOGY 2.4 HYDROLOGY AND BIOLOGY 2.5 GEOLOGY AND SEISMOLOGY 2.6 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SECTION 3 3.0 FACILITY DESIGN AND OPERATION 3.1 DESIGN CRITERIA 3.2 FACILITY STRUCTURES 3.3 SYSTEMS SECTION 4 4.0 RADIOACTIVE WASTE MANAGEMENT 4.1 SOURCE TERMS 4.2 RADIATION SHIELDING 4.3 HEALTH PHYSICS INSTRUMENTATION 4.4 RADIATION PROTECTION PROGRAM 4.5 LIQUID WASTE MANAGEMENT SYSTEMS 4.6 SOLID WASTE MANAGEMENT 4.7 EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS (Continued) VYNPS DSAR Revision 0

TOC-2 of 2 SECTION 5 5.0 CONDUCT OF OPERATIONS 5.1 ORGANIZATION AND RESPONSIBILITY 5.2 TRAINING 5.3 EMERGENCY PLAN 5.4 QUALITY ASSURANCE PROGRAM 5.5 REVIEW AND AUDIT OF OPERATIONS 5.6 TECHNICAL REQUIREMENTS MANUAL SECTION 6 6.0 SAFETY ANALYSIS

6.1 INTRODUCTION

6.2 ACCEPTANCE CRITERIA 6.3 ACCIDENTS EVALUATED 6.4 SITE EVENTS EVALUATED

6.5 REFERENCES

6.6 APPENDICES SECTION 7 7.0 AGING MANAGEMENT 7.1 SUPPLEMENT FOR RENEWED OPERATING LICENSE 7.2 AGING MANAGEMENT PROGRAMS AND ACTIVITIES

7.3 REFERENCES

7.4 LIST OF LICENSE RENEWAL COMMITMENTS APPENDICESA SEISMIC ANALYSIS G.2 CURRENT ON-SITE METEOROLOGICAL PROGRAM VYNPS DSAR Revision 0 1.0-1 of 17 INTRODUCTION AND

SUMMARY

TABLE OF CONTENTS

Section Title Page

1.1 INTRODUCTION

.......................................................... 3 1.2 DESIGN CRITERIA ....................................................... 5 1.3 FACILITY DESCRIPTION .................................................. 7 1.3.1 General ...................................................... 7 1.3.1.1 Site and Environs ............................... 7 1.3.1.2 Facility Arrangement ............................ 9 1.3.2 Fuel Storage and Handling .................................... 9 1.3.2.1 Nuclear Fuel and Control Rods ................... 9 1.3.2.2 Irradiated Fuel Storage ......................... 9 1.3.2.3 Standby Fuel Pool Cooling and Demineralizer System ............................ 9 1.3.3 Radioactive Waste Management ................................ 10 1.3.3.1 Equipment and Floor Drainage Systems ........... 10 1.3.3.2 Liquid Radwaste System ......................... 10 1.3.3.3 Solid Radwaste System .......................... 11 1.3.4 Radiation Monitoring and Control ............................ 11 1.3.4.1 Reactor Building Ventilation Radiation Monitoring System .............................. 11 1.3.4.2 Process Radiation Monitoring ................... 11 1.3.4.3 Area Radiation Monitors ........................ 12 1.3.5 Auxiliary Systems ........................................... 12 1.3.5.1 Electrical Power Systems ....................... 12 1.3.5.2 Service Water System ........................... 12 1.3.5.3 Fire Protection System ......................... 13 1.3.5.4 Heating, Ventilating, and Air Conditioning Systems ........................... 13 1.3.5.5 Service and Instrument Air Systems ............. 13 1.3.5.6 Process Sampling System ........................ 14

VYNPS DSAR Revision 0 1.0-2 of 17 1.3.6 Communications Systems ...................................... 14 1.3.6.1 Facility Communications System ................. 14 1.3.7 Station Water Purification, Treatment and Storage ........... 14 1.3.7.1 Makeup Water Treatment System .................. 14 1.3.7.2 Potable and Sanitary Water System .............. 15 1.3.8 Shielding, Access Control, and Radiation Protection Procedures .................................................. 15 1.3.8.1 General ........................................ 15 1.3.9 Structural Loading Criteria ................................. 16 1.4

SUMMARY

OF RADIATION EFFECTS ......................................... 17 1.4.1 Fuel Storage and Handling and Waste Management .............. 17 1.4.2 Accidents and Events ........................................ 17 1.5 GENERAL CONCLUSIONS .................................................. 17

VYNPS DSAR Revision 0 1.0-3 of 17

1.1 INTRODUCTION

On January 12, 2015, Entergy Nuclear Operations (ENO) certified to the Nuclear

Regulatory Commission (NRC) that a determination to permanently cease operation

at the Vermont Yankee Nuclear Power Station (VYNPS) was made on December 29, 2014

which was the date on which operation ceased at VYNPS. ENO also certified that

the fuel has been permanently removed from the VYNPS reactor vessel and placed in

the spent fuel pool. ENO acknowledged that, following docketing, the VYNPS

license no longer authorized operation of the reactor or emplacement or retention

of fuel into the reactor vessel.

This Defueled Safety Analysis Report (DSAR) is derived from Revision 26 of the

VYNPS Updated Final Safety Analysis Report (UFSAR). The DSAR has been developed

as a licensing basis document that reflects the permanently defueled condition of

VYNPS. The DSAR serves the same function during SAFSTOR and decommissioning that

the UFSAR served during operation of the facility. An evaluation of the systems, structures and components (SSCs) described in the UFSAR was performed to

determine the function, if any, these SSCs would perform in a defueled condition.

The criteria used to evaluate the major SSCs and the conclusions of the

evaluations are provided in appropriate station documents.

ENO acknowledged that the 10CFR50 operating license continues to remain in effect

until the Nuclear Regulatory Commission terminates the license.

The Vermont Yankee Nuclear Power Corporation was originally organized by ten New

England utilities in August, 1966, for the purpose of building and operating a

nuclear generating station in Vermont. At the time of application, Vermont

Yankee was similar in organization to the Yankee Atomic Electric Co. and the

Connecticut Yankee Atomic Power Co. Nine of the twelve Vermont Yankee sponsors

were also sponsors of Yankee and Connecticut Yankee. Thus, Vermont Yankee had

the benefit of the experience gained from the operation of these two plants.

The Vermont Yankee Nuclear Power Corporation was the sole applicant for an

operating license for a nuclear power station, located at the Vernon site in

Windham County, Vermont, for initial power levels up to 1593 MWt under Section

104(b) of the Atomic Energy Act of 1954, as amended, and the regulations of the

NRC set forth in Part 50 of Title 10 of the Code and Federal Regulations

(10CFR50).

The facility was designated as the Vermont Yankee Nuclear Power Station.

The Vermont Yankee Nuclear Power Corporation, as owner, was responsible for the

design, construction, operation and decommissioning of the station.

VYNPS DSAR Revision 0 1.0-4 of 17 EBASCO Services, Inc. designed and constructed the station exclusive of the nuclear steam supply system.

General Electric Company was awarded a contract to design, fabricate, and deliver

the nuclear steam supply system and nuclear fuel for the station, as well as to

provide technical direction for installation and startup of this equipment.

General Electric Company was also contracted to design, fabricate, deliver, and

install the turbine generator as well as to provide technical assistance for the

startup of this equipment.

In July 2002, the operating license was transferred to Entergy Nuclear Vermont

Yankee, LLC, a limited liability company and wholly owned subsidiary of Entergy

Nuclear Operations, Inc. VYNPS DSAR Revision 0 1.0-5 of 17 1.2 DESIGN CRITERIA

The principal architectural and engineering criteria for the design and

construction of the station, applicable in the permanently defueled state, are

summarized below.

General The station design shall be in accordance with applicable codes and

regulations.

The station shall be designed in such a way that the release of radioactive

materials to the environment is limited so that the limits and guideline values

of Title 10 of the Code of Federal Regulations pertaining to the release of

radioactive materials are not exceeded.

Structural

Adequate strength and stiffness with appropriate safety factors shall be

provided so that a hazardous release of radioactive material shall not occur.

Nuclear Fuel

The fuel cladding shall be designed to retain integrity as a radioactive

material barrier

The fuel cladding shall be designed to accommodate without loss of integrity

the pressures generated by the fission gases released from the fuel material

throughout the design life of the fuel.

The fuel cladding, in conjunction with other facility systems, shall be

designed to retain integrity throughout any abnormal operational transient.

Fuel Handling and Storage

Fuel handling and storage facilities shall be designed to maintain adequate

shielding and cooling for spent fuel.

Fuel handling and storage facilities shall be designed to preclude inadvertent

criticality.

VYNPS DSAR Revision 0 1.0-6 of 17 Electrical Power Systems The electric power system shall be designed to provide sufficient normal and

standby electrical power to assure proper operation of the spent fuel pool

cooling and support systems.

Transformers, switchgear, buses, and cables shall be designed to have adequate

current carrying capacity without exceeding the acceptable voltage drop of the

electrical loads.

Switchgear protective devices shall be provided to detect and interrupt

electrical malfunctions.

The rated capacity of interrupting devices shall exceed the maximum available

fault current.

Radioactive Waste Disposal Systems

Liquid and solid waste disposal facilities shall be designed so that the

discharge and off-site shipment of radioactive effluents can be made in

accordance with applicable regulations.

The design shall provide means to inform station operating personnel of an

approach to limits on the release of radioactive material.

Shielding and Access Control Radiation shielding shall be provided and access control patterns shall be

established to allow the staff to control radiation doses within the limits of

10CFR20.

VYNPS DSAR Revision 0 1.0-7 of 17 1.3 FACILITY DESCRIPTION

1.3.1 General

1.3.1.1 Site and Environs

1.3.1.1.1 Location and Size of Site

The site is located on the west shore of the Connecticut River immediately

upstream of the Vernon Hydroelectric Station, in the town of Vernon, Vermont, which is in Windham County. Site coordinates are approximately 42 °47' north, 72°31' west. The facility is located on about 125 acres which are bounded by privately owned land on the north, south, and west and by the Connecticut River on the east. The site plot plan is shown on Drawing G-191142.

1.3.1.1.2 Site Ownership

Entergy Nuclear Vermont Yankee, LLC is the owner of the site, with the

exception of a narrow strip of land between the Connecticut River and the VYNPS

property for which it has perpetual rights and easements from its owner.

1.3.1.1.3 Activities at Site

All activities at the facility site will be under the control of Entergy

Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. at all times.

1.3.1.1.4 Access to the Site

The immediate area around the facility is completely enclosed by a fence with

access to the facility controlled at a security gate. Access to the site is

possible from either Governor Hunt Road, a local road, or from a spur of the

Central Vermont Railroad. Site boundaries are posted.

1.3.1.1.5 Description of Environs

The area adjacent to the facility is primarily farm and pasture land. Downstream of the facility are the Vernon Hydroelectric Station and the town of Vernon, Vermont. The area within a 5-mile radius is predominantly rural with the exception of a portion of the city of Brattleboro, Vermont and the town of Hinsdale, New Hampshire. Between 75% and 80% of the area within 5 miles of the

facility is wooded. The remainder is occupied by farms and small industries.

VYNPS DSAR Revision 0 1.0-8 of 17 1.3.1.1.6 Geology The major structures at the site are supported by bedrock. Compression tests

indicated minimum failure of the bedrock to be 16,000 psi (1,152 tons per

square foot). An allowable bearing pressure has been established at 50 tons

per square foot; however, actual loadings do not exceed 20 tons per square

foot.

1.3.1.1.7 Seismology

Based on a three-fold seismic evaluation, the site was found to be relatively

quiescent from a seismic standpoint. From these studies the design earthquake

has been established at 0.07g horizontal ground acceleration and the maximum

hypothetical earthquake at 0.14g horizontal ground acceleration. The seismic

evaluation consisted of a review of historical data from the New England area, an analysis of instrument and historical records for the Vermont area, and a

study of earthquake intensity attenuation with distance for the northeast

United States.

1.3.1.1.8 Hydrology

The facility is on the Connecticut River in Vernon, Vermont, some 138.3 miles

from the river mouth. The river in the vicinity of the facility is comprised

of a series of ponds formed by dams constructed for the generation of

hydroelectric power. All local surface streams drain to the Connecticut River, and the site is in the direct path of natural drainage to the east of the local

watershed. In the vicinity of the site there is also a considerable amount of

groundwater which several municipalities utilize as one source of water supply.

1.3.1.1.9 Regional and Site Meteorology

The general climatic regime is that of a continental type with some

modification from the maritime climate which prevails nearer the coast. For

the one-year period between August 1967 and July 1968, temperature inversions

occurred 39% of the total time. Seasonal inversion frequencies ranged between

36% and 42%. Wind distribution is biased in the direction of the river due to

the channeling effect of the valley.

Historical records show that annual snowfall varies between 30 inches and 118 inches. Temperature range is about 133 °F. Occasional heavy rains and ice storms occur in the area.

VYNPS DSAR Revision 0 1.0-9 of 17 1.3.1.2 Facility Arrangement The facility arrangement is shown on Drawing G-191142. The principal

structures of the station are the reactor building and primary containment, turbine building, control building, radwaste building, intake structure, cooling towers, main stack, and Interim Spent Fuel Storage Installation (ISFSI)

storage pad.

1.3.2 Fuel Storage and Handling

1.3.2.1 Nuclear Fuel and Control Rods

Nuclear fuel previously used for power generation consists of slightly enriched

uranium dioxide pellets contained in sealed Zircaloy tubes. These fuel rods

are assembled into individual fuel assemblies. On January 12, 2015, VYNPS

certified to the NRC that all nuclear fuel had been permanently removed from

the reactor vessel and placed in the spent fuel pool. Therefore, all nuclear

fuel is stored either in the Spent Fuel Pool (SFP) or at the Independent Spent

Fuel Storage Facility (ISFSI).

Gross control of the reactor core was achieved through movement of control

rods. The control rods are cruciform shaped and were dispersed throughout the

lattice of fuel assemblies. Following certification of permanent defueling, control rods are stored either in the reactor vessel or in the spent fuel pool.

1.3.2.2 Irradiated Fuel Storage

Irradiated fuel is stored underwater in a spent fuel pool in the Reactor

Building or transferred to the ISFSI until prepared for shipment from the site.

1.3.2.3 Standby Fuel Pool Cooling and Demineralizer System

The Standby Fuel Pool Cooling (SFPCS) removes decay heat released from the

spent fuel to maintain fuel pool temperature within specified limits. The Fuel

Pool Demineralizer System (FPDS) maintains water clarity.

VYNPS DSAR Revision 0 1.0-10 of 17 1.3.3 Radioactive Waste Management The Radioactive Waste Systems are designed to control the release of

radioactive material to within the limits specified in 10CFR20 and within the

limits specified in technical specifications and the Off-Site Dose Calculation

Manual (ODCM). The methods employed for the controlled release of these

contaminants depends primarily upon the state of the material.

1.3.3.1 Equipment and Floor Drainage Systems

Drains and sumps are provided to ensure proper drainage and collection of all

reject liquids throughout the facility. The drain systems fall into four basic

groupings determined by the type of waste they will collect:

1. Radioactive liquids are drained through floor drains or equipment sumps and are pumped to collector tanks in the Radwaste Building.

2. Uncontaminated liquids are drained to storm sewers or other areas where they can be discharged to the river.

3. Oil is drained into main waste lines and is discharged to oil separator manholes.

4. Liquid chemicals are discharged to neutralizing equipment in the Radwaste Building.

1.3.3.2 Liquid Radwaste System

The Liquid Radioactive Waste Control System collects, treats, stores, and

disposes of all radioactive liquid wastes. These wastes are collected in sumps

at various locations throughout the facility and then transferred to the

appropriate collection tanks in the Radwaste Building for treatment, storage, and disposal. Wastes that would be discharged from the system would be

processed on a batch basis with each batch being processed by such method or

methods appropriate for the quality and quantity of radioactive materials

determined to be present. Processed liquid wastes are normally returned to the

Condensate System but may be discharged to the environs in accordance with

applicable permits.

VYNPS DSAR Revision 0 1.0-11 of 17 Equipment is selected, arranged, and shielded to permit operation, inspection, and maintenance with minimum personnel exposure. For example, tanks and

processing equipment which will contain significant radiation sources are located

behind shielding; and sumps, pumps, instruments, and valves are located in

controlled access rooms or spaces. Processing equipment is selected and designed

to require a minimum of maintenance.

Protection against accidental discharge of liquid radioactive waste is provided

by valving redundancy, instrumentation for detection, alarms of abnormal

conditions, and procedural controls.

1.3.3.3 Solid Radwaste System Solid radioactive wastes are collected, processed, and packaged for storage and

subsequent off-site burial. Generally, these wastes are stored on-site until

the short half-lived activities are insignificant. Solid wastes from equipment

originating in the Nuclear System are stored for radioactive decay in the fuel

storage pool and prepared for reprocessing or off-site burial in approved

shipping containers. Examples of these wastes are spent fuel, spent control

rods, in-core ion chambers, etc. Process solid wastes, such as resins or

filter material, are collected, dewatered, and prepared for storage in shielded

casks. Dry active waste such as paper, air filters, and used clothing is

collected and temporarily stored in large shipping containers before being sent

to a disposal site or to an off-site waste processor for volume reduction prior

to disposal. The processed waste may be returned to VYNPS in strong tight

packages, or sent directly to burial.

1.3.4 Radiation Monitoring and Control

1.3.4.1 Reactor Building Ventilation Radiation Monitoring System

The Reactor Building Ventilation Radiation Monitoring System consists of

radiation monitors arranged to monitor the activity level of the ventilation

exhaust from the Reactor Building.

1.3.4.2 Process Radiation Monitoring

Radiation monitors and monitoring systems are provided on process liquid and

gas lines that may serve as discharge routes for radioactive materials. The

monitors include the following:

• Plant Stack Radiation Monitoring System
• Process Liquid Radiation Monitoring System
• Reactor Building Ventilation Radiation Monitoring System

VYNPS DSAR Revision 0 1.0-12 of 17 1.3.4.3 Area Radiation Monitors Radiation monitors are provided to monitor for abnormal radiation at various

locations in the Reactor Building, Turbine Building, Main Control Room

Building, Service Building, Radwaste Building, and the immediate vicinity of

the facility. These monitors annunciate alarms when abnormal radiation levels

are detected.

1.3.5 Auxiliary Systems

1.3.5.1 Electrical Power Systems

At the 345 kV switchyard, a ring bus arrangement supplies the 115 kV switchyard

through a 345 kV/115 kV autotransformer. A line from the 115 kV switchyard also

interconnects with 115 kV transmission systems in New Hampshire. Off-site

power is supplied to the facility from the 115 kV switchyard via two startup

transformers.

The Auxiliary AC Power System provides adequate power for the safe storage and

handling of irradiated fuel and support activities.

Two standby diesel generators are provided as independent, redundant power

sources. Backup power is also available from the Vernon Hydroelectric Station

and the Station Blackout Diesel.

The Main Battery System provides a reliable source of dc power for control

power to selected breakers and power to selected lighting systems.

1.3.5.2 Service Water System

The Service Water System supplies cooling water from the Connecticut River

directly to auxiliary equipment. Pumps supply the systems and equipment through

a dual header arrangement.

VYNPS DSAR Revision 0 1.0-13 of 17 1.3.5.3 Fire Protection System Water for the Fire Protection System is supplied by two vertical turbine-type

pumps, one diesel driven and one electric-motor driven, both located in the

intake structure. These pumps supply water to the facility fire loop with its

various hydrants and subsequently to the hose stations, sprinklers, spray, and

deluge systems throughout the Turbine Building, office, and service areas.

Supplementing these water systems are a CO 2 Fire Protection System for the cable vault and Switchgear Rooms and portable fire extinguishers located throughout

the facility, as well as fixed and portable foam suppression systems.

The Diesel Generator Rooms, the Heating Boiler Room and the motor generator

sets in the Reactor Building are protected by automatic fire detection devices

which alarm in the Main Control Room.

Consideration has been given to the use of noncombustible and fire-resistant

materials throughout the facility.

1.3.5.4 Heating, Ventilating, and Air Conditioning Systems

The Heating, Ventilating, and Air Conditioning (HVAC) Systems normally provide

filtered air to the facility structures.

This air provides the appropriate temperature and humidity conditions as

required in these structures for personnel and equipment protection. It

provides for the effective protection of personnel against possible airborne

radioactive contaminants by maintaining flow direction and rate so that the

gaseous or particulate contaminants are effectively prevented from entering the

cleaner zones.

1.3.5.5 Service and Instrument Air Systems

The Instrument Air System provides the facility with a continuous supply of

dry, oil-free air for pneumatic instruments and controls through a dual header

system.

The Service Air System provides the facility with a continuous supply of air

where the air quality of the Instrument Air System is not required. Four 100%

capacity air compressors and two air receiver tanks comprise the pieces of

equipment for the two systems. Additionally, the Instrument Air System has a

filter and drier in each header to ensure air quality.

VYNPS DSAR Revision 0 1.0-14 of 17 1.3.5.6 Process Sampling System

The Process Sampling System provides a means for sampling and testing various

process fluids in centralized locations, from which the performance of the

facility, items of equipment, and systems may be determined.

1.3.6 Communications Systems

1.3.6.1 Facility Communications System

The Communications System provides adequate means of communication throughout

the facility and from the facility to off-site locations. The on-site means of

communication are:

1. Intrasite dial telephone system

2. Intrastation public address system

3. Sound-powered telephone system

4. Intrastation radio communications system

Communications to off-site locations can be accomplished by means of:

1. Public telephones

2. Off-site radio communications system

3. Intersite microwave communications system 1.3.7 Station Water Purification, Treatment and Storage 1.3.7.1 Makeup Water Treatment System

The Makeup Water Treatment System processes raw river water from the

Connecticut River to maintain a supply of high quality water which may be used

as a makeup for the facility. Filters, purifiers, and ion exchangers are

provided to ensure the required water quality in sufficient quantities.

Treated water is stored in the demineralized water storage tank. VYNPS DSAR Revision 0 1.0-15 of 17 1.3.7.2 Potable and Sanitary Water System

Potable and sanitary water, filtered and treated as necessary, is provided in

sufficient quantity by this system to supply all facility drinking and sanitary

water requirements.

1.3.8 Shielding, Access Control, and Radiation Protection Procedures

1.3.8.1 General

Control of radiation exposure of facility personnel and people external to the

facility exclusion area is accomplished by a combination of radiation

shielding, control of access into certain areas, and administrative procedures.

The requirements of 10CFR20 are used as a basis for establishing the basic

criteria and objectives.

Shielding is used to reduce radiation dose rates in various parts of the

facility to acceptable limits. Access control and administrative procedure are

used to limit the integrated dose received by facility personnel to less than

that set forth in 10CFR20. Access control and procedures are also used to

limit the potential spread of contamination from various areas, particularly

areas where maintenance occurs.

Shielding is also used as necessary to protect equipment from radiation damage.

Of principal concern are organic materials such as insulation, linings, and

gaskets. The design levels are adjusted to accommodate the radiation damage

resistance of specific materials.

VYNPS DSAR Revision 0 1.0-16 of 17 1.3.9 Structural Loading Criteria Structures and equipment are designed to substantially resist mechanical damage

due to loads produced by mechanical and thermal forces. For the purpose of

categorizing mechanical strength designs for these loads, the following

definitions were established:

1. Class I Class I includes those structures, equipment, and components whose failure

or malfunction might cause or increase the severity of an accident which

would endanger the public health and safety.

2. Class II Class II includes those structures, and components which are important to

the safe storage and handling of irradiated fuel and radioactive waste, but

are not essential for preventing or mitigating the consequences of an

accident which would endanger the public health and safety.

The loading categories are generically described and their meaning is expanded

in Section 3.

VYNPS DSAR Revision 0 1.0-17 of 17 1.4

SUMMARY

OF RADIATION EFFECTS

1.4.1 Fuel Storage and Handling and Waste Management

Spent fuel storage and handling and waste management operations will be

conducted so that the dose to any off-site person, from external or internal

sources, will not exceed that permitted by 10 CFR 20.1301. It is expected that

during fuel storage and handling and waste management operations the dose to

any off-site person from gaseous waste discharge will not average more than

about 1% of the permissible dose, and that concentrations of liquid waste at

the point of discharge will average less than the concentrations permitted by

10 CFR 20. Both effects are only a small fraction of the effect of natural

background radiation.

For ISFSI operations, 10 CFR 72.106(b) defines the dose that any individual

located on or beyond the nearest boundary of the controlled area may receive

from any design basis accident associated with the ISFSI. For additional

information, see the VYNPS 10 CFR 72.212 Evaluation Report.

1.4.2 Accidents and Events

The ability of the station to withstand the consequences of accidents and

events without posing a hazard to the health and safety of the public is

evaluated by analyzing a fuel handling accident in the spent fuel pool and a

radwaste transfer cask drop event. The calculated consequences are

substantially below the dose limits given in 10 CFR 50.67 for the fuel handling

accident and 10CFR100 for the transfer cask drop event. A further description

is provided in Section 6.

1.5 GENERAL CONCLUSIONS

Based on the design of the facility and the analysis of credible events, there

is reasonable assurance that the facility can safely manage irradiated fuel and

radioactive waste without endangering the health and safety of the public.

VYNPS DSAR Revision 0 2.0-1 of 120 SITE AND ENVIRONS TABLE OF CONTENTS

Section Title Page 2.1

SUMMARY

DESCRIPTION ................................................... 9 2.2 SITE DESCRIPTION ..................................................... 10 2.2.1 Location and Area .......................................... 10 2.2.2 Population ................................................. 10 2.2.3 Land Use ................................................... 11 2.2.4 Site Area Boundaries, Exclusion Area, and Low Population Zone ............................................ 12 2.2.5 Conclusions ................................................ 15 2.3 METEOROLOGY .......................................................... 31 2.3.1 General .................................................... 31 2.3.2 On-site Meteorological Programs ............................ 31 2.3.3 Diffusion Climatology ...................................... 31 2.3.4 Winds and Wind Loading ..................................... 32 2.3.5 Temperature and Precipitation .............................. 32 2.3.5.1 Temperature .................................... 32 2.3.5.2 Precipitation .................................. 33 2.3.5.3 Snowfall, Snow and Ice Loading ................. 33 2.3.6 Storms ..................................................... 35 2.3.6.1 Thunderstorms .................................. 35 2.3.6.2 Hurricanes ..................................... 36 2.3.6.3 Tornadoes ...................................... 36 2.3.7 Conclusions ................................................ 37 2.3.8 References ................................................. 38 2.4 HYDROLOGY AND BIOLOGY ................................................ 48 2.4.1 General .................................................... 48 2.4.2 Land Area Ground Hydrology ................................. 48 2.4.2.1 Introduction ................................... 48

VYNPS DSAR Revision 0 2.0-2 of 120 2.4.2.2 Surface Water .................................. 48 2.4.2.3 Groundwater .................................... 48 2.4.3 Hydrology .................................................. 49 2.4.3.1 Introduction ................................... 49 2.4.3.2 Stream Flow .................................... 49 2.4.3.3 Temperature .................................... 50 2.4.3.4 Floods ......................................... 50 2.4.4 Uses of River .............................................. 57 2.4.4.1 Introduction ................................... 57 2.4.4.2 Industrial Use ................................. 57 2.4.4.3 Public Use ..................................... 57 2.4.5 Biology .................................................... 58 2.4.5.1 Commercial Fisheries ........................... 58 2.4.5.2 Sport Fisheries ................................ 59 2.4.5.3 Bottom Fauna ................................... 59 2.4.5.4 Aquatic Plants ................................. 60 2.4.5.5 Conclusions .................................... 60 2.4.6 Chemical and Bacteriological Quality of Water .............. 60 2.4.7 River Field Program ........................................ 61 2.4.8 Conclusions ................................................ 61 2.4.9 References ................................................. 63 2.5 GEOLOGY AND SEISMOLOGY ............................................... 88 2.5.1 General .................................................... 88 2.5.2 Geology .................................................... 88 2.5.2.1 Introduction ................................... 88 2.5.2.2 Geological Investigation Program ............... 88 2.5.2.3 Regional Geology ............................... 89 2.5.2.4 Site Geology ................................... 91 2.5.2.5 River Geology .................................. 93 2.5.3 Seismology ................................................. 95

VYNPS DSAR Revision 0 2.0-3 of 120 2.5.3.1 Introduction ................................... 95 2.5.3.2 Seismic Investigation Program .................. 95 2.5.3.3 Geologic and Tectonic Background ............... 95 2.5.3.4 Seismic History ................................ 95 2.5.3.5 Seismicity of Area ............................. 97 2.5.4 Conclusions ................................................ 98 2.6 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ....................... 116 2.6.1 Objectives ................................................ 116 2.6.2 Monitoring Network ........................................ 117 2.6.2.1 Direct Radiation .............................. 117 2.6.2.3 Waterborne .................................... 118 2.6.2.4 Ingestion ..................................... 119 2.6.3 Land Use Census ........................................... 119 2.6.4 Emergency Surveillance .................................... 119 2.6.5 Reports ................................................... 120

VYNPS DSAR Revision 0 2.0-4 of 120 STATION SITE AND ENVIRONS LIST OF TABLES

Table No. Title

2.2.1 Population Density Comparison: 1990-2000

2.2.2 2000 Population Distribution: (0-50 Miles)

2.2.3 Projected Population Distribution for Year 2010

2.2.4 Urban Centers Within 30 Miles of Site

2.2.5 Table of Land Use - Square Miles

2.2.6 Agricultural Statistics for Counties Within 50 Miles

2.2.7 Hazardous Materials Railroad Traffic Through Vernon, Vermont

2.3.1 Meteorology Record

2.3.2 Temperature Data for the Vernon Area

2.3.3 Precipitation Data for the Vernon Area

2.3.4 Winds During Thunderstorms

2.3.5 Rainfall Data from Hurricane Connie

2.4.1 Average and Extreme Values of Stream Flow Connecticut River at Vernon, Vermont Water Years 1944-1988

2.4.2 Vermont Yankee Nuclear Power Station, Daily Stream Flow for October 1964 to September 1965, Connecticut River at Vernon, Vermont 2.4.3 Municipal and Industrial Groundwater Usage Within a 10-Mile Radius of the Vernon Site

2.4.4 Public Water Supplies Within a 10-Mile Radius of the Vernon Site

2.4.5 Water Supplies Within a l-Mile Radius of the Site

2.4.6 Six-Hour PMP and Runoff Increments - Connecticut River Basin above Vernon, Vermont

2.4.7 Maximum Annual Floods on Connecticut River at Vernon, Vermont - Arranged in Descending Order (1927, 1936, 1938, 1945-1973)

VYNPS DSAR Revision 0 2.0-5 of 120 STATION SITE AND ENVIRONS LIST OF TABLES (Cont'd)

2.4.8 Time - Varying PMF Stage - Discharge Table Vermont Yankee Nuclear Plant Site

2.4.9 Time - Varying Modified PMF Stage - Discharge Table Vermont Yankee Nuclear Plant Site

2.4.10 Checklist of Connecticut River Fishes Found Near Vernon, Vermont

2.4.11 Fishes of the Connecticut River in the Vicinity of Vernon, Vermont - All Collections, 1980 2.5.1 Available Information Concerning Geology and Seismic Activity Related to the Vermont Yankee Nuclear Power Station Site

2.5.2 Vernon Pluton: Estimated Mode of the Oliverian Magma Series VYNPS DSAR Revision 0 2.0-6 of 120 STATION SITE AND ENVIRONS LIST OF FIGURES Reference Figure No. Drawing No. Title

2.2-1 Location Map Mile Radius

2.2-2 Location Map Mile Radius

2.2-3 Location Map Mile Radius

2.2-4 G-191142 Station Plan

2.2-5 5920-6245 Plan Showing Exclusion Area and Restricted Area Boundaries 2.2-6 Station Site - Area Population Distribution - 5-Mile Radius - Year-2000

2.2-7 Station Site - Area Population Distribution - 10-Mile Radius - Year-2000

2.2-8 Station Site - Area Population Distribution - 50-Mile Radius - Year-2000 2.3-1 Station Site - Westover AFB, Massachusetts Area - Annual Surface Windrose

2.3-2 Station Site - Westover AFB, Massachusetts Area - Seasonal Surface Windroses - (Winter, Spring, Summer, Fall)

2.3-3 Station Site - Concord, NH Area - Return Period of Rainfall (for extremely short intervals) 2.4-1 Station Site - Area Public Water Supplies - 10-Mile Radius

2.4-2 Station Site - Area Private Water Supplies - 1-Mile Radius

2.4-3 Enveloping Depth-Duration-Area Values of PMP for Susquehannna River Basin

2.4-4 6-Hour Unit Hydrograph

2.4-5 Total SPF Hydrograph 2.4-6 Total PMF Hydrograph (Natural and Modified)

VYNPS DSAR Revision 0 2.0-7 of 120 STATION SITE AND ENVIRONS

LIST OF FIGURES (Cont'd)

Reference Figure No. Drawing No. Title

2.4-7 Connecticut River Basin - Federal Power Commission Water Resource Appraisals for

Hydroelectric Licensing - Summary of Planning

Status 2.4-8 Vermont Yankee Nuclear Plant - Location of River Cross-Sections

2.4-9 Stage-Discharge Curve at the Vermont Yankee Nuclear Plant

2.4-10 Cross Section of the Critical Fetch

2.4-11 Vermont Yankee Sample Stations on Connecticut River 2.5-1 Not Used

2.5-2 Station Site - Geological Survey - General Plan - Location of Test Borings

2.5-3 Station Site - Geological Survey - Subsurface Profile - Log of Test Borings (1A, 2A, 3A, 4, 5, 8) 2.5-4 Station Site - Tectonic Map - State of Vermont 2.5-5 Station Site - Tectonic Map - State of New Hampshire

2.5-6 Station Site - Geological Survey - Area Bedrock Geology

2.5-7 Station Site - Geological Survey - Area Geological Section

2.5-8 Station Site - Geological Survey - Subsurface Profile (Section AA) - Log of Test Borings

(5, 8, S9, 11, and 21)

2.5-9 Station Site - Geological Survey - Subsurface Profile (Section BB) - Log of Test Borings

(2A, 3A, ST6-1/2, and S9) VYNPS DSAR Revision 0 2.0-8 of 120 STATION SITE AND ENVIRONS

LIST OF FIGURES (Cont'd)

Reference Figure No. Drawing No. Title

2.5-10 Station Site - Geological Survey - Subsurface Profile (Section CC) - Log of Test Borings

(2, 2A, 5, 7, 7A, 13, and 15)

2.5-11 Station Site - Geological Survey - Subsurface Profile (Section DD) - Log of Test Borings

(3, 3A, 4, 8, 8A, 12, and 16)

2.5-12 Station Site - Tectonic Map - New England Area 2.5-13 Station Site - Compilation of Earthquakes - New England Area

2.5-14 Station Site - Earthquake Intensity - Modified Mercalli and Rossi - Forel Scales

2.5-15 Station Site - Compilation of Earthquakes - Central New England Area VYNPS DSAR Revision 0 2.0-9 of 120 2.1

SUMMARY

DESCRIPTION

This section provides information about the site and environs of the Vermont

Yankee Nuclear Power Station (VYNPS) and summarizes the analyses and studies

which confirm the suitability of the site. Necessary information is included for

establishing safety-related portions of the facility design bases and for

analyzing facility safety under normal and abnormal conditions.

Several independent consultants were originally used by the Vermont Yankee

Nuclear Power Corporation to assist in the development of information presented

in this Section. They included:

Meteorology TRC Service Corporation Hartford, Connecticut

Geology Goldberg Zoino & Associates Cambridge, Massachusetts

Seismology Weston Geophysical Research Inc. Weston, Massachusetts

Hydrology & Marine Ecology Webster-Martin, Inc. South Burlington, Vermont

Environmental Radiation Monitoring Eberline Instrument Corporation Santa Fe, New Mexico

The site of the VYNPS at Vernon, Vermont, was thoroughly investigated and found

to be suitable in 1967 when the construction permit was issued. Since the

issuance of the construction permit, further review has been pursued in the areas

of meteorology, hydrology, and marine ecology, geology and seismology, and

environmental radiation monitoring. The results of this additional review

confirmed the suitability of Vernon as a nuclear power plant site.

VYNPS DSAR Revision 0 2.0-10 of 120 2.2 SITE DESCRIPTION 2.2.1 Location and Area The site is located in the town of Vernon, Vermont in Windham County on the west shore of the Connecticut River immediately upstream of the Vernon Hydroelectric

Station. The site contains about 125 acres owned by Entergy Nuclear Vermont

Yankee, LLC and a narrow strip of land between the Connecticut River and the east

boundary of the VYNPS property to which Entergy Nuclear Vermont Yankee, LLC has

perpetual rights and easements from its owner. This land is bounded on the

north, south, and west by privately-owned land and on the east by the Connecticut

River. Site coordinates are approximately 42 o 47' north latitude and 72 o 31' west longitude. Figures 2.2-1 through 2.2-3 locate the site. The site plot plan

is shown on Drawing G-191142. The site's exclusion area boundary and site area

boundaries for both gaseous and liquid effluents are shown on Drawing 5920-6245.

2.2.2 Population

The population density for 1990 was estimated to be about 121 people per square

mile within a five-mile radius of the site. The population density in this same

area was estimated to be 126 people per square mile in 2000, and projected to be

about 131 people per square mile by 2010. Table 2.2.1 compares the growth of the

estimated population and population density within 25 miles of the site between

1990 and 2000. In 1990, the total population within 25 miles was estimated to be

189,038, or an average density of 96 people per square mile. For 2000, the

25-mile radius population has been estimated to be about 193,746, or an average

density of 99 people per square mile. This represents a growth factor of about

2.5% for 2000 area over the ten-year period 1990 to 2000. Table 2.2.2 shows the

distribution of population in the area within a 50-mile radius of the site for

calendar year 2000. The total resident population within 50 miles for 2000 is

estimated to be about 1,467,343. Table 2.2.3 indicates the projected population

by radial distance out to 50 miles for the year 2010 based on this region's

projected growth rate of 4% over the next 10 years. The estimated 50-mile

population for the year 2010 is 1,526,037.

Figure 2.2-6 shows the estimated population in each 22-1/2 o sector around the site to a distance of 5 miles. Figure 2.2-7 shows the estimated population in

each sector out to 10 miles from the site. The greatest concentration of

population in these sectors is in the city of Brattleboro 4 to 5 miles NNW of the

site. Figure 2.2-8 shows the estimated population in each sector out to 50 miles

from the site. Table 2.2.4 shows population growth histories for urban centers

within 30 miles of the site. The nearest towns with populations of 25,000 or

more are Northampton, Massachusetts (2000 population 28,978) at about 30 miles to

the south; and Amherst, Massachusetts (2000 population 34,874) at about 28 miles

south. Accordingly, 28 miles is the population center distance.

VYNPS DSAR Revision 0 2.0-11 of 120 2.2.3 Land Use About 80% of the land within a 25-mile radius of the site is undeveloped. Most

of the developed land is used for agriculture and dairying, with homes scattered

or grouped in small villages. Table 2.2.5 shows the distribution of land within

25 miles of the site.

Table 2.2.6 describes basic annual agricultural statistics concerning milk, meat, poultry, and vegetable production by county for those counties wholly or partly

within 50 miles of the site. The primary agricultural crop in the immediate site

area is silage corn which is stored for year-round feed for milk cows.

The area within 10 miles of the site has only one urban area, the city of

Brattleboro, Vermont (2000 population 12,005), which is located about 5 miles

upriver. The remainder of this area is rural and contains several small villages

with populations between 1,000 and 3,000. The area between 10 and 25 miles has

only three urban centers with 2000 populations between 11,299 and 22,563 (see

Table 2.2.4).

The closest site boundary is 910 feet west of the Reactor Building. The nearest

homes are situated along the Governor Hunt Road just west of the site. An annual

land use census checks on the location of the nearest resident and reports this

finding as part of the Annual Radiological Environmental Operating Report. The

Vernon Elementary School, which has a pupil enrollment of about 250 is on the

other side of the road (Highway No. 4) about 1,500 feet from the Reactor

Building.

The nearest hospital, Brattleboro Memorial, is approximately five (5) miles from

the site. The nearest dairy farm is approximately 1/2-mile west-northwest of the

site and there are several others within a 5-mile radius of the plant. The

nearest railroad line runs north-south through the site area, and is

approximately 0.5 miles west of the plant at its closest approach. Table 2.2.7

lists the approximate quantities of hazardous materials which are annually

shipped past the site by the Springfield Terminal Railway and the Central Vermont

Railway which utilize this track. No other significant off-site sources of

hazardous materials have been identified within five (5) miles of the site.

The land within a 1-mile radius of the site is occupied by rural homes and is

used for dairy feed products and pasture, except for a residential area of about

75 houses located about 0.8 miles across the Connecticut River. About 30% of

this area consists of the river and undeveloped land adjacent to it.

VYNPS DSAR Revision 0 2.0-12 of 120 2.2.4 Site Area Boundaries, Exclusion Area, and Low Population Zone As defined in 10 CFR 20 and 10 CFR 100, the terms "unrestricted area," "controlled area," "restricted area," "exclusion area," and "low population zone" each refer to a specific area about the site as a result of applying different

radiological health constraints. The "unrestricted area" refers to all areas

beyond the site's outer security fence access to which is neither limited nor

controlled by the licensee. The "controlled area" refers to all plant areas

inside the site boundary, but outside of any restricted area, access to which is

limited by the licensee for any reason. Access to the controlled area can be

limited to minimize exposures to members of the public from routine radioactive

releases from the plant and fixed radiation sources. "Restricted area" refers to

the inner most areas of the plant site and facilities, access to which is limited

by the licensee for the purpose of protecting occupationally exposed individuals

against undue risks from radiation and radioactive materials. Exclusion area

means that area surrounding the reactor, as measured from the reactor center

line, in which the reactor licensee has the authority to determine all activities

including exclusion or removal of personnel and property from the area. This

area may be traversed by a highway, railroad, or waterway, provided those are not

so close to the facility as to interfere with normal operations of the facility

and provided appropriate and effective arrangements are made to control traffic

on the highway, railroad, or waterway, in case of an emergency, to protect the

public health and safety. The exclusion area also includes part of the adjacent

waterway (Connecticut River) extending across to the opposite shoreline.

Finally, the low population zone is delineated by an area about the plant which

includes residential, farming, industrial, etc., activities to some extent, but

is not so large or populated to prevent orderly, effective radiological control

or evacuation in the event of an accident of an environmentally significant

nature.

Thus, these areas and zones are delineated for different purposes and vary in the

degree of control that the licensee can exercise from a radiation protection

standpoint. The following discussion presents an analysis of each area in

relation to the plant and its operations.

VYNPS DSAR Revision 0 2.0-13 of 120

1. Controlled Area The controlled area for the VYNPS site consists of a significant portion of

the 125-acre property area owned by Entergy Nuclear Vermont Yankee, LLC. The

fenced boundaries of this area are delineated on Drawing 5920-6245. The fence

is a 6-foot high security fence topped by l foot of barbed wire. In addition

to the fence, signs are posted clearly informing an individual that the area

is private property and unauthorized entry is strictly prohibited. Access to

and activities within this area are under the direct control of Entergy

Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. Access to the

area is from the Governor Hunt Road through the main gate. The fence and

location combine to afford access and activity control to the VYNPS site.

Two normally locked gates exist in the northern corners of the Controlled Area

for access by Security Officers from the Controlled Area into the Exclusion

Area on the northern part of the property. One gate is located along the east

fence line and one gate is located along the west fence line. The gate on the

west fence will also be used for alternate access to the site for fire trucks

and as an alternate emergency evacuation route per the Emergency Plan.

For ISFSI operations, 10 CFR 72.106(b) defines the dose that any individual

located on or beyond the nearest boundary of the controlled area may receive

from any design basis accident associated with the ISFSI. For additional

information, see the VYNPS 10 CFR 72.212 Evaluation Report.

2. Effluent Boundaries In addition to the land area within the site's outer security fence, VYNPS

includes the river water area between the northern and southern boundary

fences, and extending out to the state border near the middle of the river, as

part of the site boundary for control of gaseous effluents as regulated under

the dose objectives of 10 CFR 50, Appendix I. The low exposure rates involved

and the zero or near zero occupancy factor applicable to individuals in the

river area combine to allow VYNPS to include this region for the purpose of

controlling plant releases to levels as-low-as-reasonably achievable. The

restricted area boundary for liquid discharge concentration limits (10 CFR 20)

is set at the point of discharge from the plant to the river (see Drawing

5920-6245). Thus, the overall boundary area for the plant is as shown on

Drawing 5920-6245. To ensure compliance with the constraints applicable to the unrestricted and

controlled areas as described, area dosimeter stations are provided at

strategic locations around the site. Measurements of integrated gamma

exposure are made to alert VYNPS to any condition that may produce a greater

exposure than necessary.

VYNPS DSAR Revision 0 2.0-14 of 120

3. Exclusion Area The exclusion area for the VYNPS site is also shown on Drawing 5920-6245 and

includes the controlled area defined above. The minimum distance to the

boundary of the exclusion area, as measured from the reactor center line, is

910 feet. In addition, the Connecticut River water area between Vernon Dam

and the northern VYNPS property line is included in the exclusion area since

it will be a controlled access region during an accident condition. The means

of controlling access on the river, and evacuating it if necessary, have been

worked out with the State of New Hampshire officials who will coordinate

control activities over the river. Passage on the Connecticut River to Vernon Pond is possible. The licensee

will at all times retain the complete authority to determine and maintain

sufficient control of all activities through ownership, easement, contract

and/or other legal instruments on property which is closer to the reactor

center line than 910 feet. This includes the authority to exclude or remove

personnel and property within the exclusion area. Only facility related

activities are permitted in the exclusion area. No residences will be

permitted in the exclusion area. Control over activities within, and access to, the exclusion area assume an

entirely different form immediately following a condition that produces, or

threatens to produce, a radiological hazard to the site. The VYNPS Emergency

Plan describes the types and level of emergency action that will be initiated

at the plant in order to minimize radiation exposure following an accidental

release. The only addition to that discussion is that, as previously

mentioned, evacuation and access control will be placed into effect for the

Connecticut River area included in the exclusion zone.

A normally locked gate on the northwest corner of the Exclusion Area fence is

used for access by Vermont Electric Power Co for access to their switchyards, and is also used by VYNPS as an alternate access to the site for fire trucks

and emergency equipment, or as an alternate emergency evacuation route per the

Emergency Plan.

VYNPS DSAR Revision 0 2.0-15 of 120

4. Low Population Zone The low population zone for the VYNPS is the area included within a 5-mile

radius of the site. It is outlined on Figure 2.2-2. The 2000 population

distribution for this zone is shown in Figure 2.2-6. This land area was

selected because it meets the requirements of 10 CFR 100 with respect to

radiation dose resulting from a design basis accident and orderly evacuation

capability if it is required. The methods and responsibilities for actions

taken within the low population zone during an accident condition are given in

the VYNPS Emergency Plan.

5. General The boundaries for the unrestricted area, controlled area, restricted area, exclusion area, and low population zone, as well as for control of effluents

to levels as-low-as-reasonably achievable, as described, are fully consistent

with the principles involved in ensuring the health and safety of the public, together with the plant personnel. In addition, the delineation yields an

effective arrangement with regard to efficient facility operation. The complete perimeter fence described for the protected area, together with

the fact that the only facility access point is maintained by the security

force, afford the licensee with complete, continuous access and activity

control for every component of the facility. In addition, fencing is provided

for the 115 kV and 345 kV switchyards. Thus, the responsibilities of the licensee are met from both radiological

protection and plant security standpoints.

2.2.5 Conclusions

About 80% of the land within 25 miles of the site is undeveloped. The 2000

census shows that about 489 people live within 1 mile of the site and about 9,919

live within 5 miles. The 2000 data also show that population density in the

vicinity is light, about 126 persons per square mile within a 5-mile radius and

99 persons per square mile within a 25-mile radius. Population projections to

2010 predict about a 4% increase above the 2000 figures. However, the average

population density is expected to remain low. The location of the site provides

good local isolation with light population density in the surrounding area.

In summary, the site is suitable for the facility as designed from population

distribution and land usage considerations.

VYNPS DSAR Revision 0 2.0-16 of 120 TABLE 2.2.1 Population Density Comparison (1990 -2000)

Distance from Reactor Building Miles Population Land Area Square Miles A verage Density Person per Square Mile 1990 0-1 4543.14145 1-2 2,3529.42250 2-3 1,79115.71114 3-4 1,48921.9968 4-5 3,39328.27120 0-5 9,47978.541215-10 23,510235.6210010-20 104,415942.4811020-25 51,634706.86730-25 189,0381,963.5096 2000 0-1 4893.14156 1-2 2,4969.422652-3 1,93715.711233-4 1,55621.99714-5 3,44128.27122 0-5 9,91978.541265-10 23,954235.6210210-20 111,005942.4811820-25 48,868706.86690-25 193,7461,963.5099 VYNPS DSAR Revision 0 2.0-17 of 120 Table 2.2.2

2000 Population Distribution

(0 -50 Miles)

Miles Sector 0-5 5-10 10-20 20-30 30-40 40-50 Sector Totals N 1,196 1,341 4,810 9,060 16,858 12,940 46,204 NNE 382 1,063 3,3124,8555,07821,73436,424NE 340 154 24,0922,8816,68511,07845,230ENE 1,327 787 8,4056,86910,71122,84050,939E 627 2,411 3,43012,17415,38258,36592,389ESE 161 556 1,46022,37664,29699,853188,701SE 686 956 14,6369,07113,308103,284141,941SSE 604 1,692 2,8541,56615,33038,39660,442S 536 1,448 10,82541,89369,847334,004458,553SSW 239 1,313 21,3487,72439,68949,744120,057SW 235 721 5,0213,3685,79417,47132,609WSW 204 390 1,6813,27622,25650,48178,287W 207 373 1,9054,42027,5259,85844,288WNW 338 592 2,77690515,36511,87931,855NW 659 3,350 1,5521,5245,8984,52917,512 NNW 2,180 6,809 2,898 1,885 3,503 4,637 21,912 Ring Totals 9,919 23,954 111,005 133,847 337,525 851,093 1,467,343 VYNPS DSAR Revision 0 2.0-18 of 120 TABLE 2.2.3Projected Population Distributionfor Year 2010(0 -50 Miles) Miles 0-5 5-10 10-20 20-30 30-40 40-50 Total Ring Population (people) 10,316 24,912115,445139,201 351,026885,1371,526,037 Miles 0-5 5-10 10-20 20-30 30-40 40-50 0-50 Ring Density (people/mi

2) 131 10612289160313194 Miles 0-5 0-10 0-20 0-30 0-40 0-50 Cumulative

Population (people) 10,316 35,228150,673289,874 640,9001,526,037 VYNPS DSAR Revision 0 2.0-19 of 120 TABLE 2.2.4 Urban Centers Within 30 Miles of Site City Approximate Distance From Site - Miles 1960 1970 1980 1990 2000 Brattleboro, VT 4 9,315 12,239 11,886 12,241 12,005 Greenfield, MA 12 14,389 18,116 18,415 18,666 18,168 Keene, NH 13 17,562 20,467 21,449 22,430 22,563 Athol, MA 19 10,161 11,185 10,619 11,451 11,299 Amherst, MA 28 13,718 26,331 33,210 35,228 34,874 Northampton, MA 30 30,058 29,669 29,128 29,289 28,978 VYNPS DSAR Revision 0 2.0-20 of 120 TABLE 2.2.5 Table of Land Use - Square Miles Distance From Site Land Use 0 - 10 Miles 10 - 25 Miles 0 - 25 Miles Residential 30.2 79.2 109.4 Commercial and Industrial 1.3 7.7 9.0 Agricultural 25.9 143.8 169.7 Road 6.0 21.6 27.6 Public 7.2 78.3 85.5 Undeveloped 243.4 1,318.4 1,561.8 TOTAL 314.0 1,649.0 1,963.0 VYNPS DSAR Revision 0 2.0-21 of 120 TABLE 2.2.6 Agricultural Statistics for Counties Within 50 Miles (1) Windham Vermont Windsor Vermont Rutland Vermont Bennington Vermont Sullivan New Hampshire Milk Cows Number of Head 4,484 5,708 11,858 2,355 3,024

Cattle and Calves Number of Head Sold 4,948 6,385 11,839 2,549 2,483

Sheep and Lambs Number of Head Sold 1,302 3,262 781 598 1,133

Hogs and Pigs Number of Head Sold 926 1,356 1,447

• 208 Poultry Number of Chickens 4,549 2,397 7,063 917 1,191

Vegetables, Melons, Sweet Corn Acreage 329 166 202 110 83

Orchards Acreage 890 265 318

• 53 (1) Source: 1987 Census of Agriculture, U.S. Department of Commerce, Bureau of Census (AC87-A-21; AC87-A-29; and AC87-A-45).
• Owner did not disclose.

TABLE 2.2.6 (Continued)

Agricultural Statistics for Counties Within 50 Miles (l) VYNPS DSAR Revision 0 2.0-22 of 120 Cheshire New Hampshire Hillsboro New Hampshire Franklin Massachusetts Worcester Massachusetts Hampshire Massachusetts Milk Cows Number of Head 3,067 2,203 6,475 8,933 5,271

Cattle and Calves Number of Head Sold 2,876 2,378 5,524 9,855 5,028

Sheep and Lambs Number of Head Sold 1,949 713 1,008 1,589 2,458 Hogs and Pigs Number of Head Sold 511 2,341 1,145 7,552 4,645

Poultry Number of Chickens 96,525

• 17,585 1,014,621 4,826 Vegetables, Melons, Sweet Corn Acreage 125 1,184 1,403 2,026 2,273

Orchards Acreage 85 1,879 1,157 3,443 943

• Owner did not disclose.

TABLE 2.2.6 (Continued)

Agricultural Statistics for Counties Within 50 Miles (l) VYNPS DSAR Revision 0 2.0-23 of 120 Berkshire Massachusetts Middlesex Massachusetts Rensselaer (2) New York Columbia (2) New York Washington (2) New York Milk Cows Number of Head 5,282 1,494 10,132 16,008 30,925

Cattle and Calves Number of Head Sold 5,589 2,232 9,555 16,071 27,394

Sheep and Lambs Number of Head Sold 708 2,283 281 419 713

Hogs and Pigs Number of Head Sold 388 8,739 2,685 1,364 1,240

Poultry Number of Chickens 153,423 85,074 N/A 116,694 N/A

Vegetables, Melons, Sweet Corn Acreage 510 2,273 916 1,642 232

Orchards Acreage 250 1,327 212 6,628 402

(1) Source: 1987 Census of Agriculture, U.S. Department of Commerce, Bureau of Census (AC87-A-21; AC87-A-29; and AC87-A-45).

(2) Source:  1978 Census of Agriculture, U.S. Department of Commerce, Bureau of Census (AC78-A-21; AC78-A-29; AC78-A-32; and AC78-A-45).

VYNPS DSAR Revision 0 2.0-24 of 120 TABLE 2.2.7 Hazardous Materials Railroad Traffic Through Vernon, Vermont Chemical (1) Central Vermont (2) Springfield Track (2) Total Per Year Carbon Dioxide 395 96 491 Nitrogen 248 -- 248 Propane (LPG) 60 162 222 Chlorine 60 -- 60 Sulfuric Acid -- 24 24 Anhydrous Ammonia 1 6 7 Methyl Alcohol -- 4 4 Xylene -- 2 2

(1) Listed in either Regulatory Guide 1.78 or EPA's Extremely Hazardous Substance List. (2) Railcars per year. VYNPS DSAR Revision 0 2.0-25 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Location Map Mile Radius Figure 2.2-1 VYNPS DSAR Revision 0 2.0-26 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Location Map Mile Radius Figure 2.2-2 VYNPS DSAR Revision 0 2.0-27 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Location Map Mile Radius Figure 2.2-3 VYNPS DSAR Revision 0 2.0-28 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Station Site - Area Population Distribution 5-Mile Radius - Year 2000 Figure 2.2-6 VYNPS DSAR Revision 0 2.0-29 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Station Site - Area Population Distribution 10-Mile Radius - Year 2000 Figure 2.2-7 VYNPS DSAR Revision 0 2.0-30 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Station Site - Area Population Distribution 50-Mile Radius - Year 2000 Figure 2.2-8 VYNPS DSAR Revision 0 2.0-31 of 120 2.3 METEOROLOGY 2.3.1 General The general climatic regime of the site area is that of a continental type

with some modification from the maritime climate which prevails nearer the

coast. Of special importance from an engineering standpoint is a temperature

range of 133 o F for the period of record; extremes in annual snowfall, which may be as little as 30 inches or as much as 118 inches; occasional ice storms;

occasional severe thunderstorms; occasional heavy rains due to hurricane

influences; and the possibility of an occasional tornado. These and other

pertinent meteorological data are presented in the following subsections.

Table 2.3.1 indicates the elements, station of record and lengths of record

that were utilized in the analyses.

The site meteorological monitoring program is the most important source of

additional information obtained since the submittal of the Plant Design and

Analysis Report (PDAR).

2.3.2 On-site Meteorological Programs

An initial data collection program was undertaken at the site of the Vermont

Yankee Atomic Power Station to provide information on meteorological

conditions for dispersion analysis for the PDAR. Data from one year, from

August 1, 1967 through July 31, 1968, were evaluated and formed the basis for

those analyses. Appendix G contains a discussion of the August 1967 - July

1968 data collected from the initial monitoring program.

An upgraded on-site monitoring program which meets the intent of Revision 0 to

Regulatory Guide 1.23 was installed in early 1976 and is currently in

operation. A description of this upgraded system is also presented in

Appendix G, along with wind and stability data summaries for one year of

operation.

2.3.3 Diffusion Climatology

The river valley location of the site exerts a strong influence on wind

distribution. As seen in the various wind roses of Appendix G, the channeling

effect of the valley is readily apparent. However, there is no appreciable

difference in wind distribution during poorer dispersion conditions.

VYNPS DSAR Revision 0 2.0-32 of 120 The PDAR stated inversion frequency estimates on the basis of C. R. Hosler's tabulations (1). These are repeated below together with the corresponding values determined from the site preoperational meteorological program. As

shown, the maximum difference occurred in the spring season.

Inversion Frequency (% of total hours) Hosler's Estimates Site Meteorological Program Winter 3337 Spring 2642 Summer 3137 Fall 3636

2.3.4 Winds and Wind Loading

At the time the PDAR was submitted, no continuous wind records were available

for the Vernon area. Due to the similarity in terrain, the relatively close

location, and ready availability of information, wind data from Westover, Massachusetts, was presented at that time. The annual and seasonal surface

wind roses from Westover are shown in Figures 2.3-1 and 2.3-2. The annual and

seasonal wind roses are based upon the total possible hours for each time

interval specified and, in each case, add to 100%.

The corresponding annual and seasonal wind roses obtained from the site

monitoring programs are shown in Appendix G. The several sets of wind roses

show the same channeling effect due to topographical similarities.

The minimum allowable resultant wind pressure (2) at 30 feet suggested by the National Bureau of Standards for the Vernon area is 25 lb-ft -2. This value was used as the general facility design basis. 2.3.5 Temperature and Precipitation

2.3.5.1 Temperature Temperature data (3,4,5) from the records of Vernon (one-half mile south) and Brattleboro (6 miles north) should be representative of the values for the

site and are shown in Table 2.3.2.

VYNPS DSAR Revision 0 2.0-33 of 120 The mean number of days with temperatures greater than 90°F or less than 32°F for Vernon (1951-1960) are as follows: 2.3.5.2 Precipitation

Precipitation (6) at the site averages 43 inches per year and is distributed rather evenly throughout the 12-month period. Snowfall is moderately heavy on

the average; but there is considerable variation in amounts from season to

season. Nearly all winter precipitation is in frozen form, although not

entirely as snow. Sleet and freezing rain are not uncommon.

A summary of precipitation statistics (5,7) for Vernon is shown in Table 2.3.3.

Intense rainfall will be produced by the occasional severe thunderstorm or

modified hurricane. The maximum (8,9) recorded rainfall (inches) for short time intervals at Concord, New Hampshire, is given below:

Minutes Hours 5 10 15 30 60 2 3 6 12 24

0.66 1.12 1.60 2.53 2.71 2.73 3.56 3.82 5.53 5.97

The return period of extreme short-interval rainfall is a useful

design-and-planning guide. The nearest location for which return data are

available and which should be reasonably representative for the Vernon area is

Concord, New Hampshire. These data are shown in Figure 2.3-3.

2.3.5.3 Snowfall, Snow and Ice Loading

The site being located in the northeastern part of the United States is

subjected to a wide range of snowfall, which may be as little as 30 inches or

as much as 118 inches (5,10). Average snowfall statistics for Vernon (25 years of record) are considered to be representative of the site.

• More than 0 but less than 0.5 Jan Feb Mar A pr May Jun Jul A ug Sep Oct Nov Dec A nn >90° 0 0 0 0
• 36310 0 013<32° 30 28 29 14 6
• 00213 23 30175 VYNPS DSAR Revision 0 2.0-34 of 120 The most significant departure from the historical values occurred in the amount of snowfall at Vernon between November 1968 and February 1969.

Snowfall during this period amounted to 80.2 inches compared with an average

for this period of 45.9 inches. The heaviest monthly snowfall was 42.7 inches

and occurred in February. This compares with a historical average value of

15.7 inches. However, the maximum annual snowfall of 118 inches was not

exceeded.

Average Monthly Snowfall (inches) for Vernon

Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec Ann

16.4 15.7 12.1 2.1 0.1 0.0 0.0 0.0 0.0 T 3.3 10.5 60.0

T = Trace

Snow load data (11) available from a Housing and Home Finance Agency (HHFA) study conducted in 1952 is as follows:

Weight of Seasonal Weight of Maximum Weight of Estimated

Snowpack Equaled or Snowpack Maximum Accumulation on Ground Plus

Exceeded 1 year in 10 of Record Weight of Maximum Possible Snowstorm 30 lb-ft -2 50 lb-ft -2 70 lb-ft -2 Data relating to freezing rain and resultant formation of glaze ice (12) on highways and utility lines are available from the following studies:

American Telephone and Telegraph Company, 1917-18 to 1924-25 Edison Electric Institute, 1926-27 to 1937-38 Association of American Railroads, 1928-29 to 1936-37 Quartermaster Research and Engineering Command, U.S. Army, 1959

The U.S. Weather Bureau also maintains annual summaries. The conclusions

reached from these several sources are sometimes contradictory, but the

following is probably a fairly accurate description of the glaze-ice

climatology of southern Vermont.

VYNPS DSAR Revision 0 2.0-35 of 120 The most typical synoptic condition for glaze formation or freezing rain in the northeastern United States is a polar front wave with an active warm front

moving in the north or northeasterly direction toward the region. A high

pressure area almost always is found north of New England, with the center of

the ridge or high pressure cell usually located somewhere northeast of

Newfoundland. This distribution causes a flow of cold continental-polar air

over the area from the north or east, and warm maritime-tropical air up from

the south behind the warm front.

In this situation, the over-running maritime-tropical air is frequently warmer

than 32 o F, while the cold continental-polar air beneath the front has temperatures from 20 o to 30 o F, and a situation almost ideal for the formation of freezing rain or drizzle results. The Vermont site is situated on the

northern edge of the "glaze belt" which extends from southern New England

west-southwest to Ohio and then curving down into Texas. The following data

will apply:

1. Times of occurrence - November through April,

2. Average frequency without regard to ice thickness 10 storms per year,

3. Duration of ice on utility lines - 20 hours (mean) to 55 hours (maximum of record), 4. Return periods for freezing rain storms producing ice of various thicknesses are:

Ice 0.25 inch every year 0.50 inch every year 0.75 inch at least every 3 years

A U.S. Weather Bureau summary for the years 1939-48 give the actual number of

days with freezing rain (without regard to ice formation) for Concord, New

Hampshire, as follows:

Total Days in Nov Dec Jan Feb Mar Apr 10 Years

2 24 29 23 16 1 95

2.3.6 Storms

2.3.6.1 Thunderstorms

Some localized wind damage occurring with the passage of thunderstorm line

squalls may be experienced each year. Extreme wind data (13) for Westover, Massachusetts, is shown in Table 2.3.4. VYNPS DSAR Revision 0 2.0-36 of 120 The "Index of Wind Damage Potential" (excluding tornado, hurricane, and

tropical storm and hail)(defined in units of 1000ths of 1% of residential

property value per year) for the Vernon area is 12 compared to a value of 16

for the Oklahoma-Kansas area.

Heavy precipitation is usually associated with severe thunderstorms and

modified hurricanes. The maximum in 24 hours for Vernon (62 years of record)

is listed below. (8) Maximum in 24 Hours (inches)

Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec

2.21 2.89 4.35 2.49 3.23 3.50 3.80 4.35 3.99 3.57 3.13 2.21

2.3.6.2 Hurricanes

Unusual heavy precipitation (14) was associated with hurricane Connie (August 11-14, 1955) and Diane (August 17-20, 1955). Mass rainfall tables for Birch

Hills, Massachusetts (approximately 28 miles south) are in Table 2.3.5.

In "Index of Hurricane and Tropical Storm Damage Potential" (defined in units

of 1000ths of 1% of residential property value per year) for the Vernon area

is 140 as compared to 337 for the Cape Cod area, 606 for the Cape Hatteras, North Carolina area, and 633 for the Miami, Florida area. The decrease in the

index of hurricane potential as one moves northward is indicative of the

decreased intensity of the hurricane due to several physical reasons. Being

cut off from the major source of energy (the ocean) as a hurricane proceeds

northward, it diminishes in intensity. Topography also causes frictional drag

the farther the storm travels over land, thereby reducing the storm's

magnitude.

2.3.6.3 Tornadoes

Severe storms such as tornadoes (15) are not numerous, but they do occur occasionally. Most tornadoes that occur in New England occur in

Massachusetts.

Massachusetts Vermont New Hampshire Total Number of

Tornadoes (1916-1958) 56 11 15

              (1959-1965)        23              12               22 

VYNPS DSAR Revision 0 2.0-37 of 120 The apparent increase in tornado activity is probably due to increased population and more and better observing and reporting facilities and

techniques.

The monthly distribution (1916-1965) of tornadoes for the tri-state area is as

follows: Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec Total

Massachusetts 12 15 26 9 6 6 3 2 79

Vermont 1 1 3 7 8 2 1 23

New Hampshire 6 8 14 6 2 1 37

In the period 1916 through 1965, Bennington County, Vermont, has reported only

2 tornadoes. Cheshire County, New Hampshire, reported 8, and Franklin County, Massachusetts reported 9 for a total of 19 tornadoes for the immediate area.

The "Index of Tornado Damage Potential" (defined in units of 1000ths of 1%

residential property values per year) for the tri-county area is 1 as compared

to a value of 33 in "tornado alley" (Oklahoma-Kansas-Nebraska).

Thom (16) divides the United States into 1-degree squares and determines the tornado frequency for each square. Using data from 1953-62, Thom records 12

tornadoes occurring within a 1-degree square (about 3 million acres)

encompassing the Vernon site. A mean recurrence interval for a tornado

striking a point within this 1-degree square was calculated to be 1040 years.

This seems reasonable if one considers that only 12 tornadoes were reported in

about 3 million acres in a 10-year period.

Even though the probability of a tornado at the site is small, all structures

and equipment necessary for the safe storage of irradiated fuel are designed

to withstand short-term loadings resulting from 300 mph tornadic winds and an

external pressure drop of 3 psi in 5 seconds.

2.3.7 Conclusions

The meteorology of the site is basically that of a continental type with some

modification from the maritime climate which prevails nearer the coast. The

annual frequency of inversion was determined to be 39%, within the 30% to 40%

range predicted in the PDAR.

The average annual wind speed for the site is 7.5 mph and the most frequent

direction is NNW, the downriver direction. The river valley location leads to

a channeling of the winds.

VYNPS DSAR Revision 0 2.0-38 of 120 In summary, the site meteorological program substantiates the preliminary conclusions. No changes in the previously described protection features were

necessary as a result of meteorological considerations.

2.3.8 References

1. "Low-Level Inversion Frequency in the Contiguous United States," Charles R. Hosler, Monthly Weather Review, Vol. 89, No. 9, September 1961, pp.

319-339.

2. "Wind Pressures in Various Areas of the United States," Building Materials and Structures, Report 152. National Bureau of Standards, 1959.

3. Climatological Data, New England, July Issue for 1962-1965 (four publications), U.S. Weather Bureau.

4. Climatic Summary of the United States - Supplement for 1931 through 1952, New England, U.S. Weather Bureau.

5. Climatic Summary of the United States - Supplement for 1951 through 1960, New England, U.S. Weather Bureau.

6. "Rainfall Intensity - Duration - Frequency Curves", Technical Paper No.

25, U.S. Weather Bureau, 1955.

7. Climatological Data - New England 1961-65, Monthly Issues, U.S. Weather Bureau (60 publications).

8. "Maximum 24-Hour Precipitation in the United States," Technical Paper No.

16, U.S. Weather Bureau.

9. "Maximum Recorded United States Point Rainfall for 5 Minutes to 24 Hours," Technical Paper No. 2, U.S. Weather Bureau.

10. Climatological Data, New England, July Issues 1961-1965, U.S. Weather Bureau (five publications).

11. "Snow Load Studies," Housing Research Paper 19, Housing and Home Finance Agency, 1952.

12. "Glaze, Its Meteorology and Climatology, Geographical Distribution, and Economic Effects," Quartermaster Research and Engineering Center, 1959.

VYNPS DSAR Revision 0 2.0-39 of 120 13. Climatological Data, National Summaries (1959-60-61-62-63-64-65), U.S. Weather Bureau.

14. "Hurricane Rains and Floods of August 1955, Carolinas to New England," Technical Paper No. 26, U.S. Weather Bureau.

15. "Tornado Occurrences in the United States," Technical Paper No. 20, U.S.

Weather Bureau.

16. "Tornado Probabilities," H.C.S. Thom, Monthly Weather Review, U.S.

Weather Bureau, Washington, D.C., October-December 1963, pp. 730-736 VYNPS DSAR Revision 0 2.0-40 of 120 TABLE 2.3.1

METEOROLOGY RECORD

Weather Element Station Record Period

Temperature Brattleboro 11 years

Vernon 10 years

Precipitation Vernon 62 years

Snowfall Vernon 25 years

Surface Wind Westover, MA 22 years

Surface Wind Vernon 1 year

Stability Class Vernon 1 year VYNPS DSAR Revision 0 2.0-41 of 120 TABLE 2.3.2

TEMPERATURE DATA FOR THE VERNON AREA

Vernon (1951-1960)

Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec

Mean Daily

Maximum 34 37 43 58 70 78 83 82 74 63 50 37

Mean Daily

Minimum 11 13 22 33 41 52 56 54 47 36 27 15

Mean 22 25 33 46 56 65 70 68 60 50 39 26

Extreme Maximum 55 63 70 87 91 98 99 100 100 88 76 62

Extreme Minimum -30 -33 -11 14 22 31 41 36 24 19 0 -24

Brattleboro (1931-1941)

Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec

Mean Daily

Maximum 33 34 42 57 70 78 83 81 73 61 48 35

Mean Daily

Minimum 11 10 20 32 42 52 57 55 47 36 27 18

Mean 22 22 31 44 56 65 70 68 60 49 37 26

Extreme Maximum 67 61 77 90 94 97 98 98 94 87 75 64

Extreme Minimum -22 -22 -16 13 26 35 37 34 26 11 -6 -20

NOTE: Temperatures in degrees Fahrenheit. VYNPS DSAR Revision 0 2.0-42 of 120 TABLE 2.3.3

PRECIPITATION DATA FOR THE VERNON AREA

         (1)Mean No. of (2)Mean No. of (2)Extreme (2)Extreme            Days with      Days with                  Monthly     Monthly (4) Maximum            0.10 Inch      0.50 Inch (3)Mean       Min.        Max.          in 24 Hrs.

Month or More or More (inches) (inches) (inches) (inches)

December 6 3 3.44 1.03 6.38 2.21

January 7 2 3.37 0.58 5.73 2.21

February 6 2 2.71 1.28 4.75 2.89 Winter 19 7 March 7 3 3.57 1.71 8.35 4.35

April 8 3 3.86 2.27 6.14 2.49

May 8 2 3.85 1.30 5.84 3.23 Spring 23 8

June 7 2 3.70 1.55 6.73 3.50 July 7 2 3.78 1.32 5.81 3.80

August 6 2 3.68 0.83 8.99 4.35 Summer 20 6

September 7 3 4.19 1.65 7.60 3.99

October 6 3 3.01 1.42 8.63 3.57 November 8 4 4.01 1.99 6.61 3.13 Fall 21 10

Annual 83 31 43.17

(1) 7 years of record (2) 10 years of record (3) 1931-1960 (4) 62 years of record

VYNPS DSAR Revision 0 2.0-43 of 120 TABLE 2.3.4

WINDS DURING THUNDERSTORMS

Westover, MA

Max. Winds (1) Peak Gusts (2) (from hourly obs.) (from daily obs.)

Speed Speed

Month Direction (knots) Direction (knots)

Jan S 52 NW 55 Feb NNW 40 S 58 Mar NE 63 NNW 50 Apr S 36 ENE 61 May SSE 38 NNW 48 June S 28 NNW 49 July NNW 28 W 47 Aug NNE 47 NNE 62 Sept NNE 52 N 60 Oct NW 37 SSE 49 Nov N 39 ENE 69 Dec N 39 NNW 54

(1) For period Apr 1941 through Dec 1963. (One minute sustained wind)

(2) For period Jan-Apr 1946, Jan, Feb, Apr, June, July 1949, Jan, Apr 1950 through Dec 1963 VYNPS DSAR  Revision 0 2.0-44 of 120 TABLE 2.3.5 RAINFALL DATA FROM HURRICANE CONNIE 

Birch Hill, MA Amherst, MA

Time (Accumulative Inches) (Accumulative Inches)

Aug. 11 - 6 AM 0.05 12 N 0.06 6 PM 0.14 2.15 12 M 0.24 2.25

Aug. 12 - 6 AM 1.00 3.35 12 N 1.40 3.90 6 PM 1.45 4.07 12 M 1.60 4.40

Aug. 13 - 6 AM 2.15 4.90 12 N 2.30 5.91 6 PM 4.30 7.65 12 M 6.30 7.70

Aug. 14 - 6 AM 6.39 7.70 12 N 6.40 7.70 6 PM 6.48 7.70 12 M 6.48 7.70

Aug. 15 - 6 AM 7.72 12 N 7.72 6 PM 7.72 12 M 7.73

VYNPS DSAR Revision 0 2.0-45 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Station Site - Westover AFB, Massachusetts Area- Annual Surface Windrose Figure 2.3-1 VYNPS DSAR Revision 0 2.0-46 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Station Site - Westover AFB,Massachusetts Area-Seasonal Surface Windroses (Winter - Spring - Summer - Fall) Figure 2.3-2 VYNPS DSAR Revision 0 2.0-47 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Station Site - Concord, New Hampshire Area-Return Seasonal Surface Windroses Period of Rainfall - (For extremely short intervals) Figure 2.3-3 VYNPS DSAR Revision 0 2.0-48 of 120 2.4 HYDROLOGY AND BIOLOGY 2.4.1 General

The site is at mile 138.3 above the mouth of the Connecticut River, located on the west bank of the river, on the pond formed by the Vernon Dam and

Hydroelectric Station, licensed by the Federal Energy Regulatory Commission as

Project No. 1094. The site is about 3,500 feet upstream from the Vernon

Hydroelectric Station, on the same side of the river. The Vernon

Hydroelectric Station is the furthest downstream of a series of six

hydroelectric projects totaling over 456,000 kW on the river. Storage

reservoirs, whose contents total over 330,000 acre-feet, are also usable for

power generation.

Three of the dams, at 32, 75, and 132 miles above the site, are relatively low structures developing heads of from 29 to 62 feet, with small amounts of

pondage. The large storage reservoirs are from 150 to 260 miles upstream from

Vernon. 2.4.2 Land Area Ground Hydrology

2.4.2.1 Introduction

The river in this general reach comprises a series of ponds formed by several dams constructed for the generation of hydroelectric power.

There is sufficient groundwater in the area to provide wells for public as well as private use.

2.4.2.2 Surface Water

All local streams in the area drain to the Connecticut River, and the site is in the direct path of natural drainage to the east from the local watershed.

Surface drainage will flow toward the river.

2.4.2.3 Groundwater

2.4.2.3.1 Regional Area

There are several municipalities in the vicinity in which groundwater is utilized as one source of water supply. These are listed in Tables 2.4.3 and

2.4.4 and shown in Figure 2.4-1. Private wells in the vicinity of the site

are listed in Table 2.4.5 and shown in Figure 2.4-2.

VYNPS DSAR Revision 0 2.0-49 of 120 2.4.2.3.2 Site Area The local water table level fluctuates differentially depending on the amount of precipitation. It is affected by level changes in the Connecticut River.

River flooding will cause a temporary reversal in the flow direction of

groundwater, so that the local water table will be considerably higher than

usual during periods when the river level is high. Natural subsurface

drainage is over the rock surface.

In 1988 and 1989, groundwater monitoring wells were established throughout the site area. Groundwater levels varied between about 5 feet to 18 feet below

ground surface in the northern portion of the site. In the vicinity of the

major plant structures, groundwater was determined to be about 20 feet below

ground surface. Along the southern portion of the site, depth to groundwater

was about 30 feet. Although these levels do vary throughout the year, they do

provide a general indication of site area groundwater levels.

Hydraulic gradients, as computed from water level elevations measured in monitoring wells, bedrock water supply wells and the river, demonstrate that

groundwater flow in the overburden and bedrock is from west to east. Vertical

hydraulic gradients indicate vertically downward groundwater flow from the

shallow soils to the underlying lower sand deposit, and vertically upward flow

from the bedrock to the overlying lower sand deposit. These data indicate

that groundwater discharges into the river.

Current groundwater monitoring requirements are specified by the VYNPS Radiological and Non-Radiological Environmental Monitoring Programs and

associated implementing procedures.

2.4.3 Hydrology

2.4.3.1 Introduction Under normal conditions, the flow of river water is largely determined by

operation of the hydroelectric stations and by the upstream reservoirs and

lakes. 2.4.3.2 Stream Flow Connecticut River flow is monitored at Vernon Dam. Records were published by

the United States Geological Survey from 1944 to 1973 when their gage at

Vernon was discontinued. Drainage area at Vernon Dam is 6,266 square miles.

Nearby gages on the Connecticut River include N. Walpole, New Hampshire and Turners Falls, Massachusetts. Their continuous periods of record are from

1942 and 1915 to the present at N. Walpole and Turners Falls, respectively.

The drainage areas at these two gages are 5,493 and 7,163 square miles.

VYNPS DSAR Revision 0 2.0-50 of 120 Table 2.4.1 shows average and extreme values of monthly stream flow plus minimum weekly flows for the Connecticut River below Vernon Dam for a 44-year

period of record (1944-1988). These stream flows were compiled using the

measured stream flows at Vernon from 1944 to 1973 and the generated stream

flows for Vernon using stream flow data from the two nearby gages for the

period 1973 to 1988.

Table 2.4.2 shows daily stream flows measured below Vernon for the period October 1964 through September 1965.

2.4.3.3 Temperature

River temperatures have been measured along the river at six sampling stations. The uppermost is at a point just downstream from Brattleboro, some

4-1/2 miles above the site, and the farthest downstream is at a point just

upstream of the Schell Bridge in Northfield, Massachusetts, some 6-1/2 miles

below the site.

Temperature measurements were made below Vernon Dam, in the general area of the permanent monitoring Station 3, to determine how and where best to

establish this station for reliable, consistent results. A similar

temperature monitoring station (Station 7) has been established upstream of

the station circulating water intake. Continual records from these two

temperature monitoring stations are submitted annually to the Vermont Agency

of Natural Resources. The locations of these monitoring stations are

presented in Section 2.4.5.

In addition to the records presently being obtained at the site, temperature records have been kept for a number of years at the Bellows Falls

Hydroelectric Station, some 32 miles upstream. River temperatures also have

been recorded over a period of several years at the Cabot Hydroelectric

Station, a unit of the Turners Falls Hydroelectric Project, downstream from

the site.

2.4.3.4 Floods

The flood of March 19, 1936, was the greatest and most destructive flood on this reach of the river. The discharge on that day was 176,000 cfs, reaching

a river stage at Vernon of 231.4 feet MSL. Other major floods were those of

November 5, 1927, 155,000 cfs at elevation 229.0 feet MSL; and

September 22, 1938, 132,500 cfs at elevation 226.6 feet MSL.

Since the floods of 1936-1938, extensive flood control works, consisting of some five projects with 247,800 acre-feet of flood storage, have been designed

and constructed by the Corps of Engineers in the Connecticut River Basin

upstream from the Vernon Dam.

VYNPS DSAR Revision 0 2.0-51 of 120 The Probable Maximum Flood (PMF) on the Connecticut River Basin above Vernon, Vermont (drainage area of 6,266 square miles) was determined using procedures

and information contained in the analytical studies for the Susquehanna River

(1). In addition, a stage-discharge curve was developed through step

backwater computations to determine the PMF elevation at the site.

In this study, the major emphasis was in the direction of conservatism. The following conservative assumptions were made:

1. The maximum persisting 12-hour, 1000-millibar (mb) dew point temperature of record is used as an index of the maximum precipitable water.

Furthermore, the 12-hour maximum persisting dew point was used throughout

the 72-hour rainfall period.

2. The unit of time selected for the unit hydrograph is 6 hours, although for a basin area of 6,266 square miles and a lag time of 75 hours, characteristic of the Connecticut River at Vernon, a more realistic unit

of time for the unit hydrograph would be 12 hours.

3. An infiltration rate of 0.05 inches per hour is assumed throughout the rainfall period, although the recorded range for this particular basin is

0.05-0.10 inches per hour.

4. A baseflow of 58,800 cfs, which is about 5.7 times the average discharge and greater than the annual peak discharge recorded in four of the 29-year

period of record, and about twice the value which is normally used. Enveloping curves of PMP for 6, 12, 24, 48, and 72 hours were obtained by adjustment of the depth-area-duration curves for the Susquehanna River Basin (Figure 2.4-3). The adjustment is based on the precipitable water in the

1,000-200-mb air column (2) for the maximum persisting 12-hour, 1,000-mb dew

point of record (3). By applying the maximum persisting 12-hour, 1,000-mb dew

point of record and assuming this condition persists for an additional

60 hours (the PMP duration is 72 hours), there is a considerable amount of

conservatism in deriving the PMP for the basin.

At Harrisburg, Pennsylvania, the record maximum persisting 12-hour, 1,000-mb dew point of 75.3 °F is equivalent to 2.93 inches of precipitable water, while at Vernon, Vermont, it is 73.3 °F or 2.62 inches of precipitable water. The 6-hour increments of PMP and runoff amounts, based on an infiltration rate of 0.05 inches per hour, are presented in Table 2.4.6.

The 6-hour unit hydrograph (Figure 2.4-4) was derived from the Standard Project Flood (SPF) hydrograph developed by the New England District of the

Corps of Engineers (Figure 2.4-5) by:

VYNPS DSAR Revision 0 2.0-52 of 120

1. Separating the base flow and snowmelt from the total flow to obtain the flood flow due to rainfall runoff.

2. Computation of the rainfall runoff (4.5 inches).

3. Dividing the ordinates of the SPF net flow by the rainfall runoff.

4. Conversion of the resulting 24-hour unit hydrograph to a six-hour unit hydrograph by the S-curve technique (4).

The resulting hydrograph is a 6-hour unit hydrograph for the entire 6,266-square mile basin. The natural PMF hydrograph was then derived by

multiplying the ordinates of the 6-hour unit hydrograph and the 6-hour values

of rainfall runoff, summing the subtotals, and adding back the base flow

(58,800 cfs). The resulting natural PMF hydrograph is presented in

Figure 2.4-6.

There are five flood control storage reservoirs in the Connecticut River Basin above Vernon. The total storage capacity of the reservoirs is

247,800 acre-feet, which represents a rainfall runoff over the total basin of

0.74 inches. Locations of these reservoirs are shown in Figure 2.4-7. The

storage capacity of each reservoir is:

1. Union Village 38,000 acre-feet

2. North Hartland 71,400 acre-feet

3. North Springfield 50,600 acre-feet

4. Ball Mountain 54,600 acre-feet

5. Townshend 33,200 acre-feet The operation of these flood control facilities has reduced the flood threat in the basin. For instance, the Corps of Engineers estimates that the SPF

natural peak discharge of 263,700 cfs at Vernon has been reduced to

225,000 cfs for a net reduction of 38,700 cfs.

As stated above, the operation of current flood control facilities has reduced the SPF at Vernon from a natural peak of 263,700 cfs to 225,000 cfs, or a

reduction of 38,700 cfs. If this same reduction were applied to the PMF, the

peak discharge would be decreased from 506,400 cfs to 469,700 cfs.

However, for conservatism, it is assumed that due to antecedent conditions, the entire 247,800 acre-feet of storage capacity upstream is not available for

regulation of the PMF. Therefore, assuming that about 68% of the SPF

reduction would go into storage, the modified PMF discharge becomes

480,100 cfs. The resulting modified PMF hydrograph is shown in Figure 2.4-6.

VYNPS DSAR Revision 0 2.0-53 of 120 The stage-discharge curve at the VYNPS site at Vernon was determined by the standard step backwater method as described by Chow (5) utilizing the Ebasco

Backwater Calculation with Bridge Loss programmed for implementation on a

Burroughs 5500 computer. The recorded water surface profiles for applicable

floods of record (Table 2.4.7) were used as a basis for selecting roughness

coefficients, "n". The following "n" values were found to yield excellent

agreement with recorded flood profiles:

River Reach Low High Channel Overbank Above Vernon Dam 0.030 0.033 0.040 At Vernon Dam 0.013 0.013 0.013 Below Vernon Dam 0.030 0.033 0.050

River and valley cross sections upstream from Vernon Dam to the plant site and downstream to the Central Vermont Railroad Bridge at Northfield, Massachusetts, which were used for the step backwater computation are located

in Figure 2.4-8. The final rating curve for the plant site is shown in

Figure 2.4-9.

The time-varying PMF stage-discharge relationships are listed in Table 2.4.8 for the natural flood hydrograph and in Table 2.4.9 for the hydrograph as

modified by existing flood storage. Based on the PMF hydrograph modified for

existing flood storage, the PMF stillwater level at the site is

252.5 feet MSL.

As a check on the design flood for the site, failure of the largest upstream flood control reservoir, Townshend Reservoir, was postulated to occur as a

result of an earthquake, which, in turn, occurs simultaneously with the SPF.

For conservatism, the maximum inflow of 71,000 cfs for this reservoir, which

is located about 22 miles upstream from Vernon, as shown in Figure 2.4-7, was

considered to be translated downstream and directly added onto the SPF peak

discharge. This coincident dam failure with the SPF modified peak discharge

of 225,000 cfs would produce a peak discharge of 296,000 cfs. From

Figure 2.4-9, a peak discharge of 296,000 cfs would produce a maximum

stillwater elevation at the site of 240.8 feet MSL.

The dam failure analysis described above was originally developed as a check to ensure that the controlling flood for the site was the

precipitation-induced PMF. Since completion of the above upstream dam failure

analysis, additional information on flooding at the site due to failure of

upstream flood control and hydropower dams has been developed by the dam

owners and is summarized below. These more recent studies are based on

different criteria and analysis techniques than the previously described

analysis.

VYNPS DSAR Revision 0 2.0-54 of 120 There are several large dams on the Connecticut River upstream of the VYNPS site. The owners of these dams are required by the Federal Energy Regulatory

Commission to perform dam failure analysis as input to the development of

Emergency Action Plans. The only upstream dam failure flood that reaches the

VYNPS site for these Connecticut River dams is that for the Moore Dam. The

impacts for the other dam failures terminate well upstream of the site.

The hypothetical failure of Moore Dam was assumed to coincide with the peak of the PMF inflow hydrograph. The dam is about 145 miles upstream from the VYNPS

site. Four downstream dams, Comerford, McIndoes, Dodge Falls and Wilder, were

assumed to fail in cascade. The results of the Moore Dam failure analyses at

Vernon Dam are a peak inflow of 305,600 cfs and a peak flood elevation of

240.1 feet MSL. The VYNPS site is subject to the same flood elevation as the

Vernon Dam. The arrival time at the site for the leading edge of the Moore

Dam failure flood wave is about 22 hours after the postulated failure of the

dam. The time of the peak flood at the site is about 47 hours after the

postulated dam failure.

There are also five flood control reservoirs on Connecticut River tributaries, upstream of the VYNPS site. The owners have developed dam breach profiles for

each of the five dams. A review of these analyses showed that the impacts of

dam failure for three of the dams, Union Village, North Hartland, and North

Springfield do not reach the VYNPS site. Two of the dams, Townshend and Ball

Mountain, do produce flood levels downstream that reach the site. Both of

these dams are located on the West River, which is a tributary of the

Connecticut River.

For an assumed failure of Townshend Dam, the peak stage at Vernon Dam is elevation 230 feet MSL. The time from the start of dam failure until the peak

stage is reached at the VYNPS site is 9.2 hours. The time from the start of

dam failure until the initial rise at the site is 5.2 hours. This analysis

used assumed pre-breach high flows in both the West and Connecticut Rivers.

For an assumed failure of Ball Mountain Dam, the peak stage at Vernon Dam is elevation 235 feet MSL. The Ball Mountain Dam is upstream of the Townshend

Dam. The Townshend Dam fails as a result of the assumed failure of the Ball

Mountain Dam. The time from the start of dam failure until the peak stage is

reached at the VYNPS site is 10.0 hours. The time from the start of dam

failure until the initial rise at the site is 7.6 hours. This analysis also

assumed pre-breach high flows in both the West and Connecticut Rivers.

In summary, the flood levels at the VYNPS site due to upstream dam failures are well below the PMF level at the site.

VYNPS DSAR Revision 0 2.0-55 of 120 The maximum PMF stillwater level at the VYNPS site at Vernon, Vermont was computed to be 252.5 feet MSL occurring 96 hours after the beginning of the

72-hour probable maximum precipitation period. Additional consideration is

now given to the problem of wave runup.

Atomic Energy Commission Safety Evaluation Docket No. 50-271 dated June 1, 1971 has been reviewed during the NEI 12-07 Fukushima Flooding

evaluation and is considered the governing document. Page 12 of this document

concludes, "The PMF will produce a maximum discharge of 480,000 cfs at the

site and a corresponding stage of 252 feet 6 inches MSL. This maximum occurs

eight days after the start of the rainfall causing the flood. We consider it

possible that another storm or synoptic weather system with sustained winds of

at least 45 mph could follow the original storm and be at the site at the same

time that the peak discharge occurs. If the winds came from the most

effective direction, waves two to four feet high could result. These waves

would break at the river bank, but could produce plant flooding at elevations

as high as 254 feet MSL.

Nominal plant grade is 252.0 feet MSL. Accesses to the Turbine, Reactor, Radwaste, and Control Buildings from out of doors are at grade 252.5 feet MSL.

In addition, direct access to the Reactor Building from out of doors is

through a pair of leak-tight doors.

Normal fuel pool cooling will be maintained during a maximum probable flood

until service water is lost due to river water leakage into the intake

structure or normal power is lost.

If normal electrical power is unavailable, diesel generators are available to

supply back-up power.

If normal fuel pool cooling cannot be maintained during a maximum probable

flood, alternate fuel cooling strategies are available and will be implemented

in accordance with applicable facility procedures.

The PMF stillwater level is essentially equal to the top of most yard electrical manholes. A potential avenue of water intrusion into the

Switchgear Room, Elevation 248.5 feet MSL exists through underground conduits

routed from manholes and handholes to the Switchgear Room floor. Should water

enter these manholes, the underground conduits could provide a path for water

to enter the Switchgear Room manholes. If the water level gets high enough, flooding in the Switchgear Room and lower levels of the administration and

Turbine Building could occur. This flooding could affect the operability of

switchgear.

VYNPS DSAR Revision 0 2.0-56 of 120 To preclude, or reduce the amount of water entering the Switchgear Room manholes through the underground conduits which extend from the yard manholes, these conduits have been sealed. In conjunction with the conduit sealing, portable pumping capacity is available on-site to remove water which may enter

the Switchgear Room manholes. Additionally, facility procedures direct

personnel to remove this water as part of the site flood procedures.

Based on our review of these results of the flood analysis, we conclude that

acceptable measures will and can be taken to assure safe storage of irradiated

fuel even in the unlikely event that floods as large as the PMP should occur."

The facility is, therefore, suitably protected against the maximum probable

flood and all lesser floods, including those due to the failure of upstream

dams.

VYNPS DSAR Revision 0 2.0-57 of 120 2.4.4 Uses of River 2.4.4.1 Introduction

The Connecticut River and its ponds are used by industry, chiefly for hydroelectric power generation, and to some extent, by the public for

recreational purposes.

2.4.4.2 Industrial Use

The series of hydroelectric stations and their associated reservoirs on the Connecticut River have been operated for many years to obtain maximum power

benefits for the power consumers of the New England region. This has required

operation of the river's hydroelectric stations as peak load facilities which

were shut down during the low load hours of each day and on weekends.

When river flow rates are less than 10,000 cfs, the Vernon Hydroelectric Station is operated as a peak load facility. Often at such times, only one

hydroelectric unit is utilized during off-peak hours.

VYNPS's NPDES permit defines the maximum allowable thermal limits on the Connecticut River.

Turners Falls Hydroelectric Project, FERC License No. 1889, is located 19.8 miles below Vernon Dam. This project, which utilizes water released from

the Vernon project, is owned and operated by the Western Massachusetts

Electric Company.

2.4.4.3 Public Use

Both Vernon Pond and Turners Falls Pond, next downstream, are used to some extent for canoeing, boating, water skiing, and fishing. The utilization of

fishes resident in the Connecticut River has grown over the past few years.

Finfish have been studied in the Connecticut River in the area near VYNPS

since 1967. Fish were collected by various methods, including seining, gill

netting, minnow traps, fish traps (fyke nets), and electrofishing.

Table 2.4.10 lists, by scientific and common names, all of the species of

finfish taken through 1980 at Stations 2, 3, 4, and 5 on Figure 2.4-11. With

few exceptions, all specimens collected were identified, weighed, measured, and released. Scale samples were taken from selected species for age-growth

studies. During the open cycle testing programs, similar data were collected

on all fish impinged on the traveling screens at the cooling water intake.

Fish data are presented in VYNPS's preoperational report (9), in subsequent

annual reports and in the reports of the open cycle testing programs.

A fish passage facility became operational at the Vernon Hydroelectric Station in May 1981. VYNPS DSAR Revision 0 2.0-58 of 120 There are no direct municipal water intakes downstream of the VYNPS site. Northeast Utilities operates a 1,000,000 kW pumped storage hydroelectric generating plant at Northfield and Erving, Massachusetts. This plant obtains

water from the Connecticut River at a point 14 miles downstream from the

Vermont Hydroelectric Station and pumps the water to an upper reservoir.

During hours of peak electrical demand, this water is allowed to return to the

river through the reversible turbine pump units to provide peaking electrical

generating capacity.

The Metropolitan District Commission of Massachusetts has investigated the feasibility of taking water from the Northfield Mountain Reservoir and

diverting it through a penstock and canal system to the Quabbin Reservoir, approximately 10 miles distant. This reservoir supplies water to Metropolitan

Boston, Clinton, Marlboro, Southboro, Worcester, and other communities in

Massachusetts. No action has been taken on this proposal to date.

2.4.5 Biology

The location of VYNPS biological monitoring stations in the Connecticut River are depicted in Figure 2.4-11. The approximate location of the eight

monitoring stations in river miles north and south of Vernon Dam are shown

below: Station No. Location Relative to Vernon Dam 1 6.45 miles south 2 4.70 miles south 3 0.65 miles south 4 0.55 miles north 5 1.25 miles north 6 4.10 miles north 7 4.25 miles north 8 8.70 miles north 2.4.5.1 Commercial Fisheries

There are no commercial fisheries in the Connecticut River in the Vernon Pool area.

VYNPS DSAR Revision 0 2.0-59 of 120 2.4.5.2 Sport Fisheries Thirty-three species of fishes have been found in the Connecticut River in the vicinity of Vernon Dam. Some of the species, such as smallmouth bass and

yellow perch, are generally considered to be game fishes. Two anadromous

species, the Atlantic salmon and the American shad, have recently been

reintroduced into the Vernon area, as a direct result of the construction of

fish ladders at the Turners Falls and Vernon Dams. Successful passage of

these and other species have been recorded over the last few years. Other

species in the Connecticut River are either forage, coarse food, or "trash" fish and are not generally sought by anglers. These species include the white

sucker and carp. Nearly all of the fish species present are warm-water

tolerant.

A 1980 survey shows that perch (both white and yellow), minnows, white sucker, and bass are the most abundant fish species in the vicinity of the Vernon Dam

as seen in Table 2.4.11. These species comprised about 89% of the fish

population. The average weight of the smallmouth bass captured was

approximately 0.5 pounds, while the weight of the average sucker was nearly

1.5 pounds. Carp, white suckers, and minnows were shown to comprise

approximately 1/3 of the total number of all fishes caught, but accounted for

over 1/2 of the total weight.

2.4.5.3 Bottom Fauna

Monthly samples of Connecticut River benthic fauna were collected at Stations 2, 3, 4, and 5 of Figure 2.4-11, from May through November with a

9-inch Ekman dredge and Henson traps (wire cages filled with 2 to 3-inch

diameter rocks). The following compares the number of samples and number of

genera of benthos collected by Ekman dredges over the years.

COMPARISON OF NUMBER OF SAMPLES AND NUMBER OF GENERA OF BENTHOS COLLECTED BY EKMAN DREDGE Station Number of Samples/Number of Genera Number 1969 1977 1978 1979 1980

2 6/33 8/20 8/22 7/27 7/36 3 6/24 8/25 8/13 7/26 7/39 4 7/16 8/19 8/17 7/26 7/30 5 8/18 8/20 6/14 7/28 7/25

VYNPS DSAR Revision 0 2.0-60 of 120 As has been found in earlier years, caddis fly and chironomid larvae were the predominant organisms in most of the spring and summer samples. Fall samples

showed a greater variety of dominant forms - fingernail clams, planarians, oligochaetes. Chironomids and caddis flies were again dominant in the

November Henson trap samples. The very low Station 2 and 3 diversity indices

in that sample set were attributable to large percentages of a single

chironomid species, Tanytarsus sp., which accounted for 90% of the Station 2 sample and 94% of the Station 3 sample. Large percentages of the chironomid, Glyptotendipes sp., in all three Henson trap samples of July and the Station 5 sample of September are evidenced in the relatively low diversity indices of

those samples.

2.4.5.4 Aquatic Plants Few species of aquatic plants are found in the waters of the Connecticut River in the Vernon Pool area. Marshes adjacent to the river are, however, rich in

vegetation. Cattails are the predominate vascular plant found in these

wetlands; other abundant species are rushes, sedges, grasses, horsetails, and

sweetflag.

2.4.5.5 Conclusions

The waters of the Connecticut River in the Vernon Pool area support a variety of aquatic organisms. The fishes found in these waters are predominantly

those generally referred to as "warm-water" species. The benthic fauna are

generally sparse due to the silty nature of the river bottom. Marshes

adjacent to the river are rich in aquatic vegetation. Safe storage and

handling of irradiated fuel and radwaste management at the VYNPS does not

adversely affect the ecology of the Vernon Pool adjacent to the site.

2.4.6 Chemical and Bacteriological Quality of Water Water quality monitoring requirements are established by the current National Pollutant Discharge Elimination System (NPDES) Permit.

The water above Vernon Dam, as determined by the Vermont Water Resources Board (effective July 2, 2000), has been classified as Class B waters and can be

described as follows:

Class B: The designated uses of Class B waters include aquatic biota, wildlife, aquatic habitat, aesthetics, public water supply, irrigation of crops and other agricultural uses, swimming and other primary contact

recreation, boating, fishing and other recreational uses.

VYNPS DSAR Revision 0 2.0-61 of 120 2.4.7 River Field Program Water quality parameters were monitored continuously from 1968 at Station 3, downstream of VYNPS, and from 1970 at Station 7, upstream of the plant (station locations are described in Section 2.4.5). In February 1980, the

requirement that conductivity and turbidity be monitored continuously was

deleted. The current ecological monitoring program is set forth in the NPDES

Permit which is issued every 5 years. Parameters to be monitored and

associated limits are ultimately established by the State of Vermont and may

be subject to revision within the course of each 5-year period. Data and

analysis from this monitoring program are presented in annual ecological

studies reports.

Biological studies, both qualitative and quantitative, are made to establish the presence and amount of fish and benthic fauna. A comprehensive

environmental assessment is presented in Reference 11.

2.4.8 Conclusions The station site nominal grade level is at elevation 252 feet Mean Sea Level (MSL). The maximum river level that has occurred at the site was

elevation 231.4 feet MSL. The maximum Probable Maximum Flood Level at the

site is 252.5 feet MSL.

The PMF stillwater level is 6 inches above most yard electrical manholes. If flood waters enter these manholes, potential flood pathways through conduits

which extend from the manholes into the Switchgear Room and the Fuel Oil

Transfer Pump Building exist. This potential water pathway through conduits

is significantly reduced by the inclusion and inspection of seals in conduits

entering the Switchgear Room and the Fuel Oil Transfer Pump Building and

measures in the flooding procedure which monitor and address any leakage into

these rooms.

Because the river is the natural low point and drainage channel for the region, the groundwater table can be expected to slope toward the river.

Surface drainage also will flow toward the river. Thus, it is unlikely that

any liquids discharged to the river from the site would mix with domestic

water supplies in the area.

The Federal Energy Regulatory Commission (FERC) requires, under Order No. 122, issued January 21, 1981, that the dams and related structures of all Licensed

Projects be inspected once every five years by an independent consulting

engineer and that they be certified as safe in their construction and

operation. In the event unsafe conditions of any nature are found, under the

order they must be called to the attention of the owner and the FERC and

necessary corrective measures must be carried out. In response to an

exception request dated June 26, 1997, FERC issued an exemption from filing an VYNPS DSAR Revision 0 2.0-62 of 120 Independent Consultant's Safety Inspection Report pursuant to the above regulation for the Vernon, VT and Bellows Falls, VT dams by FERC letter dated

August 6, 1997, on the basis that those projects are "low hazard potential" facilities. The dam owner retains the responsibility to provide an emergency

action plan and an inspection by an independent consultant in the event the

upstream or downstream circumstances of either project change such that

failure of a project structure would present a hazard to the public. The dams

operated by the Corps of Engineers are also subject to periodic safety

inspections. It is believed that these actions will assure the safety of all

dams on the river.

VYNPS DSAR Revision 0 2.0-63 of 120 2.4.9 References

1. Probable Maximum Precipitation Susquehanna River Drainage above Harrisburg, Pennsylvania, Hydrometeorological Report No. 40, U.S. Weather

Bureau, Washington, D.C., May 1965.

2. Seasonal Variation of the Probable Maximum Precipitation East of the 105th Meridian for Areas from 10 to 1,000 Square Miles and Durations of

6, 12, 24, and 48 hours, Hydrometeorological Report No. 33, U.S. Weather

Bureau, Washington, D.C., 1956.

3. Climatic Atlas of the United States, ESSA, U.S. Department of Commerce, 1968. 4. Flood Hydrograph Analysis and Computations, EM 1110-2-1405, U.S. Army Corps of Engineers, 1959.

5. Chow, Ven Te, Open-Channel Hydraulics, Civil Engineering Series, McGraw-Hill, 1959.

6. Technical Paper No. 55, Tropical Cyclones of the North Atlantic Ocean, U.S. Department of Commerce, U.S. Weather Bureau, Washington, D.C., 1965.

7. Shore Protection, Planning and Design, Technical Report No. 4, Third Edition, U.S. Army Coastal Engineering Research Center, Department of

Army, Corps of Engineers.

8. Computing Freeboard Allowances for Waves in Reservoirs, ETL No. 1210-2-8, Department of Army, Corps of Engineers.

9. Webster-Martin, Incorporated, 1971. Ecological Studies of the Connecticut River, Vernon, Vermont. Preoperational Report. Report

prepared for Vermont Yankee Nuclear Power Corporation.

10. U.S. Department of Commerce, 1978. "Tropical Cyclones of the North Atlantic Ocean, 1871-1977," National Climatic Center, NOAA, Asheville, N.C.

11. Aquatec, Incorporated, 1978. "316 Demonstration - Engineering, Hydrological and Biological Information." 12. Aquatec, Incorporated, 1981. Ecological Studies of the Connecticut River, Vernon, Vermont. Report X, January-December 1980. Report

prepared for Vermont Yankee Nuclear Power Corporation.

13. GZA GeoEnvironmental, Inc., 2011. Hydrogeologic Investigation of Tritium in Groundwater, Vermont Yankee Nuclear Power Station Vernon, VT. Report

Prepared for Entergy Nuclear Operations, Vermont Yankee Nuclear Power

Station. VYNPS DSAR Revision 0 2.0-64 of 120 TABLE 2.4.1 Average and Extreme Values of Stream Flow Connecticut River at Vernon, Vermont Water Years 1944 - 1988 Highest Lowest Lowest Average Average Average Average Monthly Monthly Monthly Weekly Flow Flow Flow Flow Cfs Cfs Cfs Cfs

October 6,571 20,201 1,646 1,475 November 9,033 20,450 3,366 2,159 December 9,486 24,326 2,934 2,494 January 7,655 17,338 2,589 2,283 February 8,187 24,428 2,935 2,135 March 15,544 36,245 5,308 4,373 April 30,799 51,210 14,980 11,523 May 18,047 38,790 7,262 3,118 June 8,768 21,890 3,387 2,424 July 4,911 21,790 1,841 1,033 August 4,005 13,615 1,805 1,223 September 4,159 15,610 1,650 1,138

NOTES:

1. All flows reflect regulation of upstream reservoirs for power purposes.

2. Flows measured at Vernon for 1944 - 1973, flows generated for Vernon from United States Geological Survey gages at N. Walpole, New Hampshire, and

Turners Falls, Massachusetts for 1973 - 1988. VYNPS DSAR Revision 0 2.0-65 of 120 TABLE 2.4.2 Vermont Yankee Nuclear Power Station, Daily Stream Flow for October 1964 to September 1965, Connecticut River at Vernon, Vermont

Connecticut River Basin 1-1565. Connecticut River at Vernon, Vermont

Location Lat 42°46'10", long 72 °30'50", on right bank just downstream from Vernon Dam at Vernon, Windham County, and 2 miles upstream from Ashuelot River.

Drainage area

6,266 sq mi.

Records available February to April 1936 (in WSP 798). September and October 1938 (in WSP 867), October 1944 to September 1965.

Gage Water-stage recorder (digital). Datum of gage is at mean sea level, datum of 1929. Prior to January 20, 1948, at datum 94.13 ft higher.

Average discharge

21 years (1944-65), 10,170 cfs (adjusted for storage).

Extremes Maximum discharge during year, 32,000 cfs April 17 (gage height, 190.94 ft):

minimum daily, 108 cfs September 6, 1936, 1938, 1944-65: Maximum discharge, 176,000 cfs March 19, 20, 1936 (gage height, 128.8 ft. datum then in use), from rating curve extended above 86,000 cfs: minimum daily, 99 cfs

October 8, 1944.

Remarks Records good except those below 1,000 cfs, which are fair. Flow regulated by

powerplants and by First Connecticut and Second Connecticut Lakes, Lake

Francis. Moore Reservoir and Comerford Station Pond (see Page 196), and other

reservoirs (combined usable capacity, about 29 billion cubic feet). VYNPS DSAR Revision 0 2.0-66 of 120 TABLE 2.4.2 (Continued) DISCHARGE IN CUBIC FEET PER SECOND. WATER YEAR OCTOBER 1964 TO SEPTEMBER 1965 DAY OCT. NOV. DEC. JAN. FEB. MAR. APR. MAY JUNE JULY AUG. SEPT.

1. 1,220 164 10,900 6,730 4,500 7,000

• 4,700 15,000 4,560 2,950 125 5,790

2. 1,900 3,070 5,580 5,450 3,500 6, 600 4,390 13,600 7,620 2,830 1,710 8,810

3. 167 3,140 4,610 1,180 3,300 7,000 3,290 13,500 4,530 128 1,780 11,100

4. 157 3,050 6,740 4,890 3,400 7,400 826 13,000 *4,370 125 1,320 2,790

5. 3,270 2,210 1,760 7,300 4,900 8,000 4,970 12,300 1,100 125 1,900 134

6. 3,420 2,890 196 6,950 1,100 8,700 7,410 11,900 125 4,120 1,320 108

7. 3,720 725 4,150 6,420 1,000 9,800 8,910 10,200 4,650 2,650 296 3,760

8. 2,550 167 3,840 5,560 7,600 12,500 7,830 9,710 4,790 3,070 128 4,220

9. 2,620 3,400 4,190 5,250 6,800 13,000 10,200 4,630 5,600 3,270 2,640 2,860

10. 167 3,570 3,840 2,850 7,600 13,700 9,650 8,560 5,110 917 3,460 3,470

11. 160 811 3,990 6,770 8,400 13,800 11,900 9,660 4,890 567 3,460 2,210

12. 157 3,820 2,060 *6,760 9,000 12,800 14,300 10,400 1,060 2,290 2,390 628

13. 2,800 3,580 194 7,010 7,800 11,500 24,800 10,700 2,440 1,910 1,920 5,740

14. 2,850 2,920 4,340 6,960 7,400 9,650 23,600 8,840 10,800 2,390 125 6,540

15. 2,900 654 5,100 7,000 6,000 7,670 21,300 4,130 6,310 2,770 125 3,680

16. 2,200 4,990 5,330 2,800 6,300 9,260 24,100 2,400 7,400 2,720 3,900 2,270

17. 759 4,750 5,100 131 6,000 6,870 29,900 7,400 5,310 128 3,020 1,780

18. 737 4,750 5,630 4,300 6,500 7,500 25,400 8,730 5,300 125 4,690 125

19. 2,720 4,790 1,500 4,000 7,000 7,990 20,800 7,870 880 4,030 3,580 125

20. 3,550 5,250 174 3,700 3,500 7,250 19,400 7,050 2,550 3,810 3,080 1,940

21. 5,320 1,080 6,870 3,000 1,750 894 19,400 6,870 4,960 3,670 125 2,530

22. 3,580 167 5,120 3,800 5,000 7,320 *21,000 1,060 5,180 2,300 125 2,850

23. 5,470 5,260 6,980 2,000 5,900 6,610 24,400 134 5,450 2,530 2,280 3,050

24. 174 4,050 3,650 1,500 5,900 6,960 23,600 5,300 5,730 128 1,100 2,900

25. 167 4,460 3,660 4,600 7,000 6,440 19,100 4,910 4,450 125 1,430 9,360

26. 3,440 5,030 7,990 5,000 6,600 6,500 14,000 5,610 1,660 2,200 1,760 5,010

27. 3,620 8,580 15,500 4,100 1,850 3,920 15,400 5,960 759 1,730 1,740 7,960

28. 4,140 11,900 20,000 4,200 800 872 16,800 4,320 5,410 1,450 973 8,410

29. *3,870 12,100 17,400 4,500 ------ 6,770 16,300 1,120 4,940 859 885 7,040

30. 3,970 8,670 13,900 2,100 ------ 6,190 16,700 128 3,740 1,010 2,150 6,960

31. 969 ------ 12,400 600 ------ 5,530 ------ 125 ------ 131 2,420 ------

TOTAL 72,744 119,998 192,694 137,411 146,400 245,996 464,446 225,117 131,674 57,058 55,957 124,150 MEAN 2,347 4,000 6,216 4,433 5,229 7,935 15,480 7,262 4,389 1,841 1,805 4,138

MEAN** 1,942 4,335 6,040 3,867 3,834 6,385 16,490 8,924 4,518 1,809 2,174 4,383 CFSM** .310 .692 .964 .617 .612 1.02 2.63 1.42 .721 .289 .347 .699 IN** .36 .77 1.11 .71 .64 1.17 2.94 1.64 .80 .33 .40 .78 CALENDAR YEAR 1964 MAX 63,200 MIN 143 MEAN 7,840 MEAN** 7,822 CFSM** 1.25 IN 17.00

WATER YEAR 1964-65 MAX 29,900 MIN 108 MEAN 5,407 MEAN** 5,382 CFSM** .859 IN 11.65 Peak discharge (base, 50,000 cfs). - No peak above base. Note: Stage-discharge relation affected by ice January 15, 16, January 18 to March 9.

3* Discharge measurement made on this day.

** Adjusted for change in contents in all reservoirs from First Connecticut and Second Connecticut Lakes to reservoirs in West Riv er basin listed on Page 196.

VYNPS DSAR Revision 0 2.0-67 of 120 TABLE 2.4.3 Municipal and Industrial Groundwater Usage Within a 10-Mile Radius of the Vernon Site

Minimum and Minimum and Static Number Maximum Depth Maximum Yield Level Town of Wells Feet GPM Feet

Brattleboro, VT 49 10-540 1-40 9-35 Guilford, VT 30 50-540 1-50 8-41 Halifax, VT 12 34-345 1-30 8-10 Vernon, VT 14 36-565 1-75 25-115 Chesterfield, NH 26 41-470 0.2-25 12-50 Hinsdale, NH 5 72-280 3-30 -30 Winchester, NH 3 155-473 1.5-20 - Bernardston, MA 3 29-145 0.5-20 - Gill, MA 15 40-345 0.5-15 10-20 Leyden, MA 4 95-208 3-100 - Northfield, MA 19 52-345 0.3-40 14-85 Warwick, MA 5 52-372 1-9 -

This information was obtained from records of the Green Mountain Well Company, Putney, Vermont. VYNPS DSAR Revision 0 2.0-68 of 120 TABLE 2.4.4

Public Water Supplies Within a 10-Mile Radius of the Vernon Site

Location Town on Map Source Treatment Capacity

Brattleboro, VT A Lake and Reservoir Filtration 3 MGD Plant

Brattleboro B 3 Gravel-Packed Wells Filtration 970 GPM Supplementary 30 feet deep Plant 465 GPM 800 GPM

Hinsdale, NH C-1 2 Gravel-Packed Wells Chlorine 220 GPM 74 feet and Sodium Phosphate 68 feet deep Sodium Hydroxide

C-2 2 Gravel-Packed Wells Chlorine 500 GPM 47 feet and Sodium Phosphate 64 feet deep Sodium Hydroxide

Winchester, NH D 3 Gravel-Packed Wells Phosphate 0.6 MGD Approx. 60 feet deep

Northfield, MA No. 1 E 1 Gravel-Packed Well Sodium Hydroxide 100 GPM No. 2 F Reservoir Chlorine 0.1 MGD

Bernardston, MA G 2 Gravel-Packed Wells Potassium Hydroxide 480 GPM 69 feet and 74 feet 250 GPM deep

G Sand and Gravel Well Chlorine 0.07 MGD 24 feet deep Sodium Hydroxide

From 1963 Inventory of Municipal Water Facilities, U.S. Department of HEW, updated by

local municipal water departments/town offices (2000). VYNPS DSAR Revision 0 2.0-69 of 120 TABLE 2.4.5 Water Supplies Within a 1-Mile Radius of the Site

HINSDALE, NEW HAMPSHIRE Ground

Feet Yield Elevation (ft)

No. Depth GPM U.S.G.S. Use

1. 200+/- 5 300 Domestic

2. 72 30 280

+/- Domestic

VERNON, VERMONT

1 198 15 295 Domestic 2 189 5.5 295 Domestic 3 368 11.5 275 School 4 356 25 208 Domestic & Homes 5 Spring Fed Supply Miller Farm Spring 6 125 30 320 Domestic 7 288 3.75 360 Domestic 8 19 275 Domestic & Farm 8A 18 Now Dry 275 9 17 3-4 275 Domestic 9A 25 Now Dry 275 Domestic 10 20 - 275 Domestic & Farm 10A 18 275 Domestic 11 30.5 4-5 275 Domestic 12 18 6 275 Domestic 13 16 275 Domestic 14 16 275 Domestic 15 14 275 Domestic 16 275 17 Spring 275 18 28 275 Domestic 19 20 15 275 Domestic 20 Spring 275 21 23 10 270 Domestic 22 19 270 Domestic 23 Spring 270 24 Spring 270 25 Spring 270 26 Spring 240 27 Spring 240 28 Spring 220 29 Spring 230 30 Going to 240 Drill 31 17 10 240 Domestic 32 20 240 Domestic 33 165 4 270 Domestic 34 70 20 270 Domestic 35 20 270 Domestic 36 31 270 Domestic & Farm 37 235 16.5 270 Domestic 38 23 275 Domestic 39 12 270 Domestic 40 18 270 Domestic 41 30 280 Domestic 42 20 280 Domestic 43 24 6 280 Domestic 44 35 Dry 280 VYNPS DSAR Revision 0 2.0-70 of 120 Table 2.4.5 (Continued)

Water Supplies Within a 1-Mile Radius of the Site

Ground Feet Yield Elevation (ft)

No. Depth GPM U.S.G.S. Use

45 275 Domestic 46 270 47 24 270 Domestic 48 23 275 Domestic 49 24 275 Domestic 50 22 275 Domestic 51 23 275 Domestic 52 20 275 Grange & Domestic 53 175 6 270 Domestic 54 18 270 VYNPS DSAR Revision 0 2.0-71 of 120 TABLE 2.4.6 Six-Hour PMP and Runoff Increments Connecticut River Basin Above Vernon, Vermont

Basin Arranged

Shape Cumulative Incremental in PMP

Time PMS Red. PMP PMP Critical Losses Runoff (hrs) (ins) Factor (ins) (ins) Order .05"/hr (ins)

6 6.1 .95 5.8 5.8 0.3 0.3 0.0 12 8.3 .95 7.9 2.1 0.7 0.3 0.4 18 9.7 .95 9.2 1.3 0.7 0.3 0.4 24 10.7 .95 10.2 1.0 0.9 0.3 0.6 30 11.7 .95 11.1 0.9 1.0 0.3 0.7 36 12.4 .95 11.8 0.7 2.1 0.3 1.8 42 13.2 .95 12.5 0.7 5.8 0.3 5.5 48 13.5 .95 12.8 0.3 1.3 0.3 1.0 54 13.8 .95 13.1 0.3 0.3 0.3 0.0 60 14.0 .95 13.3 0.2 0.2 0.3 0.0 66 14.2 .95 13.5 0.2 0.2 0.3 0.0 72 14.4 .95 13.7 0.2 0.2 0.3 0.0

TOTAL = 13.7 TOTAL = 10.4 VYNPS DSAR Revision 0 2.0-72 of 120 TABLE 2.4.7 Maximum Annual Floods on Connecticut River at Vernon, Vermont Arranged in Descending Order (1927, 1936, 1938, 1945-1973)

Year Peak Discharge, CFS Stage, Ft MSL

1936 176,000 222.9 1927 155,000 220.5 1938 132,500 214.8 1960 107,000 209.6 1973 102,000 - 1948 101,000 208.5 1953 98,800 208.0 1949 88,600 205.4 1968 88,100 205.3 1952 86,600 204.9 1958 84,200 204.3 1951 81,200 203.6 1969 81,200 203.5 1959 80,600 203.4 1947 79,600 203.3 1956 79,600 203.1 1972 78,600 201.7 1962 73,900 201.7 1955 70,500 200.9 1945 69,700 200.8 1950 68,300 200.3 1967 66,800 200.0 1964 66,100 199.8 1971 65,500 199.6 1954 65,000 199.5 1970 63,400 199.1 1946 62,700 198.8 1963 61,600 198.6 1961 57,900 197.7 1966 46,700 194.8 1965 32,000 190.9 1957 30,000 190.1 VYNPS DSAR Revision 0 2.0-73 of 120 TABLE 2.4.8 Time-Varying PMF Stage-Discharge Table Vermont Yankee Nuclear Plant Site

Time Discharge Stage Time Discharge Stage (hrs) (cfs) (ft MSL) (hrs) (cfs) (ft MSL)

0 58,800 221.3 156 247,000 237.0 6 58,800 221.3 162 231,000 235.8 12 58,900 221.4 168 217,000 234.7 18 59,000 221.5 174 205,300 233.8 24 59,300 221.5 180 192,300 233.0 30 59,800 221.6 186 182,000 232.1 36 60,800 221.8 192 172,000 231.4 42 66,300 222.0 198 162,400 230.7 48 78,300 223.4 204 154,000 230.1 54 95,200 225.0 210 146,000 229.4 60 117,500 227.2 216 139,400 229.0 66 158,200 230.4 222 133,000 228.5 72 255,000 237.5 228 128,400 228.1 78 367,000 245.7 234 122,000 227.5 84 417,000 248.8 240 117,300 227.2 90 464,000 251.5 246 112,000 226.6 96 506,400 253.9 252 107,500 226.2 102 465,000 251.6 258 104,000 226.0 108 430,000 249.5 264 100,000 225.5 114 403,000 248.0 270 97,000 225.2 120 380,000 246.5 276 93,700 225.0 126 351,000 244.8 282 91,400 224.7 132 321,000 242.7 288 88,500 224.5 138 294,000 240.8 294 86,000 224.2 144 283,000 239.6 300 83,700 224.0 150 265,000 238.3 VYNPS DSAR Revision 0 2.0-74 of 120 TABLE 2.4.9 Time-Varying Modified PMF Stage-Discharge Table Vermont Yankee Nuclear Plant Site

Time Discharge Stage Time Discharge Stage (hrs) (cfs) (ft MSL) (hrs) (cfs) (ft MSL)

0 58,800 221.3 156 235,100 236.1 6 58,800 221.3 162 220,600 235.0 12 58,900 221.4 168 207,400 234.1 18 59,000 221.5 174 196,500 233.2 24 59,200 221.5 180 184,200 232.4 30 59,600 221.6 186 174,300 231.5 36 60,500 221.9 192 164,900 230.9 42 65,900 222.3 198 156,000 230.3 48 74,100 223.0 204 148,200 229.6 54 87,900 224.4 210 140,600 229.1 60 106,400 226.3 216 134,500 228.6 66 142,300 229.2 222 128,200 228.1 72 235,000 236.1 228 122,300 227.5 78 343,800 244.1 234 118,100 227.2 84 390,700 247.2 240 113,800 226.8 90 437,700 250.0 246 108,800 226.4 96 480,100 252.5 252 104,500 226.0 102 439,700 250.1 258 101,200 225.6 108 407,300 248.2 264 97,700 225.3 114 381,900 246.8 270 94,500 225.0 120 360,500 245.5 276 91,500 224.6 126 333,500 243.5 282 89,460 224.4 132 305,100 241.4 288 86,630 224.2 138 283,900 239.6 294 84,280 224.0 144 269,000 238.5 300 82,140 223.9 150 252,900 237.4 VYNPS DSAR Revision 0 2.0-75 of 120 TABLE 2.4.10 Checklist of Connecticut River Fishes Found Near Vernon, Vermont

• Salmonidae - Trouts

Salmo salar Linnaeus Atlantic Salmon Salmo trutta Linnaeus Brown Trout Salmo gairdneri Richardson Rainbow Trout Salvelinus fontinalis (Mitchill) Brook Trout

Osmeridae - Smelts

Osmerus mordax (Mitchill) Rainbow Smelt

Catostomidae - Suckers

Catostomus commersoni (Lacepede) White Sucker Catostomus catostomus (Forster) Longnose Sucker

Cyprinidae - Minnows and Carps

Cyprinus carpio Linnaeus Carp Semotilus corporalis (Mitchill) Fallfish Semotilus atromaculatus (Mitchill) Creek Chub Couesius plumbeus (Agassiz) Lake Chub Notemigonus crysoleucas (Mitchill) Golden Shiner Notropis cornutus (Mitchill) Common Shiner Notropis hudsonius (Clinton) Spottail Shiner Hybognathus nuchalis Agassiz Silvery Minnow

Ictaluridae - Freshwater Catfishes

Ictalurus nebulosus (LeSueur) Brown Bullhead Ictalurus natalis (LeSueur) Yellow Bullhead

Esocidae - Pikes

Esox lucius Linnaeus Northern Pike Esox niger LeSueur Chain Pickerel

Anguillidae - Freshwater Eels

Anguilla rostrata (LeSueur) American Eel

Cyprinodontidae - Killifishes

Fundulus diaphanus (LeSueur) Banded Killifish

Percichthyidae - Temperate Basses

Morone americana (Gmelin) White Perch

• Common names used in this checklist are those proposed by Bailey, Reeve M., et al., 1970. "A List of Common Scientific Names of Fishes from the United States and Canada." Special Publication No. 6, American Fisheries Society, Washington.

VYNPS DSAR Revision 0 2.0-76 of 120 TABLE 2.4.11 Fishes of the Connecticut River in the Vicinity of Vernon, Vermont All Collections - 1980

Total Total Weight Length

Number Weight Extremes Extremes in

Species Captured In Grams In Grams Millimeters

Catostomus commersoni (Lacepede) White Sucker 190 129,514 0.5-1408 33-507

Cyprinus carpio Linnaeus Carp 19 91,765 138-8500 195-740

Semotilus corporalis (Mitchill) Fallfish 1 473 473 332

Notemigonus crysoleucas (Mitchill) Golden Shiner 12 913 35-170 137-225

Notropis hudsonius (Clinton) Spottail Shiner 195 2,062 7-15 73-128

Hybognathus nuchalis Agassiz Silvery Minnow 1 16 16 112 Juvenile Cyprinidae 133 208 0.05-2.6 17-68

Ictalurus nebulosus (LeSueur) Brown Bullhead 20 6,500 32-733 140-375

Ictalurus natalis (LeSueur) Yellow Bullhead 1 76 76 180

Esox lucius Linnaeus Northern Pike 1 400 400 400

Esox niger LeSueur Chain Pickerel 12 5,818 162-846 282-508

Anguilla rostrata (LeSueur) American Eel 1 1,360 1,360 750

Morone americana (Gmelin) White Perch 494 58,551 4-410 64-308

Perca flavescens (Mitchill) Yellow Perch 229 25,338 7-350 90-290

Stizostedion vitreum (Mitchill) Walleye 48 30,522 51-1156 185-490

Micropterus dolomieui Lacepede Smallmouth Bass 70 16,693 4-1470 70-490

Micropterus salmoides (Lacepede) Largemouth Bass 8 3,457 23-2040 110-507

Lepomis gibbosus (Linnaeus) Pumpkinseed 48 4,490 2.7-843 57-420

Lepomis macrochirus Rafinesque Bluegill 16 3,585 3-383 56-240

Ambloplites rupestris (Rafinesque) Rock Bass 103 13,246 2.1-302 51-250

TOTALS 1602 394,987

Source: See Reference 12. VYNPS DSAR Revision 0 2.0-77 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Station Site - Area Public Water Supplies 10 Mile Radius Figure 2.4-1 VYNPS DSAR Revision 0 2.0-78 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Station Site - Area Private Water Supplies 1 Mile Radius

VYNPS DSAR Revision 0 2.0-79 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Enveloping Depth-Duration-A rea Values Of PMP for Susquehanna River Basin Figure 2.4-3

VYNPS DSAR Revision 0 2.0-80 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 6-Hour Unit Hydrograph Figure 2.4-4

VYNPS DSAR Revision 0 2.0-81 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Total SPF Hydrograph Figure 2.4-5

VYNPS DSAR Revision 0 2.0-82 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Total PMF Hydrograph (Natural and Modified) Figure 2.4-6

VYNPS DSAR Revision 0 2.0-83 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Connecticut River Basin Federal Power Commission Water Resource Appraisals for Hydroelectric Licensing Summary of Planning Status Figure 2.4-7

VYNPS DSAR Revision 0 2.0-84 of 120 Vermont Yankee Defueled Safety Analysis Report Revision0 Vermont Yankee Nuclear Plant Location of River Cross-Sections Figure 2.4-8

VYNPS DSAR Revision 0 2.0-85 of 120 Vermont Yankee Defueled Safety Analysis Report Revision0 Stage Discharge Curve at The Vermont Yankee Nuclear Plant Fi g ure 2.4-9 VYNPS DSAR Revision 0 2.0-86 of 120 Vermont Yankee Defueled Safety Analysis Report Revision0 Cross Section of the Critical Fetch Fi g ure 2.4-10

VYNPS DSAR Revision 0 2.0-87 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Vermont Yankee Sample Stations Conn. River Figure 2.4-11

VYNPS DSAR Revision 0 2.0-88 of 120 2.5 GEOLOGY AND SEISMOLOGY 2.5.1 General This subsection provides information related to geological and seismological considerations at the site. Detailed site and laboratory investigations were

undertaken by independent consulting firms to obtain necessary design data

contained in this section. The consulting firms were Goldberg-Zoino and

Associates, Cambridge, Massachusetts, and Weston Geophysical Research, Inc., in

conjunction with the Vermont Yankee Nuclear Power Corporation and Ebasco Services

Incorporated. Evaluations of the investigations indicate that the proposed site

was adequate from the geological and seismological viewpoints and could safely

support the nuclear station installation. 2.5.2 Geology The site is located on the west bank of the Connecticut River in the town of Vernon, Vermont, which is in Windham County. Site coordinates are approximately 42° 47' north latitude and 72 ° 31' west longitude, in the extreme southeastern corner of the state of Vermont. 2.5.2.1 Introduction All but one of the major structures of the facility, including the reactor building and turbine building, are supported on rock. The storage pad for the

Interim Spent Fuel Storage Installation (ISFSI) is supported on engineered fill

placed on existing soils. Sixteen of the 93 borings at the site were made in the

immediate vicinity of the reactor building (see Figure 2.5-2). These borings show

that the area is overlaid by glacial deposits from the Pleistocene Age, with an

average 30 feet of glacial overburden above the local bedrock, which consists of

hard biotite gneiss. Rock outcroppings near the site are found along the river

bank. Bedrock exists at or near the foundation grades for the structures, namely

elevation 206 feet MSL for the reactor building, elevation 217 feet MSL for the

turbine building, elevation 227 feet MSL for the radwaste building, and elevation

187 feet MSL for the circulating water intake structure. 2.5.2.2 Geological Investigation Program Standard geologic procedures were employed during the site investigation, beginning with a complete search of available literature concerning geology and

seismic activity in the area including unpublished and published material (refer

to Table 2.5.1). A complete geologic field reconnaissance of the general area and

the immediate site was performed, employing United States Geodetic Survey

topographic maps, aerial photographs, and the state of Vermont geologic and

tectonic maps.

VYNPS DSAR Revision 0 2.0-89 of 120 An extensive subsurface exploration project was undertaken at the site. Ninety-three borings were made, 35 of which were from 32 to 100 feet in depth.

Thirty of these borings were AX (1-3/8 in. cores) and 5 were NX (2-1/8 in.

cores)(see Figure 2.5-2). All NX-size holes were logged in detail (see NX core

logs in Figure 2.5-3). The other cores were examined carefully to determine

general features and characteristics. Representative cores were taken at and

immediately below foundation grade in all NX core holes, and were submitted to

intense laboratory testing for determination of specific physical properties of

bedrock at the site. Several petrographic sections were made and analyzed to

ascertain the mineral composition and structure of bedrock.

A thorough seismic survey program was carried out to determine several of the

in-place physical properties of the site bedrock as it relates to earthquake

criteria for design of structures - such as compressional wave velocities (V c), shear wave velocities (V s), subsurface rock contours, and the possibility of extensive faulting and jointing.

The results of this investigation program are summarized in the following

paragraphs.

2.5.2.3 Regional Geology

Geologic structure of the region is complex, in that there are several sequences

of anticlinoria and synclinoria trending essentially in a northerly direction (see

Figures 2.5-4 and 2.5-5). The site is located geologically within the so-called

Brattleboro syncline, which is part of the Connecticut Valley-Gaspe Synclinorium.

Most of the region is underlain by Paleozoic metamorphic rocks and by a narrow

band of Triassic sedimentary rocks south of the site. The general outcrop pattern

of local Paleozoic formations indicates the presence of a major recumbent fold, overturned to the west. The entire region has undergone extensive metamorphism (mostly of the regional type) which apparently ranges from low-grade west of the

Vernon area to relatively high-grade east of the Vernon area.

Foliated igneous rocks of middle- and late-Devonian age underlie a large portion

of the region. These include three fairly large plutons of the Oliverian Magma

Series (Billings, 1935; Skehan, 1961), one of which is below the site, in the

towns of Vernon, Vermont and Hinsdale, New Hampshire - the Vernon Dome.

VYNPS DSAR Revision 0 2.0-90 of 120 2.5.2.3.1 Regional and Local Stratigraphy The Vernon pluton is a narrow elongate mass approximately 8 miles long and 2 miles

wide, striking approximately 10 degrees to the northwest and dipping steeply to

the east (see Figures 2.5-6 and 2.5-7). The Connecticut River flows southeasterly

across its central portion. Gneisses of the Oliverian Plutonic Series (middle-Devonian) make up the pluton core (Billings, 1935)(see Figure 2.5-7).

Both the Vernon Dome and the Westmoreland Dome north of Brattleboro (see Figure

2.5-5), are part of the Bronson Hill Anticlinorium, a series of echelon gneissic

domes that extend northward into northern New Hampshire and southward into the

State of Connecticut. Except where the Connecticut River crosses the Vernon Dome, the local topography reveals strikingly the distribution of the lithologic units.

2.5.2.3.2 Geological History

There appears to have been at least two distinct tectonic periods of folding after

formation of the Vernon Dome (late Paleozoic to pre-Triassic - over 70 million years ago). Most normal faults on the flanks of the domes strike N30 °E. The two faults at the northern terminus strike N10 °W. In all cases, the faults dip steeply, and appear to be Triassic or younger in age. The Clough Quartzite and the Littleton Formation have been intensely folded at the northern end of the

domal structure. Folds in the immediate area indicate differential movement with

reverse drag folds occurring along with recumbent structures (see Figure 2.5-4).

Faulting took place over 70 million years ago along the southeastern boundary of

the Vernon-Chesterfield area, particularly in the state of New Hampshire. The

Triassic Border Fault (see Figure 2.5-4) is the only fault structure of major

significance related to the site and there has been no apparent movement in it

during the last several million years. Rocks of the Oliverian Plutonic Series and

the mantle of a gneiss dome comprising the Ammonoosuc, Clough, and Littleton

Formations adjoin the fault on the east (Robinson, 1963). There has been relative

movement down on the west side of the fault. A crushed zone in gneiss on the east

side of the fault near Gill Station at the southern end of the Vernon area may be

associated with the fault (Bolk, 1956). The fault is exposed 4 miles south of the

Vernon-Chesterfield area where it dips steeply to the west (Keller and Brainard, 1940). It is difficult to match structures across the fault, and this prevents an

accurate estimate of throw on the fault. Recent movement along the fault is not

indicated. All minor faults in the region appear to be high-angle and Triassic or

younger in age.

VYNPS DSAR Revision 0 2.0-91 of 120 2.5.2.3.3 Regional Structure The three main structural features at the site are a relatively large, heavily

mantled gneissic dome, a major recumbent fold (Bernardson Nappe), and the eastern

margin of a regional syncline (Brattleboro Syncline).

The Bernardson Nappe apparently first formed during the main sequence of

deformation. No minor folds or lineations can be ascribed specifically to

movements that produced the nappe. The tectonic transport direction of the nappe

was generally from east to west. Formation of the gneissic domes followed

emplacement of the nappe and produced an early set of minor folds and lineations

on the mantle of the domes and on the rocks of the nappe (Trask, 1964). Early

folds at the north end of the Vernon Dome are overturned to the northeast and

northwest, with an apparent reverse drag. These early folds and lineations were

deformed then by still later folding in the synclinal area between the two domes.

Local isograds are essentially parallel to the regional structural trend. The

development of slip-cleavage in the Brattleboro Syncline, adjacent to the Vernon

Dome, was accompanied by extensive retrograde metamorphism (Moore, 1949).

Rocks of the Oliverian Plutonic Series, intrusive into the surrounding metamorphic

rocks, form the cores of elongate domes uplifts (Billings, 1935). The Vernon Dome

represents a southern counterpart of the Oliverian Magma Series. Foliation is

well developed around the margins of the Vernon pluton, but decreases slightly

toward its central portion. According to Moore (1949) and Skehan (1961), the

foliation is essentially parallel to the contact between the gneiss and the

overlying Ammonoosuc Volcanics. Available data (Moore, 1949), indicate that the

contact between the gneiss and the overlying Ammonoosuc Volcanics is concordant.

2.5.2.4 Site Geology

2.5.2.4.1 Physiography

The Connecticut River traverses the area near the site from north to south, along

the eastern side of the Vernon, Vermont area, geographically separating the states

of Vermont and New Hampshire at this point.

A strip of lowlands and terraces, about 1 mile in width, borders the river in the

area. There are naturally dissected uplands with an average local relief of

several hundred feet east and west of the lowlands. Wantastiguet Mountain, 0.5

mile east of Brattleboro, is the highest point in the area with an elevation of

1351 feet MSL. The lowest point is on the Connecticut River near Northfield, Massachusetts, with an elevation of 175 feet MSL.

VYNPS DSAR Revision 0 2.0-92 of 120 Rock composition of Vernon pluton is essentially a gneiss, grading from a granodiorite (quartz-diorite) to granite. It is essentially light-gray to

pink-gray, slightly to moderately foliated, medium-grained, subporphyritic

quartz-diorite with hypidiomorphic to granoblastic texture. As tabulated in Table

2.5.2, the gneiss contains plagioclase feldspar (An 12 to An 44), quartz and biotite mica, as essential minerals. Epidote, muscovite mica, K-feldspar, hornblende, garnet, and magnetite are included as accessory minerals. Sericite, some

chlorite, and calcite are present also, as alteration products. Granulated and

flattened "quartz eyes" are present, and individual grains of these aggregates

show sutured and mortar textures. The quartz eyes and strings of biotite flakes

produce prominent lineation. As much as 1% zircon has been found in rocks of the

central region of the Vernon Dome.

An extensive subsurface exploration project was undertaken at the site. The

drilling program was carried out by the Raymond Concrete Pile Company during the

fall of 1966. Subsurface profiles of the borings in the vicinity of the station

structures are illustrated in Figures 2.5-8, 2.5-9, 2.5-10 and 2.5-11. Detailed

logs of deep borings in the reactor building area are provided in Figure 2.5-3.

During formation of the metamorphic plutonic body, various joints and minor

slippages occurred. Many of these joints served as avenues for solutions to

travel with subsequent mineralized fillings. Visual examination of rock cores

indicates that many joints were filled by hydrothermal solutions. A few joint

surfaces have a drusy appearance, some with crystal growth, and others with

mineral stainings left by ground water. Some fractures of joint surfaces appeared

weathered to highly weathered.

Pegmatitic quartzite veins were encountered in borings 1 at elevation 198 feet

MSL, 4 at elevation 208 feet MSL, and 5 at elevation 169 feet MSL. The approximate strike of this vein is N60 °E and it dips 40 °NW. Apparent thickness of this vein is 1-1/2 feet. Pegmatitic veins were encountered also in borings 6 and

21.

Dike or sill-like bodies of a dark green, fine grain diorite were found in boring

6 at elevation 194.5 feet MSL and elevation 191.7 feet MSL. They were found also

in boring 2A at elevation 210.1 feet MSL. Both units are approximately 5 inches

thick. Hard milky quartzite bands or veinlets with accessory magnetite were found

in many of the borings at various depths. Geologic relationships of these bands

have not been made, but it may be determined that they belong to a particular

joint set.

The rock is extensively jointed. Three or more joint sets may be present. These

joint sets appear reasonably tight.

VYNPS DSAR Revision 0 2.0-93 of 120 2.5.2.4.2 Bedrock Bedrock, although extremely hard and structurally competent, appears to be

fractured sufficiently to present occasional hydrostatic conditions in zones of

fracturing, as was observed in several of the drill holes. Water pressure tests

at the site were conducted to determine the "tightness" or permeability of certain

fractured zones. Tests proved that the formation is very tight.

2.5.2.4.3 Surficial Deposits

Rock types at the site are considered to be a metamorphosed igneous intrusive of

the Oliverian Plutonic Series. Generally, the rock is a quartzofeld

spathic-biotite gneiss, with variable amounts of orthoclase and plagioclase

feldspar. The attitude of this gneissic plutonic body is considered to trend

slightly west of north and dip to the east near the site.

At the site, the exposed outcrops along the edge of the river are massive, in some

instances intensively jointed due to mechanical weathering, and without any

visible gneissic structure. Foliation and lineation of the rock has been obscured

due to surface weathering. Rock cores reveal the gneissic structure. Foliation

is fairly well developed. The attitude of the rocks can be determined from the

cores by noting dip of foliation planes.

The gneiss is medium-grained, light-gray to slightly pinkish-gray rock, and its

texture somewhat approaches granoblastic. It is slightly subporphyritic and

rarely has a flaser fabric. Grains of white to gray glassy quartz, white to pink

feldspar, black biotite, with some muscovite and amphibole can be recognized.

Feldspar is quite variable. Minor constituents noted in the rock types are

magnetite, garnet, and possibly zircon and sphene.

2.5.2.5 River Geology

2.5.2.5.1 General

The Connecticut River at site lies within the New England upland. The basin is

maturely dissected with the river flowing throughout most of its course in an open

valley with well-developed flood plains above which rise glacial terraces tiered

on the valley walls. The main river in the upland section winds between rounded, irregular hills and ridges.

VYNPS DSAR Revision 0 2.0-94 of 120 The topography of the entire basin has been modified by glaciation which scraped the tops from the bedrock hills and filled the valleys with glacial detritus with, however, little actual diversion of drainage. The major effect of the glacial

fill was to raise the streams from their old beds, thereby permitting the

development of present channels which may or may not be related to the underlying

configuration of the old valleys in the bedrock.

2.5.2.5.2 Seismic Survey

The seismic tests resulted in conclusions as follows:

1. Seismic velocity measurements in boreholes and from surface studies are high and indicate hard massive bedrock. Deeply weathered zones or faults were not

detected.

2. Bedrock surface is slightly irregular as evidenced by the borings and the lines of seismic refraction investigations.

3. Elastic moduli values, based on seismic velocities, are high values of approximately 4.16 x 10 6 lb/in 2 for Young's Modulus and 1.53 x 10 6 lb/in 2 for the shear or rigidity modulus. Poisson's ratio is 0.347.

4. Compressional wave velocity was found to be 13,800 fps and the shear wave velocity 6,500 fps.

2.5.2.5.3 Shoreline Retreat

The presence of natural outcrops of the bedrock at the various dam-sites such as

Vernon, Turners Falls, and Bellows Falls, coupled with the construction of dams at

these sites, have restricted the river's velocity and concentrated its potential

erosion at the sites themselves. The river banks, as a result, are relatively

stable, with erosion, if any, manifest only as a result of major floods. The long

intervening periods of placid flows provide ample opportunity for inspection and

stabilization of the river banks, should this be required.

At the site, the natural river banks have become well stabilized during the

60-year existence of the Vernon Hydroelectic Project immediately downstream.

There is little evidence of bank erosion.

VYNPS DSAR Revision 0 2.0-95 of 120 2.5.3 Seismology 2.5.3.1 Introduction

The evaluation of a nuclear power station site from a seismic standpoint is based

upon a combination of historical and instrumental data. Historical records before

1900 are somewhat misleading since observations are limited to population centers

and the untrained observer appears to sometimes exaggerate. Later historical

records, such as those of the early 1900's, appear to be more reliable.

Instrumentation for the detection of local earthquakes, which may or may not be

felt, has been operating in the New England area since the mid-1930's.

2.5.3.2 Seismic Investigation Program

The seismic evaluation of the station is a threefold study consisting of a review

of historical data from the New England area, an analysis of instrumental and

historical records for the Vernon area, and a study of earthquake intensity

attenuation with distance for northeastern United States.

2.5.3.3 Geologic and Tectonic Background

As described in detail in Subsection 2.5.2, "Geology", the southern parts of

Vermont and New Hampshire are composed of early Paleozoic sediments which have

been metamorphosed through intense folding. Some middle and late Paleozoic

igneous intrusives and extrusives are also present. The site itself is located on

the Vernon Dome, a middle Ordovician intrusive body of quartz-diorite gneiss. The

only post-Paleozoic tectonic feature present in the area is the eastern border

fault of the Triassic Basin of Massachusetts. The fault is present in extreme

southwestern New Hampshire where it strikes in a northeasterly direction passing

about 6 miles to the southwest of the station site. A tectonic map of the New

England area is shown in Figure 2.5-12.

2.5.3.4 Seismic History

Those earthquakes which have been strongly felt or have produced some damage in

the New England area are shown in Figure 2.5-13. Areas of some seismic activity

are noted in the vicinity of the following locations: Lake George, New York;

Concord, New Hampshire; Ossipee Mountains, New Hampshire; southeastern New

Hampshire and northeastern Massachusetts; and Haddam Connecticut. All of these

areas lie between 50 and 100 miles from the plant site and have experienced at

least one historical earthquake which has produced some minor damage (Modified

Mercalli Intensity VI or greater). VYNPS DSAR Revision 0 2.0-96 of 120 The nearest of these areas to the site is the Concord, New Hampshire area, about 50 miles to the northeast of the site. The earthquake of November 23, 1884, of

Intensity VI on the Modified Mercalli Scale, is the largest to have occurred at

Concord. This earthquake was felt over an 8000-square mile area which did not

include the Vernon, Vermont area.

The largest earthquake to have originated in the vicinity of Lake George, New

York, occurred on April 20, 1931. The epicenter of this earthquake is about 75

miles northwest of the site. It was reported to have been felt at Bellows Falls, Vermont; Greenfield, Massachusetts; and Hinsdale, New Hampshire. The intensity at

Vernon can be estimated at about IV on the Modified Mercalli Scale (see Figure

2.5-14).

Modified Mercalli isoseismal lines for the Ossippe, New Hampshire, earthquakes of

December 20 and 24, 1940, which were of epicentral Intensity VII, show that the

intensity at Vernon, Vermont, was about IV. The epicenter of these earthquakes

was about 95 miles northeast of the site. Although the isoseismal lines show an

intensity of IV, reports from various localities in the area show that intensities

range from III to VI. In Keene, New Hampshire, a great part of which is located

on alluvium, an intensity of VI was noted, although just outside the town in the

surrounding highlands, the intensity was IV. Brattleboro, Vermont, reported an

Intensity V; Bellows Falls, Vermont, reported an Intensity IV; and Hinsdale, New

Hampshire across the Connecticut River from Vernon, reported an Intensity of III.

The earthquake of October 5, 1817, whose epicenter was near Woburn, Massachusetts, was listed as Modified Mercalli Intensity VII by the United States Coast and

Geodetic Survey. The only report of damage is that "walls were thrown down at

Woburn". Since no other reports concerning this earthquake could be found, it is

doubtful that this earthquake had any effect on a site located 70 miles to the

west-northwest of Woburn, Massachusetts.

The earthquakes of November 9, 1727, at Newburyport, Massachusetts, and May 18, 1791, at East Haddam, Connecticut, are both listed by the United States Coast and

Geodetic Survey as Intensity VIII (Modified Mercalli). Both earthquakes occurred

between 85 and 90 miles from the plant site. Historical evidence shows that these

earthquakes were felt over wide areas of the northeastern United States, probably

including the Vernon, Vermont area. Although there is evidence that these

earthquakes were less than Intensity VIII, attenuation of earthquake intensity

with distance would probably have reduced these (even if they were of intensity

VIII) to Intensity IV or V at the plant site.

VYNPS DSAR Revision 0 2.0-97 of 120 2.5.3.5 Seismicity of Area A more detailed picture of the seismicity of the central New England area

surrounding Vernon, Vermont, is shown in Figure 2.5-15. This figure shows the

approximate epicentral location of all the earthquakes of record.

The nearest earthquake to Vernon, Vermont, for which instrumental records were

obtained, took place on June 1, 1963. The epicenter of this earthquake was

located near Shelburne Falls, Massachusetts, about 15 to 20 miles southwest of

Vernon, Vermont. The intensity was listed as Modified Mercalli Intensity II or

less.

The earthquake catalogue of Henry Fielding Reid lists some local activity in the

Keene, New Hampshire area on October 10, 1854, and December 1, 1875. No newspaper

accounts of these earthquakes could be found in the New Hampshire Sentinel, published in Keene.

A report of the earthquake which was observed in Vernon, Vermont, on June 11, 1898, appeared in the Monthly Weather Review of June 1898. "Vernon, Vermont, reports an earthquake on the 11th, at 1:25 a.m., which was distinctly felt and

jarred the house. This seems to be quite an isolated case, and it is worth

inquiring whether this jar was not due to something else than a true earthquake".

Local newspapers were studied for any accounts of this earthquake. The earthquake

was observed in Brattleboro, but apparently not observed in Keene, or Hinsdale, New Hampshire or Greenfield, Massachusetts. The newspaper account of the

earthquake which appeared in the Brattleboro Evening Phoenix of June 13, 1898, is as follows: "An earthquake shock was felt distinctly in Brattleboro at 1:45

Saturday morning. People who were awake say that houses were shaken, and that

doors were slammed by the shock. The nervous shock which one woman sustained was

sufficient to cause illness. Mr. Pratt, the night watchman at S. A. Smith and

Company's factory, says that almost the same moment that the earthquake was felt, a brilliant meteor flashed across the sky and exploded with a loud report. People

who felt the earthquake also heard the report, but few saw the meteor".

It is possible that this event was a meteor, but in evaluating the seismic

history, we must consider it as a local earthquake of Modified Mercalli Intensity

IV in the Vernon-Brattleboro area (see Figure 2.5-14).

VYNPS DSAR Revision 0 2.0-98 of 120 2.5.4 Conclusions The nuclear installation is located on the west side of the Connecticut River near

Vernon, Vermont. The station is supported on rock at the site. Bedrock in the

area is hard, strong, competent gneiss with unconfined compressive strengths that

generally exceed 15,000 psi. The rock is moderately to highly jointed. The mass

of the rock has not been weakened structurally to any important degree by the

jointing. Seismic velocity measurements at the site verify the hard massive

nature of the bedrock. Deeply weathered or faulted zones were not detected at the

site. Geologic considerations do not preclude utilization of the site for a

nuclear station location.

The effects at the site resulting from a significant seismic disturbance have been

considered based upon local and regional geology, tectonics, and historical and

instrumental seismology.

It is indicated from geologic and tectonic history that the region is relatively

quiescent. Low magnitude seismic events can occur, but should be relatively

infrequent.

The seismic activity of the area is depicted on Figure 2.5-13. Some concentrated

areas of seismic activity may be noted 50 miles to the northeast of the site in

the Concord, New Hampshire area and at other localities 75 miles to 100 miles

distance from the site. The nearest earthquake to the site which produced damage

occurred near Concord, New Hampshire, and was of Intensity VI on the Modified

Mercalli Scale. Concord, New Hampshire, is 50 miles from the site.

Based on intensity attenuation with distance, the largest New England earthquakes, which occurred some 85 to 90 miles from the site, would have been observed as

Modified Mercalli Intensity IV or V at the plant site.

The nearest earthquake to the site, which occurred during the instrumental

recordings of the last 30 years, had an epicentral location of approximately 15 to

20 miles from the site. The probable maximum intensity from an earthquake which

has been observed in the Vernon area is that of Intensity V on the Modified

Mercalli Scale. Based on extrapolated data from earthquakes which occurred on the

west coast of the United States, the maximum acceleration to be expected at the

bedrock surface of the plant site in Vernon, Vermont, would be from an earthquake

of Intensity V to low Intensity VI on the Modified Mercalli Scale. This

earthquake would produce an acceleration of approximately 0.03g to 0.04g.

VYNPS DSAR Revision 0 2.0-99 of 120 It is believed that the earthquake accelerations developed for this site are conservative. They result from detailed studies of the site and region by

consultants knowledgeable in the field of seismology. However, for design

purposes, a minimum ground acceleration of 0.07g was used. In addition, structures and equipment have been examined for an acceleration of 0.14g to

ascertain that no failure could occur that would prevent safe storage of

irradiated fuel. VYNPS DSAR Revision 0 2.0-100 of 120 TABLE 2.5.1 AVAILABLE INFORMATION CONCERNING GEOLOGY AND SEISMIC ACTIVITY RELATED TO THE VERMONT YANKEE NUCLEAR POWER STATION SITE

REFERENCES

Balk, R., 1956. Bedrock Geology of the Massachusetts Portion of the Northfield Quadrangle, Massachusetts - New Hampshire -

Vermont: U.S. Geol. Survey, Geol. Quad. Map GQ92. Billings, M.P., 1935. Geology of the Littleton - Moosilauke Quadrangles, New Hampshire: New Hampshire State Planning and

Development Comm., Concord, New Hampshire

Keeler, J., and Faulted Phyllite East of Greenfield, Brainard, C., 1940. Massachusetts: Am. Jour. Sci., V. 238, pp. 354-365.

Moore, G. Em. Jr., 1949. Structure and Metamorphism of the Keene - Brattleboro Area, New Hampshire - Vermont: Geol. Soc. Am., Bull., Vol. 60, pp. 1613-1670.

Robinson, P., 1963. Gneiss Domes of the Orange Area, Massachusetts and New Hampshire: Doctoral Thesis, Harvard University.

Skehan, J. W., 1961. The Green Mountain Anticlinorium in the Vicinity of Wilmington and Woodford, Vermont: Bull: 17, Vermont Geol. Survey, Vermont Development Dept.

Trask, N.J., Jr., 1964. Stratigraphy and Structure in the Vernon - Chesterfield Area, Massachusetts; New Hampshire;

Vermont: Doctoral Thesis, Harvard University.

Unpublished.

MAP REFERENCES

1. Centennial Geologic Map of Vermont, 1961, Compiled and Edited under the direction of Dr. Charles G. Doll, State Geologist.

2. Geologic Map of New Hampshire, Marland P. Billings, Dept. of Geology, Harvard University in cooperation with New Hampshire Planning and Development

Commission and U.S. Geological Survey.

3. Geology Map of the Keene - Brattleboro Area; G.E. Moore, Jr., 1949, for New Hampshire Planning and Development Commission.

VYNPS DSAR Revision 0 2.0-101 of 120 TABLE 2.5.2 VERNON PLUTON: ESTIMATED MODE OF THE OLIVERIAN MAGMA SERIES

1 2 3 4 5

Phenocrysts

Quartz - 2 5 8 - Plagioclase 2 - Hornblende - -

Groundmass Plagioclase 54 52 58 54 53 K-feldspar tr tr - - 7 Quartz 32 36 31 30 36 Biotite 5 3 - - 3 Chlorite 2 1 2 tr - Muscovite 3 2 4 - tr Epidote 4 1 - - 1 Magnetite - tr tr tr tr Garnet - tr tr - - Zircon tr tr tr tr tr Apatite tr tr tr tr tr Sphene - - - - tr Pyrite - - tr - - Hematite - - tr - - Leucoxene - - tr - - Tourmaline - - tr - - Carbonate tr 1 - - -

% of Anorthite in Plagioclase 41 37 12 44 41 

Size of Groundmass (mm) 0.25-1.0 0.1-0.25 0.05-0.2 0.15-0.3 0.05-0.5

Size of Phenocrysts (mm) - 2.0-3.0 2.0-4.0 1.0-4.0 -

Texture Gr* Gr Gr Gr Gr Subp" Subp M** Por' M**

1. Quartz-diorite 4. Hornblende quartz diorite *GR = Granoblastic

2. Quartz-diorite 5. Granodiorite "Subp = Subporphyritic

3. Quartz-diorite **M = Mortar

  'Por =  Porphyritic (After Moore, 1949)

VYNPS DSAR Revision 0 2.0-102 of 120 Vermont Yankee Defueled Safety

Analysis Report Station Site - Geological Survey - General Plan-Location of Test Borings VYNPS DSAR Revision 0 2.0-103 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Station Site - Geological Survey - Subsurface Profile - Log of Test Borings (1A, 2A, 3A, 4, 5, 8) Fi g ure 2.5-3 VYNPS DSAR Revision 0 2.0-104 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Station Site - Tectonic Map - State of Vermont Figure 2.5-4

VYNPS DSAR Revision 0 2.0-105 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Station Site - Tectonic Map - State of New Hampshire Figure 2.5-5 EXPLANATION -+-.=.::*:::*.

1. --

   ....,_,._

.... , _.%,.-f"'"" fT* ........ .. ) -... ... ;-. (lt-.. er--'**-OC -* .. . L.;_._.J ............. ... tiff ..... . N'OTE: THIS MA.P IS TAKEN FROM "GEOL..OGIC MAP OF NEW HAMPSHIRE" MARYLAND P BILLtNGS, DEPT. OF GEOLOGY HARVARD UNIV. IN COOPERA .. TION WITH N.H. PLANNING AND' DEVELOPMENT tSSION AND U.S. GfOLOGfCAL SURVEY. *-*--*-... .. -t-*--. I ! i I .. I I "" z I :.:.. -----+----. -----;---* -*-** -_j *-1 0 l M A S S A c A N GOLDBERG *,ZOINO AN=:> ASSOCIATES, tNC. SOIL. AND FOUNDATION ENGINEERS -**-.. -.. i l I 1 --t *-., --z ,...; :----I i . . -* ---,-----*-I VYNPS DSAR Revision 0 2.0-106 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Station Site - Geological Survey Area Bedrock Geology Figure 2.5-6

VYNPS DSAR Revision 0 2.0-107 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Station Site - Geological Survey Area Geological Section Figure 2.5-7

VYNPS DSAR Revision 0 2.0-108 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Station Site - Geological Survey Subsurface Profile (Section AA) Log of Test Borings (5, 8, S9, 11 and 21) Figure 2.5-8

VYNPS DSAR Revision 0 2.0-109 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Station Site - Geological Survey Subsurface Profile (Section BB) Log of Test Borings (2A, 3A, ST6-1/2 and S9) Figure 2.5-9

VYNPS DSAR Revision 0 2.0-110 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Station Site - Geological Survey Subsurface Profile (Section CC) Log of Test Borings (2, 2A, 5, 7, 7A, 13, 15) Figure 2.5-10

VYNPS DSAR Revision 0 2.0-111 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Station Site - Geological Survey Subsurface Profile (Section BB) Log of Test Borings (3, 3A, 4, 8, 8A, 12 and 16) VYNPS DSAR Revision 0 2.0-112 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Station Site - Tectonic Map - New England Area Figure 2.5-12 VYNPS DSAR Revision 0 2.0-113 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Station Site - Compilation of Earthquakes-New England Area Fi g ure 2.5-13 VYNPS DSAR Revision 0 2.0-114 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Station Site - Earthquake Intensity Modified Mercalli and Rossi-Forel Scales Figure 2.5-14 VYNPS DSAR Revision 0 2.0-115 of 120 Vermont Yankee Defueled Safety Analysis Report Revision 0 Station Site - Compilation of Earthquakes Central New England Area Figure 2.5-15 VYNPS DSAR Revision 0 2.0-116 of 120 2.6 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 2.6.1 Objectives

The radiological environmental monitoring program is designed to demonstrate

the adequacy of environmental safeguards inherent in station design, the

effectiveness of the Process Radiation and Area Radiation Monitoring Systems

in measuring the controlled releases of low levels of radioactive materials

and the impact, if any, on the environment as a result of facility operation.

Emphasis is placed on control at the source with follow-up and confirmation by

environmental radiological surveillance.

The program consists of two phases, preoperational and operational, each

having specific objectives. The preoperational phase was conducted over the

two-year (approximate) period preceding station operation to establish

background radiation levels and radioactivity concentrations at selected

locations, to assess the variability between sample locations, and to observe

any cyclical or seasonal trends in the environmental sample media. Although

VYNPS has certified permanent cessation of operation and permanent defueling

in accordance with 10 CFR 50.82, the facility will continue in the operational

phase of the radiological environmental monitoring program. The operational

phase of the program has the following objectives:

1. To assure that radiation levels and radioactivity concentrations in the environment resulting from facility operation meet the applicable

regulatory and license requirements.

2. To make possible the prompt recognition of any significant increase in environmental radiation or radioactivity levels and to identify the cause

of the change, whether it be station effluents, effluents from other

nuclear facilities, fallout from atmospheric nuclear weapons tests,

seasonal changes in natural background, or other sources.

3. Obtain information on the critical radionuclides and pathways leading to the quantitative evaluation of the dose to man resulting from the

operation of the station.

VYNPS DSAR Revision 0 2.0-117 of 120 2.6.2 Monitoring Network The radiological environmental monitoring program compares measured radiation

levels and levels of radioactivity in samples from the area possibly

influenced by the station to levels found in areas not influenced by the

station. Sampling in both areas is done in accordance with the requirements

of Off-Site Dose Calculation Manual (ODCM) and Technical Specifications, with

the area outside the influence of the station serving as a background or

control for the area in the immediate vicinity of the station. A comparison

of survey data collected at control locations and locations within the range

of influence of the station (indicator locations) allows the determination of

any significant difference between the two areas. This method of

environmental sampling makes it possible to differentiate between facility

releases and other fluctuations in environmental radioactivity due to

atmospheric nuclear weapons test fallout, seasonal variations in natural

background, and other causes.

In addition to the control and indicator stations, as described above, the

direct radiation monitoring network is further grouped into an inner ring and

an outer ring for emergency response purposes. The inner ring is located in

the general vicinity of the station (0-4 km), while the outer ring is located

within a range of 2-8 km from the station. Additional stations are situated

at special interest and control locations.

The types of sample media used for environmental surveillance are divided into

four categories, based on exposure pathways. These categories are direct

radiation, airborne, waterborne, and ingestion. Each of these is described

below. Specific and more detailed monitoring requirements may be found in

ODCM Section 3/4.5.1, and the identification of specific monitoring locations

may be found in Table 7.1 of the ODCM. The number of sampling locations and

the frequency of sampling discussed below reflect minimum ODCM requirements.

The actual sampling program may exceed these requirements.

2.6.2.1 Direct Radiation

Environmental direct radiation (gamma) measurements are continuously monitored

at approximately 40 locations. Either pressurized ion chambers or

Thermoluminescent Dosimeters (TLDs) are used to obtain an integrated gamma

radiation exposure at frequencies as prescribed in the ODCM. However, the

frequency of analysis readout is based upon the specific system used as

discussed in the ODCM. TLDs in the outer ring are collected at frequencies as

prescribed in the ODCM, but need only be processed for exposure measurements

if a gaseous radioactive effluent control as identified in the ODCM was

exceeded during the exposure period. VYNPS DSAR Revision 0 2.0-118 of 120 2.6.2.2 Airborne Air is sampled for particulates and Iodine-131 at a minimum of five locations (including one control). The samples are collected by passing the air through

a glass fiber filter in series with a charcoal cartridge. The sampling pumps

operate continuously, and a meter is incorporated into the sampling stream to

measure the total volume of air sampled during a given interval.

The air particulate filters are collected and analyzed weekly for gross beta

radioactivity. These filters are composited for each sampling station and are

analyzed quarterly for gamma-emitting radionuclides. Charcoal cartridges are

collected weekly and are analyzed for Iodine-131.

Increased sampling frequency or additional analyses may be required on air

particulate filters or charcoal cartridges if conditions warrant, pursuant to

the footnotes to Off-Site Dose Calculation Manual Table 3.5.1.

2.6.2.3 Waterborne

2.6.2.3.1 Surface Water

River water samples are collected from one upstream and one downstream

location. At the upstream (control) location, a grab sample is collected

monthly. At the downstream location, an automatic compositing water sampler

collects an aliquot of river water at time intervals that are very short

relative to the compositing period (monthly). These composited samples are

collected monthly.

A gamma isotopic analysis is required on each monthly sample. These samples

are also composited, by station, for a quarterly tritium analysis.

2.6.2.3.2 Ground Water

Grab samples of ground water are collected and analyzed in accordance with the

requirements of the Off-Site Dose Calculation Manual.

2.6.2.3.3 Sediment from Shoreline

Sediment grab samples are collected semiannually from two locations, one

downstream from the station and one at the North Storm Drain Outfall. Each

sample is analyzed for gamma-emitting radionuclides.

VYNPS DSAR Revision 0 2.0-119 of 120 2.6.2.4 Ingestion 2.6.2.4.1 Milk

Milk samples are collected monthly from four locations, including a control

location. When milk animals are identified as feeding on pasture vegetation, sampling is increased to semimonthly. Each sample is analyzed for

gamma-emitting radionuclides. A separate radiochemical separation is

performed on each milk sample, followed by an Iodine-131 analysis.

2.6.2.4.2 Fish

Recreationally important species of fish are collected semiannually from two

locations, one upstream and one in the vicinity of the station discharge. The

edible portions of each sample are analyzed for gamma-emitting radionuclides.

2.6.2.4.3 Vegetation

A mixed grass sample is collected at each air sampling station on a quarterly

schedule, as available. Each sample is analyzed for gamma-emitting

radionuclides.

A silage sample is collected from each milk sampling station at the time of

harvest, as available. Each sample is analyzed for gamma-emitting

radionuclides.

2.6.3 Land Use Census

A Land Use Census is performed annually according to the Off-Site Dose

Calculation Manual 3/4.5.2. Analyses are done to choose the optimum milk

sampling locations and to ensure that the receptors used for calculations done

in accordance with the Off-Site Dose Calculation Manual 3/4.3.3 are

conservative.

2.6.4 Emergency Surveillance

The environmental monitoring program is designed to supplement emergency

monitoring functions as well as perform the routine surveillance activities.

The monitoring stations are strategically located and equipped to provide

radiation monitoring data essential to the rapid assessment of any accidental

radioactivity release.

VYNPS DSAR Revision 0 2.0-120 of 120 2.6.5 Reports An Annual Radiological Environmental Operating Report is submitted to the NRC.

The report contains a summary, interpretations, and an analysis of trends for

the results of the radiological environmental surveillance activities for the

report period. Included are comparisons with operational controls and

previous environmental surveillance reports, plus a description of the

radiological environmental program and a map of all sampling locations. An

assessment of the impact of the station operation on the environment is also

included.

VYNPS DSAR Revision 0 3.0-1 of 98 FACILITY DESIGN AND OPERATION TABLE OF CONTENTS

Section Title Page 3.1 DESIGN CRITERIA ....................................................... 9 3.1.1 Conformance with 10 CFR 50 Appendix A General Design Criteria .................................................... 9 3.1.2 Classification of Structures, Systems and Components ....... 11 3.1.3 Loading Considerations for Structures, Foundations, Equipment and Systems ...................................... 12 3.1.3.1. Seismic Classification ......................... 17 3.1.3.2 Seismic Design ................................. 19 3.1.4 References ................................................. 23 3.2 FACILITY STRUCTURES .................................................. 26 3.2.1 Reactor Building ........................................... 26 3.2.1.1 Function ....................................... 26 3.2.1.2 Description .................................... 26 3.2.1.3 Seismic Analysis ............................... 28 3.2.2 Turbine Building ........................................... 30 3.2.2.1 Function ....................................... 30 3.2.2.2 Description .................................... 30 3.2.2.3 Seismic Analysis ............................... 30 3.2.3 Plant Stack ................................................ 31 3.2.3.1 Description .................................... 31 3.2.3.2 Seismic Analysis ............................... 31 3.2.4 Control Room Building ...................................... 32 3.2.4.1 Description .................................... 32 3.2.4.2 Seismic Analysis ............................... 32 3.2.5 Circulating Water Intake and Discharge Structures .......... 32 3.2.5.1 Intake Structure ............................... 32 3.2.5.2 Discharge and Aerating Structure ............... 33

VYNPS DSAR Revision 0 3.0-2 of 98 3.2.6 Cooling Tower Deep Basin ................................... 33 3.2.7 Interim Spent Fuel Storage Installation .................... 34 3.2.7.1 Description .................................... 34 3.2.7.2 Seismic Analysis ............................... 34 3.2.8 References ................................................. 36 3.3 SYSTEMS .............................................................. 38 3.3.1 Fuel Storage and Handling .................................. 38 3.3.1.1 Nuclear Fuel ................................... 38 3.3.1.2 Spent Fuel Storage ............................. 41 3.3.1.3 Standby Fuel Pool Cooling and Demineralizer Systems ........................................ 47 3.3.1.4 Tools and Servicing Equipment .................. 50 3.3.1.5 References ..................................... 53 3.3.2 Service Water System ....................................... 57 3.3.2.1 Objective ...................................... 57 3.3.2.2 Design Bases ................................... 57 3.3.2.3 Description .................................... 57 3.3.2.4 Evaluation ..................................... 59 3.3.2.5 Inspection and Testing ......................... 59 3.3.3 Electrical Power Systems ................................... 59 3.3.3.1 Transmission System ............................ 59 3.3.3.2 Auxiliary Power System ......................... 61 3.3.3.3 Diesel Generator Systems ....................... 64 3.3.3.4 125 V DC System ................................ 67 3.3.3.5 +/-24 V DC Power System .......................... 70 3.3.4 Fire Protection System ..................................... 71 3.3.4.1 Objective ...................................... 71 3.3.4.2 Design Basis ................................... 71 3.3.4.3 Description .................................... 72 3.3.4.4 Inspection and Testing ......................... 75 3.3.4.5 References ..................................... 75

VYNPS DSAR Revision 0 3.0-3 of 98 3.3.5 Heating, Ventilating and Air Conditioning Systems .......... 75 3.3.5.1 Objective ...................................... 75 3.3.5.2 Design Bases ................................... 76 3.3.5.3 Description .................................... 76 3.3.5.4 Inspection and Testing ......................... 84 3.3.6 Instrument and Service Air Systems ......................... 84 3.3.6.1 Objective ...................................... 84 3.3.6.2 Design Basis ................................... 84 3.3.6.3 Description .................................... 84 3.3.6.4 Inspection and Testing ......................... 85 3.3.7 Process Sampling ........................................... 85 3.3.7.1 Objective ...................................... 85 3.3.7.2 Design Basis ................................... 85 3.3.7.3 Description .................................... 86 3.3.8 Station Water Purification, Treatment and Storage .......... 88 3.3.8.1 Station Makeup Water System .................... 88 3.3.8.2 Potable and Sanitary Water System .............. 89 3.3.9 Lighting Systems ........................................... 91 3.3.9.1 Objective ...................................... 91 3.3.9.2 Design Basis ................................... 91 3.3.9.3 Description .................................... 91 3.3.9.4 Inspection and Testing ......................... 92 3.3.10 Communication Systems ...................................... 92 3.3.10.1 Objective ...................................... 92 3.3.10.2 Design Basis ................................... 92 3.3.10.3 Description .................................... 93 3.3.10.4 Inspection and Testing ......................... 94 3.3.11 Process Computer System .................................... 94 3.3.11.1 Objectives ..................................... 94 3.3.11.2 Design Bases ................................... 94

VYNPS DSAR Revision 0 3.0-4 of 98 3.3.11.3 Description .................................... 95 3.3.11.4 Inspection and Testing ......................... 97 3.3.11.5 Cyber Security ................................. 97 3.3.11.6 Process Computer Data Feed to the Plant Data Server (PDS) ................................... 98

VYNPS DSAR Revision 0 3.0-5 of 98 FACILITY DESIGN AND OPERATION LIST OF TABLES

Table No. Title

3.1.1 Allowable Stresses for Class I Structures

3.1-2 Safety Margins for Several Critical Portions of Major Class I Structures VYNPS DSAR Revision 0 3.0-6 of 98 FACILITY DESIGN AND OPERATION LIST OF FIGURES

Reference Figure No. Drawing No. Title

3.1-1 G-191529 Reactor Building Reactor Vessel Pedestal Mat-M+R 3.1-2 G-191483 Reactor Building Foundation Mat Plan-M+R

3.2-1 G-191148 GEN ARRGT REACTOR BLDG PLANS - SHEET 1

3.2-2 G-191149 GEN ARRGT REACTOR BLDG PLANS - SHEET 2

3.2-3 G-191150 GEN ARRGT REACTOR BLDG SECT

3.2-4 G-191143 GEN ARRGT TURB BLDG BASE FLR PLN 3.2-5 G-191144 GEN ARRGT TURB BLDG GND FLR PLN

3.2-6 G-191145 GEN ARRGT TURB BLDG OPR FLR PLN

3.2-7 G-191146 GEN ARRGT TURB BLDG SECT - SHEET 1

3.2-8 G-191147 GEN ARRGT TURB BLDG SECT - SHEET 2

3.2-9 G-191142 PLOT PLAN

3.2-10 G-191592 CONTROL ROOM BLDG FLOOR PLAN & FDN PLAN R SH1 3.2-11 G-191595 CONTROL RM BLDG EXT WALL ELEVS M

3.2-12 G-191451 Intake Structure Masonry

3.2-13 G-191452 Intake Structure Masonry

3.2-14 G-191453 Intake Structure Masonry

3.2-15 G-191463 Discharge Structure Masonry

3.2-16 G-191461, Sh1 Discharge Structure Masonry

3.2-17 G-200347 Aerating Structure Masonry and Reinforcing

3.2-18 Main Stack Geometry

3.3.1-1 Fuel Storage-Arrangement

VYNPS DSAR Revision 0 3.0-7 of 98 FACILITY DESIGN AND OPERATION LIST OF FIGURES (Cont'd)

Reference Figure No. Drawing No. Title 3.3.1-2 5920-6893 POOL FUEL STORAGE RACK ARRANGEMENT

3.3.1-3 5920-12795 12th Rack installed in cask laydown area

3.3.1-4 Fuel Storage Rack Assembly

3.3.1-5 HOLTEC Fuel Storage Rack Assembly (Partial)

3.3.1-6 G-191173 Sh1 Fuel Pool Cooling System

3.3.1-7 G-191173 Sh2 Fuel Pool Cooling System

3.3.2-1 G-191159 Sh1 F/D SERVICE WATER SYSTEM

3.3.2-2 G-191159 Sh2 F/D SERVICE WATER SYSTEM

3.3.3-2 G-191299 4KV AUX ONE LINE

3.3.3-3 G-191300 Sheet 1, 480V AUX ONE SWITCHGEAR BUS 8, MCC-8A, 8C 3.3.3-4 G-191300 Sh2 480V AUX ONE LINEMCC-8B, 8E, 89B

3.3.3-5 G-191301 Sheet 1, 480V AUX ONE W/D SWGR BUS 9, MCC-9A, 9C 3.3.3-6 G-191301 Sh2 480V AUX ONE LINE MCC-9B, 9D, 89A

3.3.3-7 G-191372 Sh1 125V DC ONE LINE W/D

3.3.3-8 G-191372 Sh2 125V DC ONE LINE W/D

3.3.3-9 G-191372 Sh3 125V DC ONE LINE W/D

3.3.3-10 G-191372 Sh4 120/240 Volt Instrumentation One-Line Diagram

3.3.3-11 G-191372 Sh5 One Line Diagram, +24 Volt DC Power System

3.3.4-1 G-191163 Sh1 FIRE PROTECT SYS INNER LOOP

3.3.4-2 G-191163 Sh2 FIRE PROT SYS OUTER LAYOUT 3.3.5-1 G-191237 Sh1 HVAC - Flow Diagram Turbine, Service, and Control Room Buildings 3.3.5-2 G-191237 Sh1 HVAC - Flow Diagram Turbine, Service, and Control Room Buildings 3.3.5-3 G-191236 HVAC - Flow Diagram Radwaste Building

VYNPS DSAR Revision 0 3.0-8 of 98 FACILITY DESIGN AND OPERATION LIST OF FIGURES (Cont'd)

Reference Figure No. Drawing No. Title 3.3.5-4 G-191254 Station Heating Boiler Systems

3.3.5-5 G-191238 Reactor Building Heating Ventilation and Air- Conditioning 3.3.6-1 G-191160 Sh1 F/D INSTRUMENT AIR SYS

3.3.6-2 G-191160 Sh2 F/D INSTRUMENT AIR SYS

3.3.6-3 G-191160 Sh3 F/D INSTRUMENT AIR SYS

3.3.6-4 G-191160 Sh4 F/D INSTRUMENT AIR SYS

3.3.6-5 G-191160 Sh5 F/D SERVICE AIR SYS

3.3.6-6 G-191160 Sh6 F/D SERVICE AIR SYS

3.3.6-7 G-191160 Sh7 F/D DIESEL GEN START AIR SYS

3.3.6-8 G-191160 Sh8 FLOW DIAG INSTRUMENT AIR SYSTEM

3.3.7-1 G-191164 Station Process Sampling System

3.3.7-2 G-191165 Station Process Sampling Systems

3.3.8-1 G-191161 Make-Up Water Treatment System VYNPS DSAR Revision 0 3.0-9 of 98 3.1 DESIGN CRITERIA 3.1.1 Conformance with 10 CFR 50 Appendix A General Design Criteria

The final version of the General Design Criteria was published in the Federal

Register February 20, 1971 as 10CFR50 Appendix A. Differences between the proposed

and final versions of the criteria included a consolidation from 70 to 64 criteria

and general elaboration of design requirement details. At the time of issuance, the

Commission stressed that the final version of the criteria were not new requirements

and were promulgated to more clearly articulate the licensing requirements and

practices in effect at the time.

In a Staff Requirements Memorandum on SECY-92-223, the NRC approved a proposal in

which it was recognized that plants with construction permits issued before May 21, 1971 were not licensed to meet the final General Design Criteria. The memo

recognized that while compliance with the intent of the final General Design

Criteria was important, back fitting of these requirements to older plants would

provide little or no safety benefit.

Although VYNPS was not required to comply with the General Design Criteria, the

design and construction of VYNPS was reviewed against the intent of the General

Design Criteria proposed in July, 1967. That review was documented in the VYNPS

UFSAR, Appendix F.2, Revision 17, is historical, and is not included in the DSAR.

Although changes were made to the facility over the life of the plant that may have

invoked the final General Design Criteria as design criteria, such invocation was

not intended to constitute a regulatory commitment, unless specifically docketed as

such.

The original Appendix F information, except cross-reference to applicable FSAR

Sections, is retained here for historical significance. Indications of the present

or future tense should be understood as being related to the time frame during which

this Appendix was originally written. Refer to information elsewhere in the DSAR and

in other design basis documentation to determine current design configuration.

The proposed General Design Criteria that are considered to remain applicable in the

defueled condition include the following:

Criterion 1--Quality Standards The quality assurance program is presented in the VY Quality Assurance Program Manual (VY QAPM). The description of the various systems and components includes

the codes and standards that are met in the design and their adequacy.

VYNPS DSAR Revision 0 3.0-10 of 98 Criterion 2--Performance Standards Conformance to the applicable structural loading criteria ensures that those systems and components affected by this criterion are designed and built to withstand the

forces that might be imposed by the occurrence of the various natural phenomena

mentioned in the criterion, and this presents no risk to the health and safety of

the public. The phenomena considered and margins of safety are also given.

Criterion 3--Fire Protection The materials and layout used in the station design have been chosen to minimize the possibility and to mitigate the effects of fire. Sufficient fire protection

equipment is provided in the unlikely event of a fire. Criterion 5--Records Requirement

Complete records of the as-built design of the station, changes during operation and

quality assurance records will be maintained throughout the life of the station. Criterion 11--Control Room The facility is provided with a centralized control room having adequate shielding to permit access and continuous occupancy under 10CFR20 dose limits during the

design basis accident situation.

Criterion 12--Instrumentation and Control Systems The necessary controls, instrumentation, and alarms for safe and orderly facility

operation are located in the control room. These instruments and systems allow appropriate monitoring control of the facility. Sufficient instrumentation is provided to allow monitoring of all variables necessary for effective facility

control. Criterion 17--Monitoring Radioactive Releases The station process and area radiation monitoring systems are provided for monitoring significant parameters from specific station process systems and specific

areas including the station effluents to the site environs and to provide alarms and

signals for appropriate corrective actions.

Criterion 18--Monitoring Fuel and Waste Storage The spent fuel storage areas have been analyzed to determine their safety, and instrumentation is provided for monitoring where needed.

Criterion 66--Prevention of Fuel Storage Criticality Appropriate facility fuel handling and storage facilities are provided to preclude VYNPS DSAR Revision 0 3.0-11 of 98 accidental criticality for spent fuel. Criterion 67--Fuel and Waste Storage Decay Heat The system used to cool the spent fuel pool is designed to remove sufficient decay heat to maintain the pool water temperature. The fuel storage pool contains sufficient water so that in the event of the failure of an active system component, sufficient time is available to either repair the component or provide alternate means of cooling the storage pool.

Criterion 68--Fuel and Waste Storage Radiation Shielding The handling and storage of spent fuel is done in the spent fuel storage pool. Water

depth in the pool is maintained at a level to provide sufficient shielding for normal reactor building occupancy by facility personnel. A demineralizer system is used to control water clarity and to reduce water radioactivity. Accessible portions of the reactor and radwaste buildings have sufficient shielding to maintain dose rates within the limits of 10CFR20.

Criterion 69--Protection Against Radioactivity Release From Spent Fuel and Waste Storage The consequences of a fuel handling accident in the spent fuel pool are presented

elsewhere in the DSAR. In this analysis, it is demonstrated that undue amounts of

radioactivity are not released to the public.

All spent fuel and waste storage systems are conservatively designed with ample margin to prevent the possibility of gross mechanical failure which could release significant amounts of radioactivity. Backup systems such as floor and trench drains

are provided to collect potential leakages. Appropriate facility personnel are

rigorously trained and administrative procedures are strictly followed to reduce the

potential for human error.

The radiation monitoring system is designed to provide facility personnel with early indication of possible malfunctions.

Criterion 70--Control of Releases of Radioactivity to the Environment

The station radioactive waste control systems (which include the liquid and solid radwaste systems) are designed to limit the off-site radiation exposure to levels below limits set forth in 10CFR20. 3.1.2 Classification of Structures, Systems and Components

Following certification of permanent defueling, VY is no longer authorized to

emplace or retain fuel in the reactor vessel in accordance with 10CFR50.82(a)(2).

Since it is no longer possible to load a nuclear core, power operations can no

longer occur and reactor related design basis accidents are no longer possible.

VYNPS DSAR Revision 0 3.0-12 of 98 Consequently, it was determined that the remaining design basis accident possible at VY is a fuel handling accident (FHA) consisting of a dropped fuel bundle in the

Spent Fuel Pool. As presented in the Safety Analysis chapter of the DSAR, the dose

consequences of the fuel handling accident are well within acceptance criteria, with

no reliance on either the Standby Gas Treatment System or the Secondary Containment

System.

Based on the changed conditions described above, an evaluation of the systems, structures and components (SSCs) described in the UFSAR was performed to determine

the SSC safety classification based on the function, if any, each SSC would perform

in the permanently defueled condition. The process and criteria used to classify

the SSCs and the conclusions of the evaluation are provided in appropriate station

documents.

3.1.3 Loading Considerations for Structures, Foundations, Equipment and Systems

All structures have been designed to withstand the combinations of dead and live

loads which give the severest credible conditions of loading. Loading, including

seismic, wind, and impact loading, are in accordance with the applicable codes, and

incorporate the applicable provisions of the Uniform Building Code, Zone II, 1967

Edition; ACI Standard Building Code Requirements for Reinforced Concrete (ACI

318-63); ACI Standard Specification for the Design and Construction of Reinforced

Concrete Chimneys (ACI 505-54); AISC Specification for the Design, Fabrication, and

Erection of Structural Steel for Buildings (1963); American Water Works Association, "AWWA Standard for Steel Tanks, Standpipes, Reservoirs, and Elevated Tanks for Water

Storage," AWWA D-100 (1967); USA Standards Institute ASA B96.1, "Welded Aluminum

Alloy Field-Erected Storage Tanks; National Fire Protection Association Standard

NFPA No. 30, "Flammable and Combustible Liquids Codes" (1966); Section III of the

ASME Boiler and Pressure Vessel Code, "Nuclear Vessels" (1968); and Section VIII of

the ASME Boiler and Pressure Vessel Code, "Unfired Pressure Vessels" (1968).

The Reactor Building and all other Class I structures except the main stack and

ISFSI storage pad are founded on firm bedrock. The main stack rests on end-bearing

steel piles which transfer stack loads to the bedrock. The ISFSI storage pad is

founded on engineered fill placed on existing soil.

The maximum allowable bearing pressure is 50 tons per square foot. The maximum

loading on the bedrock does not exceed 20 tons per square foot.

The maximum anticipated earthquake at the site would result in a maximum horizontal

ground acceleration of 0.07g. Facility design ensures that appropriate functions

remain available during or following a ground horizontal acceleration of 0.14g.

A strong motion, solid state digital accelerograph is floor mounted in the Reactor

Building southwest corner room, floor Elevation 213'-9". The unit is designed to VYNPS DSAR Revision 0 3.0-13 of 98 provide continuous monitoring for earthquakes by means of three orthogonal accelerometers, two horizontal and one vertical, which sense earthquake ground

motion. Once triggered, the accelerograph will record the seismic event or a series

of events as long as the trigger levels are exceeded. The primary function of the

strong motion accelerograph is to provide data which will be of value in promptly

assessing the condition of the plant subsequent to an earthquake per 10CFR100, Appendix A.

The maximum anticipated wind velocity that is anticipated at the site is 80 mph with

gusts to 100 mph. The station structures are designed to withstand the anticipated

wind loadings. The site is located in a geographic area which has a small

probability of being subjected to tornadic wind conditions.

Live loads, including construction loads, which are greater than the loads

prescribed under code, and loads from operating pressures and/or temperatures which

increase the stresses, have also been used in the design. Standard practice of use

and application in power plants determined the selection of the materials used in

the various structures and supports.

The loadings considered were as follows:

D is the dead load of structure and equipment plus any other permanent loads contributing stress such as soil, hydrostatic pressure, temperature

loading, or operating pressures.

L is the live load from any nonpermanent loads such as equipment not fixed in place, roof snow load, etc.

R is the jet force or pressure on the structure due to rupture of any one pipe. H is the force on the structure due to thermal expansion of pipes.

E is the design earthquake load.

E' is the maximum hypothetical earthquake load.

W is the load due to wind.

W' is the load due to tornado.

The loading considerations, using the postulated events, which have been followed

for all Class I structures and equipment to determine the controlling stress levels

to be used in design are:

VYNPS DSAR Revision 0 3.0-14 of 98 Loading Consideration Allowable Stress A. Reactor Building and All Other Class I Structures, excluding the primary containment

1. D+L+R+E Normal allowable code stresses are used.

The customary increase in design stresses

for the loading combinations considered is

not permitted.

2. D+L+R+E' Stresses are allowed to approach the

3. D+L+W' yield point for ductile materials, and 0.85 times the ultimate strength for concrete.

4. D+L+W Normal allowable code stresses and customary increases in stresses are used

for these load combinations.

B. Class I Tanks

1. D+L+H+E Normal allowable code stresses are

2. D+L+H+W used. The customary increases in design stresses for the load combinations

considered are not permitted.

3. D+L+H+E' Stresses are allowed to approach the yield point for ductile materials and 0.85 times

the ultimate strength for concrete.

The load combination equations listed above are based on allowable stress design. No plastic strength design for steel structures or ultimate strength design for

concrete was used for Vermont Yankee; therefore, no load factors were applied to the

subject equations. VYNPS DSAR Revision 0 3.0-15 of 98 To assure the required properties of concrete poured during cold weather, placing of

concrete with ambient temperatures around 15 o F was done with several requirements that included temperature control during the mixing, placing, and curing of the

concrete. The mixing water was heated to a temperature range of 100°F to 175°F

which was adequate to maintain a concrete temperature of +/-65°F at the point of

discharge from the mixer. This temperature is within allowable limits for proper

concrete placement. No frozen lumps of material were allowed in the charging hopper

of the batching plant. When necessary, the area of concrete placement was sheltered

for protection against the weather and preheated. This precaution was taken to

assure that no concrete would be placed against frozen surfaces. During placement

of concrete floors, heat was provided for the underside as well as the top surface.

The ambient temperature in the area of the placement was maintained at a minimum

temperature of 45°F for at least 5 days, and special coverings or enclosures were

provided to permit proper curing conditions for the concrete. Concrete specialists

were retained to design the concrete mixes, perform testing as required, and to

assist in developing an overall concrete program for the project. They also

witnessed and reported on concrete placements and were encouraged to comment on all

phases of the program including cold weather concreting.

Loading criteria and combinations for the primary containment are discussed in

Appendix A. Structural loading criteria, load combinations, allowable and actual

stresses for Class I systems, equipment, piping, supports, etc., are also discussed

in Appendix A.

Table 3.1.1 gives the maximum allowable stresses used for the various loading

conditions for Class I structures.

Floor live loads are based on equipment and operating loads and applied in

accordance with the Uniform Building Code Zone II (UBC), 1967 Edition. Roof live

loads are 40 psf applied as specified in the UBC to obtain the worst condition of

stress.

The 40 psf design roof live loads (snow loading) was determined as follows:

VYNPS DSAR Revision 0 3.0-16 of 98 The American National Standards Institute (formerly the American Standards Institute) in their "Minimum Design Loads in Buildings and other Structures," specify the weight of seasonal snowpack equaled or exceeded 1 year in 10 as the

minimum snow load for design purposes. This figure for the Vermont Yankee Nuclear

Power Station is equal to 30 pounds per square foot. Forty pounds per square foot

or 10 psf more than specified, was conservatively used for the design of the

structures. The weight of the estimated maximum accumulation on the ground plus the

weight of the maximum possible snowstorm of 70 psf, as shown in Section 2.3.5.3, is

interpreted as applicable for the drifts on the ground where accumulation is

permitted by the terrain. Winds will not permit such accumulations to occur on

building roofs of the station; therefore, the 40 psf used in the design is

considered a conservative loading.

The station masonry wall design for Class I structures is analyzed to meet the NRC

Bulletin 80-11 guidelines. The design approach and analysis used were approved in

References 1 and 2.

Floor dead loads include the weight of the structural components and the

architectural appurtenances. Operating loads consist of gravity loads from all

equipment and piping. All structures satisfy the requirements of the UBC, Zone II, 35 psf basic wind as per American Standards Association (ASA) A58.1, 1955. In

addition, the following Class I structures have been designed to withstand

short-term tornado winds up to 300 mph: Control Room Building, Reactor Building

below the refueling level, intake structure (service bay area), Turbine Building

self-contained Diesel Generator Rooms, tornado walls around outdoor condensate

storage and fuel oil storage tanks. The effect of a 300 mph wind on a Class I

structure was analyzed by applying a uniformly distributed positive pressure of

185 psf on the windward side of the structure and a negative pressure of 115 psf on

the leeward side in accordance with ASCE Wind Forces on Structures. It is assumed

that there is a 3 psi pressure drop associated with the passage of a tornado. Only

those structures which are enclosed require design against the effect of this

pressure drop. In the Reactor Building, the internal overpressure is relieved by

providing that specified areas of the siding enclosure (blowout panels) above the

refueling level will fail at an overpressure falling in the designated range of 0.35

psi to 0.60 psi (Reference 3). Subsequent pressure equalization is obtained at each

successive level below the refueling floor by means of large open hatch areas on

each floor. In the Diesel Generator Rooms of the Turbine Building, dedicated

tornado pressure relief dampers are provided which will allow the room to vent

through the intake air supply to the exterior of the building. Since the siding on

the Turbine Building will blow off with winds, such as those associated with a

tornado, the Daytank Rooms are vented into the Turbine Building by the open space

that is provided beneath their doors. These dampers and openings will provide

adequate venting capacity to limit the pressure differential on the enclosure walls.

The Control Room Building has been designed to withstand a 3 psi pressure drop

without venting. VYNPS DSAR Revision 0 3.0-17 of 98 Class I structures are also designed against penetration by tornado-created

missiles. The missiles which have been considered are 4 x 4 inch x 16 foot-long

wood posts and 2 x 12 inch x 16 foot-long wood planks. For an analysis of the

effects of a tornado on the spent fuel storage pool, see APED-5696, "Tornado

Protection for the Spent Fuel Storage Pool."

For tornado loading, metals are allowed to approach their yield point, and concrete, its ultimate strength.

3.1.3.1. Seismic Classification

The two classes of structures applicable to the earthquake design requirements are

as follows:

Class I - Structures and equipment whose failure could cause significant release of

radioactivity in excess of 10CFR100 for a low probability event or which are vital

to the removal of decay and sensible heat from the spent fuel pool.

The ISFSI storage pad is classified as Important to Safety as defined in 10CFR72.3.

The Important to Safety features of the storage pad are to maintain the conditions

required to store spent fuel safely and prevent damage to the spent fuel container

during storage.

Class II - Structures and equipment which may be essential or even nonessential to

the operation of the facility.

An analysis of the consequences of failure of several structures was performed.

This analysis showed that a condensate storage tank rupture could result in the

release of radioactivity resulting in potential doses in excess of the limits of

10CFR20 for unrestricted areas. It should be recognized, however, that this failure

constitutes an accident and that 10CFR50.67 rather than 10CFR20 applies. Within the

scope and bases used in this analysis, no Class I or II structures or equipment were

found which, upon failure, could result in doses in excess of the limits of

10CFR50.67 at the site boundary.

VYNPS DSAR Revision 0 3.0-18 of 98 3.1.3.1.1 Class I Structures The following is a complete listing of all the Class I structures of the station:

Drywell and Suppression Chamber, including vents and penetrations Reactor Building Control Room Building Plant Stack Intake Structure (service water pump area) Turbine Building (diesel generator and oil day tank areas) Cooling Tower Deep Basin Interim Spent Fuel Storage Installation (ISFSI) pad

3.1.3.1.2 Class I Equipment

The following is a complete list of all Class I equipment:

Nuclear Steam Supply Systems

- Reactor Primary Vessel 

- Reactor Primary Vessel Supports  - Fuel Assemblies 

Station Service Water System (up to the main condenser discharge block)

Fuel Storage Facilities, to include spent fuel storage equipment

Standby Electrical Power Systems

- Standby Diesel Generator System 

Instrumentation and Control Systems - Radiation Monitoring System (partial)

Diesel Oil Day Storage Tanks

Fuel Oil Storage Tank

Fuel Oil Transfer System

VYNPS DSAR Revision 0 3.0-19 of 98 3.1.3.1.3 Class II Structures Turbine Building (except as noted under Class I structures) Administration Building Intake and Discharge Structures (except as noted under Class I structures) Radwaste Building All Other Structures, not listed in Paragraph 3.1.3.1.2, that have seismic design requirements

3.1.3.1.4 Class II Equipment

Reactor Building Cranes Condensate Storage Transfer System Station Auxiliary Power Busses Electrical Controls and Instrumentation (for above systems) Radwaste System All Other Piping and Equipment, not listed in Paragraph 3.1.3.1.3, that have seismic design requirements

3.1.3.2 Seismic Design

All Class I structures were designed conservatively so that under the worst loading

conditions the allowable stresses will not be exceeded. Several critical portions

of major Class I structures are listed in Table 3.1.2, showing the margins of safety

for the controlling loads for the listed structural member.

No. 1 shows (a) the circumferential stresses in the reactor pedestal due to a jet force and (b) the vertical stresses at the base of the pedestal due to

direct load plus earthquake plus jet force.

No. 2 shows the stresses due to dead and live load plus design base or maximum hypothetical earthquake at the face of the biological wall and at midspan of an

important beam in the Reactor Building. This beam supports part of the floor

deck at Elevation 280 plus an interior column which extends up to the refueling

floor. No. 3 shows the stresses under the same loading conditions as in No. 2 in a footing supporting a column of the Control Room.

No. 4 shows the stresses in the south wall of the housing of the service water pumps in the intake structure under tornado wind load.

VYNPS DSAR Revision 0 3.0-20 of 98 Based on seismological investigations, response spectra and dynamic analyses established for the station, envelopes of maximum acceleration, displacement, shear, and overturning moment versus height have been developed. The horizontal ground

acceleration for the design earthquake is 0.07 times gravity (0.07g), and the

vertical motion 2/3 that of the horizontal. Both motions are assumed to occur

simultaneously.

3.1.3.2.1 Class I Structures

Mathematical models whose properties correspond to those of the structures or

equipment were formulated. The seismic design for the Class I structures and

equipment is based on dynamic analysis using the acceleration response spectrum

curves. The design is such that safe shutdown can be made during a ground motion of

0.14g, combined with the vertical accelerations assumed to be 2/3 of the horizontal

ground acceleration, with no variation of the vertical coefficients with height.

For the dynamic analysis of Class I structures, the damping factors used for

vibrations below the elastic limit are as follows:

VYNPS DSAR Revision 0 3.0-21 of 98 Item Percent of Critical Damping Reinforced Concrete Structures 5.0 Steel Frame Structure 2.0 Bolted or Riveted Assembly 2.0 Welded Assembly (equipment and supports) 1.0 Vital Piping System See Appendix A Wood Structures with Bolted Joints 5.0

Summaries of the seismic analyses for the Reactor Building, Control Room Building, plant stack, Turbine Building, intake structure, and deep basin are given in the

Facility Structures section under the respective structure. The detailed analyses

for these structures are contained in Appendix A. Detailed analyses for the ISFSI

storage pad are contained in References 4, 5, 6, 7 and 8.

3.1.3.2.2 Class II Structures

Design was in accordance with the provisions of the Uniform Building Code, Zone II.

Alternately, all such structures were designed to resist a minimum horizontal

seismic coefficient of 0.05, with a 1/3 allowable increase in basic stresses.

3.1.3.2.3 Equipment Seismic Design

Class I equipment analysis considers vertical and horizontal ground motions. The

coefficients for horizontal motion were adjusted to correct for equipment elevation

above grade, and also consider the stiffness of the equipment supports. The

magnitude of the vertical acceleration used was 2/3 of the horizontal ground

acceleration with no variation of the vertical coefficients with height. Allowable

stresses are in accordance with Table 3.1.1.

Stresses have also been checked for an earthquake with two times the seismic

coefficients. Class I equipment is bolted or fastened so that it will not be

displaced.

All Class I tanks have been analyzed for forces resulting from a horizontal

acceleration of 0.22g acting simultaneously with a vertical acceleration of 0.05g.

These accelerations take into account the height above grade of the various Class I

tanks. Stresses have been kept within the basic code allowables with no increase

for short-term loading. Further analysis was performed using twice the horizontal

and vertical accelerations, and for this condition of loading, stresses in the

ductile materials have been permitted to go to 0.90 of yield.

VYNPS DSAR Revision 0 3.0-22 of 98 The selection of the horizontal seismic loading coefficient for the Class I tanks was based on the maximum acceleration at the elevation of the tank in the supporting

structure. This permits maximum flexibility in arrangement of the vital tanks

within the structures and ensures no condition of overstress due to seismic loading.

For Class II equipment, the seismic analysis has assumed there is no vertical ground

motion. This is in accordance with the Uniform Building Code, Zone II. The

horizontal motion has been adjusted to correct for equipment elevation above grade.

Code allowable stresses with increase for short-term loading have been maintained.

Class II tanks are analyzed for forces resulting from a horizontal acceleration of

0.09g, with allowable stresses increased by 25% in accordance with the provisions of

the Uniform Building Code.

The selection of the seismic acceleration coefficients for the Class II tanks also

reflects the upper elevations of the structures where these tanks are located.

Class I equipment is principally supported on reinforced concrete. In the Reactor

Building, all supporting concrete has a minimum 4,000 psi, 28-day ultimate

compressive strength. All other supporting concrete has a minimum 3,000 psi, 28-day

ultimate compressive strength. All reinforcing has a minimum yield stress of 40,000

psi. Where structural steel is used to support Class I equipment, ASTM A36 standard

rolled shapes or other material analyzed to meet the requirements of this section

are used. The allowable stresses are as listed in Table 3.1.1.

The temperature differential used in the design calculations for the induced thermal

stresses in the biological wall was 45°F. It was derived as follows. The

temperature inside the drywell was assumed to be 135°F. The temperature at the

outside face of the steel shell was calculated to be 106°F. The temperature at the

interior face of the biological wall was calculated to be 115.7°F, and the

temperature at the outside face of the biological wall was calculated to be 70.7°F.

The same thermal stresses used in the design of the biological wall were also

applied at the top of the structural beads which support the lower portion of the

drywell and reactor operating floor at Elevation 252'-6".

Reactor and drywell supports are shown on Drawings G-191529 and G-191483. VYNPS DSAR Revision 0 3.0-23 of 98 3.1.4 References

1. Letter, G. Lainas (USNRC) to J. B. Sinclair (VYNPC), "Masonry Wall Design, IE Bulletin 80-11," NVY 83-262, dated November 15, 1983.

2. Letter, D. B. Vassallo (USNRC) to R. W. Capstick (VYNPC), "Masonry Wall Design Supplement - Inspection and Enforcement Bulletin 80-11," NVY 85-240, dated

November 18, 1985.

3. Calculation VYC-1828, "Reactor Building Masonry Wall Review for HELB Loadings."

4. Calculation VYC-2427, "Development of Acceleration Time Histories for Vermont Yankee ISFSI Analysis."

5. Calculation VYC-2428, "Development of Strain Compatible Soil Properties for Vermont Yankee ISFSI Analysis."

6. Calculation VYC-2433, "Soil Structure Interaction Analysis of the Vermont Yankee ISFSI."

7. Calculation VYC-2435, "Vermont Yankee Nuclear Power Plant ISFSI Facility Concrete Storage Pad Design"

8. Calculation VYC-2434, "Vermont Yankee ISFSI Cask Sliding Analysis."

VYNPS DSAR Revision 0 3.0-24 of 98 TABLE 3.1.1 Allowable Stresses for Class I Structures Loading Conditions Reinforcing Steel Maximum Allowable Stress Concrete Maximum Allowable Compressive Stress Concrete Maximum Allowable Shear Stress Concrete Maximum Allowable Bearing Stress Structural Steel Tension on Net Section Structural Steel Shear on Gross Section Structural Steel Compression on Gross Section Structural Steel Bending 1. Loading as defined without E', W and W' 0.50 Fy 0.45 f c 1.10 f c 0.25 f c 0.60 Fy 0.40 Fy Varies with slenderness ratio 0.66 Fy to 0.60 Fy 2. Loading as defined excluding E, E' and W' 0.667 Fy 0.60 f c 1.467 f c 0.333 f c 0.80 Fy 0.53 Fy Varies with slenderness ratio 0.88 Fy to 0.80 Fy 3. Loading as defined with E' or W' present Seismic load (0.14g) See Note A 0.85 f c -- -- See Note A 0.60 Fy Varies with slenderness ratio See Note A

• 25% Live Load is considered concurrent with seismic load.

Fy is the minimum yield point of the steel used. f c is the compressive strength of concrete. Note A: Stresses permitted to approach but not exceed yield stress of the material.

VYNPS DSAR Revision 0 3.0-25 of 98 TABLE 3.1.2 Safety Margins for Several Critical Portions of Major Class I Structures Controlling Loading Allowable Stress in psi Actual Stresses in psi Safety Margins (Allowable/Actual) Structure Condition Concrete Reinforcing Concrete Reinforcing Concrete Reinforcing

1. RPV Pedestal a. Circumferential Stresses .........

b. Vertical Stresses ................

R D+L+E+R 3400 1800 36,000 20,000 1720 877 34,000 19,644 1.97 2.06 1.06 1.02 2. RB Biological Wall Beam a. At face .......................... b. At face .......................... c. At midspan ....................... d. At midspan ....................... D+L+E D+L+E' D+L+E D+L+E' 1800 3400 1800 400 20,000 36,000 20,000 36,000 950 1080 1640 2380 18,100 20,800 16,400 23,600 1.90 3.14 1.16 1.43 1.10 1.73 1.22 1.53 3. Control Room Footing a. Face of column ................... b. Face of column ................... D+L+E D+L+E' 1350 2550 20,000 36,000 450 810 15,600 28,200 3.00 3.15 1.28 1.28 4. Intake Structure Service Bay a. South enclosure .................. D+L+W' 2550 36,000 300 15,000 8.50 2.40

NOTE: Loads as defined in Section 3.1.3.

VYNPS DSAR Revision 0 3.0-26 of 98 3.2 FACILITY STRUCTURES 3.2.1 Reactor Building

3.2.1.1 Function

The Reactor Building encloses the spent fuel storage pool.

3.2.1.2 Description

The Reactor Building is constructed of monolithic reinforced concrete floors and

walls to the refueling level. Above the refueling level, the structure consists of

steel framing covered by insulated sealed siding and roof decking. The siding and

roofing can withstand a limited internal overpressure before pressure relief is

obtained by venting through the refuel floor blowout panels designed to release at

an overpressure falling in the designated range of 0.35 psi to 0.60 psi (Reference

1).

A 110/7.73 ton capacity overhead bridge crane provides services for the reactor and

refueling area. The crane is designed to remain on the rails and retain its load

with a 0.2g seismic loading. The Reactor Building bridge crane is of Class II

seismic design. Accordingly, the coefficient of 0.20g was specified based on the

building response of the level of crane supports under 0.07g minimum ground

acceleration. The crane supports are of Class I seismic design. The crane bridge

and trolley wheels are provided with seismic hold-down lugs to assure crane

stability in the event of a maximum hypothetical earthquake.

Reference 2 details the commitments to control the handling of heavy loads, including the specific commitments made during the submittal process to the NRC, as

input to their Safety Evaluation Report, and how they are implemented at Vermont

Yankee. The Reactor Building overhead bridge crane trolley was modified to provide

redundancy in the load carrying path from the load to the crane itself, so that no

single failure would allow the load to drop. All components in the load path of the

main hoist are either redundant or designed with a large factor of safety, and are

structurally adequate to maintain the load capacity, as well as any transfer loads

should one path fail. Each load path for the main hook consists of a hook or

attachment point, load block, cable, reversing sheaves, drum, gear drive, and

brakes. Sheaves and blocks are captured so that failure would not result in

uncontrolled descent of the load. Redundant limit switches, of different types, are

provided to prevent over-hoisting, and a load indicating/limiting device prevents

overloading. An overspeed switch is provided on each load path to prevent runaway

lowering. Operating power and control for all crane motions are provided by a

control system which incorporates a torque limiter on the main hoist for additional

overload protection. VYNPS DSAR Revision 0 3.0-27 of 98 When moving a spent fuel shipping cask, the crane speeds are reduced and the travel path limited to prevent the cask from passing over the stored spent fuel.

The crane was designed in accordance with the Electric Overhead Crane Institute (EOCI) Specification No. 61 and, with minor exceptions, meets all requirements of

the Crane Manufacturers Association of America (CMAA) Specification No. 70.

The primary containment structure is an integral part of the Reactor Building and

occupies the core of the building. The spent fuel storage pool is located in the

Reactor Building. Access to the drywell and reactor head space is obtained by

removing a large segmented concrete plug in the refueling level floor by means of

the bridge crane. The crane also handles the drywell head, the reactor vessel head, the segmented pool plugs, and the spent fuel shipping cask. A refueling platform, with the requisite handling and grappling fixtures, services the spent fuel storage

pool. A passenger-freight elevator is provided for access to the various floors

above grade level.

The steel drywell vessel is fixed to the building along its lower portion, and is

laterally supported by the building along its upper portion. Within the drywell, a

cylindrical sacrificial shield structure surrounds the reactor vessel.

There is a remote possibility that the height of ground water during a given period

could exceed the elevation of the extreme lower portion of the drywell.

Nevertheless, it is not considered possible for this ground water to reach the steel

plating, assuming a crack in the foundation concrete. The bases for this conclusion

are as follows:

1. The monolithic foundation concrete structure is greater than 18 feet thick below the drywell and is divided into three separate pours in the horizontal

plane. It is considered almost impossible for a crack to propagate completely

through any given pour because of the thicknesses involved and the bedrock

foundation. Even if this were to occur, it is not considered possible for any

given crack to propagate beyond the joint between pours.

2. Water-stop material is used at all foundation concrete joints between pours, both in the horizontal and vertical planes. This design assures that water

will not propagate along any given joint in the concrete.

The possible effects of a given thermal gradient through the foundation concrete

have been considered. Based on the concrete thicknesses and possible temperature

differentials, it is not considered possible for any thermal gradients to exist

which would damage or otherwise affect the structural integrity of the concrete.

Therefore, thermal gradients are not considered a factor in the above discussion on

foundation cracking. VYNPS DSAR Revision 0 3.0-28 of 98 The reinforced concrete portion of the Reactor Building has been designed against tornado missiles. Pressure relief below the refueling level is obtained through

large open hatches.

The general arrangement of the Reactor Building and the principal equipment is shown

on Drawings G-191148, G-191149 and G-191150.

3.2.1.3 Seismic Analysis

Dynamic earthquake analysis was made of the coupled Drywell/Reactor Building System

for an empty and flooded condition of the drywell. A separate analysis was made for

the pressure suppression chamber.

The effect of the adjacent Class II Turbine Building has been considered, and the

analysis shows that failure of the adjacent Turbine Building will not compromise the

integrity of the Class I Reactor Building in the event of a design basis or maximum

hypothetical earthquake.

The sacrificial shield wall and reactor pedestal are hollow cylinders of uniform

thickness connected by anchor bolts embedded in the top of the pedestal.

The pedestal carries the vertical load of the sacrificial shield wall including the

loads transmitted to it. The pedestal is supported at Elevation 238.0' by a

concrete foundation which rests on the lower part of the containment vessel.

Moments, vertical loads, and horizontal forces from the reactor pressure vessel, pedestal, and drywell are transmitted to the supporting drywell foundations in the

following manner:

The reactor pressure vessel transmits vertical loads and shears directly to the

drywell foundation through the vessel skirt into the reactor pedestal via shear

rings welded to the inner skirt. The vertical and horizontal loads from the

pedestal are transferred to the interior and exterior surface of the drywell by a

combination of bond and friction forces between steel and concrete contact surfaces.

The contact between the exterior surface of the drywell and the supporting concrete

foundation is assured by the pressure grouting method used for the concreting of the

foundation itself. Additional resistance to shear is afforded by the physical

characteristics of the drywell which, in its lower portion, can be considered as a

bowl embedded in the supporting reinforced concrete foundation. No increase in

allowable stresses was permitted in any of the above considerations.

The stresses resulting from the maximum hypothetical earthquake were also checked to

make sure that their value was below allowable limits.

VYNPS DSAR Revision 0 3.0-29 of 98 The interaction of the drywell base with the exterior concrete is comprised of bonding and friction, and it is a result of these phenomena that the relative shears

are handled.

The phenomenon of bonding, although a significant contributory factor, is ignored

for conservatism. Extreme care is exercised in placing the grout between the

drywell base and the exterior concrete. This provides adequate assurance that there

are no significant voids in this area and that the actual drywell contact area is

high. In addition to providing significant bonding, this surface area also provides

a large contact area to resist relative shears through friction.

The vertical load transmitted through the drywell is approximately 8,230k. The

horizontal load resulting from a maximum hypothetical earthquake is 3,165k. To be

conservative, the calculations assume that vertical, horizontal and moment forces

are transmitted from the drywell to the foundation mat by the reactor vessel skirt

alone. It is further assumed that the reactor vessel skirt, welded to the drywell, will transmit the horizontal forces by bearing against the fill concrete surrounding

it. For conservatism, only the top two feet of the skirt were considered as

transmitting the load.

The concrete stresses and welding stresses were checked against the allowable

stresses to determine if the skirt and the surrounding concrete can withstand the

horizontal forces. The concrete stress is 638 psi, which is less than the 1,000 psi

allowed by ACI 318, 1963. The unit shear stress on the skirt weld is 488 psi, which

is small in comparison with the load-carrying capability of the weld.

The ability of the foundation mat to resist shear forces was also investigated. No

credit was taken for the anchor bolts which fasten the skirt to the foundation mat, and friction alone is assumed to resist shear forces. A coefficient of friction was

conservatively assumed to be 0.4, which results in a shear resisting force

capability of 3,292k. As the maximum horizontal load is 3,165k, the adequacy of the

foundation mat is demonstrated. VYNPS DSAR Revision 0 3.0-30 of 98 3.2.2 Turbine Building 3.2.2.1 Function

The Turbine Building provides space for auxiliary equipment.

3.2.2.2 Description

The Turbine Building is a Class II structure except for the portions that support

and protect the diesel generators and oil day tank areas which are Class I seismic

design and strengthened where required to meet Class I requirements. The equipment

layout and principal dimensions are shown on Drawings G-191143, G-191144, G-191145, G-191146 and G-191147.

The superstructure portion of the building consists of steel framing with insulated

metal siding and roof decking. A 140-ton capacity main lift bridge crane provides

service for the generator and related equipment.

The Turbine Building siding contains blowout panels that vent the Turbine Building

during tornado conditions. The panels are designed to blow out at 0.5 psid and are

held in place by shear pins.

3.2.2.3 Seismic Analysis

A seismic investigation of the Class I portion of the Turbine Building was performed

analyzing the Diesel Generator Room and those areas of the Turbine Building which

could affect the Diesel Generator Room in the event of an earthquake.

The emergency diesel generators are supported on an isolated Class I foundation.

The emergency diesel generators were assigned horizontal and vertical seismic

coefficients developed from the ground response spectra for 1% damping, as described

in Appendix A. The walls of the Diesel Generator Room were designed and analyzed as

freestanding cantilevers fixed at the base mat. The enclosure was then dynamically

analyzed using 0.07g and 0.14g horizontal ground accelerations and 0.05g and 0.10g

vertical accelerations for the design basis and maximum hypothetical earthquakes, respectively. Stresses were within the allowable values shown in Table 3.1.1. The

function of diesel generators is therefore not compromised by structural failure, even in the event of a maximum hypothetical earthquake, and their functional

integrity is assured as discussed above.

VYNPS DSAR Revision 0 3.0-31 of 98 3.2.3 Plant Stack 3.2.3.1 Description

The plant stack provides an elevated point for the release of gases to the

atmosphere from portions of the Turbine Building, Reactor Building, and Radwaste

Building. Stack drainage is routed to the Liquid Radwaste Collection System via

loop seals.

The plant stack is designed for dead load, wind load, seismic load, and effects of

exhaust gas temperature. The plant stack is provided with appurtenances such as

aviation obstruction lights and isokinetic samplers for radiation monitoring, and is

designed in accordance with all applicable codes.

The unlined, freestanding, tapered, reinforced concrete stack has the following

dimensions:

Overall height above foundation 318 ft Inside diameter at top 7 ft Outside diameter at base 27.5 ft Thickness at top 0.67 ft

A schematic of the stack geometry appears in Figure 3.2-18.

3.2.3.2 Seismic Analysis

Dynamic analysis was made of the reinforced concrete ventilation stack.

The foundation material for the site is such that rocking effects are small and were

neglected for the dynamic analysis of the stack. Due to its geometry, the stack is

very flexible, with the natural period of vibration of the first mode equal to about

1.5 seconds. The spectral accelerations for periods higher than this decrease with

an increase in period. The model used for the dynamic analysis of the stack

conservatively assumed a fixed base. The damping value used in the analysis for the

responses to both the design basis and maximum hypothetical earthquakes was 5%.

The controlling loading conditions were the maximum hypothetical earthquake in the

region approximately 120 feet from top and the wind loading for the remaining

portion of the stack. For the maximum hypothetical earthquake, the maximum

calculated stress in the reinforcing steel was 35.4 ksi or 0.86 Fy, and the calculated maximum stress in concrete was 1.32 ksi or 0.377 f c. VYNPS DSAR Revision 0 3.0-32 of 98 3.2.4 Control Room Building 3.2.4.1 Description

The Control Room Building houses all required instrumentation and controls. The

instrumentation is located in the Main Control Room. The cable vault and Switchgear

Room occupy the lower levels of the building. The location of the Control Room

Building is shown on Drawing G-191142. The building is a reinforced concrete

structure and is entirely of Class I seismic design.

Plan and elevation views of the building are shown on Drawings G-191592 and

G-191595.

3.2.4.2 Seismic Analysis

A dynamic earthquake analysis was performed on the Control Room Building utilizing a

four-mass analytical model. The effect of the adjacent Class II Turbine Building

has been considered, and the results of the analysis show that failure of the

adjacent Class II structure will not compromise the integrity of the Class I Control

Room Building in the event of a design basis or maximum hypothetical earthquake.

3.2.5 Circulating Water Intake and Discharge Structures

3.2.5.1 Intake Structure

3.2.5.1.1 Description

A reinforced concrete, single unit intake structure on the riverbank east of the

station, is supported on rock. A partial enclosure is provided at the pumps. The

following equipment is provided at the intake: manually raked coarse trash racks, regulating sluice gates; traveling screens; provisions for stoplogs; two fire water

pumps; two service water pumps and two radwaste dilution pumps.

The deck of the structure is at Elevation 237' MSL and the invert at Elevation 190'

MSL. The intake has service water bays for two service water pumps, two fire water

pumps, and two radwaste dilution pumps. Bays are provided with trash rack and

stoplog guides, traveling screens, and fine screen guides.

Water from the pond flows into service water bays at the north end of the intake

structure. These bays furnish water for the fire pumps, intake service water pumps, and radwaste dilution pumps.

Retaining walls are provided at the front face of the intake structure to retain

fill.

VYNPS DSAR Revision 0 3.0-33 of 98 The intake structure is shown on Drawings G-191451, G-191452 and G-191453.

3.2.5.1.2 Seismic Analysis

A dynamic earthquake analysis has been made of the intake structure. This analysis

verifies the adequacy of the design of the intake structure to withstand seismic

forces. The effect of adjacent Class II intake structures has been considered, and

the results of the analysis show that failure of an adjacent Class II structure will

not compromise the integrity of the Class I bay housing the service water pumps in

the event of a design basis or maximum hypothetical earthquake.

3.2.5.2 Discharge and Aerating Structure

A reinforced concrete discharge-aerating structure supported on rock and piles is

located near the riverbank south-southeast of the station. It is approximately 188

feet long by 108 feet wide by 46 feet deep. The top of the deck is at Elevation

248' MSL. Water elevation for siphon operation will be maintained by a reinforced

concrete weir. The top of the weir is at Elevation 225' MSL. An aerating spillway

concrete structure is adjacent and downstream of the discharge structure to provide

air entrainment, energy dissipation, and warm water dispersion of discharged water.

Sheet piling is used to prevent scour of the aerating apron.

The discharge and aerating structure is shown on Drawings G-191463, G-191461, Sh. 1

and G-200347.

3.2.6 Cooling Tower Deep Basin

The basin has been dynamically analyzed for 0.07g and 0.14g horizontal ground

accelerations; vertical accelerations were taken as 0.05g and 0.10g for the design

basis and maximum hypothetical earthquake, respectively.

The effect of adjacent Class II structures has been considered, and the analysis

show that a failure of the Class II adjacent cooling tower structures will not

compromise the integrity of the deep basin in the event of a design basis or maximum

hypothetical earthquake.

VYNPS DSAR Revision 0 3.0-34 of 98 3.2.7 Interim Spent Fuel Storage Installation 3.2.7.1 Description The ISFSI Storage Pad is a monolithic reinforced concrete slab supported by compacted structural fill placed on existing soils. The storage pad is sized to

provide structural support for up to 36 spent fuel storage casks arranged in a 5 X 8

array. Four extra positions are provided to include sufficient room to be able to

access and move any individual cask should the need arise. The spent fuel storage

casks are free standing on the pad. Conduits for temperature monitoring are run on

top of the slab to serve each cask as they are placed. Each cask will be grounded

to plates embedded in the storage pad. The top of the pad elevation is established

at El. 254'-0" to ensure that the ventilation inlets at the bottom of the spent fuel

storage casks remain above the Probable Maximum Flood (PMF) elevation including wave

run-up. The spent fuel cask manufacturer's Final Safety Analysis Report (Reference 3) requires that for free standing casks several criteria must be met to ensure that

the design features of the cask that protect the spent fuel from a cask drop or non-

mechanistic tip-over event are not jeopardized. These criteria are that the

thickness of the pad does not exceed 36 inches, the 28 day concrete compressive

strength must not be less than 3000 psi and must not exceed 4200 psi, the specified

minimum yield strength for the reinforcing steel be 60 ksi, and that the subgrade

modulus of elasticity not exceed 28,000 psi. 3.2.7.2 Seismic Analysis A dynamic analysis of the ISFSI storage pad was performed. This analysis is composed of several parts. A subsurface investigation was performed to establish

bedrock elevations and soil properties beneath the pad (Reference 4). A set of

three artificial time histories for the Design Basis Earthquake were developed for

input to the seismic analysis (Reference 5). These time histories envelope the

design response spectra for the site, the North 69º West component of the Taft

Earthquake, normalized to 0.14g for the Design Basis Earthquake. The earthquake is

applied at the bedrock elevation under the storage pad. Analysis was then performed

to obtain strain compatible soil properties and to propagate the earthquake motion

from the bedrock to the ground surface. Since the bedrock under the storage pad is

sloping, this analysis was performed for two profiles, one with a bedrock depth of

22' and one with a bedrock depth of 32'. This analysis is further described and

provided in Reference 6. A soil structure interaction (SSI) analysis was then

performed to determine the acceleration at the center of gravity and at the base of

the casks. This analysis was performed using three separate soil cases (upper

bound, best estimate, and lower bound). The analysis also considered two soil

profiles to represent the sloping bedrock. The SSI analysis evaluates multiple cask

configurations to insure the maximum effect on the storage pad is enveloped. The

soil structure interaction analysis is further described and presented in Reference

4.

VYNPS DSAR Revision 0 3.0-35 of 98 The results of the soil structure interaction analysis are used to perform a sliding analysis and the storage pad design. The sliding analysis determines the potential

for the casks to:

(1) slide into each other, and (2) uplift a seismic event. 

The sliding analysis evaluated coefficients of friction ranging from 0.0 to

stimulate icing conditions on the pad up to a maximum of 0.8. The results of the

analysis show that the maximum horizontal displacements of the casks for any

condition are much smaller than half the free distance between the casks and much

less than the distance between the edge of the external casks and the edge of the

pad. This analysis also shows that the casks are stable and remain upright. The

sliding analysis is provided in Reference 8. Reference 9 provides the analysis to

determine the internal forces on the storage pad for all loading conditions, including seismic, and the design of the reinforcement for the storage pad.

VYNPS DSAR Revision 0 3.0-36 of 98 3.2.8 References

1. Calculation VYC-1828, "Reactor Building Masonry Wall Review for HELB Loadings."

2. PP 7023, "Control of Heavy Loads Program Document."

3. Final Safety Analysis Report for the Holtec International Storage and Transfer Operation Reinforced Module Cask System (HI-STORM 100 Cask System), NRC Docket

No. 72-1014, Holtec Report HI-2002444, Volume I and II of II, prepared by Holtec

International, Marlton, New Jersey.

4. Geotechnical Engineering Report, Proposed ISFSI Pad and Haul Path - Vermont Yankee, prepared by GZA GeoEnvironmental, Inc., Manchester, New Hampshire, January 2004

5. Calculation VYC-2427, "Development of Acceleration Time Histories for Vermont Yankee ISFSI Analysis."

6. Calculation VYC-2428, "Development of Strain Compatible Soil Properties for Vermont Yankee ISFSI Analysis."

7. Calculation VYC-2433, "Soil Structure Interaction Analysis of the Vermont Yankee ISFSI."

8. Calculation VYC-2434, "Vermont Yankee ISFSI Cask Sliding Analysis."

9. Calculation VYC-2435, "Vermont Yankee Nuclear Power Plant ISFSI Facility Concrete Storage Pad Design"

VYNPS DSAR Revision 0 3.0-37 of 98 Vermont Yankee Defueled Safety Analysis Report Revision 0 Main Stack Geometry Figure 3.2-18 VYNPS DSAR Revision 0 3.0-38 of 98 3.3 SYSTEMS 3.3.1 Fuel Storage and Handling 3.3.1.1 Nuclear Fuel 3.3.1.1.1 Objective The nuclear fuel provides a high integrity assembly containing fissionable material which could be arranged in a critical array. The assembly efficiently transfers

decay heat to the spent fuel pool water while maintaining structural integrity and

containing the fission products.

3.3.1.1.2 Description

A fuel assembly consists of a fuel bundle, channel fastener, and the channel which

surrounded it. Each fuel assembly was designed as Class I seismic design equipment.

A fuel bundle contains fuel rods and water rods, spaced and supported in a square

array by a lower tie plate, spacers, and an upper tie plate. The lower tie plate was

formed and machined to fit into the fuel support piece. The lower tie plate for the

GE13, GE14 and GNF2 fuel bundles also includes a debris filter. The upper tie plate

has a handle for transferring the fuel bundle from one location to another. The

identifying assembly number is engraved on the top of the handle and a boss projects

from one side of the handle to aid in assuring proper fuel assembly orientation.

The tie plates were fabricated from corrosion resistant materials. The fuel spacer

grids, which are positioned along the length of the fuel bundle, are made of

Zircaloy with Inconel springs. The GE13 and GE14 fuel spacer grids, which are

positioned along the length of the fuel bundle, are made of Zircaloy with alloy X750

springs. The GNF2 spacer is made entirely from alloy X750. The primary function of

the spacer grid is to provide lateral support and spacing of the fuel rods.

Each fuel rod consists of fuel pellets stacked in a Zircaloy cladding tube which is

evacuated, pressurized with helium, and sealed by welding Zircaloy end plugs in each

end. The fuel rod cladding thickness is adequate to be "free-standing", i.e., capable of withstanding external reactor pressure without collapsing onto the

pellets within. Although most fission products were retained within the UO2, a

fraction of the gaseous products were released from the pellet and accumulated in a

plenum and the gap between the pellet stack and the clad. Sufficient plenum volume

was provided to prevent excessive internal pressure from these fission gases or

other gases liberated over the design life of the fuel. A plenum spring, or

retainer, is provided in the top plenum space to minimize movement of the fuel

column during handling or shipping. Rigid precautions are taken to prevent cladding

damage due to excessive hydrogen bearing materials. These precautions may include a

hydrogen getter in the plenum to absorb hydrogen accidentally admitted during the

fabrication process. VYNPS DSAR Revision 0 3.0-39 of 98 Eight fuel rods (called tie rods) in each bundle have end plugs which thread into

the lower tie plate and extend through the upper tie plate. Stainless steel nuts

and locking tab washers are installed on the upper end plugs to hold the assembly

together. These tie rods support the weight of the assembly only during fuel

handling operations when the assembly hangs by the handle. The remaining fuel rods

in a bundle have end plug shanks which fit into locating holes in the tie plates.

An Inconel-expansion spring located over the top end plug shank of each full length

fuel rod keeps the fuel rods seated in the lower tie plate and allows them to expand

axially by sliding within the holes in the upper tie plate to accommodate

differential axial expansion. Part length rods use a threaded lower end plug which

screws into the lower tie plate. These rods terminate near one of the spacer grids

short of the upper tie plate.

Each fuel bundle may contain one or more empty Zircaloy tubes called water rods.

Perforations at each end of the water rod(s) permit coolant flow through the tube.

Tabs are fixed at axial intervals on one or more water rods to locate the spacer

grids. Water rods provide additional moderator throughout the height of the

assembly.

The fuel is in the form of cylindrical pellets manufactured by cold pressing and

sintering uranium dioxide powder. The average density of the pellets in the core is

approximately 96.5% of the theoretical density of UO2. Ceramic uranium dioxide is

chemically inert to the cladding at operating temperatures and is resistant to

attack by water.

Several different U-235 enrichments may be used in each fuel assembly. Fuel design, manufacturing, and inspection procedures have been developed to prevent errors in

enrichment location within the fuel assembly. The fuel rods have unique

identification numbers. Rigid inspection techniques utilized during and following

assembly ensure that each fuel rod is in the correct position within the bundle.

Selected fuel rods contain gadolinia as a burnable poison for reactivity control.

The gadolinia is uniformly dispersed within the fuel pellets. However, the

gadolinia-bearing pellets are not uniformly distributed within the fuel rods, but

are grouped together into axial zones. These axially zoned regions of varying

gadolinia content provide reactivity control which enhances shutdown margin and/or

power distribution control to reduce axial peaking. U-235 enrichment is also zoned

axially to compliment the function of the gadolinia, and provide a more economical

fuel cycle.

VYNPS DSAR Revision 0 3.0-40 of 98 The fuel channel enclosing the fuel bundle is fabricated from Zircaloy and, if installed, performs the following functions:

1. Provides structural stiffness to the fuel bundle during lateral loading applied from fuel rods through the fuel spacers.

2. Transmits fuel assembly seismic loadings to the top guide and fuel support of the core internal structures.

The channel makes a sliding seal fit over finger springs attached to the lower tie

plate. The channel is attached to the upper tie plate by the channel fastener

assembly which is secured by a cap screw. Spacer buttons are located on the two

sides of the channel adjacent to the channel fastener assembly to maintain bundle

separation and form a path for the control blades in the core cell.

GNF2 fuel assemblies are arranged in a 10X10 array with two central water rods, as

well as both short and long partial length rods. Some of the design features

include the following:

• Improved part-length rod configuration for improved Cold Shutdown Margin (CSDM) and efficiency.
• Modified fuel rod clad thickness to diameter ratio (T/D) with increased uranium mass for increased bundle energy.
• Modified channel that interacts with the LTP to control leakage flow while eliminating finger springs for ease of channeling operations.
• Improved Inconel X-750 grid type spacer with Flow Wings for increased margin to Boiling Transition and reduced pressure drop.
• Defender Debris Filter Lower Tie Plate for improved resistance to the intrusion of foreign material.
• High volume pellet for increased uranium mass and manufacturing quality control.
• Locking retainer spring that restrains the fuel column during shipping and supports a wide range of column lengths.
• A non-Zircaloy 2 zirconium alloy, Ziron, is used for the fuel cladding material for 24 rods in 2 of the 4 GNF2 LUAs.

VYNPS DSAR Revision 0 3.0-41 of 98 The external envelope of GNF2 is virtually identical to GE14 and the nuclear characteristics of the GNF2 are compatible with current vintage GE14. The thermal

hydraulic characteristics of GNF2 design closely match the overall pressure drop of

previous designs.

Licensing analyses of the GNF2 LUAs have been conducted using NRC approved methods, which are capable of evaluating/analyzing all of the LUA features.

3.3.1.2 Spent Fuel Storage

3.3.1.2.1 Objective

The spent fuel storage arrangement provides specially designed underwater storage

space for the spent fuel assemblies which require shielding during storage and

handling.

Storage of spent fuel in dry casks at the Independent Spent Fuel Storage

Installation facility is licensed in accordance with 10CFR72 and is not within the

scope of the 10CFR50 Updated Final Safety Analysis Report.

3.3.1.2.2 Design Bases

1. The spent fuel pool is designed for a maximum of twelve spent fuel storage racks with a maximum capacity of 3,353 spent fuel assemblies.

2. Spent fuel storage racks shall be designed and arranged so that the fuel assemblies can be efficiently handled.

3. The fuel array in the fully loaded spent fuel racks shall be substantially subcritical such that k eff is less than or equal to 0.95.

4. Each spent fuel storage rack shall be designed to withstand earthquake loading to prevent significant distortion of spent fuel storage arrangement when empty, half-full, or fully loaded with fuel.

3.3.1.2.3 Description

The spent fuel storage racks provide storage at the bottom of the fuel pool for the

spent fuel received from the reactor vessel, as shown in Figure 3.3.1-1. The racks

are full length, top entry, and designed to maintain the spent fuel in a space

geometry which precludes the possibility of criticality under normal and abnormal

conditions. Normal conditions exist when the spent fuel is stored at the bottom of

the fuel pool in the design storage position. Abnormal conditions may result from

an earthquake or mishandling of equipment.

VYNPS DSAR Revision 0 3.0-42 of 98 The normal arrangement of the spent fuel storage racks consists of nine NES manufactured racks (Drawing 5920-6893) and two Holtec manufactured racks (Drawing

5920-12795), giving a total capacity of 3087 assemblies. A twelfth rack can be

installed in the cask lay-down area as shown on Drawing 5920-12795 to provide

additional full core discharge capacity and a total pool capacity of 3,353

assemblies. The control rod blade (CRB) storage rack shown on Drawing 5920-6893

will be unloaded and removed if a twelfth rack needs to be installed or when the

cask pad must be used, such as for an irradiated hardware disposal campaign.

Partial plans depicting the Boral loading are provided in Figures 3.3.1-4 and 3.3.1-

5. The spent fuel storage racks are designated Safety Class 2.

Each rack consists of a welded assembly of individual storage cells in a staggered

checkerboard array. The storage cells are comprised of Type 304L stainless steel

boxes (5.922 inches square ID) welded to each other with corner angles to maintain a

pitch of 6.218 inches. The rack dimensions are 178.50 inches tall, 87.43 inches to

125.27 inches long, and 74.99 inches to 112.83 inches wide. Each storage cell has

an interior height of 168 inches. The construction of the storage cells provides

four vented (open to the pool) compartments in which B 4 C neutron absorber elements are placed for criticality control. The neutron absorber elements are positioned on

the side of the storage cell at an elevation corresponding to the fuel region of a

spent fuel assembly placed within the cell. The bottom of each storage cell sits

on, and is welded to, the rack base plate which provides the level seating surface

required for each fuel assembly and also contains the openings necessary for

adequate cooling flow. Drawing 5920-6893 shows a schematic drawing of a typical

rack.

All materials used in the construction of the rack are specified in accordance with

the applicable ASME or equivalent ASTM specification, and all welds are in

accordance with ASME Section II, for materials used, and ASME Section IX. Materials

selected are corrosion-resistant or treated to provide the necessary corrosion

resistance.

The maximum number of assemblies stored in the pool cannot exceed 3,353.

Each rack is freestanding with no lateral restraints to the wall, and is supported

by a minimum of four steel feet that transfer load to the pool floor. Any lateral

loads on the racks will be transferred by friction between the feet and the pool

floor. The racks are designed such that a fuel assembly or grappling device cannot

become fouled during removal and, thereby, generate significant uplift loads.

No spaces exist between normal fuel storage positions so that it is not possible to

insert a fuel assembly, either deliberately or by accidental drop, in any position

not intended as a fuel storage position. VYNPS DSAR Revision 0 3.0-43 of 98 Each spent fuel storage rack loaded with fuel has been analyzed to determine its

continued operability during and after both design basis and safe shutdown

earthquakes. It has been determined that under the most severe seismic loading

condition, the rack will slide a maximum of 0.56 inches. A clear distance of 2

inches (minimum) is maintained between spent fuel storage racks, spent fuel storage

racks and walls, and spent fuel storage racks and any other objects in the pool. A

clear distance of 5 inches (minimum) is maintained between the Control Rod Blade (CRB) storage racks and any other large or fixed objects in the pool. A clear

distance of 6.24 inches and 10.0 inches (minimum) is maintained between the CRB

storage racks and the NES and Holtec spent fuel storage racks respectively.

The fuel storage pool is designed so that no single failure of structures or

equipment will cause inability to (1) maintain irradiated fuel submerged in water, (2) re-establish normal fuel pool water level, or (3) safely remove fuel from the

plant. In order to limit the possibility of pool leakage around pool penetrations, the pool is lined with stainless steel. In addition to providing a high degree of

integrity, the lining is designed to withstand abuse that might occur when the

transport cask is moved about. No inlets, outlets, or drains are provided that

might permit the pool to be drained below approximately 10 feet above the top of the

active fuel. Lines extending below this level are equipped with valving.

Interconnected drainage paths are provided behind the liner welds. These paths are

designed to (1) prevent pressure buildup behind the liner plate, (2) prevent the

uncontrolled loss of contaminated pool water to other relatively cleaner locations

within the secondary containment, and (3) provide expedient liner leak detection and

measurement. These drainage paths are formed by welding channels behind the liner

weld joints and are designed to permit determination of liner weld leakage.

The spent fuel pool is 26 feet-0 inches wide by 40 feet-0 inches long by 39 feet-3/4

inches deep. The pool is completely lined with seam-welded ASTM-A240, Type 304

stainless steel. The floor plate is 1/4-inch thick and the wall plate is 3/16-inch

thick. Pipe sleeves are welded to the liner plate by full circumferential fillet

welds on both sides of the plate.

All welds above the waterline were visually examined. Those welds which could be

exposed to water were examined by liquid penetrant tests. In addition, all joint

welds and welds at penetrations in plates were tested for leaks using a vacuum box

and soap solution tests.

VYNPS DSAR Revision 0 3.0-44 of 98 3.3.1.2.4 Safety Evaluation Administrative controls ensure a sufficient level of water is maintained to ensure

shielding and/or cooling.

The design of the spent fuel storage racks provides for a subcritical multiplication

factor (k eff) for both normal and abnormal storage conditions.

For all conditions, k eff is equal to or less than 0.95. Normal conditions exist when the fuel storage racks are located at the bottom of the pool covered with a normal

depth of water (about 23 feet above the stored fuel) for radiation shielding and

with the maximum number of fuel assemblies in their design storage position. The

spent fuel is covered with water at all times by a minimum depth required to provide

sufficient shielding. Abnormal conditions may result from an earthquake, accidental

dropping of equipment, or damage caused by the horizontal movement of fuel handling

equipment without first disengaging the fuel from the hoisting equipment.

Accidental dropping of large pieces of equipment, such as a spent fuel shipping

cask, is prevented by the use of an overhead bridge crane with redundant load

bearing equipment on the main hoist.

Criticality calculations were done using a two-dimensional, two-group diffusion

theory code with a water temperature of 39°F. Water temperatures were varied

between 39°F and 248°F to assure that 39°F was the more reactive under normal

conditions. Monte Carlo calculations and verifications assured the adequacy of the

diffusion theory representation.

In order to ensure that the design criteria stated above are met, the following

loading conditions have been analyzed. The results include allowance for

calculational uncertainty.

The off-normal conditions evaluated are:

1. Normal positioning in the NES and Holtec spent k eff = 0.9469 fuel storage array

2. Eccentric positioning in the NES and Holtec spent less than 0.9469 fuel storage array

3. An assembly was placed tightly in the corner formed less than 0.9267 by an L-shaped junction of three racks (NES racks only)

VYNPS DSAR Revision 0 3.0-45 of 98 Stress in a fully loaded rack will not exceed applicable stress limits for Seismic Category I structures per requirements of the NRC Standard Review Plan, Section 3.8.4. Horizontal acceleration time history data derived from a Bechtel

calculation and maximum vertical seismic acceleration were applied simultaneously.

Maximum vertical acceleration was taken from the applicable vertical spectra at the

fundamental vertical frequency of the rack. The stresses, due to partially loaded

and empty rack conditions, are smaller than the full loaded condition.

The storage rack structure is designed to absorb the vertical impact force imposed

by a fuel assembly dropped from a height of 36 inches above a rack onto any location

on the rack. Under this impact force, those members will remain intact whose

function it is to physically maintain the normal design subcritical spacing to

assure k eff is less than 0.95.

GE topical report "Tornado Protection for the Spent Storage Pool," APED-5696, November, 1968 investigated the potential effects of a tornado striking the fuel

storage pool of a boiling water reactor (BWR). Two key concerns were examined; (1)

whether sufficient water could be removed from the pool to prevent cooling of the

fuel, and (2) whether missiles could potentially enter the pool and damage the

stored fuel.

The fuel pool was designed with substantial capability for withstanding the effects

of a tornado. The design of the fuel pool makes the removal of five feet of water

due to tornado action highly improbable. With 25 feet of water covering the fuel

racks, the removal of five feet of water is of no concern. Protection against a

wide spectrum of tornado-generated missiles is provided by the water which covers

the fuel racks.

Protection is provided against all tornado-generated missiles having a probability

of hitting the pool greater than one per 1.4 billion reactor lifetimes. Typical

potential missiles in this category include a spectrum ranging up to a 3-inch

diameter steel cylinder 7 feet long or a 14-inch diameter wooden pole 12 feet long.

The General Electric Company concluded that adequate protection for the fuel pool

against the effects of a tornado was provided and no additional protection was

required.

VYNPS DSAR Revision 0 3.0-46 of 98 NUREG-1738, "Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants" (Reference 1) contains the results of an NRC staff evaluation

of the potential accident risk in spent fuel pools at decommissioning plants in the

United States. The study was undertaken to support development of a risk-informed

technical basis for reviewing exemption requests and a regulatory framework for

integrated rulemaking. The NRC staff performed analyses and sensitivity studies on

evacuation timing to assess the risk significance of relaxed offsite emergency

preparedness requirements during decommissioning. The staff based its sensitivity

assessment on the guidance in Regulatory Guide 1.174, "An Approach for Using

Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes

to the Licensing Basis" (Reference 2). The staff's analyses and conclusions apply to

decommissioning facilities with SFPs that meet the design and operational

characteristics assumed in the risk analysis.

The study found that the risk at decommissioning plants is low and well within the

Commission's Safety Goals. The risk is low because of the very low likelihood of a

zirconium fire (resulting from a postulated irrecoverable loss of SFP cooling water

inventory) even though the consequences from a zirconium fire could be serious.

NUREG-1738, Executive Summary, states in part, "the staff's analyses and conclusions

apply to decommissioning facilities with SFPs that meet the design and operational

characteristics assumed in the risk analysis. These characteristics are identified

in the study as IDCs and SDAs. Provisions for confirmation of these characteristics

would need to be an integral part of rulemaking."

Design and operation of the VY SFP has been evaluated against and confirmed to

comply with the industry decommissioning commitments (IDCs) and staff

decommissioning assumptions (SDAs) contained in NUREG-1738. The evaluation is

documented in BVY 14-009, Request for Exemptions from Portions of 10CFR50.47 and

10CFR50, Appendix E. (Reference 3)

3.3.1.2.5 Inspection and Testing

The spent fuel storage racks were tested at the plant site or visually inspected

during rack fabrication to ensure that the Boral sheets are in place and free of

voids. Since Boral absorbs neutrons, a neutron source and proportional counters

were used to verify the integrity of the Boral sheet.

An inspection mandrel was used to test each storage cell location. The insertion

and withdrawal of the mandrel was monitored over the entire length of the cell to

ensure that acceptable drag forces were not exceeded.

VYNPS DSAR Revision 0 3.0-47 of 98 3.3.1.3 Standby Fuel Pool Cooling and Demineralizer Systems 3.3.1.3.1 Objective

The Standby Fuel Pool Cooling (SFPCS) removes decay heat released from the spent

fuel to maintain fuel pool temperature within specified limits. The Fuel Pool

Demineralizer System (FPDS) maintains water clarity.

3.3.1.3.2 Design Bases

1. The FPDS shall minimize corrosion product buildup within the spent fuel pool and shall maintain proper water clarity so that the fuel assemblies can be

efficiently handled underwater.

2. The FPDS shall minimize fission product concentration in the spent fuel pool water, thereby minimizing the radioactivity which could be released from the

pool to the Reactor Building environment.

3. The Fuel pool water level shall be maintained at a level above the fuel sufficient to provide shielding for normal building occupancy.

4. The Standby Fuel Pool Cooling System shall be capable of maintaining the spent fuel pool temperature below 150°F.

3.3.1.3.3 Description

The SFPCS is shown on Drawing G-191173, Sheets 1 and 2.

Fuel Pool Structure

The fuel pool concrete structure, metal liner, spent fuel storage racks, and the

SFPCS are designed to withstand Seismic Class I earthquake loads.

FPDS Fuel Pool clarity is maintained by the FPDS. The FPDS consists of submerged

underwater units which will be operated as required to minimize fission product

concentration and maintain water clarity through demineralization and filtration. VYNPS DSAR Revision 0 3.0-48 of 98 Fuel Pool Makeup and Letdown

Makeup to the pool is supplied by the Condensate Transfer System.

Water may be removed from the fuel pool, if required, via the normal fuel pool

cooling system, through the fuel pool filter-demineralizer units, to the condensate

storage tank.

Fuel Pool Skimmers

Two skimmer pumps are provided which take suction from the top of the pool to remove

surface debris. These pumps circulate fuel pool water through cartridge filters and

return it to the pool through service boxes located around the pool.

SFPCS The SFPCS functions to maintain pool temperature within specified limits.

An administrative limit of 125°F has been established for maximum fuel pool

temperature during normal cooling and filtering.

The operating temperature of the fuel pool is permitted to rise up to 25°F above the

administrative temperature limit (125°F) as specified in applicable procedures.

The SFPCS is a two train, Seismic Class I, non-safety related system, designed to be

remotely placed in operation from the control room. The SFPCS circulates the pool

water in a closed loop, taking suction from the spent fuel storage pool, through

heat exchangers and discharging the water back into the fuel pool. The SFPCS heat

exchangers transfer the spent fuel decay heat to the seismic Class I, non-

safety-related Station Service Water System (SWS).

The SFPCS includes two seismic Class I centrifugal pumps. All the parts of the pump

in contact with water are corrosion-resistant. A pump low discharge pressure alarm

annunciates in the Control Room. In addition, the pumps trip automatically on low

suction pressure.

The heat exchangers are shell and tube design; all parts in contact with water are

corrosion resistant. These heat exchangers are each sized to maintain fuel pool

water temperature below 150°F.

VYNPS DSAR Revision 0 3.0-49 of 98 To minimize the potential for fuel pool water leakage into the Station SWS, service water pressure is normally maintained greater than SFPCS pressure. The fuel pool water

side of the heat exchangers has a maximum operating pressure equivalent to the static

pressure head from the pool surface to the heat exchanger. The Station SWS side of the

heat exchangers has a minimum operating pressure which is normally greater than the

maximum pressure on the fuel pool side of the heat exchangers as long as the operating

SW pumps can maintain system header pressure above the pressure which results in NNS SW

header isolation valve closure. During events which result in low service water header

pressure, service water pressure may be lower than SFPCS pressure until the SW header

isolation valves are closed. By maintaining a positive differential pressure, leakage

of fuel pool water to the environment is prevented. The differential pressure across

each heat exchanger is monitored by a differential pressure sensor and displayed in the

control room.

Two motor operated throttling valves, V70-257A and 257B, provide service water flow

control through the respective SFPCS heat exchanger to control both pool temperature

and service water to SFPCS differential pressure.

Two motor operated isolation valves, V19-220 and 221, close on low pool level, providing automatic pool isolation in case of a line break in the non-seismic

portion of the system.

SFPCS heat exchanger supply and return service water piping and SFPCS piping is

corrosion-resistant. The piping meets the requirements of ANSI B31.1-77.

Indication is provided in the control room and/or locally near the equipment.

Control Room indication for each train includes direct pool temperature, fuel pool

water temperature out of the heat exchangers (taken downstream of the pumps), pump

run lights, pump discharge pressures, service water flow, SWS to SFPCS heat

exchanger DP and valve position lights. Local indication includes fuel pool water

temperature into the heat exchangers, pump suction and discharge pressures, and heat

exchanger DP. Pool temperature is provided by redundant thermocouples located

within the pool. Pool level is provided by redundant transmitters located near the

pool. All other transmitters and sensors are located in or near the Fuel Pool

Cooling System cubicle.

Controls for the pumps and four MOVs are provided in the control room. Control room

controls include pump on/off switches, service water throttle valves control

switches, and V19-220 and V19-221 isolation valves control switches.

VYNPS DSAR Revision 0 3.0-50 of 98 3.3.1.3.4 Evaluation The SFPCS has a heat removal capability of 11 MBtu/hr with one pump and one heat

exchanger in service, and 22 MBtu/hr with both pumps and heat exchangers in service (assuming 2% plugging).

Both trains of the SFPCS in operation have sufficient capacity to maintain fuel pool

temperature within specified limits with the maximum number of fuel assemblies in

the pool after a full core offload. Under these conditions, after a period of

approximately 40 days of fuel decay time following reactor shutdown, one train of

the SFPCS has sufficient capacity to maintain fuel pool temperature within specified

limits.

3.3.1.3.5 Inspection and Testing

The SFPCS is normally in operation during all modes of facility operation.

Satisfactory operation is demonstrated continuously without the need for special

testing or inspection.

3.3.1.4 Tools and Servicing Equipment

3.3.1.4.1 Objective

To provide and use tools and servicing equipment in a way that ensures the bounds of

the design basis fuel handling accident are not exceeded.

3.3.1.4.2 Design Bases

1. The refueling platform shall withstand a seismic event without gross failure or overturning.

2. Fuel handling equipment shall be classified in accordance with its potential for damaging irradiated fuel.

3. Equipment weighing more than 700 pounds shall be classified as a heavy load and handled in accordance with appropriate facility procedures.

3.3.1.4.3 Description

3.3.1.4.3.1 Introduction

All tools and servicing equipment necessary are supplied for efficiency and safe

serviceability. The following is a listing of tools and servicing equipment. VYNPS DSAR Revision 0 3.0-51 of 98 Tools and Servicing Equipment Quantity Fuel Servicing Equipment Fuel Preparation Machines 2 Channel Bolt Wrenches 2 Channel Handling Tool 1 Fuel Inspection Fixture 1 Channel Gauging Fixture 1 General Purpose Grapples 3 Channel Transfer Grapple 1

Fuel Pool Gates 2 Channel Handling Boom

1 Servicing Aids Actuating Poles 3 General Area Underwater Lights 4 Local Area Underwater Lights 4

Drop Lights 4 Underwater TV Monitoring System 1 Underwater Vacuum Cleaner 1

Viewing Aids 4 Light Support Brackets 4 Jib Cranes

2 Refueling Equipment Refueling Equipment Servicing Tools 1 Refueling Platform, main fuel grapple and contents 1

Storage Equipment Control Curtain Transfer Basket 1 Spent Fuel Storage Racks 11 Channel Storage Rack 1 Control Rod Blade Storage Rack (30 Cavities per Rack) 1 Defective Fuel Storage Containers 8

3.3.1.4.3.2 Fuel Servicing Equipment

Two fuel preparation machines are used to remove the channels from and install

channels on fuel assemblies. These machines are designed to be removed from the

pool for servicing. A channel transfer grapple is provided for inserting or

withdrawing channels from storage racks.

VYNPS DSAR Revision 0 3.0-52 of 98 An equipment support railing is provided around the pool periphery in order to tie off miscellaneous equipment such as the fuel leak detector (sipper) and service

tools. Equipment lugs fabricated as part of the pool liner are required for

fixtures that might later be desired by facility personnel. In addition, a

4 x 4-inch curb with a 4-inch wide plate of 1-inch thick stainless steel on top is

provided around the entire periphery of the refueling volume. The plate provides a

suitable welding and drilling surface for mounting additional equipment. The curb

may be used as an additional support or tie-off area. Cable ways are recessed into

the floor around the pool periphery with openings to pass cables into the pool from

underneath this curbing. A number of different grapples are available at Vermont Yankee for use during maintenance activities. Grapples can be attached to the Reactor Building auxiliary

hoist, or the auxiliary hoists on the refueling platform. Grapples can be used to

shuffle fuel in the pool and to handle fuel during channeling. A channel-handling boom with an electric hoist is used to assist the operator in supporting the weight of a channel after the channel is removed from the fuel

assembly. The boom is set between the two fuel preparation machines. With the

channel-handling tool attached to the hoist, the channel may be conveniently moved

between the fuel preparation machines. 3.3.1.4.3.3 Servicing Aids General area underwater lights are provided with a suitable reflector for general

downward illumination. A portable underwater vacuum cleaner is provided to assist in removing crud and miscellaneous objects from the pool floor. The pump and the filter unit are

completely submersible for extended periods.

3.3.1.4.3.4 Fuel Handling Equipment

The refueling platform is used as the principal means of transporting fuel

assemblies in the storage pool. The platform travels on tracks extending along each

side of fuel pool. The platform supports the main hoist and fuel grapple and two

auxiliary hoists. The grapple is suspended from a trolley system that can traverse

the width of the platform. Platform operations are controlled from either the

operator station on the trolley or auxiliary stations on the auxiliary hoist control

boxes. Refueling grapple operation and platform movement are controlled through a

Programmable Logic Controller (PLC). The PLC also limits platform velocity and

movement when the grapple is in close proximity to the perimeter of the storage

pool. A Personal Computer (PC), which is also part of the system, implements the

optional automatic mode of operation to allow a preprogrammed series of platform

movements corresponding to planned fuel assembly moves. The platform contains a

position-indicating system that indicates the position of the fuel grapple. VYNPS DSAR Revision 0 3.0-53 of 98 Mounted on both the reactor well side of the refueling platform and on the platform trolley monorail are one-half-ton auxiliary hoists. These hoists normally can be

used with appropriate grapples to handle control rods, detectors, sources, and other

equipment. The auxiliary hoist can also serve as a means of shifting fuel elements

and other equipment within the pool.

All motions of the platform required to handle fuel assemblies may be controlled

from a single location. 3.3.1.4.3.5 Storage Equipment A channel storage rack is located between the fuel preparation machines to permit a logical work flow during channeling and de-channeling operations.

Racks are arranged so that fuel assemblies and control rod blades can be

conveniently positioned for storage. The racks can be removed without draining the

pool to allow inspection or replacement, should it become necessary. Capacity is

provided for a maximum of 38 control rod blades. One CRB storage rack provides a

capacity of 30 and one spent fuel storage rack provides an optional capacity of

eight. 3.3.1.4.4 Evaluation The refueling platform can withstand a seismic event without gross failure or overturning.

The safety classification of fuel handling equipment and tools is determined based

on their potential for damaging irradiated fuel and, as a result, exceeding

appropriate radiological dose criteria. Equipment weighing more than 700 pounds is

classified as a heavy load and is handled in accordance with Reference 4.

The design basis fuel handling accident is discussed in the Station Safety Analysis

Section of the DSAR.

3.3.1.5 References

1. NUREG-1738, "Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants"

2. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis"

3. BVY 14-009, Request for Exemptions from Portions of 10CFR50.47 and 10CFR50, Appendix E.

4. PP 7023, "Control of Heavy Loads Program Document" VYNPS DSAR Revision 0 3.0-54 of 98 Vermont Yankee Defueled Safety Analysis Report Revision 0 Fuel Storage-Arrangement Figure 3.3.1-1 VYNPS DSAR Revision 0 3.0-55 of 98 Vermont Yankee Defueled Safety Analysis Report Revision 0 Fuel Storage Rack Assembly (Partial)

Figure 3.3.1-4 VYNPS DSAR Revision 0 3.0-56 of 98 Vermont Yankee Defueled Safety Analysis Report Revision 0 HOLTEC Fuel Storage Rack Assembly (Partial) Figure 3.3.1-5 VYNPS DSAR Revision 0 3.0-57 of 98 3.3.2 Service Water System 3.3.2.1 Objective

The objective of the Service Water System (SWS) is to provide water from the

Connecticut River for spent fuel pool cooling and other miscellaneous services.

3.3.2.2 Design Bases

The design bases of the Station SWS are:

1. To provide water for spent fuel pool cooling.

2. To minimize the probability of a release of radioactive contaminants to the environs by monitoring the system discharge and maintaining sufficient

pressures at specific areas in the system.

3. To supply a source of cooling water for the station standby diesel generators.

3.3.2.3 Description

The flow diagram for the SWS is shown on Drawing G-191159, Shs. 1 and 2.

Two pumps located in the intake structure are provided to supply the SWS requirements. The pumps are normally started and stopped by controls on the main

control board. With a design river water temperature of 85°F, one of the two pumps

is required to supply normal station cooling demands. Operating pumps will continue to run until stopped from the Main Control Room. Each service water pump is connected to a 4160 V bus that can be supplied by a

standby diesel generator.

The SWS is a dual header system using two parallel 24-inch supply headers. Two

automatic self-cleaning strainers, for removal of suspended matter from the river

water, serve the headers. The headers include cross-connect lines 20"SW-3, 3"SW-28

and 3"SW-28A, and 24"SW-8. Normally, the valves in the interconnecting lines are

open, permitting either pump to supply the cooling water through both strainers and

headers. In the event of a major malfunction in either header, it is possible to

isolate the portion of the system affected (using electrically-operated valves V70-

19A and 19B in line SW-8 and manual valves in the other cross-connect lines) and

maintain all essential cooling water services.

VYNPS DSAR Revision 0 3.0-58 of 98 Each service water header supplies cooling water to a diesel-generator cooler. A pressurizing line to the Fire Water System and water for the Chlorination System is

supplied from lines tied into both headers. There is also a line, 12"SW-4A, tying

into the Fire Water System that can be supplied from either header. This line

contains a manual valve that is maintained closed and can only be opened under

specific procedural direction. The SFPCS heat exchangers and the water for the

backwash function of the traveling screens are supplied from the "B" service water

header.

The discharge piping from the diesel generator coolers is piped such that flow from

the units can be discharged via individual lines, through one of two common

discharge lines, or through both common discharge lines simultaneously.

The SWS discharges to the cooling tower deep basin.

A process radiation monitor is located in the station service water discharge

header. To prevent the release of radioactivity from the SFPC System to the SWS, the system is designed such that the fuel pool side of the heat exchangers has a

maximum operating pressure equal to the static head developed by the difference in

elevation between the heat exchanger and the fuel pool surface. The minimum

operating service water pressure is normally greater than the pressure in the fuel

pool side of the heat exchanger. This positive differential pressure in the heat

exchangers will protect against any possible fuel pool water leakage into the

Station SWS.

SWS piping meets code standards, including ANSI B31.1, as described in Section A.9

of Appendix A. Those portions of the Station SWS supplying spent fuel cooling are

of Class I seismic design. Other portions of the system, whose failure due to a

seismic event could cause unacceptable flooding, negatively impact flow to

essential equipment, or impact the ability to establish secondary containment

closure, are either isolable via automatic or manual valves or have been evaluated

and determined to meet Class I seismic requirements.

Vacuum breakers are installed in the following pipe lines to prevent water hammer

when service water flow is restored following a loss: supply line (8" SW-800) to

the standby fuel pool cooling heat exchanger, the supply line to the refuel floor

cooler lines, and the supply line (8" SW-33) to the Control Room chiller.

Operation of these valves will ensure adequate flow to essential components

following the events identified above and will prevent flooding in the Reactor

Building due to a water hammer event.

That portion of the intake structure, which houses the Station SWS equipment, is

Class I seismic design.

VYNPS DSAR Revision 0 3.0-59 of 98 3.3.2.4 Evaluation

Service water piping meets code standards, including ANSI B31.1. Those portions

supplying fuel pool cooling equipment are of Class I seismic design. In addition, the piping whose failure could (a) cause unacceptable flooding, or (b) negatively

impact flow to essential equipment, or (c) impact the ability to establish

secondary containment closure, is isolable via automatic or manual valves, has been

designed such that loss of water through failed piping is within the capability of

the system, or has been evaluated and determined to meet Class I seismic

requirements. Vacuum breakers have been installed on those pipe lines susceptible

to a water hammer event to ensure adequate flow to essential equipment and prevent

flooding in the Reactor Building. The maximum operating pressure of the service

water within these lines is typically on the order of 110 psig. This piping is

routed in the building such that it is not in the vicinity of any heavy equipment

movement during maintenance nor in the vicinity of vehicle traffic, and therefore, is not vulnerable to damage from collisions.

Based upon the above discussion, it is highly improbable that the piping could fail

in such a manner as to cause flooding or interrupt SWS flow for spent fuel pool or

diesel generator cooling.

Maintaining a positive differential pressure between the service water side and the

fuel pool side of the SFPC heat exchanger protects against any possible fuel pool

water leakage into the SWS.

3.3.2.5 Inspection and Testing

The Station SWS is normally in operation during all modes of station operation.

Satisfactory operation is demonstrated continuously without the need for special

testing or inspection.

3.3.3 Electrical Power Systems

3.3.3.1 Transmission System

3.3.3.1.1 Objective

The objective of the Transmission System is to provide reliable power from off-site

to the facility Auxiliary Power System to facilitate the safe storage and handling

of irradiated fuel.

VYNPS DSAR Revision 0 3.0-60 of 98 3.3.3.1.2 Design Basis The Transmission System provides a reliable source of power from off-site to the

facility to facilitate the safe storage and handling of irradiated fuel.

3.3.3.1.3 Description

There are two 345 kV switchyards and two 115 kV switchyards on site at VY. The

original 345 kV and 115 kV switchyards are now called the VY switchyards. New

Vernon 345 kV and Vernon 115 kV switchyards were installed by Vermont Electric

Power Company (VELCO) as part of their Southern Loop Project. These two

switchyards are on VY property north of the existing VY switchyards.

The VY 345 kV switchyard consists of four circuit breakers in a ring bus

configuration as shown on Drawing G-191298, Sh.3.

Electric power is supplied from off-site via the transmission network to the

on-site electric distribution system through either of two 345 kV/115 kV

autotransformers to the VY 115 kV switchyard. The VY 115 kV switchyard powers the

station startup transformers.

The VY 345 kV switchyard north bus powers a 400 MVA autotransformer which supplies

power to the VY 115 kV switchyard.

The Vernon 115 kV switchyard supplies a second source of normal power to the VY 115

kV switchyard via a K-40 tie line to a second autotransformer.

The auto-transformers are operated in parallel; the loss of either source will not

cause the VY 115 kV switchyard to lose power.

An alternate circuit through the 115 kV K-186 transmission line may be made

available.

A 13.2 kV underground power line runs from the adjacent Vernon Hydroelectric

Station (VHS) to a 13.2-4.16 kV transformer near the cooling towers. From there, a

4160 V underground power line connects to the Station Blackout (SBO) Diesel

Generator (DG) switchgear and then goes on to the station switchgear. The SBO DG

or the VHS can be connected to selected 4160 V buses through manually operated

circuit breakers.

VYNPS DSAR Revision 0 3.0-61 of 98 3.3.3.1.4 Evaluation of System Protection Transmission system protection design meets the objectives of the Northeast Power

Coordinating Council, "Bulk Power System Protection Criteria."

The two tie lines between the VY switchyards and the Vernon switchyards each have

two redundant and diverse channels of the line differential, directional over-

current line and round impedance protection.

Four 345 kV and one 115 kV transmission lines are terminated in the Vernon

Switchyards. Protection in the Vernon Switchyards for the transmission lines

consists of Primary and Secondary (or backup) protection.

The 345 kV and 115 kV switchyards each have a primary and secondary bus

differential relay system. These systems are independent of one another and the

tripping of one will not cause tripping of breakers in the other substation, with

the exception that the 115 kV secondary side breaker on the 400 MVA autotransformer

will be tripped for a fault on the 345 kV switchyard's north bus.

In the unlikely event that the two sources to the 115 kV primary of the station

startup transformers, that is, the 345/115 kV autotransformer supply from the VY

345 kV switchyard, or the tie line to the Vernon 115 kV switchyard become

disconnected, two emergency diesels described in Section 3.3.3.3, the Station

Blackout Diesel Generator Alternate ac source, and the line to Vernon Hydroelectric

Station, would also be available.

The 115KV switchyard contains three capacitor banks, one 30MVAr bank and two 15MVAr

banks. Each bank has its own breaker connecting it to the 115KV bus. These

breakers are individually controlled by the system operator via SCADA. Phase and

ground overcurrent, unbalance, over-voltage and breaker failure protection is

provided for each bank breaker.

3.3.3.2 Auxiliary Power System

3.3.3.2.1 Objective

The objective of the Auxiliary Power System is to provide a reliable power supply

to all station loads required for the safe storage and handling of irradiated fuel.

3.3.3.2.2 Design Basis

The Station Auxiliary Power System shall have the capacity and capability to supply

the required facility loads. Protective, control, and instrumentation devices

shall be provided to insure reliability and availability of the system. VYNPS DSAR Revision 0 3.0-62 of 98 3.3.3.2.3 Description

The Station Auxiliary Power System is shown on Drawings G-191299, G-191300, Sh. 1

and 2, G-191301, Sheets 1 and 2. The system consists of six 4160 V buses, which

supply power to all 4000 V motors, and to the 4160-480 V station service

transformers, which supply power to the 480 V buses.

3.3.3.2.3.1 4160 V Switchgear

The normal supply for the 4160 V load is the startup transformers (T-3A and T-3B)

which are supplied from the 345/115 kV Transmission System.

The startup transformers have adequate capacity for all loads required for the safe

storage and handling of irradiated fuel.

The switchgear for the 4160 V Auxiliary System is of the metal-clad indoor type, except 4160 V Buses 5A and 5B which are outdoor metal-clad units. Circuit breakers

are three pole, air break type, electrically-operated with control power supplied

from batteries.

3.3.3.2.3.2 480 V Buses

480 V auxiliary power is supplied from the 4160 V Auxiliary System through 4160-480

V station service transformers. The 480 V system consists of switchgear buses and

motor control centers.

The 480 V switchgear buses are self-supporting, metal-clad structures with draw-out

circuit breakers.

3.3.3.2.3.3 120/240 V Instrumentation Distribution System

A 120/240 V Single Phase Instrumentation Distribution System supplies selected

instrumentation and other loads. The system consists of the 120/240 V

uninterruptible ac bus and its subpanel and the 120/240 V instrumentation

distribution panel and its subpanels.

The bus arrangement is shown on Drawing G-191372, Sheets 4 and 5.

VYNPS DSAR Revision 0 3.0-63 of 98 3.3.3.2.4 Cable Installation and Separation Criteria

1. Intermixing of Cables

Low-level instrumentation cables are routed in separate trays from control

cables.

The definition of "low level instrumentation cable" is:

A cable used for data, control, or instrumentation service. In

general, this service includes cable from thermocouples, resistance

temperature detectors, process instruments, and computer signals. As a

general rule, anything less than 50 V is considered low level.

The definition of "control cable" is:

A cable used for control, metering, relaying, and alarm circuits. In general, these services include 125 V dc and 120 V ac control leads, annunciator cables, PT cables, CT cables, and solenoid cables. The following exception to the criteria for intermixing cables has been

justified as acceptable:

1. Instrumentation cable which only provides a Control Room indication function may be run in the same tray as control cable.

2. Tray Loading and Cable Sizing

The general rules of 50% derating outside the drywell area was used in calculating

power cable sizes. The following design conditions were considered in arriving at

the 50% derating criteria for cables in tray: (1) load factor, (2) tray loading, (3) short circuit capacity of cable, (4) ambient temperature, (5) grouping factor, and (6) voltage drop.

For cable tray loading, the design is based on the IPCEA Code Bulletin

No. P-46-426, 1962.

For cables that are not routed in tray, derating is based on ampacities taken from

the appropriate tables in the IPCEA Code Bulletin No. P-46-426, 1962, which are a

function of the number of conductors in the conduit or duct bank.

VYNPS DSAR Revision 0 3.0-64 of 98

3. Fire Protection Criteria

The criteria used for fire protection of cable installations are as follows:

Fire stops are provided in vertical tray runs at Reactor Building fire zone

boundary designations.

In areas other than the cable vault, tray covers are utilized on cable trays to

prevent fire from migrating from tray to tray in a vertical bank.

In the area where a high concentration of cables exists, such as the cable vault, an automatic Fire Protection System is provided; and flame resistant cable

constructions are used to minimize the propagation of fire along horizontal runs of

cable trays. Cable tray covers are not used in the cable vault except where

required to achieve physical separation. This allows fire extinguishment by the

Fire Protection System or by manual means.

3.3.3.2.5 Inspection and Testing

Periodic equipment tests are performed at scheduled intervals to detect any

deterioration of the system towards an unacceptable condition. The specific tests

and the frequency at which they are performed depend upon the specific components

installed, their function, and their environment.

3.3.3.3 Diesel Generator Systems

3.3.3.3.1 Objective

To assure the availability of on-site back-up power, two diesel-driven generators

are provided.

3.3.3.3.2 Design Basis

Two standby diesel-driven generators shall be provided. Each of the diesel

generator units shall be capable of supplying 100% of the loads required for the

safe storage and handling of irradiated fuel. Each unit shall be physically and

electrically independent of the other and of any off-site power source. No single

failure shall cause accidental paralleling of the two diesel generators.

VYNPS DSAR Revision 0 3.0-65 of 98 3.3.3.3.3 Description Each of the two standby diesel driven generator sets has a continuous rating of

2750 kW, an overload rating of 3000 kW for seven days, and a short-time (two-hour)

rating of 3025 kW in any 24-hour period. The generator rating is 3750 kVA at 0.8

power factor (continuous). Each of the two is capable of starting and carrying the

loads required for the safe storage and handling of irradiated fuel. The

generators can be manually loaded, within their rated capacity, at the discretion

of the operator.

Each generator is direct-driven by a twelve cylinder, turbo-charged diesel engine

which is rated 4182 bhp at 900 rpm.

The diesel-engines are started by direct air injection into the cylinders from the

diesel's own compressed air system. A diesel generator will automatically start on

experiencing a loss of voltage on its 4160 V bus and is automatically connected to

the bus. Each unit has its own independent starting circuitry.

The Air Starting Recharging Subsystem for each unit consists of a 460 V ac

motor-driven compressor which is capable of recharging either of the two empty dual

air receivers in 60 minutes. The Air Starting System is normally operated with an

air pressure range which ensures a minimum of three independent cold diesel engine

starts without recharging while ensuring that the receiver safety valves do not

lift. The diesel engines are designed to start with starting air pressure of 150

to 250 psig. Two receivers are provided for each diesel.

The engines are started through two independent air start solenoids branching off

the main air header. One of the two solenoid valves is provided with manual override. The engines are capable of starting at 0 °C and can run without any cooling for one minute at rated load. The jacket coolant and lube oil of both engines are kept at keep-warm temperatures by electric heaters and recirculated

during standby operation. Each diesel engine is cooled by a separate service water

loop. Since a service water pump is among the initial loads applied to each

generator, service water cooling is available well within the one minute allowed

for uncooled operation.

Each driver has an independent fuel tank. These independent fuel tanks can be

filled from the fuel oil storage tank. One ac motor-driven fuel oil transfer pump

is provided for each diesel generator unit to assure that an operating generator

will have a continuous supply of fuel.

Each diesel generator can be loaded sequentially with the necessary diesel cooling

and auxiliary loads. Each unit has its own independent sequencing circuitry.

VYNPS DSAR Revision 0 3.0-66 of 98 The self-cooled generator is provided with a 13 kW, 250 V dc static exciter regulator with a 0.05 to 0.1 second response characteristic. The equipment will

restore voltage effectively even when the heaviest load in the loading cycle is

added to the engine (already loaded to 75%) without jeopardizing the running loads.

The standby diesel-driven generators will be available to feed loads on Buses 3 and

4 of the 4160 V system and the associated lower voltage systems by manual control.

The loading of each generator is monitored and connection of loads will be made at

operator discretion with consideration for load limitations.

The two diesel generators cannot be paralleled at any time, as breaker interlocks

are provided to prevent closure of breakers which could tie the generator supplied

buses together when both diesel generator breakers are closed.

Fuel Oil System

Each diesel engine's fuel oil supply consists of an 800-gallon day tank piped

directly (through one locked-open valve) to the diesel fuel block. A fuel pump, mechanically powered from the diesel engine, supplies the injector valves. Excess

flow is recirculated back to the day tank. The day tank capacity is sized for

three hours of continuous full load operation.

Makeup to each diesel day tank is accomplished automatically from a 75,000 gallon

storage tank located in the yard adjacent to the Turbine Building. The storage

tank is maintained with sufficient fuel to provide seven days fuel oil supply to a

diesel generator operating at the continuous rating of 2750 kW. The tank is

partially below grade for missile protection. The protected section of the tank

provides a minimum of 5 days fuel oil supply to a diesel generator operating at the

continuous rating of 2750 kW. Two fuel oil pumps, located below grade, transfer

oil from the storage tank to the day tanks. The transfer pump motor and automatic

control circuits for the associated day tank receive electrical power from the

diesel generator supplied from the respective day tank.

Failure of any active component (i.e., transfer pump or control circuitry) would

not result in the loss of the associated diesel. Operator action would be taken to

refill the day tank. The diesel will continue to operate for up to three hours

until the day tank capacity is replenished.

VYNPS DSAR Revision 0 3.0-67 of 98 Service Water Cooling System The diesel generators are cooled via separate service water headers and each is

supplied by service water pumps located at the intake structure. The pump motors

and the control circuits receive electrical power from the diesel generator buses

and the station batteries. The service water headers are cross-connected on the

upstream and downstream side of the in-line mechanical strainers. Active components

considered necessary to provide cooling water to a diesel generator are a service

water pump and a control valve (FCV-28A, 28B) located downstream of the diesel

coolers.

A worst case consequence as a result of a single failure to the above arrangement

is the loss of one diesel generator. Failure of one service water pump or one

strainer would not result in loss of cooling to either generator. Each strainer is

designed to pass full flow from four service water pumps, and the service water

headers are cross-connected thereby permitting one strainer to pass water for both

headers. Failure of a control valve in the closed position would result in loss of

cooling to the respective diesel generator. The most significant result from a

single failure would be the loss of one diesel generator.

3.3.3.3.4 Inspection and Testing

The testing program is designed to test the ability of the diesel generator to

start and run under load for sufficient time so as to bring all components and

auxiliary systems of the Diesel Generator System into operation.

The system design allows each generator to be manually synchronized and connected

to its bus, manually loaded, and run under load.

3.3.3.4 125 V DC System

3.3.3.4.1 Objective

The objective of the 125 V DC System is to provide a supply of 125 V dc power for

the operation of equipment.

3.3.3.4.2 Design Basis

The 125 V DC System shall consist of two dc systems, each capable of supplying its

required loads.

VYNPS DSAR Revision 0 3.0-68 of 98 3.3.3.4.3 Description Two 125 V dc systems are provided to supply the station 125 V dc loads. One system

includes Main Station Battery A-1. The second system includes Main Station Battery

B-1 and Alternate Shutdown Battery AS-2. Each battery is of the central power

station type, designed for continuous duty at an operating voltage of 125 V dc. The

loads for Alternate Shutdown Battery AS-2 can also be supplied by Main Station

Battery B-1.

Main Station Batteries A-1 and B-1 are located in a single, ventilated Battery

Room. To limit the effect of any single event upon both batteries, an 8-inch thick

concrete block wall separates the two batteries.

Alternate Shutdown Battery AS-2 is located in Diesel Generator 1A Room.

Each main station battery has a pair of constant voltage, current limiting

silicon-controlled rectifier type battery chargers, which are capable of supplying

normal continuous dc load and maintaining a float charge on the battery. Each

charger is also capable of recharging its associated battery to full charge if it

should become discharged to its minimum voltage. For Alternate Shutdown Battery

AS-2, a standby battery charger is installed to substitute for the normal charger

in the event of normal charger failure or maintenance outage. This standby charger

is of the same capacity as Alternate Shutdown Battery AS-2 normal charger and can

be manually connected to the battery bus.

The redundant distribution divisions of the 125 V dc systems are designated as the

DI and DII divisions. Division DI includes Distribution Panels DC-1, DC-1A, DC-1B, and DC-1C connected to Main Station Battery A-1. Division DII includes

(1) Distribution Panels DC-2, DC-2A, DC-2B, and DC-2C connected to Main Station

Battery B-1; and (2) Distribution Panel DC-2AS connected to Alternate Shutdown

Battery AS-2. Distribution Panel DC-2D is powered from DII power supply DC-2.

Manual transfer switches are installed to provide a backup supply of power to

selected panels. Use of the transfer switches is administratively controlled by

procedure and limited to emergency situations or planned maintenance per

administrative control. Panel DC-3 is normally fed from Panel DC-2, but can be

connected to an alternative feed from Panel DC-1, by means of a manual transfer

switch. Other dc panels with manual transfer switches are DC-2A and DC-3A which

also have alternate feeds from Panel DC-1. Panel DC-2B is normally fed from Panel

DC-2, but can be connected to an alternative feed from Panel DC-1AS, by means of

manual transfer switch MTS-13-1. The electrical arrangement of the batteries, chargers, buses and switchgear is shown on Drawing G-191372, Sheets 1, 2 and 3.

VYNPS DSAR Revision 0 3.0-69 of 98 The batteries, panels, and power feeds associated with Division DI are physically isolated from the batteries, panels, and power feeds associated with Division DII

by a minimum of 15 feet, except in the Battery Room where the two batteries are

separated by an 8-inch thick concrete block wall. Where distribution circuits are

separated by less than 15 feet, the cables are routed in rigid steel conduits, flexible steel conduits, or enclosed in steel wireways.

The DC System is ungrounded and has a ground detection alarm system.

3.3.3.4.4 Evaluation

During normal operation, the continuous dc load is supplied by the battery chargers

which are connected to the dc buses. The batteries normally float on the system, supplying any momentary high current control requirements. Upon loss of a charger, the associated battery supplies its total dc load requirements. The normal ac

sources for the battery chargers are the 480 V emergency buses. These buses are

energized through transformers by normal auxiliary ac power or by the emergency

diesel generators.

Feeders from redundant dc sources are provided to control circuits for the diesel

generators, the 4160 V bus switchgear (except for buses 5A and 5B), the 480 V

emergency bus switchgear, and certain 125 V dc distribution panels. These

alternate feeders are connected through manual transfer switches such that only one

dc source can be connected at a time.

A common ventilation system is supplied for the rooms in which Main Station Batteries A-1 and B-1 are located. The normal Battery Room temperature is

approximately 70°F. On loss of this ventilation system, the temperature in the battery rooms would not exceed 100 °F in summer and would not fall below 60 °F in winter. The accumulation of hydrogen from the batteries located in the Battery Room would not exceed 4% concentration in the Battery Room in 21/2 days with a complete loss of

the Ventilation System.

Equipment in the Battery Rooms is limited to batteries, battery racks, necessary

lighting, safety equipment, and outgoing power cable which for the most part is

installed in conduits. The nonmetallic portions of this equipment, such as the

battery cases, will burn but will not support combustion. The wall between the two

batteries limits the effect of a fire or battery cell explosion to one battery.

Alternate Shutdown Battery AS-2 hydrogen gassing rates is sufficiently low such

that either normal ventilation and/or room infiltration provides sufficient

dilution to prevent hydrogen concentration from exceeding 2%. VYNPS DSAR Revision 0 3.0-70 of 98 3.3.3.4.5 Additional DC Systems

In addition to the above dc systems, two 125 V DC Systems in each of the switchyard

control houses which provide power for breaker operation and control and protective

relaying circuitry.

3.3.3.4.6 Inspection and Testing

The batteries and other equipment associated with the 125 V DC System are easily

accessible for inspection and testing. Service and testing is performed on a

routine basis in accordance with approved station procedures/programs. Typical

periodic inspections will include visual examination for leaks and corrosion, and

check of all batteries for voltage, specific gravity of electrolyte, and

electrolyte level.

3.3.3.5 +/-24 V DC Power System 3.3.3.5.1 Objective

The objective of the +/-24 V DC Power System is to provide a supply of +/-24 V DC power

for the operation of various process radiation monitoring instrumentation. 3.3.3.5.2 Design Basis

The +/-24 V DC Power System shall supply all +/-24 V dc power requirements.

3.3.3.5.3 Description

Two +/-24 V dc systems, as shown on Drawing G-191372, Sheet 5, are provided for

operation of various process radiation monitoring instruments. The systems are

identified as System A and System B. The 24 V dc systems are designed with no

automatic or manual transfers between the two systems.

Each +/-24 V dc system consists of two batteries and two battery chargers. The

rating of each battery is the same and is based on the maximum load required by any

one battery.

The batteries are housed in a room which is adequately ventilated.

VYNPS DSAR Revision 0 3.0-71 of 98 Each battery has an associated charger which is capable of supplying the continuous dc load and maintaining a float charge on the battery. The charger is also capable

of recharging the battery to full charge.

Each system is protected from high voltage by a charger output circuit breaker and

an overvoltage monitoring device which operates the charger input circuit breaker.

Any indication of high voltage, low voltage, low current, or rectifier failure, on

any charger, will activate a single alarm in the Main Control Room, with specific

annunciation provided on the local alarm panel located on the chargers themselves.

During normal operation, the continuous dc load is supplied by the battery charger.

The batteries normally float on the system. Upon loss of a charger, the associated

battery supplies its total dc load requirements.

3.3.3.5.4 Inspection and Testing

The components of this system are inspected and tested in accordance with approved

station procedures/programs.

3.3.4 Fire Protection System

3.3.4.1 Objective

This system is designed to provide fire protection for the station through the use

of water; CO 2; FM-200; dry chemicals; foam; detection and alarm systems; and rated fire barriers, doors, and dampers.

3.3.4.2 Design Basis

The Fire Protection System shall prevent propagation of fire and isolate the areas

of the fire by:

1. Providing a reliable supply of fresh water for firefighting purposes.

2. Providing a reliable system for delivery of the water to potential fire locations.

3. Providing automatic fire detection in those areas where the danger of fire is more pronounced.

4. Providing fire extinguishment by fixed equipment activated automatically or manually in those areas where danger of fire is most pronounced.

VYNPS DSAR Revision 0 3.0-72 of 98

5. Providing manually operated fire extinguishing equipment for use by station personnel at selected locations.

6. Providing means to isolate areas so that fires are prevented from propagating from one area to another.

3.3.4.3 Description

The Vermont Yankee Fire Protection Program makes use of detection and suppression

systems, separation criteria, rated fire barriers and seals, fire stops, procedures

and fire watches, manual hose stations, and training.

The fire protection program for the permanently defueled state has been developed

based on the applicable requirements of 10CFR50.48 and BTP APCSB 9.5-1, Appendix A.

The Fire Hazards Analysis (FHA) documents existing plant configurations and defines

the resources available for the prevention and limitation of damage from fire (Reference 1). In addition to plans and physical configurations for fire

protection, fire detection, fire suppression and limitation of fire damage, the FHA

also provides an overall description of the fire protection program.

The Fire Protection System is illustrated on Drawing G-191163, Sheets 1 and 2.

Water-type fire protection equipment has been limited in those areas where the

potential spread of radioactive contamination due to release of water for the

firefighting would result in more severe consequences than the results of a fire.

Fires in these areas will be primarily fought using portable dry chemical or carbon

dioxide extinguishers.

Water for the Fire Protection System is provided by two vertical turbine-type

pumps, one electric motor-driven and one diesel-driven. Each pump has a capacity

of 2,500 gpm at 125 psi discharge pressure. The pumps and drivers are located in

the intake structure. They discharge to an underground piping system which serves

the exterior and interior Fire Protection Systems.

The motor-driven pump is supplied from a 480 V bus. The diesel engine drive is

approved for fire pump service and is provided with its own fuel oil supply and

starting equipment.

The pressure in the Fire Main System is maintained at approximately 100 psig by an

interconnection to the Service Water System. An orifice in the 1.5 inch

pressurizing line limits pressure maintenance flow from the Service Water System to

30 gpm during normal operation. A check valve in the connecting pipe prevents

backflow.

VYNPS DSAR Revision 0 3.0-73 of 98 Operation of the fire pumps is controlled from pressure switches in the discharge piping. The motor-driven pump starts at a predesignated system pressure (typically

85 psig). The diesel-driven pump starts if the pressure continues to drop (typically 75 psig). The motor-driven pump automatically shuts down when the Fire

System pressure is restored to the normal range (typically 100 psig) for

approximately seven minutes. The diesel-driven pump continues to operate until

shut down manually.

The yard piping consists of a 12-inch underground piping loop around the

entire station, with valved branches serving 10 fire hydrants. Hose houses, located at these hydrants, contain standard hose house equipment. Valved branches

from the piping loop supply water for interior fire protection and transformer fire

protection purposes. Sectionalizing valves in the yard piping loop permit

isolation of portions of the loop, without interruption of service to the entire

system.

A main fire protection header in the Turbine Building supplies the following fire

protection services:

1. Automatic dry-pipe deluge systems with fixed water spray nozzles for the startup, the main, and the auxiliary

• transformers as well as a Turbine Building water curtain
• . These systems are operated by heat detectors.

2. Preaction Fire Protection System for the H 2 seal oil area and the Turbine Building condenser and heater bay area

• . Heat actuated devices initiate the opening of the deluge valve. The system utilizes sealed sprinkler heads, thus

sprinkling only those areas where the heads have been melted.

3. Automatic wet pipe sprinkler system for the condensate demineralizer resin storage area

• and the Turbine Lube Oil Room.

4. Automatic dry-pipe deluge system with fixed water spray nozzles for the Turbine Building loading bay area. This system is operated by ultraviolet and infrared

detectors.

5. An interior Fire Loop System in the Turbine Building and office area. This loop services nineteen hose stations located within these areas. In addition, three

hose stations in the service areas are served from a separate header.

6. The interior Turbine Building fire loop also serves a manual foam station which provides protection for the Diesel Day Tank Rooms, the Diesel Rooms, transformers, and the Turbine Lube Oil Storage Rooms.

Not required to satisfy the requirements of 10CFR50.48 or BTP APCSB9.5-1, Appendix A. VYNPS DSAR Revision 0 3.0-74 of 98

7. An interior Reactor Building loop. This loop services two standpipes and fifteen hose stations.

8. The interior Reactor Building loop serves a preaction Fire Protection System for the Reactor Building to cable vault cable penetration area at the Elevation 252'

and the upper northwest corner room at the Elevation 232' of the Reactor

Building. The system utilizes sealed sprinkler heads and an

automatically-actuated deluge valve, thus requiring the trip of both an

ionization detector and melting of the heads for system operation.

9. The interior Reactor Building loop also serves a Reactor Recirculation Pump Motor Generator Set Foam System. This Foam System is a fully automatic, open

nozzle suppression system, actuated by a two-zone detection system. An

actuation signal to the system is provided when both a thermal detector and an

ionization detector are tripped.

10. The interior Turbine Building loop also services a wet pipe spray system for the con-demin storage area. This system includes a remote flow alarm.

The cable vault and Switchgear Rooms are protected by fully automatic total

flooding CO 2 suppression systems. The Cable Vault CO 2 suppression system is initiated by ionization detectors. The Switchgear Room CO 2 suppression system is initiated by ionization detectors coincident with thermal detection. Bottles

located in the West Switchgear Room System may also provide a backup or second shot

to the cable vault if desired. The Diesel Fire Pump Fuel Oil Storage Tank Room is

protected by a total flooding FM-200 suppression system initiated by an ionization

detector coincident with a thermal detector.

The diesel fuel oil storage tank is protected by a manual foam station onboard a

tow behind foam tote trailer requiring connection to a site fire hydrant for a

water source.

The yard loop supplies a wet pipe sprinkler system for the warehouse

• and the house-heating Boiler Room
• . These systems are equipped with alarm check valves.

Not required to satisfy the requirements of 10CFR50.48 or BTP APCSB9.5-1, Appendix A.

VYNPS DSAR Revision 0 3.0-75 of 98 Fire detection devices are provided in areas which are not normally occupied, in areas where substantial quantities of combustible materials are present, or in

other areas determined to be highly sensitive. These detection systems provide

local and remote alarms, as well as annunciation in the Main Control Room. In some

instances trip signals are provided directly to deluge systems or electrically

operated fire dampers.

Portable fire extinguishers are located throughout the buildings at the site.

Portable fire extinguishers use dry chemical, CO 2 and water.

Buildings are constructed of steel and concrete with fire walls and/or shield walls

which isolate separate areas. Consideration has been given to the use of

noncombustible and fire-resistant materials throughout the facility, particularly

in the containment, Control Room, and areas containing critical portions of the

plant, such as components of the Engineered Safeguards Systems.

Fire barriers have been identified and their integrity assured by self-closing

doors (exception: RHR corner room doors at El. 213'-6" are not self-closing), normally locked doors, alarmed doors, doors checked daily, automatic fire dampers, and controlled procedures for penetration sealing and fire barrier repair. This

includes the northwest stairwell's ability to function as a fire exit.

Water flow alarms are provided in critical locations and annunciate in the Control

Room to provide positive indication of Fire Water System operation.

3.3.4.4 Inspection and Testing

The fire pumps, water suppression systems, CO 2 systems, FM-200 system, foam systems (manual and automatic), fire barriers, fire doors, fire dampers, detection and

alarm systems, and portable extinguishers are inspected and tested periodically in

accordance with approved station procedures/programs. All equipment is accessible

for periodic inspection.

3.3.4.5 References

1. Vermont Yankee Nuclear Power Station Fire Hazards Analysis

3.3.5 Heating, Ventilating and Air Conditioning Systems

3.3.5.1 Objective

The objective of the Heating, Ventilating, and Air Conditioning Systems is to

provide suitable environmental conditions for facility personnel and equipment.

VYNPS DSAR Revision 0 3.0-76 of 98 3.3.5.2 Design Bases The design bases of the Heating, Ventilating, and Air Conditioning Systems are as

follows:

1. Provide appropriate temperature and humidity conditions for personnel and equipment.

2. Limit exposure of personnel to airborne contaminants by controlled migration of air from radioactively clean areas to areas of progressively higher

contamination.

3. Normally, filter outside air to limit the introduction of particulate matter to the plant. During winter operation, certain filter media may be removed to

prevent freezing.

4. Vent potentially contaminated tankage through systems that exhaust to the plant stack.

3.3.5.3 Description

Flow diagrams for the Heating and Ventilation Systems are shown on Drawings G-

191237, Sheet 1 and 2, G-191236, G-19138 and G-191254.

The design temperatures used for the Heating and Ventilation Systems are provided

as follows:

VYNPS DSAR Revision 0 3.0-77 of 98 Outdoor Summer: 90 °F dry bulb, 75 °F wet bulb Winter: -12 °F dry bulb Indoor Reactor Building: Maximum: 100 °F (occupied areas) Minimum: 65 °F (refueling floor) 55°F (occupied areas other than refueling floor) Turbine Building: Maximum: 105 °F Minimum: 50 °F Radwaste Building: Maximum: 100 °F (occupied areas and Collector Tank Room) 120°F (equipment cells) Minimum: 65 °F (occupied areas) 50°F (equipment cells and Collector Tank Room) Control Room and Service Building: Maximum: 78 °F dry bulb, 50% relative humidity Minimum: 72 °F dry bulb

3.3.5.3.1 Reactor Building

The Reactor Building normal Heating, Ventilating, and Air Conditioning System

limits exposure of personnel to airborne contaminants and maintains appropriate

temperature conditions for personnel and equipment.

VYNPS DSAR Revision 0 3.0-78 of 98 The Reactor Building normal HVAC System migrates air from clean accessible areas to areas of progressively higher contamination or potential contamination, removes the

normal heat losses from all equipment and piping in the Reactor Building, limiting

the temperatures to approximately 100ºF, filters outside air to limit the

introduction of airborne particulate matter to the station, and exhausts

potentially contaminated air to the stack. During winter operation, certain filter

media may be removed to prevent freezing.

The Reactor Building normal HVAC System consists of a supply and exhaust side. See

Drawing G-191238.

The supply side includes in the direction of air flow, outside louvers, automatic

dampers, automatic roll-type filters, steam heating coils, and two double-width

centrifugal fans each sized for the full system capacity of 53,800 cfm. This

capacity provides approximately 1.5 net Reactor Building air changes per hour.

The exhaust side consists of two paralleled single-width centrifugal fans, each

having full system capacity of 55,800 cfm.

The excess of exhaust fan capacity over the supply fan capacity ensures against

building out leakages during normal operation. The main supply and exhaust ducts

penetrate the Reactor Building, each through two butterfly isolating valves in

series. The valves in the main supply duct are powered from different buses. This

is also true of the valves in the main exhaust duct. All four isolating valves

fail closed.

To permit maintenance of one fan while the other is in service without danger of

contamination, an isolation damper is provided at the inlet. Also, an isolation

outlet damper is provided to minimize the possibility of contamination through the

idle fan due to either stack backflow or recirculation from the active fan.

In addition, gravity dampers, i.e., non-return or backdraft dampers, are provided

to prevent reverse flow at all ventilating supply openings for areas having

contamination potential and in all branch exhaust ducts connecting with main ducts

which carry exhaust from areas having contamination potential.

Failure of a gravity damper to operate in a branch exhaust line will not result in

cross contamination. Each branch exhaust line consists of two 100% capacity

exhaust fans, a gravity damper on the discharge side of each fan, and a third

gravity damper in the branch line just prior to entering the main exhaust duct.

With the above arrangement, no backflow will occur through the branch exhaust lines

even in the event a gravity damper fails open and both exhaust fans are

inoperative.

VYNPS DSAR Revision 0 3.0-79 of 98 Failure of a gravity damper to operate in a supply line could result in some cross-contamination only if both redundant branch exhaust transfer fans are

inoperative. This is extremely unlikely.

Axial booster fans, each supported by an automatically cut-in standby unit, are

provided throughout the exhaust system to overcome air circuit losses.

A single fan is provided in the RCIC Room that circulates air between the RCIC Room

and the torus area.

In general, duct work is of galvanized steel. Duct work under positive pressure

exhausting to the main stack is of welded construction to minimize outleakage.

Local fan coolers are located in selected areas. Each fan cooler consists of a

centrifugal fan section and an air-cooling coil supplied with water from the

Service Water System. During operation of the cooler, air is recirculated and

cooled with little or no duct attachments.

A purge exhaust fan permits exhausting the drywell or the suppression chamber. The

upstream end of the purge exhaust fan is connected to the Primary Containment

Atmospheric Control System through butterfly valves which are remotely actuated

from the Main Control Room panel.

The downstream end of the purge exhaust fan discharges into the Reactor Building

normal Exhaust System where exhaust fans direct the purged air to the main stack.

All equipment and components are accessible for inspection, adjustment, and

testing. The only moving parts in a backdraft (gravity) damper are pinned joints

and bearings (dry, oil-impregnated porous metal, Teflon, or Zytel). Periodic

maintenance is performed on backdraft dampers.

Air flow quantities through ducts and air outlets were measured and balanced to

correspond to design quantities. The correct direction of space-to-space air

migration was thereby ascertained.

Proper sequences of operation, as well as correct control point adjustments were

determined during station pre-operational tests to assure conformity to the

requirements and intent of the specifications and drawings.

VYNPS DSAR Revision 0 3.0-80 of 98 3.3.5.3.2 Turbine Building The Turbine Building is ventilated by means of two Central Air Supply Systems.

Each system includes, in the direction of flow, outside air louvers, automatic

dampers, filters, steam heating coils, an air washer with recirculating spray water

pump, and two fans. Each fan is sized for full system capacity.

The primary exhaust system consists of two parallel full capacity single-width fans

with automatic isolating dampers. One fan serves as a standby, to be placed in

service upon failure of the lead fan.

Air is migrated from clean areas to areas of maximum contamination. To assure

positive exhaust and avoid cross contamination through the Duct System, a separate

duct exhausts the spaces for the gland steam jet air ejector, condenser vacuum

pump, and condensate backwash receiving tank.

A system of automatic dampers is provided for the condensate demineralizer cells to

assure maintenance of negative pressure at all times.

The operating floor is supplied with 15,000 cfm from the main system, and this air

migrates to the exhausts located at the mezzanine level. Additional ventilation is

provided by two centrifugal fans that induce air flow through electrically

interlocked wall louvers and exhaust the air to the main stack. Each fan exhausts

25,000 cfm of air. They are designed to operate together for a total capacity of

50,000 cfm. The system may be operated either with both fans running in parallel

or with a single fan running alone, depending on seasonal ventilation requirements.

A 4,000 cfm centrifugal fan exhausts the welding booth through a prefilter and

filter, thereby trapping any particulate matter before discharge back into the

machine shop.

The emergency diesel generator enclosures are independently cooled by an exhaust

ventilation system interlocked to operate simultaneously with the diesel

generators.

The office and warehouse extension at the south end of the Turbine Building has

ventilation to the classroom area and large office area only. Roof units provide

cooling and ventilation makeup and exhaust taken at the unit. Heating is from the

house boilers, or is electric baseboard.

VYNPS DSAR Revision 0 3.0-81 of 98 3.3.5.3.3 Main Control Room The system serving the Main Control Room is designed to provide summer air

conditioning and heating during the winter.

The Supply System has a 12,500 cfm capacity and includes, in the direction of flow, a wall louver, automatic outside air damper, filters, chilled water cooling coil, steam heating coil, centrifugal fan section, a system of duct work, and air

outlets.

The Supply System chilled water coil is serviced by a double circuit refrigeration

plant to assure continuity of cooling. Refrigeration plant components are one

double circuit water chiller with chilled water pumps, two air-cooled condensers, piping, and controls.

A remote manual switch located in the Main Control Room permits closure of the

outside air damper, Control Room kitchen and bathroom exhaust dampers, and Computer

Room supply damper, in order to isolate the Control Room, if required.

SAC-1, which supplies the Control Room, contains a humidifier in the air supply

duct after the Computer Room duct. This unit is controlled by a humidity sensor in

the Control Room and has an alarm for high humidity level.

Upon a loss of the Control Room Ventilation System the SAC 1A/B dampers could fail

to the closed position. Operator actions, including manual control of appropriate

dampers, can be taken to restore system flow as discussed in plant procedures

The Control Room can be isolated by manually closing the fresh air inlet branch

damper, cable vault damper, and Control Room vent paths. This also puts the

Control Room ventilation in the recirculation mode of operation.

3.3.5.3.4 Service Building

The original portion of the Service Building is entirely air conditioned by an air

handling unit having 15,000 cfm capacity and which in the direction of flow, consists of dampers, mixing box, filters, a chilled water cooling coil, and a fan

section. Electric zone reheat coils compensate for cooling load variations in

different areas.

The chilled water coil is served by a service water-cooled package chiller and

chilled water pump.

Air from spaces having potential contamination, such as the chemistry laboratory, is not recirculated back to the air handling unit -- it is directly exhausted to

the plant stack by one of two full capacity fans. VYNPS DSAR Revision 0 3.0-82 of 98 The added portion of the Service Building on the north side is cooled and

ventilated by packaged units on the roof. Makeup and exhaust is local at each

unit. Heat is from the house boilers.

3.3.5.3.5 Radwaste Building

Drawing G-191236 shows the air flow through the Radwaste Building. The Ventilation

System is designed to supply filtered and heated air at approximately 11,700 cfm

and exhaust it through high efficiency particulate air filters. With the exception

of controlled activities, no air is discharged or relieved from the building except

through the plant stack.

In general, air is supplied to clean accessible areas and migrated into the

contaminated equipment cells through dampered openings. Gravity dampers located in

these transfer openings prevent backflow of contaminated air. The openings are

located and configured to minimize the danger of radiation "shine".

A supply unit and exhaust unit are located on the Radwaste Building roof, Elevation

264 feet, 6 inches.

In the direction of flow, the Supply System includes outside air louvers, automatic

dampers, filters, heating coils, two 50% capacity double width centrifugal fans, duct distribution, and air outlets. The automatic outside air damper is modulated

to maintain constant flow to assure maintenance of air balance despite wind

conditions.

An exhaust air register which includes a balancing damper is located in each space

where radioactive contamination in the form of dust, gas, or vapor could be

released. Air exhausted from these contaminated spaces is ducted to the high

efficiency filtration assemblies described below.

The Exhaust System includes two 50% capacity fan-filtration trains, each consisting

of prefilters, absolute filters, filter section isolating dampers, and an axial

fan. A gravity damper at each fan outlet prevents recirculation from the operating

fan or backflow from the Exhaust Systems.

Opening and using doors and hatches located in the Radwaste Building is controlled

by administrative controls to assure that the release of any airborne material is

controlled and assured to be minimal and as-low-as-reasonably achievable.

Samples of main exhaust duct air are continuously monitored for radioactivity.

High activity is annunciated in the Main Control Room.

VYNPS DSAR Revision 0 3.0-83 of 98 3.3.5.3.6 Heating Boiler System Drawing G-191254 shows the process flow diagram for the Station Heating Boiler

System. The Heating Boiler System is designed to provide a source of steam for

space heating and process requirements.

Each of the two boilers is rated for approximately 50% of the calculated heating

load.

All pressure containing parts have been designed to the ASME Boiler and Pressure

Vessel Code, Section I.

The boiler plant consists of two forced draft, four pass, 50% capacity, No. 2

oil-fired fire tube boilers, supplemented by a condensate return tank and three

cross-connected 50% capacity feedwater pumps. Each boiler is equipped with a

locally mounted control panel. Indicating lights will show "flame failure," "low

water," "fuel valve open," and "load demand."

Also provided is a Fuel Oil Pumping System, blowdown tank, and remotely located

condensate pump, receiver sets, Sampling System, and Chemical Addition Systems.

The Station Heating Boiler System is located in the south end of the Turbine

Building. The heating boiler feedwater is monitored by a process liquid radiation

monitor (see Drawing G-191254).

The unit is equipped with the necessary controls and safety devices to operate

automatically. The combustion safety control is provided with a safety lockout in

the event of flame failure or failure to start, which requires manual reset before

the automatic cycling can continue. The draft fan controls are interlocked with

the burner controls to prevent operation of the burner under improper draft

conditions. The flame requirements are designed to meet FIA and NEPIA

requirements.

3.3.5.3.7 Advanced Off-Gas Building HVAC

The building is heated, cooled, and ventilated by rooftop units that have intakes

at the units. Exhaust air from the building is filtered and the exhaust fans

discharge to the plant stack.

VYNPS DSAR Revision 0 3.0-84 of 98 3.3.5.4 Inspection and Testing All equipment and components are accessible for inspection, adjustment, and

testing.

Absolute particulate filters will be factory tested with 0.3 micron monodisperse

thermally generated dioctyl phthalate (DOP) aerosol. Minimum acceptable efficiency

is 99.97% as measured by a light-scattering photometer.

To assure that gaskets and seals are properly installed and that no damage has

occurred to the filter during shipment or handling, in-place tests using a

polydisperse cold generated aerosol will be performed initially and then

periodically as required.

3.3.6 Instrument and Service Air Systems

3.3.6.1 Objective

The objective of the Instrument and Service Air Systems is to provide the station

with the compressed air requirements for pneumatic instruments and controls and

general station services.

3.3.6.2 Design Basis

The Instrument and Service Air Systems shall provide the plant with a continuous

supply of oil-free compressed air. Dry air shall be supplied to plant instruments

and controls as required. Undried air shall be provided for various station

services.

3.3.6.3 Description

The Compressed Air Systems are shown on Drawing G-191160, Shs.1 through 8. The

systems include nonlubricated air compressors connected in parallel, each with a

built-in intake filter-silencer, after-cooler, and moisture separator. The

compressors discharge to two vertically mounted air receivers.

Each compressor will function in either the lead or lag mode. Normally, compressors which are selected to the Lead position will maintain pressure between

100 and 105 psi. The compressors which are in the lag position will start when

header pressure drops to a predetermined value below the normal operating range.

If the backup compressors run unloaded for a preset period of time, they will

automatically shut down and remain shutdown unless header pressure drops to the

predetermined value.

VYNPS DSAR Revision 0 3.0-85 of 98 Separate piping is provided at the discharge of the air receivers for the Instrument Air System and the Service Air System. The compressed air of the

Instrument Air System passes through two parallel branches both of which contain

the following equipment:

1. Prefilter - This unit filters the air to remove moisture droplets, particles of dirt, rust, and scale of approximately 3 microns and larger through the use of automatic traps.

2. Dryer - This unit regenerates (dries out) its desiccant by a heater-less pressure-swing process. Each dryer is sized to provide 450 SCFM.

3. After Filter - This unit, like the prefilter, filters the air to remove moisture droplets, particles of dirt, rust, and scale of approximately 3 microns and larger.

The Service and Instrument Air Systems meet all appropriate seismic criteria.

All original piping is designed in accordance with USAS B31.1, 1967. The air

compressor discharge piping to Valve V72-2B is designed to USAS B31.1, 1977.

The Service and Instrument Air Systems are designed to operate at a pressure of 100

psig and supply 322 scfm +/-5% compressed air with one compressor operating.

3.3.6.4 Inspection and Testing

The Instrument and Service Air Systems are normally in continuous operation.

Satisfactory performance of these systems is demonstrated continuously without the

need for any special inspection or testing.

3.3.7 Process Sampling

3.3.7.1 Objective

The process sampling systems provide representative samples for analysis.

3.3.7.2 Design Basis

The sampling systems shall be designed to ensure accuracy and sensitivity of

measurement of process fluids.

VYNPS DSAR Revision 0 3.0-86 of 98 3.3.7.3 Description 3.3.7.3.1 General

For flow diagrams of the station liquid sampling system, refer to Drawings G-191164

and G-191165.

Fluids and gases are sampled continuously or periodically from selected equipment

or systems. Samples are taken either as grab samples or continuously. Grab

samples are taken from the collection area to the laboratory for analysis. The

continuous samples pass through analyzers and the results are recorded.

The following table lists the description, location, and purpose of the various

monitoring points associated with sampling process fluids.

VYNPS DSAR Revision 0 3.0-87 of 98

3.3.7.3.2 Radwaste Building Sample Panel

This panel includes conductivity elements, conductivity indicating transmitters, and individual grab sample connections with the outlets enclosed in a hooded sink

provided with exhaust ventilation.

3.3.7.3.3 Gas Sampling and Monitoring

A list of gas samples, their locations, and purpose is provided below. Description Location Purpose Stack sample Stack Particulate and gaseous activity and

iodine release Ventilation gases

a) Reactor Building b) Radwaste Building

Fan discharge Fan discharge

Activity release Activity release The capability exists to sample the ventilation gases, but these locations are not routinely sampled. The fan discharges from the Reactor Building and the Radwaste

Building are routed to the stack which is sampled continuously. Description Location Purpose Waste disposal

a) Waste surge tank b) Waste collection tank c) Floor drain collection tank d) Chemical waste tank e) Waste sample tank f) Floor drain sample tank g) Fuel pool filter demineralizer influent h) Fuel pool filter demineralizer effluent i) Floor drain filter effluent j) Waste filter demineralizer k) Waste demineralizer

Outlet pipe

Pump discharge

Pump discharge

Pump discharge

Pump discharge

Pump discharge

Inlet pipe

Outlet pipe

Outlet pipe Outlet pipe

Outlet pipe

Process data

Process data

Process data

Process data

Discharge suitability

Discharge suitability

Fuel pool quality

Filter demineralizer

efficiency

Filter efficiency

Filter demineralizer

efficiency

Demineralizer efficiency Makeup a) Cation effluent b) Anion effluent c) Mixed bed effluent d) Demineralized water storage tank e) Condensate storage tank

Outlet pipe

Outlet pipe

Outlet pipe

Pump discharge

Pump discharge

Demineralizer efficiency

Demineralizer efficiency

Demineralizer efficiency

Water quality

Water quality

VYNPS DSAR Revision 0 3.0-88 of 98 3.3.8 Station Water Purification, Treatment and Storage 3.3.8.1 Station Makeup Water System

3.3.8.1.1 Objective

The objective of the Makeup Water System is to maintain a supply of treated water

that may be used as makeup for the facility.

3.3.8.1.2 Design Basis

The Makeup Water Treatment System shall be designed:

1. To produce water at a rate of 75 gpm, as needed.

2. To store up to 50,000 gallons of processed water

3. To provide water of a quality required to support the safe storage and handling of irradiated fuel.

3.3.8.1.3 Description

3.3.8.1.3.1 General

The makeup water for the facility is provided by processing water from the West

Well. A backup source is river water from the Connecticut River. The makeup water

treatment system consists of a pretreatment subsystem and a demineralization

subsystem.

Demineralizer equipment will typically produce water of the following quality when

required to maintain level in the demineralizer water storage tank.

Specific conductivity at 25 °C 1 micromho/cm pH at 25 °C 7 Silica (as Si0

2) 0.10 ppm Chloride (as Cl-) 0.01 ppm

3.3.8.1.3.2 Pretreatment Subsystem

The purpose of the Pretreatment System is to process water by means of chemical

coagulation and filtration and to provide water of suitable quality to the

Demineralization System. Refer to Drawing G-191161. The normal supply is well

water. Raw river water may also be connected to the Pretreatment System inlet

through the Service Water System.

VYNPS DSAR Revision 0 3.0-89 of 98 The pretreatment subsystem is composed of a raw water heating tank and an in-line coagulation chemical treatment system utilizing alum, hypochlorite, and coagulant

aid wet chemical feeders followed by a retention tank, two dual media pressure

filters, a clearwell, and two carbon purifiers. When processing well water, only

the dual media pressure filters, clearwell, and carbon filters are used.

The Pretreatment System is normally initiated manually to fill the clearwell just

prior to and during batch processing of the clearwell to make demineralized water.

Control of the chemical feeders is manual and adjustable.

Backwashing and rinsing of the pressure filters and carbon purifiers is

accomplished manually, with backwashing and rinse water being taken from the

clearwell.

3.3.8.1.3.3 Demineralizer Subsystem

The demineralizer subsystem consists of one train that consists of a cation, anion, and a mixed bed ion exchanger.

The system is a manual system that is aligned by appropriate facility personnel

when required to process water. The demineralizer shuts down on high level in the

demineralized water storage tank and upon a high conductivity from the mixed bed

effluent.

A control panel is provided to contain the switches, indicating lights, instrumentation, etc., necessary for control of the demineralization system.

3.3.8.1.4 Inspection and Testing

Operation of the Station Makeup Water System is on demand at intermittent intervals

to replenish water in the demineralized water storage tank. The equipment can be

visually inspected periodically. Sampling of the effluent of the Pretreatment

System and demineralizer is performed regularly, as required, to verify operational

performance.

3.3.8.2 Potable and Sanitary Water System

3.3.8.2.1 Objective

The objective of the potable and sanitary water system is to provide treated water

suitable for drinking and for sanitary purposes.

VYNPS DSAR Revision 0 3.0-90 of 98 3.3.8.2.2 Design Basis

A sufficient supply of water, filtered and purified as required, shall be provided

for drinking and sanitary uses.

3.3.8.2.3 Description

Water from two (2) on-site wells, serving two (2) systems (Main and PSB) is

supplied to all water closets, electric water coolers, lavatories, urinals, service

sinks, combination emergency showers and eyewash, kitchen sinks, bench sinks, showers and wall hydrants. A 1,000-gallon potable water storage tank provides

added assurance of a continuous potable water supply to the main system.

Combination emergency showers and eyewash are provided in potentially contaminated

areas for the safety of facility personnel. A 1-1/2 inch potable water supply line

provides a deluge effect to quickly dilute any acids or radioactive material.

Emergency showers and eyewash are located in the chemical laboratory and the water

treatment area of the turbine building where facility personnel are exposed to

concentrated acids. Another emergency shower and eyewash is centrally located in

the radwaste building providing protection from acids and radioactivity.

Potable and sanitary water piping is copper with solder joint fittings. Vacuum

breakers are provided for water closets and urinals while all other sanitary

fixtures have air gaps in order to prevent any wastes from backing up and

contaminating the potable and sanitary water system. Main branch lines are valved

so various parts of the building can be isolated and shut down without affecting

other areas.

Electric hot water heaters of sufficient heating and storage capacity provide hot

water for all lavatories, service sinks, kitchen sinks, bench sinks and showers. A

hot water circulating pump and piping is provided to assure that hot water is

always readily available.

Thermostatic master mixing valves provide tempered water to all shower heads.

Tempered water is hot enough for showering and at the same time safely protects

facility personnel from possible scalding. A mixing valve is also provided at each

shower head to mix cold water and tempered water.

VYNPS DSAR Revision 0 3.0-91 of 98 3.3.9 Lighting Systems 3.3.9.1 Objective

The lighting system provides adequate lighting in all areas of the facility where

lighting is required.

During loss of power, the lighting in all areas essential to the safe storage of

irradiated fuel is provided by the following lighting systems:

1) The normal emergency lighting system - ac supplied.

2) The emergency lighting system - dc supplied.

3) Local emergency lights with self-contained rechargeable batteries.

3.3.9.2 Design Basis

The lighting systems shall provide adequate lighting for the safe storage of

irradiated fuel under all conditions.

3.3.9.3 Description

The safe storage and handling of irradiated fuel requires that adequate lighting be

available for the operation, control, and maintenance of equipment.

Buses which supply power for lighting are normally powered from the startup

transformers. Upon a loss of normal power, the buses are powered from a standby ac

power source, thereby maintaining lighting.

Critical areas and access routes are illuminated by DC lighting during the power

transition.

Portable, battery-operated lighting fixtures are available to permit maintenance of

standby power sources. They are located at the standby ac supply equipment. These

units are wall-mounted and are connected to normal ac circuits which keep the

self-contained batteries in a fully charged condition. Upon loss of ac, the lights

automatically become lighted, using the self-contained batteries as a source.

In the control room, battery-operated emergency lights are available. These units

are wall mounted with remote lights mounted in the ceiling panels and are connected

to a normal emergency ac circuit. Upon the loss of the normal emergency circuits, the lights automatically become lighted using self-contained batteries as a source.

VYNPS DSAR Revision 0 3.0-92 of 98 The level of lighting intensity is consistent with the nature of the work likely to be performed for maintenance and operation upon loss of normal ac power. Under

conditions where standby sources of power are limited, a minimum level of lighting

intensity is maintained in the access routes between the main control room and the

diesel generators. The following table shows minimum levels of lighting intensity.

3.3.9.4 Inspection and Testing

Normal lighting is used continuously.

Normal emergency and emergency lighting systems are periodically tested to check

operation of standby sources. Portable lighting sets are tested periodically to

demonstrate their functional performance.

3.3.10 Communication Systems

3.3.10.1 Objective

The objective of the communications system is to provide a reliable, convenient, and audible communication system which meets the requirements of facility operation

and maintenance.

3.3.10.2 Design Basis

The communications system shall consist of an intra-site operation and public

address system, a sound-powered telephone system, a dial telephone system, an

inter-site communication system, and an off-site radio communication system. To

ensure continued intra-site and off-site communications, power for these systems

shall be provided from the vital ac bus.

Min. Standby Light Intensity, Foot Candles Location AC Supplied DC Supplied Main Control Room 5 3 (Local) Standby AC Equipment 3 3 (Local) Access Routes 2 Silhouette VYNPS DSAR Revision 0 3.0-93 of 98 3.3.10.3 Description The communications system consists of several types of communication media

including intra-site operation and public address communication, sound-powered

telephone communication, dial telephone communication, inter-site microwave

communication, and off-site radio communication.

The function of the communications system is to permit convenient and dependable

communications between all areas of the facility vital to operation and maintenance

and protection of personnel. The systems are as follows:

1. Intra-Site Operation and Public Address System This system consists of speakers and microphones located throughout the

facility. The system has four transistorized channels and provides separate and independent page and party line channels. The page channel may be used to call

personnel over the speakers as well as issue facility-wide instructions. The

party line channels may be used to carry on inter-communication after the page

call is completed, thereby making the page channel available to others.

Simultaneous conversations can take place, one on each of the channels, without

interference. The system has an output adequate to be clearly audible in all

appropriate facility areas.

2. Sound-Powered Telephone System This system allows private communications between specific areas and pieces of

equipment for maintenance purposes of either a routine or non-routine nature.

Two independent channels are provided at each location, and the system can be

used as a back-up communication system.

3. Dial Telephone System A dial telephone system is provided for normal communications between offices

and work areas which are routinely occupied. Units are also located

conveniently within the reactor building for use by the facility personnel.

This system is connected to points outside the facility through both the

commercial telephone system and a microwave network.

VYNPS DSAR Revision 0 3.0-94 of 98

4. Microwave Communication System The microwave system provides for the interchange of information between the

facility and the electrical dispatcher. Microwave equipment is located in the

switchyard control house and, with its battery and charger, is independent of

the other plant communications systems.

5. Off-Site Radio Communication System Radios located in the control room provide contact with the State Police, the

Utility Emergency Radio Network, and Mutual Aid.

3.3.10.4 Inspection and Testing

Operational tests are frequently made as a result of constant use of the

communications systems.

3.3.11 Process Computer System

3.3.11.1 Objectives

The objectives of the process computer are to aid facility personnel by

continuously assessing the readout of instrumentation relative to permissible

limits, to provide data accumulation and logging functions, and to serve as a

Safety Parameter Display System (SPDS) which shall provide a display of selected

variables to aid facility personnel in determining the status of the plant

3.3.11.2 Design Bases

The Process Computer System (PCS) shall support the following SPDS functions:

• Perform meaningful conversions and monitoring of radiation release paths;
• Monitor, calculate, and display meteorological, plume arrival and dose projection data for emergency response personnel.
• Provide Data to the Plant Data System for use at the TSC and EOF for emergency response

VYNPS DSAR Revision 0 3.0-95 of 98 3.3.11.3 Description 3.3.11.3.1 Computer System Components

3.3.11.3.1.1 Central Processor

The central processor performs various calculations, makes necessary

interpretations, and provides for general input/output (I/O) device control and

buffered transmission between I/O devices and memory. To ensure data integrity, the computer system has built-in testing checks and diagnostic facilities, such as

parity and error detection and correction in the processor, memories, and the

system bus, and automatic self-test at power-up. Real-time processing capability is

provided with battery backup to facilitate a rapid restart without loss of memory

or loss of processor clock time.

Power for the computer is supplied from an uninterruptible power source (UPS-2A)

which can supply power for a minimum of three hours while off-site power is not

available. 3.3.11.3.1.2 Auxiliary Memory Subsystem

Auxiliary memory consists of fixed disk drives.

The Auxiliary Memory Subsystem is designed for and provides the capability for

further expansion. The disk drives incorporate outstanding data reliability

characteristics, including Error Correction Code (ECC), microprocessor-controlled

diagnostics, and a modular design for easy maintenance.

3.3.11.3.1.3 Peripheral Input/Output Subsystem

The peripheral I/O equipment used to read programming data into and out of the

computer consists of a system console, terminals, printers, alarm typer, magnetic

tape and disk subsystems. The system console, magnetic tape and disk subsystems

are located in the Computer Room. The terminals, printers and alarm typer are

located in the Control Room.

VYNPS DSAR Revision 0 3.0-96 of 98 3.3.11.3.1.4 Process Input/Output Subsystem The process I/O hardware is a real time distributed, microprocessor based, intelligent industrial I/O processor. The I/O processor with its processing

capability reduces the host computer signal linearization and raw data scaling

tasks. The process I/O hardware consists of analog/digital input cards, pulse/sequence of events input cards, and analog/digital output cards all under

microprocessor control. The analog inputs accept analog signals from plant

instrumentation and provide signal conditioning for use in the computer system.

The digital input cards provide signal conditioning and filtering. Intermittent

signals and pulse-type inputs are handled by SOE/pulse input cards. These cards

have a programmable mode of operation, including interrupt on a specific count and

continuous count. This allows immediate response for processing of information

which otherwise might be lost if digital scanning techniques were used. The

process I/O hardware supports one second scan rates for digital inputs and sub-

second scan rates for analog inputs.

3.3.11.3.4 Monitor Alarm and Logging Functions

3.3.11.3.4.1 Analog Monitor and Alarm

The processor is capable of checking each analog input variable against three types

of limits for alarming purposes: (a) process alarm limits as determined by the

computer during computation or as preprogrammed at some fixed value by the user, (b) a reasonableness limit of the analog -input signal level as determined and

programmed by the user, and (c) a rate of change alarm limit as determined and

programmed by the user.

The alarming sequence consists of a one-line message on the alarm typer for each

point exceeding process alarm limits. Alarm messages may also be displayed on a

video display as selected by the user. A variable that is returning to normal is

signified by a one-line message on the alarm typer. Actuation of the alarm typer

provides facility personnel with an audible cue that an alarm message has printed.

The processor provides the capability to alarm the Main Control Room Annunciator

System in the event of abnormal PCS operation. Abnormal conditions for alarm

include loss of power and stall conditions. Stall conditions can be caused by

software failure, hardware failure or PCS over-temperature conditions.

VYNPS DSAR Revision 0 3.0-97 of 98 3.3.11.3.4.2 Digital Inputs Status Monitoring

1. Digital Input Status Logging The status alarm function scans digital inputs at regular intervals and

provides a printed record of system alarms. The record includes point

description, state, and time of occurrence.

2. Sequence Annunciator Detection and Logging Digital inputs associated with status changes of major plant equipment and

instrumentation are terminated on change-of-state detection sequence of

events input hardware. To aid facility personnel in analysis of any plant event, this function

archives SOE messages into multiple files. The SOE messages will be logged

in the order of their detection, with one millisecond (msec) resolution

accuracy. The time logs will be synchronized to the PCS (ERFIS) internal

clock at the time of occurrence.

3.3.11.3.4.3 Alarm Logging

The alarm logs required by the associated process programs are typed by the alarm

typer. Alarm printouts, as well as alarm summary displays, are used to inform

facility personnel of computer system malfunction; system operation exceeding

acceptable limits; and potentially unreasonable, off-normal, or failed input

sensors.

3.3.11.4 Inspection and Testing

The Process Computer System is self-checking. It performs diagnostic checks to

determine the operability of certain portions of the system hardware, and it

performs internal programming checks to verify that input signals and selected

program computations are either within specific limits or within reasonable bounds.

3.3.11.5 Cyber Security

The PCS is deterministically isolated from less secure digital components and

systems by a data diode. This device can transmit PCS data to the less secure

general user community, but there is no physical channel for data flow in the

reverse direction. This prevents malicious computer code from migrating to the

PCS. VYNPS DSAR Revision 0 3.0-98 of 98 3.3.11.6 Process Computer Data Feed to the Plant Data Server (PDS)

A datalink sends all data variables from the PPC, through the Data Diode, to PDS.

This is primarily for TSC and EOF Emergency Plan users.

VYNPS DSAR Revision 0 4.0-1 of 54 RADIOACTIVE WASTE MANAGEMENT TABLE OF CONTENTS Section Title Page 4.2 RADIATION SHIELDING ................................................... 5 4.2.1 Objective .................................................... 5 4.2.2 Design Basis ................................................. 5 4.2.3 Description .................................................. 6 4.2.3.1 Materials Description ........................... 6 4.2.3.2 Reactor Building ................................ 6 4.2.3.3 Main Control Room and Technical Support Center (TSC) .................................... 6 4.2.4 Surveillance and Testing ..................................... 6 4.3 HEALTH PHYSICS INSTRUMENTATION ........................................ 7 4.3.1 Objective .................................................... 7 4.3.2 Description .................................................. 7 4.4 RADIATION PROTECTION .................................................. 9 4.4.1 Health Physics ............................................... 9 4.4.1.1 Personnel Monitoring Systems .................... 9 4.4.1.2 Personnel Protective Equipment .................. 9 4.4.1.3 Change Area and Shower Facilities ............... 9 4.4.1.4 Access Control .................................. 9 4.4.1.5 Laboratory Facilities .......................... 10 4.4.1.6 Bioassay Program ............................... 10 4.4.2 Radioactive Materials Safety Program ........................ 10 4.4.2.1 Facilities and Equipment ....................... 11 4.4.2.2 Personnel and Procedures ....................... 11 4.4.2.3 Required Materials ............................. 12 4.5 LIQUID WASTE MANAGEMENT SYSTEMS ...................................... 12

VYNPS DSAR Revision 0 4.0-2 of 54 4.5.1 Equipment and Floor Drainage Systems ........................ 12 4.5.1.1 Objective ...................................... 12 4.5.1.2 Design Basis ................................... 12 4.5.1.3 Description .................................... 13 4.5.1.4 Inspection and Testing ......................... 19 4.5.2 Liquid Radwaste System ...................................... 19 4.5.2.1 Objective ...................................... 19 4.5.2.2 Design Bases ................................... 19 4.5.2.3 Description .................................... 20 4.5.2.4 Evaluation ..................................... 24 4.5.2.5 Inspection and Testing ......................... 25 4.6 SOLID WASTE MANAGEMENT ............................................... 35 4.6.1 Solid Radwaste System ....................................... 35 4.6.1.1 Objective ...................................... 35 4.6.1.2 Design Basis ................................... 35 4.6.1.3 Description .................................... 35 4.6.1.4 Inspection and Testing ......................... 37 4.7 EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING ........................ 39 4.7.1 Process Radiation Monitoring Instrumentation ................ 39 4.7.1.1 Plant Stack Radiation Monitoring System ........ 39 4.7.1.2 Process Liquid Radiation Monitoring System ......................................... 41 4.7.1.3 Reactor Building Ventilation Radiation Monitoring System .............................. 42 4.7.2 Area Radiation Monitoring System ............................ 47 4.7.2.1 Objectives ..................................... 47 4.7.2.2 Design Basis ................................... 47 4.7.2.3 Description .................................... 47 4.7.2.4 Inspection and Testing ......................... 49

VYNPS DSAR Revision 0 4.0-3 of 54 RADIOACTIVE WASTE MANAGEMENT

LIST OF TABLES

Table No. Title

4.5.2.1 Vermont Yankee Radioactive Liquid Waste Processing Parameters 4.5.2.2 Vermont Yankee Liquid Radwaste System Tank Capacities 4.5.2.3 Vermont Yankee Liquid Effluents 4.5.2.4 Activity Input to Liquid Radwaste System (Ci/yr) 4.5.2.5 Radionuclide Discharge Concentrations 4.6.1.1 Solid Radwaste Annual Disposal History 4.7.1.1 Process Radiation Monitoring Systems Characteristics 4.7.1.2 Process Radiation Monitoring System Environmental and Power Supply Design Conditions

4.7.1.3 Plant Stack Radiation Monitoring System Characteristics 4.7.2.1 Area Radiation Monitoring System Environmental and Power Supply Design Conditions

4.7.2.2 Locations of Area Radiation Monitors 4.7.2.3 Reactor Building Area Airborne Radiation Monitoring System 4.7.2.4 Technical Support Center Area Airborne Radiation Monitoring System VYNPS DSAR Revision 0 4.0-4 of 54 RADIOACTIVE WASTE MANAGEMENT LIST OF FIGURES

Reference Figure No. Drawing No. Title

4.5.2-1 G-191151 Radioactive Waste Building General Arrangement 4.5.2-2 G-191152 Radioactive Waste Building General Arrangement 4.5.2-3 5920-644 Sheet 2 Radwaste System, Process Diagram 4.5.2-4 G-191177 Sheet 1 Liquid Radwaste System 4.5.2-5 G-191177 Sheet 2 Liquid Radwaste System 4.5.2-6 G-191177 Sheet 3 Liquid Radwaste System 4.5.2-7 G-191177 Sheet 4 Liquid Radwaste System 4.5.2-8 Radwaste Area - Plan View 4.7.1-1 5920-00526 Process Radiation Monitoring System Instrumentation Diagram 4.7.1-2 5920-03994 Plant Stack Radiation Monitoring System Diagram 4.7.2-1 5920-430 Sheet 1 Area Radiation Monitoring System, Functional Block Diagram 4.7.2-2 Reactor Building Area Airborne Radiation Monitoring System VYNPS DSAR Revision 0 4.0-5 of 54 4.1 SOURCE TERMS In the permanently defueled condition VYNPS will no longer produce fission, corrosion, or activation products from operation. The radioactive inventory

that remains is primarily attributable to activated reactor components and

structural materials and residual radioactivity. The accumulation of small

amounts of solid waste may easily be controlled. Any future planned liquid

effluent releases will be evaluated prior to release, and appropriate controls

will be established. The Offsite Dose Calculation Manual ensures that VYNPS

complies with 10 CFR 50, Appendix I. 4.2 RADIATION SHIELDING 4.2.1 Objective Radiation shielding is utilized as appropriate to limit radiation damage to

equipment and associated structures and minimize exposure of station personnel

to radiation.

4.2.2 Design Basis

Radiation shielding was provided to restrict radiation emanating from various

sources throughout the plant. Since VYNPS is permanently defueled, many

installed components are no longer required to safely store irradiated fuel.

However, many of these components continue to contain radioactive material or

remain radioactive. Shielding that was originally designed to shield these

components while they supported reactor operation continues to provide

shielding from residual radioactivity in the permanently shut down condition. Shielding is provided to maintain personnel exposures below the limits specified in 10CFR20. Compliance with these regulations is achieved through

shielding design based upon generalized occupancy requirements in various areas

of the station, and upon administrative radiological protection procedures. Continuous occupancy areas outside the controlled access area, designated Zone I, are designed to a radiation level of 0.5 mrem/hr, while those inside

the controlled access area, designated Zone II, are designed to a level of

1 mrem/hr.

Within the controlled access boundary are areas, designated Zone III, which

will allow up to 10 hours per week occupancy and are designed for 6 mrem/hr.

Controlled areas that are designed for 100 mrem/hr allowing occupancy up to

5 hours per week are designated Zone IV.

Section 6.5 of the Vermont Yankee Permanently Defueled Technical Specifications VYNPS DSAR Revision 0 4.0-6 of 54 describes the radiation protection controls for all radiation areas with dose rates exceeding 100 mrem/hr.

Select areas will be equipped with local area monitoring devices. The Area

Radiation Monitoring System detects, measures, and records the general

radiation levels in areas where personnel may be required to work. The system

will actuate alarms if radiation exceeds preset levels.

4.2.3 Description

4.2.3.1 Materials Description

The shielding materials used are primarily concrete, water, and steel. High

density concrete, lead, and neutron-absorbing materials are used as

alternatives in special applications.

4.2.3.2 Reactor Building The design dose rate in most areas outside the drywell in the Reactor Building is 1 mrem/hr. The drywell and its internal structure are shielded so that most

areas outside it are accessible.

4.2.3.3 Main Control Room and Technical Support Center (TSC)

The shielding of the Main Control Room consists of poured-in-place reinforced

concrete. Side walls and roof are 2 feet thick and 1 foot, 8 inches thick, respectively.

The TSC is located on the second floor of the Administration Building. An

additional 2 inches of concrete was added to the existing 6-inch concrete floor

above the TSC for a total concrete shielding thickness of 8 inches.

The Main Control Room and the TSC are shielded so that no individual exposure

will exceed the limits set forth in Criterion 19, Appendix A of 10CFR Part 50.

4.2.4 Surveillance and Testing

Appropriate surveillances will be conducted by trained facility personnel.

These surveys provide continuing assurance that changes which might occur and

produce significantly different radiation fields are located and appropriately

posted. VYNPS DSAR Revision 0 4.0-7 of 54 4.3 HEALTH PHYSICS INSTRUMENTATION

4.3.1 Objective

The health physics instrumentation system is a supplemental system which

provides a flexible radiation detection capability throughout the facility. It

is intended to supplement the facility process and area radiation monitoring

systems in assuring that the facility is within design limits and to supply the

required radiation control information.

4.3.2 Description

Health physics instrumentation consists of both portable and fixed equipment.

Portable Instrumentation

Portable health physics instrumentation consists of the following types of

equipment:

1. Alpha survey meters, which contain a thin "window" and an alpha sensitive detecting element that permits the location and measuring of low levels of

alpha radiation contamination.

2. Beta-Gamma survey meters, which contain a thin windowed Geiger-Mueller tube or ionization chamber, and are used for detecting low levels of surface

contamination or for making direct radiation surveys.

3. Neutron survey meters, which contain a thermal neutron sensitive BF 3 tube or tissue equivalent proportional counter. These meters are used for

locating possible shielding voids, streaming paths, etc., in the reactor

building.

4. Beta-Gamma and neutron dose rate meters are used for determining stay times for radiation workers and for posting radiation area warning signs.

High range beta-gamma meters provide dose rate information during any event

involving high levels of radiation. Neutron dose rate meters respond to

and provide an indication of the entire spectrum of neutrons encountered

around a nuclear reactor.

VYNPS DSAR Revision 0 4.0-8 of 54

5. Emergency Kits - Emergency kits are used by mobile teams during any event involving a possible release of radioactive materials. Each kit contains a

beta-gamma survey meter and an air particulate sampler, plus any other

equipment normally used by a particular survey team. Emergency kits are

provided for: off-site, on-site, control room, and technical support

center.

6. Air particulate samplers, which are air pumps which pull a known flow rate of air through filters for the purpose of sampling the atmosphere for

radioactive particulates and radioiodines. These samplers are mobile and

may be used at most parts of the plant.

7. Approved dosimeters are used in evaluating the exposure to personnel working at the site.

Fixed and Laboratory Instrumentation

In addition to the portable health physics instrumentation available, there are

a number of fixed and laboratory instruments which are used to assess or

control the spread of radioactivity throughout the facility.

1. Gamma or beta sensitive portal monitors are located in the guardhouse and several entrances to the controlled areas and monitor all outgoing

personnel for radioactive contamination.

2. Personal friskers are located at key places within the facility, and are used by facility personnel to detect surface contamination on clothing, skin, etc.

3. Dosimeter readers, which contain the equipment for measuring the dose received by personal dosimeters. These instruments are located in an

off-site dosimeter processing facility under contract with Vermont Yankee.

4. Multi-channel gamma spectrometer, which consists of a NaI, GeLi, or HpGe crystal, and analyzer circuits necessary for the identification of

individual isotopes by gamma ray energy.

5. Laboratory alpha and beta-gamma counters, which are used for measuring low levels of radioactivity in specially prepared samples such as smears, air

particulate sample filters, etc.

VYNPS DSAR Revision 0 4.0-9 of 54

6. Body-burden counters, which are used to assess internal contamination from both natural sources and from inhaled/absorbed radioactive gases or

particulates. 4.4 RADIATION PROTECTION 4.4.1 Health Physics All employees of Vermont Yankee are given training in radiological safety and

in the requirements for working in the plant.

Administrative controls are established to assure that all procedures and

requirements relating to radiation protection are followed by all station

personnel. These procedures include a radiation work permit system. All work

on systems or in locations where exposure to radiation or radioactive materials

is expected to approach prescribed limits, requires an appropriate radiation

work permit before work can begin. The radiological hazards associated with

the job are determined and evaluated prior to issuing the permit.

4.4.1.1 Personnel Monitoring Systems

Personnel monitoring equipment is assigned to Vermont Yankee personnel by the

Radiation Protection Department. Personnel monitoring equipment is also

available on a day-to-day basis for visitors not assigned to the station that

enter radiation control areas. Records of radiation exposure history and

current occupational exposure are maintained by the Radiation Protection

Department for each individual issued personnel monitoring equipment.

4.4.1.2 Personnel Protective Equipment

Special protective clothing and respiratory equipment are furnished and worn as

necessary to protect personnel from radioactive contamination.

4.4.1.3 Change Area and Shower Facilities

A change area is provided where personnel may obtain clean protective clothing

required for station work. Temporary change areas are provided when required.

Decontamination shower facilities are maintained on-site to assist in timely

personnel decontamination. Monitoring equipment is used to assess the

effectiveness of personnel decontamination efforts.

4.4.1.4 Access Control

To prevent inadvertent access to high radiation areas, warning signs, audible VYNPS DSAR Revision 0 4.0-10 of 54 and visual indicators, barricades and locked doors are used as necessary.

Procedures are also written to control access to high radiation areas.

4.4.1.5 Laboratory Facilities

The facility includes a laboratory with adequate facilities and equipment for

detecting, analyzing, and measuring radioactivity and for evaluating any

radiological problem that may be anticipated. Counting equipment, such as a

multichannel analyzer, liquid scintillation, G-M and proportional counters, and

scalars, are provided in an appropriately designed counting room.

Environmental sample analyses are conducted by outside laboratories.

4.4.1.6 Bioassay Program

In vivo bioassay counting equipment is available for quantitative and

qualitative analysis of possible internal deposition of radioactive

contaminants. Consulting laboratory services are used as backup and support

for this program. Appropriate bioassay (urine and fecal) samples are

collected, as necessary, from personnel who work in control areas as an aid in

the evaluation of internal exposure.

4.4.2 Radioactive Materials Safety Program

All Vermont Yankee personnel who work in controlled areas are given training

in radiological safety. Training Program content is specified in appropriate

training department procedures.

Additionally, those personnel in the Radiation Protection Department whose job

entails the handling of sealed and unsealed sources are given departmental

training.

Other departmental procedures detail methods of leak testing sealed sources and

receipt, handling, and storage of radioactive materials. A general calibration

procedure outlines specific techniques for the safe and expeditious handling of

all calibration sources.

Accountability of sources is maintained in inventory records that are updated

semi-annually. Accessibility control is achieved through locked storage, securing the source in place to prevent unauthorized removal, or continuous

surveillance by authorized personnel.

VYNPS DSAR Revision 0 4.0-11 of 54 Accountability of sources that are exempt from leak testing required by the TRM, but exceed the limits for licensable quantities of radioactive material

specified in Title 10, Code of Federal Regulations, is maintained in inventory

records that are updated annually. All sources of licensable quantity that are

not in use are kept in suitably shielded containers when it is necessary to

minimize personal radiation exposure. All sources of licensable quantity are

kept under the control of authorized personnel when in use.

This system of procedures, training, access control, and accountability is

periodically audited by the Vermont Yankee Quality Assurance Department and/or

one or more contracted service organization(s), collectively defined as the

Quality Assurance Department, as its authorized agent for provision of certain

quality assurance and related support services. Through this mechanism, compliance with applicable regulations is assured.

4.4.2.1 Facilities and Equipment

Station laboratory facilities and monitoring equipment are discussed in DSAR

Sections 4.3 and 4.4.1.5.

4.4.2.2 Personnel and Procedures

Implementation of the Vermont Yankee radiation protection program, including

source, special, and byproduct material safety, is accomplished by Radiation

Protection Department personnel. The qualifications of these personnel in

radioactive materials safety stem from formal and informal training and from

applied experience in the radiation protection field. Specific training of

Radiation Protection personnel in the safe handling of radioactive materials is

covered by a department training program.

VYNPS DSAR Revision 0 4.0-12 of 54 4.4.2.3 Required Materials All byproduct, source, and special nuclear materials used as reactor fuel, sealed neutron source for reactor startups, sealed sources for calibration of

reactor instruments, and radioactive monitoring equipment and fission detectors

are possessed in the amounts required for relevant use. All byproduct material

consisting of mixed fission products and corrosion products in the form of

contamination affixed to equipment used for reactor system repair, maintenance, testing, and/or surveillance may be received, possessed or used in amounts as

required without restriction to chemical or physical form.

With the permanent defueled condition of Vermont Yankee, fission, corrosion, and activation products from operation are no longer produced. The radioactive

inventory that remains is primarily attributable to sealed radioactive sources, activated reactor components, nuclear instrumentation, structural materials and

residual radioactivity. The accumulation of small amounts of solid waste as

contaminated materials may easily be controlled.

4.5 LIQUID WASTE MANAGEMENT SYSTEMS

4.5.1 Equipment and Floor Drainage Systems

4.5.1.1 Objective

The objective of the various equipment and floor drainage systems is to remove

all waste fluids from their points of origin in a controlled effective manner

and to deliver them to a suitable disposal system. Radioactive drain

collection is arranged to minimize radioactive exposure to operating personnel

and to prevent uncontrolled leakages to the environs.

4.5.1.2 Design Basis

Equipment and floor drainage systems shall operate satisfactorily and create no

danger to the health and safety of the general public. These systems shall be

designed and installed to guard against fouling, deposit of solids, and

clogging. Sumps and pumps shall be provided to preclude leakage accumulation

from preventing operation of required equipment. Nonradioactive drainage

systems shall be arranged to assure that no infiltration of radioactive waste

will occur.

VYNPS DSAR Revision 0 4.0-13 of 54 Fluids from radioactive and potentially radioactive drains will be collected, sampled, treated, stored, and/or analyzed prior to disposal in accordance with

10CFR20. Nonradioactive equipment and floor drains empty into the Storm Sewer

System and then discharge into the Circulating Water System piping at the

discharge structure or directly to the Connecticut River at the North Storm

Drain Outfall.

4.5.1.3 Description

4.5.1.3.1 General

The six basic drainage systems are:

1. Radioactive equipment drainage systems

2. Radioactive floor drainage systems

3. Radioactive liquid chemical drainage systems

4. Oil drainage systems

5. Nonradioactive water drainage systems

6. Sanitary drainage systems

The first four systems handle fluid wastes which are radioactive or potentially

radioactive. The last two systems handle fluid wastes originating in areas

which are not radioactive or potentially radioactive. Radioactive wastes are

pumped or drained to the Radwaste System. Nonradioactive wastes are drained to

either the Storm Sewer Drainage System or Sanitary Disposal System.

Radioactive drainage piping is sloped 1/4 inch per foot, and concrete floors

are pitched a minimum of 1/8 inch per foot wherever possible to remove

radioactive wastes as quickly as possible.

Accessible cleanouts are provided at each horizontal change of direction

greater than 45 degrees. Base cleanouts are provided at the base of each stack

approximately 12 inches above the finished floor. In the event a drainage line

becomes stopped or clogged, it can be quickly cleaned out.

VYNPS DSAR Revision 0 4.0-14 of 54 With the exception of that in the nonradioactive waste drainage systems and sanitary drainage systems, all drainage piping is carbon steel Schedule 80, except oil drainage piping, which is carbon steel Schedule 40, and a portion of

condensate drainage piping from the drywell cooling units, which is type 304

stainless steel, Schedule 40. Material used is ASTM A-106, Grade "B". Joints

are welded construction without backing rings. Concrete embedded piping

conforms to the USAS Code B31.1, Sections 1 and 6 for pressure piping. A

portion of condensate drainage piping from the drywell cooling unit is

stainless steel Schedule 80, ASTM A358, Grade TP304.

The Chemistry Laboratory and health physics detergent waste drains are a common

above ground polypropelene lined carbon steel flanged pipe.

Above ground drainage piping used for the Sanitary and Nonradioactive Water

Drainage System is galvanized steel Schedule 40 and galvanized cast iron

drainage fittings. Piping and fittings installed below ground are extra heavy

cast iron.

Vent piping installed above ground is galvanized steel Schedule 40 with

galvanized malleable iron fittings. Piping and fittings installed below ground

are extra heavy cast iron.

All fixtures in the health physics work area, the chemical laboratory, and

fixtures discharging into the Sanitary Drainage System are vented. Each

fixture trap is protected against siphonage and back pressure. The individual

vents collect in a main vent header and terminate full size above the roof.

The radioactive equipment drainage systems receive clean radioactive waste

which is processed and reused. These radioactive systems receive equipment

leak-offs and drains only from equipment handling radioactive liquids.

Radioactive liquids are routed to an equipment drain sump in a closed system

and then pumped to the Radwaste System waste collector tank for future

filtering and demineralization before returning to the Condensate System.

Radioactive or potentially radioactive floor drains are routed to a floor drain

sump in open-ended lines and then pumped to the Radwaste System floor drain

collector from where they are processed and reused. Nonradioactive floor

drains are routed directly to the Storm Sewer System.

Radioactive equipment drains are connected directly to the component serviced

to preclude the possibility of spillage.

VYNPS DSAR Revision 0 4.0-15 of 54 4.5.1.3.2 Radioactive Equipment Drainage Systems Drywell Equipment Drainage Systems

Equipment drains are provided for various components in the drywell and these

lines are run directly to a 500-gallon equipment drain sump. The sump is

provided with two 50 gpm pumps and a number of level switches. A sump pump

will start automatically upon the liquid reaching a pre-set high level and will

trip automatically upon the liquid being lowered to a pre-set low level. A

second sump pump starts and an alarm sounds in the Control Room upon the liquid

reaching a high-high level. Two sump pumps are provided to improve

reliability. An alternator is provided to ensure equal wear on each pump.

Remote operating capability is provided for this system. The common discharge

pipe from the two sump pumps runs through a containment penetration and has two

air-operated valves outside the containment wall. A relief valve provides

overpressure protection of the penetration and connected piping.

Reactor Building Equipment Drainage System

Various equipment drainage in the Reactor Building is piped directly to one of

the two 1000-gallon equipment drain sumps. Each sump (one on the north side

and one on the south side of the Reactor Building) is provided with two 50 gpm

pumps, which discharge to the waste collector tank in the Radwaste Building.

Each sump is provided with level switches used for automatic pump control and

sump high-high level alarm. Pump control switches are located on the radwaste

control panel.

Turbine Building Equipment Drain System

One 1000-gallon sump, located in the feedwater heater area of the Turbine

Building, is provided to collect various equipment drainage. The sump contains

two 50 gpm pumps that discharge to the waste collector tank in the Radwaste

Building.

Sump level switches are used to operate the pumps automatically and provide a

high-high level alarm.

VYNPS DSAR Revision 0 4.0-16 of 54 Radwaste Building Equipment Drain System Various radwaste pumps seal leakage and radwaste tanks, drains, and overflows

are piped directly to one 1000-gallon sump. One 50 gpm sump pump is controlled

automatically by sump level switches and discharges to the waste collector

tank. A sump high level alarm annunciates on the radwaste control panel and in

the Control Room.

4.5.1.3.3 Radioactive Floor Drainage Systems

Drywell Floor Drainage System

The Drywell Floor Drain System collects and disposes of leakage from various

systems and components. Remote operating capability is provided for this

system.

Reactor Building-Floor Drainage System

The Reactor Building Floor Drainage System collects drainage into two

1000-gallon sumps. Each sump (one on the north side and one on the south side

of the Reactor Building) is equipped with two 50 gpm sump pumps which discharge

to the floor drain collector tank.

Turbine Building Floor Drainage System

The Turbine Building Floor Drainage System consists of two 1000-gallon sumps, each provided with two 50 gpm sump pumps which discharge to the floor drain

collector tank. One sump is located in the condenser area and the other is in

the condensate pump area.

Radwaste Building Floor Drainage System

The Radwaste Building contains one 1000-gallon sump utilized to collect various

floor and tank overflow drainage. One 50 gpm sump pump is provided which

discharges to the floor drain collecting tank.

VYNPS DSAR Revision 0 4.0-17 of 54 4.5.1.3.4 Radioactive Liquid Chemical Drainage Systems The Radioactive Liquid Chemical Drainage System consists of radioactive filter sludge piping, radioactive chemical drain piping, and radioactive detergent

waste piping. The system handles drainage of radioactive contaminants and

foreign matter such as sludge, detergents, or chemicals from equipment. The

various systems begin with floor drains, direct connection equipment drains, gutter drains, shower drains, service sinks, and laboratory benches from the

Reactor, Turbine, and Radwaste Buildings. Drainage is collected in waste lines

and discharged into various items of storage and treatment equipment in the

Radwaste Building.

Special showers are provided in the health-physics work area for personnel

decontamination purposes. The drainage from the fixtures in this area is

collected in waste lines and discharged directly into the chemical waste tank

in the Radwaste Building. 4.5.1.3.5 Oil Drainage Systems Oil drain systems outside the restricted are not considered radioactive. Oil drain systems within the restricted area are treated as potentially

contaminated. Drainage from systems and equipment using oil is either

collected in sumps or drains to oil separator manholes. Separated oil is

retained while the oil-free water drains into the Storm Sewer System.

Two oil sumps are provided. One is located beneath the floor of the Reactor

Building, and the second is located in the northwest corner of the Turbine

Building.

Oil drainage not routed and collected in a sump is collected in branch lines

which empty into main lines and discharge directly into oil separator manholes

outside the Turbine Building and Control Room Building. The oil separator

manholes function to separate and retain the oil while discharging oil-free

water into the Storm Water Sewer System which drains either to the discharge

structure or the North Storm Drain Outfall to the Connecticut River.

Oil drainage systems in specific areas, which could have propagated a fire, have been modified. To ensure that spilled fluid is contained within the

respective berm areas, various transformer oil drains have been permanently

plugged.

The oil collected in the oil sumps will be pumped to suitable containers for

disposal using a portable pump. Grab samples can be taken at this time for

radioactive analysis 4.5.1.3.6 Nonradioactive Water Drainage System VYNPS DSAR Revision 0 4.0-18 of 54 The Storm and Nonradioactive Water Drainage System receives rain water, clear

liquid wastes not hotter than 140°F, and drainage from equipment which is

nonradioactive. This drainage is routed separately to the Storm Sewer System.

Heating, ventilation and air conditioning equipment in the Reactor and Turbine

Buildings was considered nonradioactive in the original plant design. Low

levels of tritium have been found in the various drains associated with this

equipment even though modifications to alleviate the condition have been

performed. The levels of contamination have been evaluated and found to be

acceptable for continued discharge to the storm drain system. The condition is

monitored through a surveillance program and reported in the "Annual

Radiological Environmental Operating Report". Funnel type equipment drains and

floor drains serving this equipment are collected in branch lines, empty into

main drain lines, and discharge into the Storm Sewer System. A separate

Nonradioactive Water Drainage System is provided in the Turbine Building for

certain items of equipment. Air handling equipment, certain water pumps on the

basement floor and miscellaneous equipment on the ground floor were considered

nonradioactive in the original plant design. Low levels of activity have been

found in the turbine building clean sump associated with this equipment. The

levels of contamination have been evaluated and found to be acceptable for

continued discharge to the discharge structure. The condition is monitored

through a surveillance program and reported in the "Annual Radiological

Environmental Operating Report". Funnel type equipment drains and floor drains

in these areas are collected in branch lines, empty into main drain lines, and

discharge into the clean equipment and floor drain sump, located below the

basement floor. To improve administrative control over the sources of

radioactive liquid entering this sump, floor drains, which are aligned to this

sump, have been permanently plugged or fitted with removable plugs. In

addition, the drain header from floor and equipment drains in the vicinity of

the demineralized water transfer pumps and the station air compressor receiver

tanks has been cut, capped and valved to allow sampling prior to release.

Other equipment drains are either permanently plugged or go directly to the

sump. Sump pumps are provided to transfer the discharge from the Turbine

Building to the service water discharge.

In addition to the low levels of tritium discussed above, surface run-off from

within the Protected Area carries low levels of particulate activity to the

Storm Sewer System. The low levels of contamination in the Storm Sewer System

have been evaluated to ensure that the calculated maximum release is a

significant percentage less than the total body and critical organ doses

allowed under the routine effluent ALARA objectives of 10CFR50, Appendix I.

VYNPS DSAR Revision 0 4.0-19 of 54 Acid resistant drains and piping are provided in areas where highly concentrated acids are present. Duriron floor drains and piping are provided

to serve the Turbine Building water treatment area. Duriron floor drains and

piping, plus Duriron funnel type equipment drain and piping, are also provided

to serve the Battery Room in the Control Building. 4.5.1.3.7 Sanitary Drainage Systems The Sanitary Drainage System is provided for the convenience, health, and safety of facility personnel. This system receives the domestic sewage from

various fixtures that are water supplied and discharges liquid wastes.

Except for fixtures in the health-physics work area, all water closets, urinals, lavatories, drinking water coolers, service sinks, kitchen units, and

showers in the Turbine and Control Buildings discharge into the Sanitary

Drainage System. Each fixture is trapped and vented, then collected in branch

lines, emptied into main soil lines, and discharged by gravity into the

Sanitary Disposal System.

4.5.1.4 Inspection and Testing

The Equipment and Floor Drainage Systems are normally in operation during all

modes of facility operation. Satisfactory operation is demonstrated

continuously without the need for special testing or inspection.

4.5.2 Liquid Radwaste System

4.5.2.1 Objective

The Liquid Radwaste System collects potentially radioactive liquid wastes, treats and returns the processed radioactive liquid wastes to the facility for

reuse. Any liquid waste which would not be suitable for reuse could be diluted

and discharged from the facility.

4.5.2.2 Design Bases

The Liquid Radwaste System is designed to assure that the release of liquid

radwaste is kept as low as reasonably achievable and is within the annual dose

limits specified in 10CFR20.1301.

The Liquid Radwaste System shall be designed to prevent the inadvertent release

of significant quantities of liquid radioactive material to unrestricted areas

so that resulting radiation exposures are within the limits of 10CFR20.1301.

VYNPS DSAR Revision 0 4.0-20 of 54 4.5.2.3 Description The Liquid Radwaste System collects, processes, stores, and disposes of all

radioactive liquid wastes. The radwaste facility is located in the Radwaste

Building, with the exception of the waste sample tanks, floor drain sample

tank, and waste surge tank, all located outdoors at grade level (see Drawings

G-191151 and G-191152).

Included in the Liquid Radwaste System are the following:

1. Floor and Equipment Drain System for handling potentially radioactive wastes. 2. Tanks, piping, pumps, process equipment, instrumentation, and auxiliaries necessary to collect, process, store and dispose of potentially radioactive

wastes. Equipment is selected, arranged, and shielded to permit operation, inspection, and maintenance with acceptable personnel exposures. For example, sumps, pumps, valves, and instruments are located in controlled access areas. Tanks

and processing equipment which can contain large quantities of liquid radwastes

are shielded. The Radwaste System equipment, equipment arrangement, capacities, flow paths, and flow rates are shown in Drawings 5920-644, Sh.2 and

G-191177, Shs. 1 through 4. Operation of the Waste System is essentially

manual start-automatic stop.

This is a batch-type system wherein the wastes are separately collected and

processed based on the most efficient methods. Cross-connections between

subsystems provide additional flexibility for processing of the wastes by

alternate methods. Treated wastes can be: (a) returned to the system for

reuse, (b) diluted and discharged from the facility, or (c) if not suitable for

either reuse or discharge, they receive additional processing. The liquid

radwastes are classified, collected, and treated as high purity, low purity, chemical or detergent wastes. The terms "high" purity and "low" purity refer

to conductivity and not radioactivity.

VYNPS DSAR Revision 0 4.0-21 of 54 The Liquid Radwaste System is operated such that any liquid radioactive waste releases would be minimized. Successful processing of all liquid wastes to

maintain a low release system calls for special plant controls. Detergent and

soap used to clean areas and equipment are kept to a minimum. The majority of

chemical wastes are neutralized and metered slowly into higher purity water for

processing. Low purity water is filtered, and combined with higher purity

water for reprocessing. The combination of low purity, chemical, and detergent

waste into higher purity waste streams allows for reprocessing and plant reuse

and reduces the need to discharge any fraction of the waste stream. Liquid

waste could be discharged from the waste sample tank to the environment through

approved discharge pathways. The maximum concentration of tritium and

dissolved noble gases at the point of discharge will not exceed applicable

limits.

The processing equipment is located within a concrete building to provide

secondary enclosures for the wastes in the event of leaks or overflows. Tanks

and equipment which contain wastes with radioactive concentrations are

shielded. Except where flanges are required for maintenance, all pipe

connections are welded to reduce the probability of leaks. Chemistry

lab/detergent waste piping is lined with plastic and cannot be welded. As a

result, these lines are flanged, and the flanges, located in the switchgear

room, are fitted with Vue-Guards to reveal and collect any leakage to minimize

the potential for flooding. Process lines which penetrate shield walls are

routed to prevent a direct radiation path from the tanks or equipment for which

shielding is required. Control of the Waste System is from a local panel in

the Radwaste Building Control Room.

Therefore, because the radioactivity concentrations in the liquid radwaste

effluent do not exceed the guideline limits of 10CFR20, the Liquid Radwaste

System fulfills the design basis.

4.5.2.3.1 High Purity Wastes

High purity (low conductivity) liquid wastes are collected in the waste

collector tank.

The high purity wastes are processed by filtration and ion exchange through the

waste collector filter or fuel pool and waste demineralizers as required.

After processing, the liquid is pumped to the waste sample tank where it is

sampled and either recycled for additional processing or transferred to the

condensate storage tank for spent fuel pool inventory makeup.

Should discharge be necessary, wastes would be sampled on a batch basis. VYNPS DSAR Revision 0 4.0-22 of 54 Samples from the waste sample tanks are analyzed for water quality and radioactivity. If high purity requirements are met, the contents are

transferred to the condensate storage tank.

If high purity requirements are not met, the liquid wastes are recycled through

the Radwaste System or could be discharged. The high purity requirements are

specified in plant procedures.

Table 4.5.2.5 lists the radionuclide discharge concentrations at original 100%

power operation, assuming an 80% plant capacity factor and a dilution flow of

20,000 gallons per minute. The total annual release values are based on output

from the BWR Gale Code (NUREG-0016). The information in Table 4.5.2.5 is

historical and is being retained to provide bounding values.

It can be seen from Column 5 (Fraction of ECL w) that the concentration for each radionuclide at the point of discharge is several orders of magnitude below

limits established in 10CFR20 for release of effluents to unrestricted areas.

The design of the Radwaste Treatment System is therefore consistent with the

policy that any radioactive effluents would be reduced to the lowest reasonably

achievable level.

Liquid effluents discharged from the plant enter a 30-inch dilution water line

which terminates in a diffuser at the head end of the aerating apron at the

discharge structure. The effluent enters the Vernon Pond at the downstream end

of the aerating apron.

4.5.2.3.2 Low Purity Wastes

Low purity (high conductivity) liquid wastes which are collected in the floor

drain collector tank are from the following sources:

1. Drywell floor drains 2. Reactor Building floor drains

3. Radwaste Building floor drains

4. Turbine Building floor drains

These wastes generally have low concentrations of radioactive impurities, and

processing consists of filtration and a combination with the high purity waste

in the waste collector tank, with subsequent processing, as high purity waste.

VYNPS DSAR Revision 0 4.0-23 of 54 Operation of the Liquid Radwaste System is such that all liquid wastes will be processed and reused without having to discharge for total system volume

control, or water purity constraints. For the purpose of analyzing future

radiological impacts during the plant's life, it is assumed that 1% of the

combined processed stream treated each year would be discharged from the

facility. Table 4.5.2.1 indicates the radioactive liquid waste sources, flow

rates, expected activities, holdup times, decontamination factors, and assumed

fraction of waste discharged from the Liquid Waste Processing System at

original 100% power operation. The information in Table 4.5.2.1 is historical

and is being retained to provide bounding values. Table 4.5.2.2 lists the

capacity of all major tanks in the Liquid Radwaste System. The plant operating

parameters and design information provides the necessary inputs for the

calculation of potential radioactive source terms by the Nuclear Regulatory

Commission's BWR Gale Computer Code (NUREG-0016). Table 4.5.2.3 lists the

calculated liquid source terms for Vermont Yankee at original 100% power

operation. The information in Table 4.5.2.3 is historical and is being

retained to provide bounding values. Based on processing parameters in Tables

4.5.2.1 and 4.5.2.2, Table 4.5.2.4 lists the activity input to the Liquid

Radwaste System at original 100% power operation for all major nuclides. The

information in Table 4.5.2.4 is historical and is being retained to provide

bounding values. The radioactivities listed represent activities prior to

treatment and will be reduced significantly due to decontamination and isotopic

decay while passing through treatment systems.

4.5.2.3.3 Chemical Wastes

Chemical wastes are collected in the chemical waste tank and are from the

following sources:

1. Chemical lab waste

2. Laboratory drains

3. Sample sinks

When the chemical concentrations are low enough, these wastes may be

neutralized and processed by filtration and dilution in the same manner and

with the same equipment as the low purity wastes. When the chemical

concentrations are too high, these wastes may receive additional processing.

VYNPS DSAR Revision 0 4.0-24 of 54 4.5.2.3.4 Detergent Wastes Detergent wastes are collected in the detergent waste tank. These wastes are

primarily from radioactive decontamination solutions which contain detergents. Detergent wastes are of low radioactivity concentration (<10 -5 Ci/cc). Because detergents will foul ion exchange resins, their use is minimized in the plant.

For initial cleanings, little or no detergent is used. The facility uses an

off-site cleaning laundry, thus minimizing the quantity of waste generated.

Detergent wastes are normally dumped to the floor drain collector tank for

processing with low purity waste.

4.5.2.4 Evaluation

The Radwaste Building is classified as a Class II seismic design structure, and

the Waste System is classified as Class II seismic design equipment, since

failure of the structure and/or the equipment will not cause a significant

release of radioactivity.

With the exception of three 10,000 gallon sample tanks and a 35,000-gallon

waste surge tank, the Radwaste System processing equipment and storage tanks

are located in the Radwaste Building. Failure of the building could be

postulated and the failure could conceivably result in damage to storage tanks

within the building. If the contents of all the tanks within the building were

released, and this is extremely unlikely because of the compartment-like

arrangement and the arrangement of shield walls, the liquid waste would

ultimately accumulate in the basement of the building. Considering the volume

with all tanks two-thirds full, and the existing basement floor space, the

accumulation would amount to approximately 18 inches of water. Considering the

low driving head of the liquid waste in the basement of the building and the

distance to the river, it is very unlikely that entrained activity would find a

leakage path to the river. It is possible that some seepage may occur through

the building foundation, but such seepage would be expected to be small in

quantity and would tend to be absorbed in the soil surrounding the foundation.

If the seepage persisted over a long period of time, the soil surrounding the

foundation would not only act as a liquid absorber, but also as a filter.

The outside storage tanks located within approximately 1.5 foot high concrete

dikes, Figure 4.5.2-8, provide a less remote potential for off-site discharge

of activity. Sumps are provided within the diked area to provide for draining

any leakage or rainwater. Although it is virtually impossible to postulate a

condition which would result in the complete discharge of the contents of the

four outside tanks into the river, the consequences of such an occurrence have

been analyzed. VYNPS DSAR Revision 0 4.0-25 of 54 The maximum gross radioactivity in the four outside tanks is limited to

3.2 curies on the basis of an accidental spill from all tanks due to a seismic

event great enough to damage them. Assuming a low river flow of 108 ft 3/sec, a one day period over which the radioactive liquid wastes are diluted in the

river, and consumption of the water by individuals at standard man consumption

rate (3,000 ml/day), the single intake by an individual would not exceed

one-third the yearly intake allowable by 10CFR20 for unidentified radioisotopes (1 x 10-6 µCi/ml). Radwaste liquids are processed on a batch basis. The design of the system

precludes direct discharge of either unprocessed or processed liquids without

first holding them up in the sample tanks where the liquid is analyzed for

activity levels. Procedural controls would be implemented to ensure that the

activity of processed liquid, after dilution, will not exceed the guideline

limits of 10CFR20, prior to liquids being released to the river.

In order to release liquid from the sample tanks to the river, the sample pumps

must be started, valves opened, and the flow controller positioned. In

addition, a dilution water pump must be put into operation prior to discharge

of the processed liquid. An interlock precludes discharge of processed liquid

to the river when dilution water is unavailable.

The process radiation monitor in the discharge line from the sample tanks to

the river is provided to back up the administrative control provided by sample

tank liquid analysis. It provides a warning to the appropriate facility

personnel that the activity of the processed liquid is approaching ten times

the annual average concentration values of Appendix B, Table 2, Column 2, of

10CFR20.1001-2402. When appropriate, facility personnel could take action to

reduce processed liquid flow or terminating flow entirely to assure that the

releases do not exceed the limits for which the facility is licensed.

Sufficient administrative and design control is provided to prevent accidental

releases of liquid effluents from the Radwaste System.

Therefore, the design basis is considered met.

4.5.2.5 Inspection and Testing

The Liquid Radwaste System is normally operating on an "as-required" basis

thereby demonstrating the ability to perform its function without special

testing. VYNPS DSAR Revision 0 4.0-26 of 54 TABLE 4.5.2.1 Vermont Yankee Radioactive Liquid Waste Processing Parameters Waste Stream Input Sources Input Flow Rates (gpd) Fraction of Primary Coolant Activity (pca) Holdup Time (Days) Available Process Decontamination Factors Assumed Fraction of Waste Stream Discharged CollectionProcess Discharge Nuclide 1st Demin (a) 2nd Demin (b) Total DF High Purity Waste 1) Drywell Equip. Drains 3400 1.0 I = 10 10 10 2 2) Reactor Bldg. Equip. Drains 3720 0.01 3) Radwaste Bldg. Equip. Drains 1060 0.01 0.75 0.15 1.11 Cs, Rb = 2 10 20 0.01 4) Turbine Bldg. Equip. Drains 2960 0.01 5) Condensate Phase Sep. 8100 2 x 10 -6 6) Cleanup Phase Sep. 640 0.002 Other Nuclides =10 10 10 2 7) Resin Rinse 5000 0.002 Low Purity Waste 1) Drywell Floor Drains 700 1.0 I = 10 10 10 2 2) Reactor Bldg. Floor Drains 2000 0.01 1.39 0.15 1.11 Cs, Rb = 2 10 20 0.01 3) Radwaste Bldg. Floor Drains 1000 0.01 Other Nuclides =10 10 10 2 4) Turbine Bldg. Floor Drains 2000 0.01 Chemical Waste 1) Chem. Lab. Waste 100 0.02 I = 10 10 10 2 2) Lab. Drains 500 0.02 2.67 0.15 1.11 Cs, Rb = 2 10 20 0.01 3) Personal Shower and Decon. Drains 900 1.4x10 -4 0.44 Other Nuclides =10 10 10 2 Detergent Waste 1) Decon. Drains 900 1.4x10 -4 0.44 0.15 1.11 I = 10 10 10 2 0.01 Cs, Rb = 2 10 20 Other Nuclides =10 10 10 2 (a) Fuel Pool (Powdered Resin) Filter-Demineralizer (b) Radwaste Deep Bed Demineralizer VYNPS DSAR Revision 0 4.0-27 of 54 NOTE: Data contained in this table is based on original 100% power operation, and is retained as historical information. VYNPS DSAR Revision 0 4.0-28 of 54 TABLE 4.5.2.2 Vermont Yankee Liquid Radwaste System Tank Capacities Tank Capacity Per Tank (Gal.) Waste Collection Tank (1) 25,000 Waste Surge Tank (1)35,000 Floor Drain Collection Tank (1)25,000 Chemical Waste Tank (1)4,000 Detergent Waste Tank (1)1,000 Floor Drain Sample Tank (1)10,000 Waste Sample Tank (2)10,000

TABLE 4.5.2.3 Vermont Yankee Liquid Effluents Assumption of Concentration Annual Releases to Discharge Canal in Primary Adjusted Detergent Total Half-Life Coolant High Purity Low Purity Chemical Total Lws. Total Wastes

Nuclide (Days) (Micro Ci/ml) (Curies) (Curies) (Curies) (Curies) (Ci/Yr) (Ci/Yr) (Ci/Yr) VYNPS DSAR Revision 0 4.0-29 of 54 CORROSION AND ACTIVATION PRODUCTS Na24 6.27E-01 8.72E-03 .00130 .00021 .00000 .00151 .01008 .00000 .01000 P32 1.43E+01 2.12E-04 .00010 .00002 .00000 .00012 .00078 .00000 .00078 Cr51 2.78E+01 5.31E-03 .00249 .00052 .00001 .00302 .02014 .00000 .02000 Mn54 3.03E+02 6.38E-05 .00003 .00001 .00000 .00004 .00025 .00000 .00025 Mn56 1.08E-01 3.90E-02 .00004 .00000 .00000 .00004 .00030 .00000 .00030 Fe55 9.49E+02 1.06E-03 .00051 .00011 .00000 .00062 .00415 .00000 .00420 Fe59 4.51E+01 3.19E-05 .00002 .00000 .00000 .00002 .00012 .00000 .00012 Co58 7.10E+01 2.13E-04 .00010 .00002 .00000 .00012 .00082 .00000 .00082 Co60 1.92E+03 4.25E-04 .00020 .00004 .00000 .00025 .00166 .00000 .00170 Cu64 5.35E-01 2.87E-02 .00352 .00054 .00001 .00407 .02713 .00000 .02700 Zn65 2.45E+02 2.13E-04 .00010 .00002 .00000 .00012 .00083 .00000 .00083 Zn69m 5.73E-01 1.92E-03 .00026 .00004 .00000 .00030 .00199 .00000 .00200 Zn69 3.95E-02 .0 .00028 .00004 .00000 .00032 .00214 .00000 .00210 W187 9.95E-01 3.00E-04 .00007 .00001 .00000 .00008 .00054 .00000 .00054 Np239 2.35E+00 7.24E-03 .00254 .00049 .00001 .00304 .02026 .00000 .02000

FISSION PRODUCTS Br83 1.00E-01 2.13E-03 .00000 .00000 .00000 .00000 .00001 .00000 .00001 Sr89 5.21E+01 1.06E-04 .00005 .00001 .00000 .00006 .00041 .00000 .00041 Sr90 1.03E+04 6.38E-06 .00000 .00000 .00000 .00000 .00002 .00000 .00003 Sr91 4.03E-01 3.72E-03 .00030 .00004 .00000 .00034 .00227 .00000 .00230 Y91m 3.47E-02 .0 .00019 .00003 .00000 .00022 .00146 .00000 .00150 Y91 5.90E+01 4.25E-05 .00003 .00001 .00000 .00004 .00025 .00000 .00025 Sr92 1.13E-01 7.86E-03 .00001 .00000 .00000 .00001 .00008 .00000 .00008 Y92 1.47E-01 4.90E-03 .00011 .00001 .00000 .00012 .00082 .00000 .00082 Y93 4.24E-01 3.74E-03 .00033 .00005 .00000 .00037 .00249 .00000 .00250 Zr95 6.52E+01 7.44E-06 .00000 .00000 .00000 .00000 .00003 .00000 .00003 Nb95 3.50E+01 7.43E-06 .00000 .00000 .00000 .00000 .00003 .00000 .00003 Mo99 2.80E+00 2.08E-03 .00077 .00015 .00000 .00092 .00613 .00000 .00610 Tc99m 2.50E-01 1.76E-02 .00117 .00020 .00000 .00138 .00919 .00000 .00920 Ru103 3.95E+01 2.12E-05 .00001 .00000 .00000 .00001 .00008 .00000 .00008 Rh103m 3.95E-02 .0 .00001 .00000 .00000 .00001 .00008 .00000 .00008 Ru105 1.85E-01 1.69E-03 .00002 .00000 .00000 .00002 .00014 .00000 .00014 Rh105m 5.21E-04 .0 .00002 .00000 .00000 .00002 .00014 .00000 .00014 Rh105 1.50E+00 .0 .00007 .00001 .00000 .00008 .00053 .00000 .00053 Ru106 3.66E+02 3.19E-06 .00000 .00000 .00000 .00000 .00001 .00000 .00001 Rh106 3.47E-04 .0 .00000 .00000 .00000 .00000 .00001 .00000 .00001 Te129m 3.40E+01 4.25E-05 .00002 .00000 .00000 .00002 .00016 .00000 .00016 TABLE 4.5.2.3 (Continued)

Vermont Yankee Liquid Effluents

Concentration Annual Releases to Discharge Canal in Primary Adjusted Detergent Total Half-Life Coolant High Purity Low Purity Chemical Total Lws. Total Wastes

Nuclide (Days) (Micro Ci/ml) (Curies) (Curies) (Curies) (Curies) (Ci/Yr) (Ci/Yr) (Ci/Yr) VYNPS DSAR Revision 0 4.0-30 of 54 Te129 4.80E-02 .0 .00001 .00000 .00000 .00002 .00010 .00000 .00010 Te131m 1.25E+00 1.01E-04 .00003 .00000 .00000 .00003 .00021 .00000 .00021 Te131 1.74E-02 .0 .00000 .00000 .00000 .00001 .00004 .00000 .00004 I131 8.05E+00 5.24E-03 .00230 .00048 .00001 .00278 .01858 .00000 .01900 Te132 3.25E+00 1.04E-05 .00000 .00000 .00000 .00000 .00003 .00000 .00003 I132 9.58E-02 2.12E-02 .00002 .00000 .00000 .00002 .00011 .00000 .00011 I133 8.75E-01 1.90E-02 .00392 .00067 .00001 .00460 .03068 .00000 .03100 Cs134 7.50E+02 3.19E-05 .00008 .00002 .00000 .00009 .00062 .00000 .00062 I135 2.80E-01 1.65E-02 .00062 .00008 .00000 .00070 .00468 .00000 .00470 Cs136 1.80E+01 2.11E-05 .00005 .00001 .00000 .00006 .00039 .00000 .00039 Cs137 1.10E+04 1.45E-05 .00018 .00004 .00000 .00022 .00145 .00000 .00150 Ba137m 1.77E-03 .0 .00017 .00004 .00000 .00020 .00136 .00000 .00140 Ba140 1.28E+01 4.23E-04 .00019 .00004 .00000 .00023 .00155 .00000 .00160 Tritium Release 4 Curies per year NOTE: Data contained in this table is based on original 100% power operation, and is retained as historical information. La140 1.67E+00 .0 .00007 .00002 .00000 .00009 .00059 .00000 .00059 La141 1.62E-01 .0 .00000 .00000 .00000 .00000 .00003 .00000 .00003 Ce141 3.25E+01 3.18E-05 .00002 .00000 .00000 .00002 .00013 .00000 .00013 Ce143 1.38E+00 3.05E-05 .00001 .00000 .00000 .00001 .00007 .00000 .00007 Pr143 1.57E+01 4.23E-05 .00002 .00000 .00000 .00002 .00016 .00000 .00016 Ce144 2.84E+02 3.19E-06 .00000 .00000 .00000 .00000 .00001 .00000 .00001 Pr144 1.20E-02 .0 .00000 .00000 .00000 .00000 .00001 .00000 .00001 Nd147 1.11E+01 3.17E-06 .00000 .00000 .00000 .00000 .00001 .00000 .00001 All Others 1.86E-01 .00000 .00000 .00000 .00001 .00004 0.0 .00004 Total (Except Tritium) 3.86E-01 02235 .00403 . .00005 02644 .17644 .00000 .18000 VYNPS DSAR Revision 0 4.0-31 of 54 TABLE 4.5.2.4 Activity Input to Liquid Radwaste System (Ci/yr)

Nuclide

µci/ml  PCA Drywell Equip. Drains  Reactor Bldg. Equip. Drain  Radwaste Bldg. Equip. Drain  Turbine Bldg Equip. Drain Condensate Phase Sep. 

Cleanup Phase Sep. Resin Rinse Drywell Floor Drains Reactor Bldg. Floor Drains Radwaste Bldg. Floor Drains Turbine Bldg. Floor Drains Chem Lab. Waste Lab. Drains Personal Shower and Decon. Drains Na24 8.72E-03 3.3E+01 3.6E-01 1.02E-01 2.8E-01 1.6E-04 1.2E-02 9.6E-02 6.7E+00 1.9E-01 9.6E-02 1.9E-01 1.9E-02 9.6E-02 1.2E-03 P32 2.12E-04 7.9E-01 8.7E-03 2.5E-03 6.9E-03 3.8E-06 3.0E-04 2.3E-03 1.6E-01 4.7E-03 2.3E-03 4.7E-03 4.7E-04 2.3E-03 2.9E-05 Cr51 5.31E-03 2.0E+01 2.2E-01 6.2E-02 1.7E-01 9.6E-05 7.5E-03 5.9E-02 4.1E+00 1.2E-01 5.9E-02 1.2E-01 1.2E-02 5.9E-02 7.4E-04 Mn54 6.38E-05 2.4E-01 2.6E-03 7.5E-04 2.1E-03 1.1E-06 9.0E-05 7.0E-04 4.9E-02 1.4E-03 7.3E-04 1.4E-03 1.4E-04 7.0E-04 8.9E-06 Mn56 3.90E-02 1.47E+0 2 1.6E+00 4.6E-01 1.3E+00 7.0E-04 5.5E-02 4.3E-01 3.0E+01 8.6E-01 4.3E-01 8.6E-01 8.6E-02 4.3E-01 5.4E-03 Fe55 1.06E-03 3.9E+00 4.4E-02 1.2E-02 3.5E-02 1.9E-05 1.5E-03 1.2E-02 8.2E-01 2.3E-02 1.2E-02 2.3E-02 2.3E-02 1.2E-02 1.5E-04 Fe59 3.19E-05 1.2E-01 1.3E-03 3.7E-04 1.0E-03 5.7E-07 4.5E-05 3.5E-04 2.5E-02 7.0E-04 3.5E-04 7.0E-04 7.0E-05 3.5E-04 4.4E-06 Co58 2.13E-04 8.0E-01 8.8E-03 2.5E-03 7.0E-03 3.8E-06 3.0E-04 2.4E-03 1.6E-01 4.7E-03 2.4E-03 4.7E-03 4.7E-04 2.4E-03 3.0E-05 Co60 4.25E-04 1.6E+00 1.7E-02 5.0E-03 1.4E-02 7.7E-06 6.0E-04 4.7E-03 3.3E-01 9.4E-03 4.7E-03 9.4E-03 9.4E-04 4.7E-03 5.9E-05 Cu64 2.87E-02 1.1E+02 1.2E+00 3.4E-01 9.4E-01 5.2E-04 4.0E-02 3.2E-01 2.2E+01 6.3E-01 3.2E-01 6.3E-01 6.3E-02 3.2E-01 4.0E-03 Zn65 2.13E-04 8.0E-01 8.8E-03 2.5E-03 7.0E-03 3.8E-06 3.0E-04 2.4E-03 1.6E-01 4.7E-03 2.4E-03 4.7E-03 4.7E-01 2.4E-03 3.0E-05 Zn69m 1.92E-03 7.2E+00 7.9E-02 2.2E-02 6.3E-02 3.5E-05 2.7E-03 2.1E-02 1.5E+00 4.2E-02 2.1E-02 4.2E-02 4.2E-03 2.1E-02 2.7E-04 W198 3.00E-04 1.1E+00 1.2E-02 2.5E-03 9.8E-03 5.4E-06 4.2E-04 3.3E-03 2.3E-01 6.6E-03 3.3E-03 6.6E-03 6.6E-04 3.3E-03 4.2E-05 Np239 7.24E-03 2.7E+01 3.0E-01 8.5E-02 2.4E-01 1.3E-04 1.0E-02 8.0E-02 5.6E+00 1.6E-01 8.0E-02 1.6E-01 1.6E-02 8.0E-02 1.0E-03 Br83 2.13E-03 8.0E+00 8.8E-02 2.5E-02 7.0E-02 3.8E-05 3.0E-03 2.4E-02 1.6E+00 4.7E-02 2.4E-02 4.7E-02 4.7E-03 2.4E-02 3.0E-04 Sr89 1.06E-04 3.9E+00 4.4E-03 1.2E-03 3.5E-03 1.9E-06 1.5E-04 1.2E-03 8.2E-02 2.3E-03 1.2E-03 2.3E-03 2.3E-04 1.2E-03 1.5E-05 Sr90 6.38E-06 2.4E-02 2.6E-04 7.5E-05 2.1E-04 1.1E-07 9.0E-06 7.0E-05 4.9E-03 1.4E-04 7.0E-05 1.4E-04 1.4E-05 7.0E-05 8.9E-07 Sr91 3.72E-03 1.4E+01 1.5E-01 4.4E-02 1.2E-01 6.7E-05 5.2E-03 4.1E-02 2.9E+00 8.2E-02 4.1E-02 8.2E-02 8.2E-03 4.1E-02 5.2E-04

Y91 4.25E-05 1.6E-01 1.7E-03 5.0E-04 1.4E-03 7.7E-07 6.0E-05 4.7E-04 3.3E-02 9.4E-04 4.7E-04 9.4E-04 9.4E-05 4.7E-04 5.9E-03 Sr92 7.86E-03 2.9E+01 3.2E-01 9.2E-02 2.6E-01 1.4E-04 1.1E-02 8.7E-02 6.1E+00 1.7E-01 8.7E-02 1.7E-01 1.7E-02 8.7E-02 1.1E-03 Y92 4.90E-03 1.8E+01 2.0E-01 5.7E-02 1.6E-01 8.8E-05 6.9E-03 5.4E-02 3.8E+00 1.1E-01 5.4E-02 1.1E-01 1.1E-02 5.4E-02 6.8E-04 Y93 3.74E-03 1.4E+01 1.5E-01 4.4E-02 1.2E-01 6.7E-05 5.3E-03 4.1E-02 2.9E+00 8.3E-02 4.1E-02 8.3E-02 8.3E-03 4.1E-02 5.2E-04 Zr95 7.44E-06 2.8E-02 3.1E-04 8.7E-05 2.4E-04 1.3E-07 1.0E-05 8.2E-05 5.8E-03 1.6E-04 8.2E-05 1.6E-04 1.6E-05 8.2E-05 1.0E-06 Nb95 7.43E-06 2.8E-01 3.1E-04 8.7E-05 2.4E-04 1.3E-07 1.0E-05 8.2E-05 5.8E-03 1.6E-04 8.2E-05 1.6E-04 1.6E-05 8.2E-05 1.0E-06 Mo99 2.08E-03 7.8E+00 8.6E-02 2.4E-02 6.8E-02 3.7E-01 2.9E-03 2.3E-02 1.6E+00 4.6E-02 2.3E-02 4.6E-02 4.6E-03 2.3E-02 2.9E-04 Tc99m 1.76E-02 6.6E+01 7.2E-01 2.1E-01 5.8E-01 3.2E-04 2.5E-02 1.9E-01 1.4E+01 3.9E-01 1.9E-01 3.9E-01 3.9E-02 1.9E-01 2.4E-03 Ru103 2.12E-05 8.0E-02 8.7E-04 2.5E-04 6.9E-04 3.8E-07 3.0E-05 2.3E-04 1.6E-02 4.7E-04 2.3E-04 4.7E-04 4.7E-05 2.3E-04 2.9E-06 Ru105 1.69E-03 6.3E+00 6.9E-02 2.0E-02 5.5E-02 3.0E-05 2.4E-03 1.9E-02 1.3E+00 3.7E-02 1.9E-02 3.7E-02 3.7E-03 1.9E-02 2.3E-04 Ru106 3.19E-06 1.2E-02 1.3E-04 3.7E-05 1.0E-04 5.7E-08 4.5E-06 3.5E-05 2.5E-03 7.0E-05 3.5E-05 7.0E-05 7.0E-06 3.5E-05 4.4E-07

VYNPS DSAR Revision 0 4.0-32 of 54 TABLE 4.5.2.4 (Continued) Activity Input to Liquid Radwaste System (Ci/yr)

Nuclide

µci/ml PCA  

Drywell Equip. Drains Reactor Bldg. Equip. Drain Radwaste Bldg. Equip. Drain Turbine Bldg. Equip Drain

Condensate

Phase Sep.

Cleanup Phase Sep.

Resin Rinse

Drywell Floor Drains Reactor Bldg. Floor Drains Radwaste Bldg. Floor Drains Turbine Bldg. Floor Drains

Chem. Lab. Waste

Lab. Drains Personal Shower and Decon. Drains Te129m 4.25E-05 1.6E-01 1.7E-03 5.0E-04 1.4E-03 7.7E-07 6.0E-06 4.7E-04 3.3E-02 9.4E-04 4.7E-04 9.4E-04 9.4E-05 4.7E-04 5.9E-06 Te131m 1.01E-04 3.8E-01 4.2E-03 1.2E-03 3.3E-03 1.8E-06 1.4E-04 1.1E-03 7.8E-02 2.2E-03 1.1E-03 2.2E-03 2.2E-04 1.1E-03 1.4E-05 I131 5.24E-03 2.0E+00 2.2E-01 6.1E-02 1.7E-01 9.4E-05 7.4E-03 5.8E-02 4.1E+00 1.2E-01 5.8E-02 1.2E-01 1.2E-02 5.8E-02 7.3E-04 Te132 1.04E-05 3.9E-02 4.3E-04 1.2E-04 3.4E-04 1.9E-07 1.5E-05 1.1E-04 8.0E-03 2.3E-04 1.1E-04 2.3E-04 2.3E-05 1.1E-04 1.4E-06 I132 2.12E-02 8.0E+01 8.7E-01 2.5E-01 6.9E-01 3.8E-04 3.0E-02 3.0E-02 1.6E+01 4.7E-01 3.0E-02 4.7E-01 4.7E-02 3.0E-02 2.9E-03 I133 1.90E-02 7.1E+01 7.8E-01 2.2E-01 6.2E-01 3.4E-04 2.7E-02 2.1E-01 1.5E+01 4.2E-01 2.1E-01 4.2E-01 4.2E-02 2.1E-01 2.6E-03 Cs134 3.19E-05 1.2E-01 1.3E-03 3.7E-04 1.0E-03 5.7E-07 4.5E-05 3.5E-04 2.5E-02 7.0E-04 3.5E-04 7.0E-04 7.0E-05 3.5E-04 4.4E-06 I135 1.65E-02 6.2E+01 6.8E-01 1.9E-01 5.4E-01 3.0E-04 2.3E-02 1.8E-01 1.3E+01 3.6E-01 1.8E-01 3.6E-01 3.6E-02 1.8E-01 2.3E-03 Cs136 2.11E-05 7.9E-02 8.7E-04 2.5E-04 6.9E-04 3.8E-07 3.0E-05 2.3E-04 1.6E-02 4.7E-04 2.3E-04 4.7E-04 4.7E-05 2.3E-04 2.9E-06 Cs137 7.45E-05 2.8E-01 3.1E-03 8.7E-04 2.4E-03 1.3E-06 1.1E-04 8.2E-04 5.8E-02 1.6E-03 8.2E-04 1.6E-03 1.6E-04 8.2E-04 1.0E-05 Ba140 4.23E-04 1.59E+00 1.7E-02 4.9E-03 1.4E-02 7.6E-06 6.0E-04 4.7E-03 3.3E-01 9.3E-03 4.7E-03 9.3E-03 9.3E-04 4.7E-03 5.9E-05 Ce141 3.18E-05 1.2E-01 1.3E-03 3.7E-04 1.0E-03 5.7E-07 4.5E-05 3.5E-04 2.5E-02 7.0E-04 3.5E-04 7.0E-04 7.0E-05 3.5E-04 4.4E-06 Ce143 3.05E-05 1.1E-01 1.3E-03 3.6E-04 1.0E-03 5.5E-07 4.3E-05 3.4E-04 2.4E-02 6.7E-04 3.4E-04 6.7E-04 6.7E-05 3.4E-04 4.2E-06 Pr143 4.23E-05 1.6E-01 1.7E-03 4.9E-04 1.4E-03 7.6E-07 6.0E-05 4.7E-04 2.3E-02 9.3E-04 4.7E-04 9.3E-04 9.3E-05 4.7E-04 5.9E-06 Ce144 3.19E-06 1.2E-02 1.3E-04 3.7E-05 1.0E-04 5.7E-08 4.5E-06 3.5E-05 2.5E-03 7.0E-05 3.5E-05 7.0E-05 7.0E-06 3.5E-05 4.4E-07 Nd147 3.17E-06 1.2E-02 1.3E-04 3.7E-05 1.0E-04 5.7E-08 4.5E-06 3.5E-05 2.5E-03 7.0E-05 3.5E-05 7.0E-05 7.0E-06 3.5E-05 4.4E-07 All Others 1.86E-01 7.0E+02 7.6E+00 2.2E+00 6.1E+00 3.3E-03 2.6E-01 2.1E+00 1.4E+02 4.1E+00 2.1E+00 4.1E+00 4.1E-01 2.1E+00 2.6E-02 NOTE: Data contained in this table is based on original 100% power operation, and is retained as historical information VYNPS DSAR Revision 0 4.0-33 of 54 TABLE 4.5.2.5 Radionuclide Discharge Concentrations Nuclide Total Annual Release (Ci/Yr) Discharge Concentration (µCi/ml) ECL w (µCi/ml) Fraction of ECL Na24 1.0x10 -2 3.1x10-10 5x10-5 6.2x10-6 P32 7.8x10 -4 2.4x10-11 9x10-6 2.7x10-7 Cr51 2.0x10 -2 6.3x10-10 5x10-4 1.3x10-6 Mn54 2.5x10 -4 7.8x10-12 3x10-5 2.6x10-7 Mn56 3.0x10 -4 9.4x10-12 7x10-5 1.3x10-7 Fe55 4.2x10 -3 1.3x10-10 1x10-4 1.3x10-6 Fe59 1.2x10 -4 3.8x10-12 1x10-5 3.8x10-7 Co58 8.2x10 -4 2.6x10-11 2x10-5 1.3x10-6 Co60 1.7x10 -3 5.3x10-11 3x10-6 1.8x10-5 Cu64 2.7x10 -2 8.5x10-10 2x10-4 4.3x10-6 Zn65 8.3x10 -4 2.6x10-11 5x10-6 5.2x10-6 Zn69m 2.0x10 -3 6.3x10-11 6x10-5 1.1x10-6 Np239 2.0x10 -2 6.3x10-10 2x10-5 3.2x10-5 Br83 1.0x10 -5 3.1x10-13 9x10-4 3.4x10-10 Sr89 4.1x10 -4 1.3x10-11 8x10-6 1.6x10-6 Sr90 3.0x10 -5 9.4x10-13 5x10-7 1.9x10-6 Sr91 2.3x10 -3 7.2x10-11 2x10-5 3.6x10-6 Y91 2.5x10 -4 7.8x10-12 8x10-6 9.8x10-7 Sr92 8.0x10 -5 2.5x10-12 4x10-5 6.3x10-8 Y92 8.2x10 -4 2.6x10-11 4x10-5 6.5x10-7 Y93 2.5x10 -3 7.8x10-11 2x10-5 3.9x10-6 Zr95 3.0x10 -5 9.4x10-13 2x10-5 4.7x10-8 Nb95 3.0x10 -5 9.4x10-13 3x10-5 3.1x10-8 Mo99 6.1x10 -3 1.9x10-10 2x10-5 9.5x10-6 Tc99m 9.2x10 -3 2.9x10-10 1x10-3 2.9x10-7 Ru103 8.0x10 -5 2.5x10-12 3x10-5 8.3x10-8 Ru105 1.4x10 -4 4.4x10-12 7x10-5 6.3x10-8 Ru106 1.0x10 -5 3.1x10-13 3x10-6 1.0x10-7 Te129m 1.6x10 -4 5.0x10-12 7x10-6 7.1x10-7 Te131m 2.1x10 -4 6.6x10-12 8x10-6 8.3x10-7 I131 1.9x10 -2 6.0x10-10 1x10-6 6.0x10-4 Te132 3.0x10 -5 9.4x10-13 9x10-6 1.0x10-7 I132 1.1x10 -4 3.4x10-12 1x10-4 3.4x10-8 I133 3.1x10 -2 9.7x10-10 7x10-6 1.4x10-4 Cs134 6.2x10 -4 1.9x10-11 9x10-7 2.1x10-5 I135 4.7x10 -3 1.5x10-10 3x10-5 5.0x10-6 Cs136 3.9x10 -4 1.2x10-11 6x10-6 2.0x10-6 Cs137 1.5x10 -3 4.7x10-11 1x10-6 4.7x10-5 Ba140 1.6x10 -3 5.0x10-11 8x10-6 6.3x10-6 Ce141 1.3x10 -4 4.1x10-12 3x10-5 1.4x10-7 Ce143 7.0x10 -5 2.2x10-12 2x10-5 1.1x10-7 Pr143 1.6x10 -4 5.0x10-12 2x10-5 2.5x10-7 Ce144 1.0x10 -5 3.1x10-13 3x10-6 1.0x10-7 Nd147 1.0x10 -5 3.1x10-13 2x10-5 1.6x10-8 All others 4.0x10 -5 1.3x10-12 1x10-6 1.3x10-6Tritium 4.0 1.3x10 -7 1x10-3 1.3x10-4 NOTE: Data contained in this table is based on original 100% power operation, and is retained as historical information. VYNPS DSAR Revision 0 4.0-34 of 54 Vermont Yankee Defueled Safety Analysis Report Revision 0 Radwaste Area - Plan View Figure 4.5.2-8

VYNPS DSAR Revision 0 4.0-35 of 54 4.6 SOLID WASTE MANAGEMENT 4.6.1 Solid Radwaste System

4.6.1.1 Objective

The Solid Radwaste System collects and processes radioactive solid wastes for

possible temporary on-site storage and off-site shipment for permanent disposal.

4.6.1.2 Design Basis

The Solid Radwaste System shall be designed to package radioactive solid wastes

for ultimate off-site shipment for disposal in accordance with applicable

published regulations.

4.6.1.3 Description

4.6.1.3.1 General

The Solid Radwaste System is a contiguous part of the Liquid Radwaste System and

is an integral part of the Radwaste Building. The system processes wet and dry

solid wastes. Because of physical differences and differences in radioactivity

or contamination levels, various methods are employed for processing and

packaging the solid radwaste. Wet solid wastes are packaged in appropriate

liners or high integrity containers for transportation within licensed shipping

casks. Dry active waste is collected in general design packages for shipment to

a licensed disposal site or a licensed processing facility for volume reduction.

Each type of waste is kept segregated to reduce shielding requirements for

storage.

Table 4.6.1 shows a history of both the wet and dry waste volumes and activity

levels that have been processed for off-site disposal. Subsequent to 1992, this

data is contained in the Radioactive Effluent Release Report.

4.6.1.3.2 Wet Wastes

Wet wastes consist of spent demineralizer resins and filter sludge. These are

pumped from the phase separators or waste sludge tanks as a slurry to disposable

liners preplaced within the licensed transportation casks. The slurry is then

dewatered from within the liner using a remote dewatering system located in the

Cask Room. The Dewatering System is kept in continuous operation as long as the

cask liner is being filled. When the cask liner is full, a high-level trip

recirculates the resin slurry to either the waste collector tank or to one of

the condensate phase separators. VYNPS DSAR Revision 0 4.0-36 of 54 The Dewatering System level instruments indicate in the Radwaste Building

Control Room.

The Dewatering System is accessible for cleaning and maintenance when not being

operated. The Dewatering System and its associated controls are arranged for

remote operation, which is manually initiated.

When feed to the Dewatering System is stopped, the feed piping is flushed in

accordance with plant procedures. External water connections are provided for

cleaning and decontamination.

The radioactive wet wastes are transported in licensed steel/lead casks. The

casks contain disposable steel liners or high integrity containers. The casks

are placed on trolleys and rolled on tracks below the Dewatering System fill

head. The solid wastes are processed through the fill head into the cask liner.

After filling, the liner is closed and the cask is rolled to a decontamination

area in the Radwaste Building where the cask is wiped or washed down to remove

surface contamination. The cask is lifted to a truck for transportation to the

on-site waste storage area or off-site to a waste disposal site. Design and use

of the cask are in accordance with 10CFR71 and 49CFR170-178 regulations of the

Department of Transportation. All resin shipments are via sole-use vehicles.

There are associated high and high-high level alarms which initiate the

following:

1. High level - reposition the three-way V20-422 valve to recirculate resin slurry.

2. High-high level - cessation of feed.

Spent resins from the various filter systems are flushed to the Radwaste

Processing System and normally combined for dewatering through the Dewatering

System. The moisture content of the processed spent resins is less than 1% by

weight.

The principal gamma-emitting radionuclides normally found in the spent resins

include Manganese-54, Cobalt-58 and 60, Cesium-134 and 137, and Zinc-65. The

volume of spent resin and filter sludge is provided in the yearly Radioactive

Effluent Release Report.

VYNPS DSAR Revision 0 4.0-37 of 54 4.6.1.3.3 Dry Wastes Dry wastes consist of air filters, miscellaneous paper, rags, shoe covers, etc., from contaminated areas; contaminated clothing, tools, and equipment parts, which cannot be effectively decontaminated; solid laboratory wastes; used

reactor equipment such as poison curtains, spent control rod blades, fuel

channels and in-core ion chambers; and large pieces of equipment.

The disposition of a particular item of waste is determined by its radiation

level and type, and the availability of disposal space. Because of high

activation and contamination level, used reactor equipment is stored in the fuel

storage pool for sufficient time to obtain optimum radioactive decay before

removal and final disposal. Most solid radwaste such as contaminated clothing, rags, and paper can be handled manually because of low radioactivity or

contamination levels.

Dry Active Waste (DAW) is collected into shipping containers to be sent to an

off-site disposal site or an off-site waste processor for volume reduction.

Table 4.6.1 indicates the volume of compacted dry waste that has been shipped

for disposal between 1985 and 1992. This includes material sent directly from

Vermont Yankee and from various vendors after processing. The dry compacted

waste comprises about 65% of volume of total waste during the time period. The

volume for years subsequent to 1992 is contained in the Radioactive Effluent

Release Report.

The principal radionuclides in the dry active waste are Cesium-134, Cesium-137, Cobalt-60, Iron-55, Manganese-54, and Zinc-65. Other nuclides which are

generally detected in the waste include Chromium-51, Cobalt-58, Barium-Lanthanum-140, Cerium-141, Iron-59, Antomony-124, and Zirconium-95. The

ratios of these isotopes vary. Samples are drawn periodically to determine

current ratios and identify trends.

4.6.1.4 Inspection and Testing

The Solid Radwaste System is normally operated on a regular basis thereby

demonstrating functionality without special testing. VYNPS DSAR Revision 0 4.0-38 of 54 TABLE 4.6.1.1 Solid Radwaste Annual Disposal History Spent Resin and Filter Sludge Dry Processed Trash Year Volume (ft 3) Activity (Ci) Volume (ft 3) Activity (Ci) 1985 3,383 254 9,940 8.9 1986 1,836 196 9,018 16.6 1987 2,892 287 4,968 12.2 1988 2,655 417 3,467 7.7 1989* 171 2 0 0.0 1990* 0 0 0 0.0 1991 7,937 1,568 7,872 40.2 1992 2,391 476 3,238 51.7 Data for subsequent years is contained in the Radioactive Effluent Release

Report pursuant to Technical Specifications.

• Vermont Yankee was denied access to disposal facilities.

VYNPS DSAR Revision 0 4.0-39 of 54 4.7 EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING 4.7.1 Process Radiation Monitoring Instrumentation

A number of radiation monitors and monitoring systems are provided on process

liquid and ventilation lines that may serve as discharge routes for radioactive

materials. The monitors include the following:

• Plant Stack Radiation Monitoring System
• Process Liquid Radiation Monitoring System
• Reactor Building Ventilation Radiation Monitoring System

These systems are described individually in the following paragraphs.

4.7.1.1 Plant Stack Radiation Monitoring System

4.7.1.1.1 Objective

The objective of the Plant Stack Radiation Monitoring System is to

representatively sample, monitor, indicate, and record the radioactivity level

of the station effluent gases being discharged from the plant stack and to alert

personnel in the event radiation levels approach or exceed pre-established

limits.

4.7.1.1.2 Design Basis

1. The Plant Stack Radiation Monitoring System shall provide a clear indication to operations personnel of the current release level of

radioactive materials to the environs.

2. The Plant Stack Radiation Monitoring System shall record the rate of release of radioactive materials to the environs so that determination of

the total amounts of activity release is possible.

4.7.1.1.3 Description

The Plant Stack Radiation Monitoring System is shown on Drawing 5920-3994, and

specifications are given in Table 4.7.1.3. The system consists of three (3)

radiation monitors (Stack Gas I, Stack Gas II, and Stack Gas III).

The primary channel provides for the continuous monitoring of radioactive gas in the plant stack effluent. It also provides filter media to be analyzed in the

plant laboratory by gamma spectroscopy to evaluate long-lived isotopic

composition of particulates and iodine composition of plant stack effluents. VYNPS DSAR Revision 0 4.0-40 of 54 The primary stack monitoring channel consists of four (4) sampling chambers and two (2) radiation monitors (Stack Gas I and II). The monitors observe the

radio-gas activity, and composites of long-lived particulates, iodine, and

tritium can be collected for laboratory analysis.

The sample flow is withdrawn from the plant stack through an isokinetic sample

probe located at elevation 464'-0", approximately 217 feet above the point where

the gases enter the stack. The sample train is branched prior to the point of

measurement. Branch I consists of one I-131 charcoal cartridge filter and one

radio-gas monitor with associated 8 cfm air pump and flow indicator. Branch II

is a duplicate of Branch I with the additional capability to sample gaseous

tritium. The fixed filters and tritium samplers can be changed on a routine

schedule. The plant radiochemistry laboratory analyzes filter media by gamma

spectroscopy to evaluate long-lived isotopic particulate and I-131 composition.

The tritium samplers are analyzed by liquid scintillation spectrometry.

Remote controls for pump motors are located in the station Main Control Room.

The sample flow is directed back to the stack at the completion of the

monitoring process.

A third radio-gas monitor provides indication of high-range discharges of

radioactive gases postulated to occur following design bases accidents. This

monitor is located within the stack at the 264'-0" elevation level.

All other monitoring equipment is located in an enclosure at the base of the

plant stack at grade level (elevation 250'-0"). Facilities for the collection

of air particulates, iodine, and radio-gas grab samples are provided at

elevation 462', several feet downstream of the isokinetic sample probe.

Facilities are also available for sampling prior to the monitoring system at the

base level at the stack.

The three monitors in the primary channel will indicate and alarm in the station

Main Control Room; no control action is provided by this system. Each monitor

is equipped with a sensor, a power supply, a logarithmic rate meter, and a trip

unit. The readout of each normal range monitor is continuously recorded in the

station Main Control Room.

Each trip unit has an adjustable trip and also signals loss of high voltage

power supply, low flow, and loss of input signal.

VYNPS DSAR Revision 0 4.0-41 of 54 4.7.1.1.4 Inspection and Testing Each monitor is inspected according to surveillance procedures and is tested in

accordance with the Off-Site Dose Calculation Manual. Stack Gas I and II are

calibrated and functionally tested per the Off-Site Dose Calculation Manual.

4.7.1.2 Process Liquid Radiation Monitoring System

4.7.1.2.1 Objective

On process streams that normally discharge to the environs, process liquid

radiation monitors are provided to indicate when pre-established limits for the

normal release of radioactive material to the environs are exceeded.

On process streams that do not discharge to the environs, process liquid

monitors are provided to indicate process system malfunctions by detecting the

accumulation of radioactive material in a normally uncontaminated system.

4.7.1.2.2 Design Basis

Process liquid radiation monitors located in streams that normally discharge to

the environs shall provide a clear indication whenever the radioactivity level

in the stream reaches or exceeds pre-established limits for the discharge of

radioactive material to the environs.

Process liquid radiation monitors located in streams that do not discharge to

the environs shall provide a clear indication whenever the radioactivity level

in the stream reaches or exceeds a pre-established limit.

4.7.1.2.3 Description

The process liquid radiation monitoring system instrumentation is shown on

Drawing 5920-00526, and characteristics are given in Table 4.7.1.1. One channel

monitors the discharge from the Liquid Radwaste System, and another channel

monitors service water discharge. All channels are connected to the +/-24 V dc

power buses.

Each channel has a scintillation detector, a radiation monitor, and strip chart

recorder. A representative sample may be continuously extracted from either of

two possible points of discharge and monitored for radioactivity. A radwaste

system recorder is located in the Radwaste Building Control Room. All monitors

and the other recorders are located in the Main Control Room.

Each channel has an upscale trip to indicate high radiation level and one VYNPS DSAR Revision 0 4.0-42 of 54 downscale trip to indicate instrument trouble. The trips give an alarm but no control action.

The Liquid Radwaste System provides for collection of waste liquids through

various drainage systems. Because of high conductivity, some of the waste

liquids may not be economically purified by demineralization. Consequently, some liquid containing radioactivity may eventually be discharged from the

system. The process liquid monitoring channel on the Liquid Radwaste System

discharge indicates discharge radiation levels.

The Service Water System serves as the heat sink for the Standby Fuel Pool

Cooling System. The water circulated through the heat exchangers by the Standby

Fuel Pool Cooling System will be spent fuel pool water, which may have a

significant activity level. Changes in the normal radiation level in the

service water discharge could indicate leakage in the Standby Fuel Pool Cooling

heat exchangers.

The environmental and power supply design conditions are given in Table 4.7.1.2.

The process liquid radiation monitors for radwaste and service water discharges

have radiation detection and monitoring characteristics sufficient to inform

facility personnel whenever radiation levels in the discharges rise above preset

limits.

4.7.1.2.4 Inspection and Testing

All alarm trip circuits can be tested by using test signals or portable gamma

sources.

Surveillances are performed as required by the Off-Site Dose Calculation Manual.

4.7.1.3 Reactor Building Ventilation Radiation Monitoring System

4.7.1.3.1 Objective

The objective of the Reactor Building Ventilation Radiation Monitoring System is

to indicate whenever abnormal amounts of radioactive material exist in the

Reactor Building ventilation exhaust.

VYNPS DSAR Revision 0 4.0-43 of 54 4.7.1.3.2 Design Basis The Reactor Building Ventilation Radiation Monitoring System shall provide a

clear indication to facility personnel whenever abnormal amounts of

radioactivity exist in the Reactor Building ventilation exhaust.

4.7.1.3.3 Description

The Reactor Building Ventilation Radiation Monitoring System is shown on Drawing

5920-00526, and characteristics are given in Table 4.7.1.1. The system consists

of two sets of exhaust system monitors with one set of detectors located in the

refueling floor zone at one half the distance between the centerline of the

reactor vessel and centerline of the fuel pool, near the wall and 10 feet above

the refueling floor. One detector is located on one side of the refueling pool

and the other on the opposite side. The other set of detectors is located in

contact with the Reactor Building exhaust duct, upstream of the exhaust

ventilation isolation valve on elevation 280 of the Reactor Building.

Each set includes two individual channels. Each channel includes a

Geiger-Muller type detector and a combined indicator and trip unit. Both

channels share a two-pen strip chart recorder. All equipment is located in the

Main Control Room except the detectors.

Power for this system is from 120 V ac buses. Two power supplies are provided

each of which supplies one channel in each set of monitors.

Each channel has two trips. The upscale trip indicates high radiation and the

downscale trip indicates instrument trouble.

The environmental and power supply design conditions are given in Table 4.7.1.2.

4.7.1.3.4 Evaluation

The physical location and monitoring characteristics of the Reactor Building

ventilation radiation monitoring channels are adequate to provide detection

capability for abnormal amounts of radioactivity in the Reactor Building

ventilation.

4.7.1.3.5 Inspection and Testing

The trip circuits are tested by using test signals or portable gamma sources. VYNPS DSAR Revision 0 4.0-44 of 54 TABLE 4.7.1.1 PROCESS RADIATION MONITORING SYSTEMS CHARACTERISTICS Monitoring System Instrument

Range (1) Instrument Scale Upscale Trips Per Channel Downscale Trips Per Channel Liquid Process (17-350) (17-359) 10-1 to 10 6 counts per

second (2) 7 Decade Log 1 1 Reactor Building Ventilation Exhaust

(17-452A, B) 0.1 mR/hr to 1 R/hr 4 Decade Log 1 1 Reactor Building Refueling Floor

(17-453A, B) 1 to 10 4 hr mR 4 Decade Log 1 1 (1) Range of measurements is dependent on items such as the source geometry, background radiation, shielding, energy levels, and method of sampling. (2) Readout is dependent upon the pulse height discriminator setting.

VYNPS DSAR Revision 0 4.0-45 of 54 TABLE 4.7.1.2 PROCESS RADIATION MONITORING SYSTEM ENVIRONMENTAL AND POWER SUPPLY DESIGN CONDITIONS Sensor Location Main Control Room Parameter Design Requirements Range Design Requirements Range Temperature 25°C 0°C to 60°C 25°C 5° to +50°C Relative Humidity 50% 20 to 98% 50% 20 to 90% Power, AC 115 V 60 H z +/-10% +/-5% 115 V 60 H z +/-10% +/-5% Power, DC +24 V dc -24 V dc +22 to +29 V dc -22 to -29 V dc +24 V dc -24 V dc +22 to +29 V dc -22 to -29 V dc VYNPS DSAR Revision 0 4.0-46 of 54 TABLE 4.7.1.3 PLANT STACK RADIATION MONITORING SYSTEM CHARACTERISTICS Monitor Type Instrument Range Instrument Scale Type Detector Remarks Radio-Gas Monitors I & II

(17-156, 157) 10 to 10 7 cpm 6 Decade Digital Beta Scintillation Radio-Gas Monitor III

(17-155) 0.1 to 10 7 mR/hr 8 Decade Linear Ion Chamber Area Monitor in Stack Base VYNPS DSAR Revision 0 4.0-47 of 54 4.7.2 Area Radiation Monitoring System 4.7.2.1 Objectives The objectives of the Area Radiation Monitoring System are:

1. To warn of abnormal gamma radiation levels in areas where radioactive material may be present, stored, handled or inadvertently introduced.

2. To warn facility personnel whenever abnormal concentrations of airborne radioactive materials exist in the Reactor Building.

4.7.2.2 Design Basis

1. The Area Radiation Monitoring System shall provide facility personnel with a record and an indication of gamma radiation levels at selected locations within

the various facility buildings and radioactive airborne concentrations within the

Reactor Building.

2. The Area Radiation Monitoring System shall provide local alarms where it is necessary to warn personnel of substantial immediate changes in radiation levels.

4.7.2.3 Description

4.7.2.3.1 Monitors

1. Area Gamma Radiation Monitoring System The Area Gamma Radiation Monitoring System is shown as a functional block

diagram on Drawing 5920-430, Sh.1. A typical channel consists of a combined

indicator and trip unit, a shared power supply, and computer points for selected

monitors. Some channels have, in addition, a local audio alarm auxiliary unit. Each monitor has an upscale trip that indicates high radiation and a downscale

trip that may indicate instrument trouble. These trips sound alarms but cause

no control action. The system is powered from the 120 V ac instrument bus. The

trip circuits are set so that loss of power causes an alarm. The environmental

and power supply design conditions are given in Table 4.7.2.1.

2. Area Airborne Radiation Monitoring System The Reactor Building Area Airborne Radiation Monitoring System is shown as a

functional block diagram in Figure 4.7.2-2. Applicable specifications are

provided in Table 4.7.2.3. The Reactor Building Area Airborne Radiation

Monitoring System is a two (2) channel system employing a continuous air

particulate monitor and an off-line radiogas monitor located within a single

enclosure.

VYNPS DSAR Revision 0 4.0-48 of 54 The air particulate monitor consists of a continuous moving tape sampler with a beta scintillation detector to provide for the continuous monitoring of air

particulates. Each off-line gas monitor contains a beta scintillation probe for

measurement of radiogas.

The sample for the Reactor Building Area Airborne Monitoring System is withdrawn

from the Reactor Building Exhaust Ventilation System through an isokinetic

sample probe located in the exhaust duct. An air pump with flow indicator and a

low flow alarm is used to obtain the required sample flow. The sample is

directed back to the ventilation duct at the completion of the monitoring

process. The air particulate and radiogas monitor will indicate an alarm in the station

Main Control Room; no control action is provided. The monitor is equipped with

a sensor, a power supply, a logarithmic ratemeter and trip unit. Trip units

have adjustable trips which may be verified by the use of a remotely operated

radioactive check source mechanism; trip units also signal loss of high voltage

power supply, low sample flow and loss of signal input. The monitor readout is

continuously recorded on a recorder located in the station Main Control Room.

3. Technical Support Center (TSC) Area Monitoring System The Technical Support Center Area Monitoring System consists of four (4)

portable area gamma radiation monitors and an air particulate monitor.

Specifications applicable to the air particulate monitor are given in Table

4.7.2.4. Each of the area monitors reads out and alarms locally. The units

alarm on high radiation. These area monitors provide no control action other

than alarms. The four (4) portable gamma radiation monitors are self-contained

units dedicated to the TSC.

4. High-Range Reactor Building Area Monitoring System The Monitoring System consists of three channels, north and west personnel

access to the Reactor Building and inside the TIP Room. Each channel consists

of a detector, an indicator trip unit in the Control Room, and a local

indicator. The detectors are the ion chamber type with internal check sources. The local

and remote indicators provide an indication in the range of 1 R/hr to 10,000

R/hr. Each monitor provides an alert alarm, a high alarm, and a fail alarm.

The TIP Room monitor also energizes a flashing red beacon in the TIP Room upon

receipt of an alert alarm.

VYNPS DSAR Revision 0 4.0-49 of 54 4.7.2.3.2 Locations Work areas where gamma monitors will be located are tabulated in Table 4.7.2.2.

Annunciation and indication are provided in the Main Control Room with the exception

of the TSC Area Monitors.

4.7.2.4 Inspection and Testing

Area Gamma Radiation Monitoring System

An internal trip test circuit, adjustable over the full range of the trip circuit, is provided. The test signal is fed into the indicator and trip unit input so that

a meter reading is provided in addition to a real trip. All trip circuits are of

the latching type and must be manually reset at the front panel.

A portable calibration unit is also provided. This is a test unit designed for use

in the adjustment procedure for the area radiation monitor sensor and converter

unit. It provides five gamma radiation levels for calibration purposes. A cavity

in the calibration unit is designed to receive the sensor and converter unit.

Located on the back wall of the cylindrical lower half of the cavity is a window

through which radiation from the source emanates. A chart on each unit indicates

the radiation levels available from the unit for the various control settings.

Reactor Building Airborne Radiation Monitoring System

The Monitoring System includes a built-in check source for each detector. The check

source is operated from the Main Control Room. Alarm circuits can be tested by

using the built-in check source.

Technical Support Center Area Monitoring System

The Monitoring System is tested with both check sources and built-in test circuits

to ascertain system operability.

High-Range Reactor Building Area Monitoring System

The Monitoring System is equipped with a built in check source to verify system

operability. The system is periodically checked with a high-range source to verify

system linearity. VYNPS DSAR Revision 0 4.0-50 of 54 TABLE 4.7.2.1 AREA RADIATION MONITORING SYSTEM ENVIRONMENTAL AND POWER SUPPLY DESIGN CONDITIONS

Sensor Location Control Room Design Design Parameter Requirements Range Requirements Range Temperature 25°C 0° to 60°C 25°C 5° to 50°C Relative 50% 20 to 100% 50% 20 to 90%

Humidity Power 120 V +/-10% 120 V +/-10% 60 H z +/-5% 60 H z +/-5% VYNPS DSAR Revision 0 4.0-51 of 54 TABLE 4.7.2.2 LOCATIONS OF AREA RADIATION MONITORS

Station Number Location

1 Reactor Building Supp. Chamber Catwalk 232' 2 Reactor Building North Pers. Access 252' 2A* Reactor Building North Pers. Access 252' 3 Reactor Building South R.R. Access 252' 3A* Reactor Building West Pers. Access 252' 4 Reactor Building TIP Room 252' 4A* Reactor Building TIP Room 252' 5 Reactor Building Reactor Pers. Acc. Hatch 252' 6 Reactor Building Elev. Ent. 280' 7 Reactor Building CRD Repair 252' 8 Reactor Building Elev. Ent. 303' 9 Reactor Building RCUW Sample Sink 303' 10 Reactor Building Elev. Ent. 318' 11 Reactor Building RCUW Panel 318' 12 Reactor Building Elev. Ent. 345' 13 Turbine Building Moist. Sep. 228' 14 Reactor Building West Refuel 345' 15 Reactor Building Spent Fuel Pool 345' 16 Reactor Building New Fuel Vault 345' 17 Radwaste Recirc. Pump Room 252' 18 Radwaste R.W. Oper. Area 252' 19 Radwaste Pump and Tank Area 230' 20 Turbine Building No. Pers. Access 248' 21 Turbine Building Steam Stops 248' 22 Turbine Building Cond. Demin. Area 232' 23 Turbine Building Machine Shop 252' 24 Turbine Building Steam Inlet 272' 25 Control Room Viewing Gallery 272' 26 Turbine Building R.R. Door 252'

• High Range Area Monitor.

VYNPS DSAR Revision 0 4.0-52 of 53 TABLE 4.7.2.3 REACTOR BUILDING AREA AIRBORNE RADIATION MONITORING SYSTEM Monitor Type Instrument Range Instrument Scale Type Detector Remarks Air Particulate

Monitor 10 to 10 7 cpm 6 Decade Log Digital Beta Scintillation Tape transport mechanism - tape speed

selectable .5, 1, 2, and 10 in/hr; alarm for

broken tape and low

flow Radio-Gas Monitor 10 to 10 7 cpm 6 Decade Log Digital Beta Scintillation

VYNPS DSAR Revision 0 4.0-53 of 54 TABLE 4.7.2.4 TECHNICAL SUPPORT CENTER AREA AIRBORNE RADIATION MONITORING SYSTEM Monitor Type Instrument Sensitivity Air Particulate Monitor 1 x 10 -11 µCi/cc Sr 90-Y 90 or Cs 137 Noble gases 1 x 10 -6 µCi/cc Xe 133 or Kr 85 Iodine 1 x 10 -11 µCi/cc VYNPS DSAR Revision 0 4.0-54 of 54 Vermont Yankee Defueled Safety Analysis Report Revision 0 Reactor Building Area Airborne Radiation Monitoring System Figure 4.7.2-2

VYNPS DSAR Revision 0 5.0-1 of 5 CONDUCT OF OPERATIONS TABLE OF CONTENTS

Section Title Page 5.1 ORGANIZATION AND RESPONSIBILITY ....................................... 2 5.2 TRAINING .............................................................. 2 5.2.1 Program Description (General) ............................... 2 5.2.2 General Employee Training ................................... 2 5.2.2.1 Access to Plant ................................. 2 5.2.3 Fire Brigade Training ....................................... 2 5.2.4 Operations Training ......................................... 2 5.2.5 Craft, Technician, and Engineering Staff Position (ESP) Training ..................................... 3 5.2.6 Training Records ............................................ 3 5.2.7 Training Program Approval and Evaluation .................... 3 5.2.8 Responsibility .............................................. 3 5.3 EMERGENCY PLAN ........................................................ 4 5.4 QUALITY ASSURANCE PROGRAM ............................................. 4 5.4.1 Scope ....................................................... 4 5.4.2 Responsibilities ............................................ 4 5.4.3 Implementation .............................................. 4 5.4.4 Management Evaluation ....................................... 5 5.5 REVIEW AND AUDIT OF OPERATIONS ........................................ 5 5.5.1 General ..................................................... 5 5.5.2 On-Site Safety Review Committee ............................. 5 5.5.3 Independent Safety Review ................................... 5 5.6 TECHNICAL REQUIREMENTS MANUAL ......................................... 5

VYNPS DSAR Revision 0 5.0-2 of 5 5.1 ORGANIZATION AND RESPONSIBILITY The Vermont Yankee Nuclear Power Station organization, including the responsibilities and duties of staff personnel, are detailed in the Vermont

Yankee Quality Assurance Program Manual. 5.2 TRAINING 5.2.1 Program Description (General) The objective of the Training Program is to provide qualified personnel to operate and maintain the permanently defueled facility in a safe manner, including the storage and handling of irradiated fuel. All operations, craft, technician, engineering staff, and general employee training requirements are

described in position-specific program descriptions or procedures. Training

programs are implemented and maintained using a Systems Approach to Training (SAT), in accordance with 10CFR50.120, Training and Qualification of Nuclear

Power Plant Personnel, and ANSI/ANS 3.1, 1978, Selection, Qualification, and

Training of Personnel for Nuclear Power Plants. 5.2.2 General Employee Training All persons permanently employed at the facility shall be trained in the applicable following areas commensurate with their job duties:

1. Chemical and Hazardous Material Program

2. Radiological Health and Safety Program

3. Site Emergency Plans

4. Industrial Safety

5. Fire Protection

6. Security 7. Quality Assurance

8. Fitness for Duty 5.2.2.1 Access to Plant Requirements to gain access to the facility protected area, including training requirements, are contained in applicable facility procedures. 5.2.3 Fire Brigade Training Fire brigade training for appropriate facility personnel meets the requirements of NFPA 600, Standard on Industrial Fire Brigades. 5.2.4 Operations Training The initial and continuing training programs for the personnel performing operator functions, including certified fuel handler and shift manager, are

based on a SAT. VYNPS DSAR Revision 0 5.0-3 of 5 5.2.5 Craft, Technician, and Engineering Staff Position (ESP) Training The initial and continuing training programs for the instrument control technician, chemistry technician, radiation protection technician, plant

mechanic (electrical and mechanical maintenance), and engineering staff

positions are based on a SAT. 5.2.6 Training Records Records of employee and contractor participation in, and completion of, training activities are maintained in accordance with the VY records retention

policy. 5.2.7 Training Program Approval and Evaluation The Vermont Yankee position-specific training program descriptions are approved by appropriate Training Department and facility management, as

specified in applicable facility procedures. This ensures that the content and

the intent of the training programs provide the necessary training for

personnel associated with the safe storage and handling of irradiated fuel and

management of radioactive waste. Training processes are controlled and

maintained in accordance with applicable Training directives.

The effectiveness of training programs is evaluated by the performance of

employees in carrying out their assigned duties, by performance on facility

evaluations, and the employment of various types of feedback mechanisms. The

results of the evaluations are maintained in accordance with applicable

records retention requirements. 5.2.8 Responsibility As delegated by the Manager, Operations, the Superintendent, Training is responsible for the conduct and administration of the specified training

activities, including:

1. Initial and continuing training programs for the non-certified operator, certified fuel handler and shift manager.

2. Fire brigade training.

3. Initial and continuing training programs for instrumentation and control, maintenance, and engineering staff positions.

4. Initial and continuing training programs for chemistry and radiation protection positions.

5. General employee training.

VYNPS DSAR Revision 0 5.0-4 of 5 5.3 EMERGENCY PLAN The emergency plan for the Vermont Yankee Nuclear Power station was originally

issued in accordance with NRC's regulations on April 1, 1981. Any information

regarding this plan should be obtained from the most current revision to that

document. 5.4 QUALITY ASSURANCE PROGRAM 5.4.1 Scope This section establishes the criteria to be applied to systems requiring Quality Assurance which prevent or mitigate the consequences of postulated

accidents which could cause undue risk to the health and safety of the public.

The structures, systems, components, and other items requiring quality

assurance are listed in the Vermont Yankee Safety Classification Program. 5.4.2 Responsibilities

1. Compliance with the requirements of the VY Quality Assurance Program Manual (VYQAPM) based on the criteria of Title 10 of the Code of Federal

Regulations, Part 50, Appendix B, and as committed to within the VYQAPM, shall be the responsibility of all personnel involved with activities

affecting operational safety. Vermont Yankee shall cross reference the

applicable criteria of 10CFR50 Appendix B in procedures that implement the

VYQAPM. The performance of quality-related activities shall be

accomplished with specified equipment under suitable environmental

conditions.

2. Individuals having direct responsibilities for establishment/distribution control/implementation of the VYQAPM are delineated in the "Organization,"

section of the VYQAPM. 5.4.3 Implementation Establishment of an effective Operational Quality Assurance Program is assured through consideration of, and conformance with, the Regulatory Position in the

Regulatory Guides listed the VYQAPM. Implementation of this program is assured

through Quality Assurance procedures, derived from Quality Assurance policies, goals, and objectives.

VYNPS DSAR Revision 0 5.0-5 of 5 5.4.4 Management Evaluation The Safety Review Committee (SRC) reports to the executive responsible for

oversight on those areas of responsibility specified in the Quality Assurance

Program Manual. The SRC conducts its function in accordance with a procedure

approved by the executive responsible for oversight. The SRC independently

monitors applicable programs and provides management with evaluations and

assessments related to the effectiveness of the nuclear program.

5.5 REVIEW AND AUDIT OF OPERATIONS

5.5.1 General

Two review bodies have been established to review operating procedures, evaluate and process changes and assure compliance and safe operation.

5.5.2 On-Site Safety Review Committee

The responsibilities and authorities of the On-Site Safety Review Committee

are described in an approved Quality Assurance Program Manual implementing

procedure.

5.5.3 Independent Safety Review

An independent safety review of activities affecting nuclear safety is

performed in accordance with an approved Quality Assurance Program Manual

implementing procedure.

5.6 TECHNICAL REQUIREMENTS MANUAL

Requirements pertinent to the permanently defueled state which have been

relocated out of Technical Specifications, as well as any other items deemed

appropriate by facility management, which do not meet the Technical

Specification screening criteria provided in 10CFR50.36(c)(2)(ii), are located

in the Technical Requirements Manual (TRM). Changes to the TRM are evaluated

per the requirements of 10CFR50.59.

VYNPS DSAR Revision 0 6.0-1 of 37 SAFETY ANALYSIS TABLE OF CONTENTS Section Title Page

6.1 INTRODUCTION

.......................................................... 4 6.2 ACCEPTANCE CRITERIA ................................................... 5 6.2.1 DBA Acceptance Criteria ...................................... 5 6.2.2 Site Event Acceptance Criteria ............................... 6 6.3 ACCIDENTS EVALUATED ................................................... 7 6.3.1 Fuel Handling Accident ....................................... 7 6.3.1.1 Analytical Methodology 7 6.3.1.2 Assembly Drop in SFP with Open Containment Scenario 7 6.3.1.3 Assembly Drop in SFP with Closed Containment Scenario 9 6.3.1.4 Software 9 6.3.1.5 Assumptions 9 6.3.1.6 Inputs 10 6.3.1.7 Impact of Water Depth on Iodine Decontamination Factor 10 6.3.1.8 Fuel Damage from Assembly Drop onto SFP Fuel Racks 12 6.3.1.9 Radiological Consequences/Results 13 6.4 SITE EVENTS EVALUATED ................................................ 25 6.4.1 High Integrity Container (HIC) Drop Event ................... 25 6.4.1.1 Analytical Methodology 25 6.4.1.2 Assumptions 25 6.4.1.3 Inputs 26 6.4.1.4 Radiological Consequences/Results 27

6.5 REFERENCES

........................................................... 29 6.6 APPENDICES ........................................................... 32

VYNPS DSAR Revision 0 6.0-2 of 37 SAFETY ANALYSIS LIST OF TABLES

Table No. Title

6.3.1 FHA Scenarios Analyzed 6.3.2 Input Conditions for FHA 6.3.3 Undecayed Core Inventory for Radionuclides Important in the Radiological Evaluation of DBAs 6.3.4 Undecayed Gap Activity Available for Release from Fuel Assembly Drop in the SFP 6.3.5 Typical Iodine Decontamination Factors and Iodine Speciation vs Water Depth above Dropped Assembly 6.3.6 Atmospheric Dispersion Factors for the Postulated FHA 6.3.7 EAB TEDE Dose vs Long Decay Time 6.4.1 HIC Drop Source Term Release Activity VYNPS DSAR Revision 0 6.0-3 of 37 SAFETY ANALYSIS LIST OF FIGURES

Reference Figure No. Drawing No. Title

6.3-1 VY FHA - EAB TEDE Dose vs Decay Time 6.3-2 VY FHA - MCR TEDE Dose vs Decay Time A-1 Dimensions of SFP, Fuel Rack and Fuel Handling Equipment VYNPS DSAR Revision 0 6.0-4 of 37 6.1 Introduction In January of 2015, the licensee certified to the NRC that Vermont Yankee had

both permanently ceased operations (final shutdown 12/29/14) and that all fuel

had been removed from the reactor vessel and placed in the spent fuel pool (SFP) (Reference 6.5-1). Since Vermont Yankee will never again enter any

operational mode, reactor related accidents are no longer a possibility.

This chapter discusses: (a) a postulated fuel handling accident (FHA)

associated with fuel movement until the fuel has been transferred to the

Independent Spent Fuel Storage Installation (ISFSI); and (b) the postulated

drop of a high integrity container (HIC) containing radioactive resins

Bounding conditions, conservatism in equipment design, conformance to high

standards of material and construction, the control of loads and strict

administrative controls over facility operations all serve to assure the

integrity of the fuel while in the spent fuel pool and during fuel transfer to

the Independent Spent Fuel Storage Installation (ISFSI).

Accidents involving fuel and the Holtec International HI-STORM system storage

casks are discussed in the HI-STORM FSAR (Reference 6.5-2).

For site events, a drop and fire of a High Integrity Container (HIC) containing

resins was evaluated.

New hazards, new initiators or new accidents that may challenge offsite

guideline exposures, may be introduced as a result of certain decommissioning

activities. These issues will be evaluated when the scope and type of

decommissioning activities are finalized. VYNPS DSAR Revision 0 6.0-5 of 37 6.2 Acceptance Criteria 6.2.1 DBA Acceptance Criteria

The radiological release acceptance criteria associated with the Alternative

Source Term (AST) methodology are identified in 10CFR50.67 (Reference 6.5-3)

and dose levels are not to exceed:

• Exclusion Area Boundary (EAB): 25 rem TEDE
• Low Population Zone (LPZ): 25 rem TEDE
• Control Room (CR): 5.0 rem TEDE

These criteria, however, are for evaluating potential reactor accidents of

exceedingly low occurrence probability and low risk of public exposure to

radiation. For events with higher probability of occurrence, such as a FHA, the acceptance criteria for the offsite receptors are more stringent, while

that for the control room operators remains the same. The applicable AST

criteria for an FHA are identified in Regulatory Guide 1.183 (Reference 6.5-4)

and 10CFR50.67 and dose levels are not to exceed:

• Exclusion Area Boundary (EAB): 6.3 rem TEDE
• Low Population Zone (LPZ): 6.3 rem TEDE
• Control Room (CR): 5.0 rem TEDE

The EAB and LPZ criteria are referred to as being "well within" the regulatory

limits (i.e., 25%).

The LPZ doses are bounded by the dose at the EAB, since the LPZ is farther

away.

Additional acceptance criteria are as follows:

• The required decay time that would preclude Evacuation as a protective action following an FHA. The limit for such an action is 1 rem TEDE (EPA

400-R-92-001, Table 2-1 (Reference 6.5-5)).

• The required decay time that would reduce the TEDE dose from gaseous releases of iodines and particulates to unrestricted areas to "well

within" (i.e., 25 %) of the 10CFR50 Appendix I (Reference 6.5-6) annual

dose limit of 15 mrem (or 3.75 mrem).

VYNPS DSAR Revision 0 6.0-6 of 37 The latter acceptance criterion was selected as a suitable basis no longer requiring the Standby Gas Treatment System (SGTS). The "well within" limit was

selected so as to accommodate other potential releases from the plant. It is

noted that, by definition, this criterion excludes the noble gas dose. There exists a separate criterion applicable to the nobles, which however is of no

interest in the present application since the noble gas release is not impacted

by the SGTS filtration.

6.2.2 Site Event Acceptance Criteria

The HIC drop acceptance criteria are based on 10% of the 10CFR100 dose

acceptance criteria.

10CFR100 Acceptance Criteria (1) (rem) 10% of 10CFR100 Acceptance Criteria (rem) EAB and LPZ 25 (whole body) 2.5 (whole body) 300 (thyroid) 30 (thyroid, critical organ)

(1)  EAB and LPZ dose acceptance criteria from 10CFR100.11 

VYNPS DSAR Revision 0 6.0-7 of 37 6.3 Accidents Evaluated 6.3.1 Fuel Handling Accident Two bounding scenarios of the FHA are considered and the scenario objectives

are summarized in Table 6.3.1. The first scenario is a drop in the SFP with an open containment (no filtration by SGTS) with an instantaneous release (ground level) to determine the required

decay time prior to fuel movement that would result in the EAB dose, and main

control room (MCR) dose within 10CFR50.67 and R.G. 1.183 limits and the EAB

TEDE dose under the EPA PAG limit of 1 rem for evacuation.

The second scenario is with a drop in the SFP, with a closed containment (not

under a negative pressure and not 'air tight') and an elevated (stack release)

instantaneous release with and without filtration by the SGTS, to determine the

required decay time prior to fuel movement that would result in 25% of the

10CFR50, Appendix I TEDE annual dose limit of 15 mrem at the EAB. The 10CFR50, Appendix I dose of 15 mrem from gaseous effluents coincides with maintaining

dose as low as reasonably achievable. A revision to the Technical

Specifications proposed by BVY 13-097 (Reference 6.5-7) and approved by NVY 15-

013 (Reference 6.5-8) contains the following commitment: "During fuel

handling/core alterations, ventilation system and radiation monitor

availability (as defined in NUMARC 91-06, Reference 6.5-9) will be assessed, with respect to filtration and monitoring of releases from the fuel. The goal

of maintaining ventilation system and radiation monitor availability is to

reduce doses even further below that provided by the natural decay." This FHA

scenario is the analysis which demonstrates that SGTS operation is not required

to maintain dose as low as reasonably achievable consistent with 10CFR50, Appendix I, thus fulfilling this VY commitment. 6.3.1.1 Analytical Methodology The postulated FHA scenarios were based on the Alternative Source Term (AST) Methodology in RG 1.183, Appendix B. Two main configurations of the reactor

building during fuel movement were considered, one with an open containment and

the other with a closed containment. 6.3.1.2 Assembly Drop in SFP with Open Containment Scenario (a) The reactor operated at full power (1950 MWt) for an extended period of time until permanent shutdown and full core offload is completed with all

fuel in the SFP. VYNPS DSAR Revision 0 6.0-8 of 37 (b) Fuel moves are in progress and an FHA takes place in the SFP with an assembly falling onto a fuel rack at various assumed decay times after

reactor shutdown, from 1 to 19 days. (c) The accident leads to the damage of 98 fuel rods (a bounding value from Table 6.3.2, Item #D1), addressing both GE14 and GNF2 10x10 assemblies

removed from the final core offload. All failed rods are peak powered, with a radial peaking factor of 1.65 (from Table 6.3.2, Item #A3). (d) All activity within the gaps of the failed fuel rods is released to the refueling cavity pool. The released activity corresponds to 8% of the

entire inventory of I-131 in the rods (i.e., within the fuel matrix and

gaps), 10% of the Kr-85, 5% of the remaining halogens and noble gases, and 12% of the alkalis (Cs and Rb), from Table 6.3.2, Item #A4. The

undecayed activity released from the damaged fuel rods is presented in

Table 6.3.4; decay correction from plant shutdown to the time of the

postulated FHA is properly accounted for by the radiological software

used in the analysis (ELISA-2). (e) All the noble gases and a fraction of the halogens (see Item (g) below) escape from the pool and are released to the refueling level. All the

alkalis are retained by the pool. The halogen speciation above the pool

is not pertinent in this scenario since there is no pre- or post-release

filtration of radioactivity. (f) The water depth above the dropped assembly in the SFP is 20.5 feet (versus the 23-foot requirement in Regulatory Guide 1.183 for full credit

of the decontamination factor (DF) of 200 for iodine retention by the

pool water), leading to a corresponding decrease in the DF to 125.4 (from

Table 6.3.5). (g) The radioactivity which becomes airborne above the SFP is instantly released to the environment without holdup (a conservative assumption

which accommodates the 2-hr release requirement in Regulatory Guide

1.183, Appendix B, Sec. 5.3). (h) The reactor building is open during the refueling operations, with all post-FHA releases to the environment assumed to be at ground level, via

the RB blowout panels. (i) Transport of the released radioactivity to the receptors of interest is dictated by the applicable atmospheric dispersion factors in Table 6.3.6. (j) The MCR ventilation configuration is in the normal operating mode during the entire exposure interval (30 days), with an intake flow of 3700 cfm, unfiltered. (k) Breathing rates and MCR occupancy factors are as given in Table 6.3.2, Items #F2 and #F3. (l) The control room operator dose point is at the base of a hemispherical cloud having a volume equal to the free air volume of the control room.

Finite-cloud correction to the submersion dose was based on the

Murphy/Campe equation in Reg. Guide 1.183 (Sec. 4.2.7).

VYNPS DSAR Revision 0 6.0-9 of 37 6.3.1.3 Assembly Drop in SFP with Closed Containment Scenario The scenario is the same as an assembly drop in the SFP with an open containment, except with the following differences:

(a) The radionuclide list is as given in Table 6.3.4, though without the noble gases. The acceptance criterion for no longer requiring the SGTS 

filtration was selected to be 25% of the 10CFR50, Appendix I annual TEDE

dose limit of 15 mrem (or. 3.75 mrem) from gaseous releases of iodines

and particulates to unrestricted areas. As such, this criterion excludes

the noble gas dose. (b) The reactor building is closed during fuel moves, such that all releases to the environment would be via the main stack, with and without credit

for filtration by the SGTS system. In-transit decay and plateout were not

credited. (c) The decay times prior to fuel movement are fairly long in this scenario because the interest is in determining the time beyond which there will

be no dose-wise beneficial purpose for maintaining the SGTS operational

after permanent plant shutdown. (d) The EAB dose consequences were evaluated without SGTS filtration, but also with SGTS filtration for informational purposes. The filtration

efficiencies for the latter case are as given in Table 6.3.2, Item #D7. (e) The MCR is of no interest in this scenario and was therefore excluded from the analysis. 6.3.1.4 Software Computation of the EAB and MCR radiological consequences for the postulated FHA

were based on the ELISA-2 computer code, Version 2.4 (Reference 6.5-10) for all

analyzed scenarios.

The dose conversions in ELISA-2 are from Federal Guidance Reports 11 (Reference

6.5-11) and 12 (Reference 6.5-12). Dose rates and cumulative doses are

computed for each organ, TEDE, skin and air. Of these, only the TEDE doses are

presented for comparison to the TEDE regulatory limits.

ELISA-2 was designed to handle the pre-FHA decay correction and the time-

release from the RB for all scenarios. Its built-in logic accounts for the

time-dependent generation and release of noble gases from the decay of halogens

retained by the pool water, and also from the halogens on the SGTS exhaust

filtration system when credited. These releases extend beyond the end of the

2-hr release from the RB. 6.3.1.5 Assumptions VYNPS DSAR Revision 0 6.0-10 of 37 Release Rate from Reactor Building In accordance with RG 1.183, Appendix B, Sec. 4.1 for an FHA in the SFP, the radioactive material that escapes the water pool is released to the environment

over a 2-hour interval. Analytically, this is conservatively accommodated by

assuming an instantaneous release to the environment, in all accident scenarios

and for all receptors (EAB and MCR). Credit for in-transit decay or plateout

was not taken.

Fuel Rod Gap Activity All activity within the gaps of the failed fuel rods is released to the

refueling cavity pool. The released activity corresponds to 8% of the entire

inventory of I-131 in the rods (i.e., within the fuel matrix and gaps), 10% of

the Kr-85, 5% of the remaining halogens and noble gases, and 12% of the alkalis (Cs and Rb), from Table 6.3.2 (Item #A4), Fuel Rod Gap Fractions. The

undecayed activity released from the damaged fuel rods is presented in Table

6.3.4.

MCR Finite Cloud Correction Doses to MCR personnel due to the external gamma radiation from airborne

radioactivity within the MCR were adjusted using the Murphy/Campe finite-cloud

correction model in R.G. 1.183, Section 4.2.7.

Modeling Simplifications There are no modeling simplifications.

Number of Failed Fuel Rods Assumptions associated with determining the number of failed fuel rods due to a

dropped fuel assembly are provided in Section 6.6, Appendix A. 6.3.1.6 Inputs Inputs for the analysis are identified in Table 6.3.2. 6.3.1.7 Impact of Water Depth on Iodine Decontamination Factor According to RG 1.183, Appendix B, Section 2, Water Depth, if the water depth above the damaged dropped assembly in a fuel handling accident is 23 feet or

greater, an overall decontamination factor of 200 may be credited for the

expected iodine retention by the pool water. As clarified in RIS 2006-004, Item 8 (Reference 6.5-13) and in the proposed revision to RG 1.183, with an

iodine speciation consisting of 99.85% elemental (including CsI) and 0.15%

organic in the fuel-rod gaps, the overall DF of 200 is achieved when the DF for

the elemental iodines is 285, with the ensuing iodine composition in air above

the pool being 70% elemental and 30% organic.

VYNPS DSAR Revision 0 6.0-11 of 37 For water depths less than 23 feet, RG 1.183 recommends the use of the Burley model (Reference 6.5-14). According to this model, the DF for a reduced water

depth is determined through use of the following formula (Reference 6.5-14, pg

26): DF inorg = exp {(6/d b)*(k eff H b / v b)} (Eq. 1) where DF inorg = elemental iodine pool retention factor (or DF) d b = bubble diameter (cm) k eff = effective mass transfer coefficient (cm/sec) H b = bubble rise height (cm, water depth above dropped assembly) v b = bubble rise velocity (cm/sec)

This equation can be rewritten by combining all of the independent

variables (excluding H b), as: DF inorg = exp (K*H b) (Eq. 2) where K = {(6/d b)*(k eff / v b)} (Eq. 3) The new variable K can now be back-calculated by using the recommended values given above for DF inorg and H b, namely 285 and 23 ft (or 701.0 cm), respectively, and is as follows: K = log e(285) / 701.0 = 0.008063 (Eq. 4)

The applicable DF equation for reduced SFP water then becomes: DF inorg = exp (0.008063*H b) (Eq. 5)

For the iodine speciation given above, namely 99.85% elemental (including CsI) and 0.15% organic, and no organic iodine retention by the pool water (i.e., DF org = 1), the overall (total iodine) DF is given by:

DF total = [(0.9985/DF inorg) + (0.0015/1)] -1 (Eq. 6)

Proceeding further, the above-water (i.e., airborne) iodine

speciation formulas (in percent) are given by:

S inorg = 99.85* (DF total / DF inorg) (Eq. 7) and S org = 0.15

• DF total (Eq. 8)

VYNPS DSAR Revision 0 6.0-12 of 37 Typical DF values and iodine speciation versus SFP water depth are presented in Table 6.3.5. The iodine speciation is of no interest since (a) there is no

filtration credit in the open containment scenario, and (b) all iodine species

were subjected to the same SGTS filtration efficiency in the closed containment

scenario. 6.3.1.8 Fuel Damage from Assembly Drop onto SFP Fuel Racks Fuel pin damage due to a drop of a fuel assembly onto a spent fuel rack within

the SFP was evaluated. A drop in the SFP is limited to a drop distance of 3

feet, rounded up from 1.9 feet for conservatism. The General Electric Standard

Application for Reactor Fuel, GESTAR II (Reference 6.5-15) is utilized to

determine the number of damaged fuel pins resulting from the drop. This

analysis is presented in Section 6.6, Appendix A. VYNPS DSAR Revision 0 6.0-13 of 37 6.3.1.9 Radiological Consequences/Results The radiological consequences of the FHA scenarios are shown below.

Scenario Limiting Dose Acceptance Criteria Open Containment Ground Level Instantaneous Release No SGTS Filtration Closed Containment Elevated Release (Main Stack) Instantaneous Release No SGTS Filtration Required Decay Time to Meet Most Restrictive Acceptance Criteria (Section 6.2.1) 15 days50 days

Dose TEDE MCR 5 rem (10CFR50.67) < 5 remN A EAB 6.3 rem (R.G. 1.183) < 6.3 remN A EAB EPA PAG 1 rem (initiation of evacuation) < 1 rem N A EAB 10CFR50 Appendix I 15 mrem annual limit N A 3.75 mrem (25% of the limit) Decay Time and Dose Details Figure 6.3-1 Figure 6.3.2 Table 6.3.7

Note that doses at the EAB bound the corresponding dose at the Low

Population Zone (LPZ), as the LPZ is farther away from the station.

The conclusion is that if there is a FHA 50 days following cessation of

power operations, the benefits of SGTS are minimal and resultant dose

without SGTS operation at the EAB is considered to be maintained as low

as reasonably achievable. VYNPS DSAR Revision 0 6.0-14 of 37 Table 6.3.1 FHA Scenarios Analyzed Scenario Open containment Closed containment Objective To determine the required decay time prior to fuel

movement in the SFP that would meet the following: (a) 90% of the 10CFR 50.67 dose acceptance criteria, and (b) an EAB TEDE dose less than the PAG limit of 1 rem for evacuation. To provide basis for no longer requiring the SGTS in support of a

VY TS commitment (contained within Reference 6.5-7) with respect to dose minimization following an FHA Containment Building

Configuration Open containment, with instantaneous atmospheric release via the blow-out panels (ground-level release) Closed containment with instantaneous atmospheric releases via main stack (with and without SGTS iodine and particulate filtration) Location of Assembly

Drop SFP (3 foot drop onto fuel racks) SFP (3 foot drop onto fuel racks) Water Depth above

Dropped Assembly

Credited for Iodine

Retention 20.5 feet [DF = 125.4] 20.5 feet

 [DF = 125.4] Fuel Damage To be determined in present calculation based on both GE14 and GNF2 assembly drops To be determined in present calculation based on both GE14 and GNF2 assembly drops Pre FHA Decay Time 

from Reactor

Shutdown Required decay time to be determined in present calculation to meet the dose consequence objectives Required decay time to meet 25% of the 10 CFR 50 Appendix I TEDE annual dose limit of 15 mrem at the EAB from iodines and particulates VYNPS DSAR Revision 0 6.0-15 of 37 Table 6.3.2 INPUT CONDITIONS FOR FHA Item No. DESCRIPTION VALUE REFERENCE FHA Source TermA1 Power level for DBA analysis [Includes 2 % measurement uncertainty] 1950 MWt 6.5-16 (VYC-2299) A2 Number of assemblies in core368A3 Maximum allowed radial peaking factor (a)1.65A4 Fuel rod gap fractions (AST Methodology) I-131

Kr-85

Other noble gases

Other halogens

Alkali metals (Cs and Rb) 0.08 0.10 0.05 0.05 0.12 6.5-4 (Reg. Guide 1.183, Table 3) A5 Undecayed core inventory for radionuclides important in the evaluation of DBAs Table 6.3.3 6.5-17 (VYC-2260, Table 4.5) A6 Post-shutdown decay time prior to postulated accident Various Assumed values for sensitivity analyses Variables for Fuel Damage Calculation for FHA in Spent Fuel Pool B1 Number of fuel rods in 10x10 assemblies (GE14 and GNF2) 92 6.5-18 (VYC-2206, Sec. 1) B2 Assembly drop height above fuel racks Bounding value 22.74" (1.9 ft) 6.5-19 (App. A, pg 17 of 20) Used in analysis (c)36" (3 ft) Conservatively assumed value B3 Wet weight of fuel assembly and channel GE14 569.5 lbm 6.5-18 (VYC-2206, pg 12 of 31) GNF2 580.0 lbm 6.5-19 (App. A, pg 8 of 20) B4 Weight of mast and grapple GE14619 lbm 6.5-19 (App. A, pg 9 of 20) GNF2619 lbm B5 Percent energy for clad

deformation GE1451%6.5-15 (GESTAR II , Sec. 5.3.1 (also applied to GNF2

assemblies) GNF2 51% B6 10x10 rod compression failure (energy required to

damage stationary fuel

rods) GE14167 ft-lb 6.5-19 (App. A., page 10 of 20) GNF2 157 ft-lb VYNPS DSAR Revision 0 6.0-16 of 37 Table 6.3.2 (Continued) INPUT CONDITIONS FOR FHA DESCRIPTIONVALUEREFERENCE Atmospheric Release Resulting from Postulated FHA in Spent Fuel Pool D1 Number of damaged fuel assemblies GE1497 6.5-19 (VYC-3187) GNF298Used in analysis 98 Bounds both GE14 and GNF2 assembly types D2 Water depth above dropped assembly (resting on top of fuel

racks) Minimum value 20.67 ft 6.5-19 (App. A, pg 2 of 19) Used in analysis 20.5 ft Conservative D3 Undecayed gap inventory available for release from 98 damaged fuel rods See Table 6.3.4 D4 Overall pool DF for given water depth Noble gases 1 6.5-4 (RG 1.183) Halogens125.4See Table 6.3.5AlkalisInfinite 6.5-4 (RG 1.183) D5 Percent of damaged-fuel rod gap activity release 100 % D6 Reactor building configuration during

refueling operations Closed Containment w/wo SGTS Instantaneous Stack Release See Table 6.3.1 Open Containment No SGTS Ground Level Release D7 Potential release point to the atmosphere (see Table 6.3.6 for the

atmospheric dispersion

factors) Closed Containment w/wo SGTS Stack Release See Table 6.3.1 Open Containment No SGTS Instantaneous RB blowout panel release 6.5-20 (VYC-2275) D8 SGTS filtration efficiency, all halogens and particulates 95% 6.5-21 (VYC-2302, Page 11 of

59) D9 Release duration to atmosphere Instantaneous Meets the RG 1.183 requirements VYNPS DSAR Revision 0 6.0-17 of 37 Table 6.3.2 (Continued)

INPUT CONDITIONS FOR FHA DESCRIPTIONVALUEREFERENCE Control Room CharacteristicsE1 Control room free air volume41534 ft 3 6.5-16 (VYC-2299) E2 MCR HVAC nominal unfiltered intake flow for accident duration and all FHA

scenarios (assumed to include fresh

air and air from surrounding areas as

a result of ingress, egress and

inleakage) 3700 cfm DESCRIPTIONVALUEREFERENCE / COMMENTS Other Variables F1 Atmospheric dispersion factors from release point to locations of interest See Table 6.3.6 6.5-16 and 6.5-20 (VYC-2299 and VYC-2275, Section 6) F2 Breathing rates Control Room 0 -720 hrs 3.5E-04 m 3/sec 6.5-4 (RG 1.183, pg 1.183-18) EAB 0 - 2 hr 3.5E-04 m 3/sec 6.5-4 (RG 1.183, pg 1.183-16) F3 Control room occupancy factors 0 -24 hrs 1.0 6.5-4 (RG 1.183, pg 1.183-18) 24 -96 hrs 0.6 96 -720 hrs 0.4 F4 Exposure Intervals (b) Control room 30 days6.5-4 (RG 1.183, Sections 4.1.3, 4.1.5

and 4.2.6) EAB 2 hrs F5 Regulatory dose limits Control room TEDE 5 rem 6.5-4 (RG 1.183, pg 1.183-19 and

10CFR50.67, Sec. (b)(2)(iii)) EAB TEDE6.3 rem 6.5-4 (RG 1.183, Table 6) LPZ TEDE 6.3 rem F6 PAG Evacuation dose limit (EAB TEDE) 1 rem 6.5-5 (EPA 400-R-92-001)

(a) In line with RG 1.183, Sec. 3.1, the radial peaking factor is applied to the average fuel-assembly inventory based on the core inventory in Table 6.3.3. This is a conservative 

approach and bounds any potential variations in the FHA source term resulting from

variations in the EFPDs and burnup in any given cycle. (b) Even though all radioactivity is released to the atmosphere within 2 hours following a design-basis FHA, the exposure intervals for the CR personnel was assumed to be 30 days.

This provides adequate time for cleanup of the airborne radioactivity still present within

the CR after termination of the 2-hr release, and also accounts for the delayed release of noble-gas decay products from the refueling pool water produced upon decay of halogens retained therein. (c) The assembly drop height within the SFP was conservatively increased to 3 ft to account for the difference in elevations between the top of the racks and the top of an assembly within

the racks, as well as any other dimensional uncertainties. VYNPS DSAR Revision 0 6.0-18 of 37 Table 6.3.3 Undecayed Core Inventory for Radionuclides Important in the Radiological Evaluation of DBAs (From VYC-2260, Table 4.5, based on 1950 MWt, an enrichment range from 3.0 to 4.65 wt % U-235, and core-average burnup from 5 to 58 GWD/MTU) Nuclide Core Ci Nuclide Core Ci Br-83 8.267E+06 I-132 7.900E+07 Kr-83m 8.265E+06 Te-133 6.602E+07 Br-85 1.874E+07 Te-133m 4.493E+07 Kr-85 9.852E+05 I-133 1.130E+08 Kr-85m 1.894E+07 Xe-133 1.128E+08 Rb-86 2.496E+05 Xe-133m 3.428E+06 Kr-87 3.788E+07 Te-134 1.036E+08 Kr-88 5.355E+07 I-134 1.254E+08 Kr-89 6.755E+07 Cs-134 2.971E+07 Sr-89 6.724E+07 I-135 1.051E+08 Sr-90 7.999E+06 Xe-135 4.540E+07 Y-90 8.363E+06 Xe-135m 2.232E+07 Sr-91 8.684E+07 Cs-136 7.602E+06 Y-91 8.270E+07 Xe-137 9.893E+07 Sr-92 8.987E+07 Cs-137 1.186E+07 Y-92 9.008E+07 Ba-137m 1.124E+07 Y-93 9.857E+07 Xe-138 9.851E+07 Zr-95 9.645E+07 Ba-139 1.043E+08 Nb-95 9.673E+07 Ba-140 1.004E+08 Zr-97 9.596E+07 La-140 1.009E+08 Mo-99 1.034E+08 La-141 9.573E+07 Tc-99m 9.051E+07 Ce-141 9.255E+07 Ru-103 9.889E+07 La-142 9.387E+07 Ru-105 7.844E+07 Ce-143 9.228E+07 Rh-105 7.183E+07 Pr-143 9.181E+07 Ru-106 5.554E+07 Ce-144 7.268E+07 Sb-127 7.194E+06 Nd-147 3.736E+07 Te-127 7.151E+06 Np-239 1.496E+09 Te-127m 9.705E+05 Pu-238 7.668E+05 Sb-129 1.976E+07 Pu-239 2.864E+04 Te-129 1.947E+07 Pu-240 6.061E+04 Te-129m 2.890E+06 Pu-241 1.281E+07 Te-131m 8.405E+06 Am-241 1.702E+04 I-131 5.564E+07 Cm-242 6.669E+06 Xe-131m 6.192E+05 Cm-244 2.358E+06 Te-132 7.739E+07 VYNPS DSAR Revision 0 6.0-19 of 37 Table 6.3.4 Undecayed Gap Activity Available for Release from Fuel Assembly Drop in the SFP Nuclide Damaged Fuel-Rod Gap Source Term for FHA (Ci Available for Release from 98 Damaged Fuel Rods) Assembly Drop in SFP Kr-83m 1.977E+03 Kr-85 4.713E+02 Kr-85m 4.530E+03 Kr-87 9.060E+03 Kr-88 1.281E+04 Kr-89 1.616E+04 Xe-131m 1.481E+02 Xe-133 2.698E+04 Xe-133m 8.199E+02 Xe-135 1.086E+04 Xe-135m 5.338E+03 Xe-137 2.366E+04 Xe-138 2.356E+04 Br-83 1.977E+03 Br-85 4.482E+03 I-131 2.129E+04 I-132 1.889E+04 I-133 2.703E+04 I-134 2.999E+04 I-135 2.514E+04 Rb-86 1.433E+02 Cs-134 1.705E+04 Cs-136 4.364E+03 Cs-137 6.808E+03 Te-131m 3.216E+03 Te-132 1.851E+04 Te-133 1.579E+04 Te-133m 1.075E+04 VYNPS DSAR Revision 0 6.0-20 of 37 Table 6.3.5 Typical Iodine Decontamination Factors and Iodine Speciation Versus Water Depth above Dropped Assembly SFP Water Depth (H b) above Dropped Assembly Iodine Decontamination Factor Iodine Speciation above Pool Water (%) (ft) (cm) Inorganic (DFinorg , Eq. 5) Total (DF total , Eq. 6) Inorganic (Sinorg , Eq. 7) Organic (S org , Eq. 8) 23 701.0 285.0 199.9 70.0 30.0 22.5 685.8 252.0 183.1 72.5 27.5 22 670.6 222.9 167.2 74.9 25.1 21.5 655.3 197.1 152.3 77.2 22.8 21 640.1 174.3 138.4 79.2 20.8 20.5 624.8 154.2 125.4 81.2 18.8 20 609.6 136.3 113.3 83.0 17.0 19 579.1 106.6 92.1 86.2 13.8 18 548.6 83.4 74.2 88.9 11.1 17 518.2 65.2 59.5 91.1 8.9 16 487.7 51.0 47.5 92.9 7.1 15 457.2 39.9 37.7 94.3 5.7 14 426.7 31.2 29.9 95.5 4.5 13 396.2 24.4 23.6 96.5 3.5 12 365.8 19.1 18.6 97.2 2.8 11 335.3 14.9 14.6 97.8 2.2 10 304.8 11.7 11.5 98.3 1.7 0 0.0 1.0 1.0 99.85 0.15

VYNPS DSAR Revision 0 6.0-21 of 37 Table 6.3.6 Atmospheric Dispersion Factors for the Postulated FHA (From VYC-2299 and VYC-2275, Section 6)

Scenario Release Point Receptor Point Post-FHA Interval(a) /Q (b) (sec/m 3) Control Room Fresh Air Intake Instantaneous release 6.04E-05 FHA Spent Fuel Pool Ground Level Release (RB blowout panel) Open Containment EAB Instantaneous release 1.69E-03 Control Room Fresh Air Intake 0 - 2 hrs 5.89E-03 2 - 8 hrs 1.53E-03 8 - 24 hrs 6.41E-04 24 - 96 hrs 6.64E-04 96 - 720 hrs 5.10E-04 FHA Spent Fuel Pool Elevated Release (Main stack) Closed Containment (w/wo SGTS) EAB Instantaneous release 1.35E-04 VYNPS DSAR Revision 0 6.0-22 of 37 Table 6.3.7 EAB TEDE Dose vs. Long Decay Time (Assembly Drop in SFP with Closed Containment)

(Instantaneous elevated release with closed containment, with and without SGTS filtration)

Post-FHA Time (days) EAB TEDE Dose (rem) Without SGTS FiltrationWith SGTS Filtration 30 2.055E-02 1.028E-03 40 8.672E-03 4.336E-04 50 3.662E-03 1.831E-04 60 1.546E-03 7.731E-05 VYNPS DSAR Revision 0 6.0-23 of 37 Figure 6.3-1 VY FHA - EAB TEDE Dose vs. Decay Time (Assembly Drop in SFP with Open Containment)

VYNPS DSAR Revision 0 6.0-24 of 37 Figure 6.3-2 VY FHA - MCR TEDE Dose vs. Decay Time (Assembly Drop in SFP with Open Containment)

15 E' Q) ..=.. Q) 10 (f) 0 0 UJ Cl UJ 1-0::: () ::E ia 5 -o I 0 ("") 0 * "' I""'L 0 " ' ...... " _roo r--......_ ..._.__ ---. 2 4 6 8 10 12 14 16 18 20 P ost-S hutdown Deca y T im e (d ays) ---+---MCR (3700 cfm in take) Limn VYNPS DSAR Revision 0 6.0-25 of 37 6.4 Site Events Evaluated 6.4.1 High Integrity Container (HIC) Drop Event The drop of a HIC containing reactor water cleanup (RWCU) resins was evaluated as taking place during normal operation of the plant, and the results are

reported in this section. Although these types of resins are no longer

expected to be on site after a period of time subsequent to cessation of power

operations (they will no longer be generated), the source term from these

resins is expected to bound source terms from other items (spent fuel pool

demineralizer resins, filter cartridges, etc.) that may be placed in containers

and moved subsequent to permanent shutdown. 6.4.1.1 Analytical Methodology The list of radionuclides released into the cloud following the postulated resin fire is provided in Table 6.4.1. The basis for this table is provided in

Section 6.4.1.2. The release was assumed to be instantaneous. Radiation doses

were calculated to the total body due to cloud submersion and a 2-hr direct

shine dose from standing on contaminated ground, and to the thyroid and

identified critical organ (lung) based on the inhalation pathway.

The whole body and organ doses were based on the standard equations for

instantaneous releases and the applicable dose conversion factors. The DCFs

were extracted from NUREG/CR-1918 (ORNL/NUREG-79) (Reference 6.5-23) for the

air submersion pathway, Regulator Guide 1.109 (Reference 6.5-24) for the

inhalation pathway and all nuclides except I-129, ICRP-30 (Reference 6.5-25)

for the inhalation pathway and I-129, and Regulator Guide 1.109 (Reference 6.5-

26) for the contaminated ground-shine pathway.

With respect to the whole body dose from ground deposition, the analysis was

based on assuming uniform dispersion of the released activity from Table 6.4.1

over the deposition area, and a 2-hr radiation exposure interval. The

deposition area (about 1400 m

2) was conservatively assumed to encompass the distance between the reactor building and the closest receptor at the site

boundary and a 2-sigma plume width for the assumed prevailing atmospheric

stability (F) at the time of the postulated incident. 6.4.1.2 Assumptions Sandia National Laboratory has conservatively estimated, for a severity Category 3 transportation accident (which includes 99% of urban and 94% of

rural accidents), no more than 1% (0.01) of any package contents would be

released. For the purposes of the analysis, it was assumed that 0.5% of the

released activity becomes aerosolized as a result of the fire.

VYNPS DSAR Revision 0 6.0-26 of 37 A HIC of 150 feet 3 capacity contains dewatered reactor water cleanup (RWCU) resins at a density of 0.8 (g/cc), and contains all radionuclides typically

found in nuclear power plant radwaste. Each radionuclide inventory in the HIC

is at the Department of Transportation (DOT) limit for Low Specific Activity (LSA) material, except for I-129, which is assumed to be at the 10CFR61 limit

for disposal. A source term of RWCU resins is considered to be the most

limiting from a radiological perspective.

The assumed liner drop occurs 250 meters from the site boundary (EAB). This is

based on original analysis performed for a drop of a HIC at the corner of the

waste storage pad (corner closest to the site boundary), built for

prefabricated concrete storage modules. This is a conservative assumption

because the radwaste loading area is farther away from the closest site

boundary than the 250 meters in the original HIC drop analysis.

Conservative dispersion conditions are assumed for a 'puff release' under

Stability Class F and a wind-speed of 1 meter/second. The puff is assumed to

travel along the ground in the direction of the nearest site boundary, at

ground level.

The dose acceptance criteria were set equal to "a small fraction" of the 10 CFR

100 dose limits of 25 rem whole body and 300 rem thyroid (i.e., to 10% of these

values, or 2.5 rem whole body and 30 rem thyroid). Because of the nature of

the source term (which consists mostly of long-lived radionuclides), the

thyroid limit of 30 rem was also applied to the critical organ (identified to

be the lung in this case).

Other assumptions are contained in the footnotes in Table 6.4.1. 6.4.1.3 Inputs The source term for the dropped container containing RWCU dewatered resins is provided in Table 6.4.1.

The atmospheric dispersion factor is based on a conservative downwind distance

of 250 meters (to the closest site boundary from the reactor building, and is

determined to be 0.079 sec/m

3.

The breathing rate for the organ dose is 8000 m 3/yr (2.537E-04 m 3/sec), from RG 1.109. VYNPS DSAR Revision 0 6.0-27 of 37 6.4.1.4 Radiological Consequences/Results 10% of 10CFR100 Dose Acceptance Criteria (rem) Calculated Dose (rem) EAB (2 hours) 2.5 rem (whole body)6.52E-03 (a) 9.59E-03 (b) 16.1E-03 (c) 30 rem (thyroid, also applied to the critical organ) 2.03E-03 (thyroid) 4.58 (lung) (a) Dose from standing on contaminated ground (2-hr exposure) (b) Dose from cloud passage overhead due to resin fire and aerosol release (c) Sum of ground plane external plus airborne from cloud VYNPS DSAR Revision 0 6.0-28 of 37 Table 6.4.1 HIC Drop Source Term Release Activity Nuclide 1 A2 Values 2 (Ci) LSA Limit 3 (mCi/gm) Total Activity 4 (Ci) Liner Drop Release Activity 5 (Ci) Cr-51 600 0.3 1020 0.051 Mn-54 20 0.3 1020 0.051 Fe-55 1000 0.3 1020 0.051 Co-58 20 0.3 1020 0.051 Co-60 7 0.3 1020 0.051 Fe-59 10 0.3 1020 0.051 Ni-59 900 0.3 1020 0.051 Ni-63 100 0.3 1020 0.051 Sb-124 5 0.3 1020 0.051 Zn-65 30 0.3 1020 0.051 Ag-110m 7 0.3 1020 0.051 Sr-89 10 0.3 1020 0.051 Sr-90 0.4 0.005 17 0.00085 Zr-95 20 0.3 1020 0.051 NB-95 20 0.3 1020 0.051 Tc-99 25 0.3 1020 0.051 I-129 6 2 NA 0.34 0.000017 Cs-134 10 0.3 1020 0.051 Cs-137 10 0.3 1020 0.051 Ce-141 25 0.3 1020 0.051 Ce-144 7 0.3 1020 0.051 Pu-238 0.003 0.0001 0.34 0.000017 Pu-239/240 0.002 0.0001 0.34 0.000017 Am-241 0.003 0.0001 0.34 0.000017 Cm-242 0.2 0.005 17 0.00085 Cm-243/244 0.01 0.0001 0.34 0.000017 19415.7 0.970785 Footnotes: 1- Nuclide Listing: A listing of radionuclides that typically are determined by laboratory analysis to be present in RWCU resin. Short lived gaseous and volatile radionuclides are not detected in typical radwaste streams. 2 - A2: For informational purposes, quantities of normal form (not special form) radionuclides, expressed in curies, permitted by DOT to be contained in a Type A disposal package. Refer to 49CFR173.435 for listing. 3 - LSA Limit: DOT determined Low Specific Activity concentration limit, expressed in units of millicuries per gram of material. Under regulations dated January 1989, LSA is a function of the tabulated A2 variable above. Refer

to 49CFR173.403(n)(4) for the relationship. 4 - Total Activity: Because concentration and distribution of radionuclides in waste are expected to vary over time, it is assumed for purposes of this radiological accident analysis that all radionuclides are at their upper limit. In reality, a small number of radionuclides might be expected to approach a limiting condition while the

majority would be at some lower level. Total activity is based on the following: A) 150 ft 3 (4.25 m3) liner waste, density of 50 lb/ft 3 = 4.248E+06 cc @ 0.8 gm/cc giving 3.40E+06 gm. B) Each nuclide is at the LSA limit. 5 - Release Activity: The quantity of each nuclide assumed to be released from the waste liner to form the source term. The release activity is based on: A) Liner drop incident results in liner failure and release of 1% total

contents. B) Of the 1% material released, 0.5% is aerosolized to form a "release cloud" source term. The

release fraction is 0.01 and the aerosol fraction is 0.00005 of the total HIC activity). 6 - I-129 is limited by 10CFR61 burial requirements rather than DOT. The class C disposal limit for I-129, as listed in 10CFR61.55, Table 1, is 0.08 Ci/m 3 (or Ci/cc). VYNPS DSAR Revision 0 6.0-29 of 37 6.5 References

1. BVY 15-001, "Certifications of Permanent Cessation of Power Operations and Permanent Removal of Fuel from the Reactor Vessel, Vermont Yankee

Nuclear Power Station", January 12, 2015.

2. Holtec International Final Safety Analysis Report for the Hi-Storm 100 Cask System, Revision 4.

3. Code of Federal Regulations Title 10 Part 50.67 (10CFR50.67), Accident Source Term

4. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Rev. 0, July

2000 5. EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents (1991)

6. 10CFR50, Appendix I, Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low as is Reasonably

Achievable" for Radioactive Material in Light-Water-Cooled Nuclear Power

Reactor Effluents

7. BVY 13-097, "Technical Specifications Proposed Change No. 306 - Eliminate Certain ESF Requirements during Movement of Irradiated Fuel", Nov. 14, 2013.

8. NVY 15-013, Vermont Yankee Nuclear Power Station - Issuance of Amendment to Renewed Facility Operating License RE: Eliminate Operability

Requirements for Secondary Containment When Handling Sufficiently Decayed

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Nuclear Power Plants Based on the Classical and Alternative-Source-Term

Methodologies" (Aug. 2008) [See also AREVA Document 2A4.26-2A4-ELISA2-

2.4_Users_Manual-000, "ELISA-2 Version 2.4 User's Manual - Revision 2".]

11. EPA 520/1-88-020, Federal Guidance Report No. 11, "Limiting Values of Radionuclide Intake and Air Concentration, and Dose Conversion Factors for VYNPS DSAR Revision 0 6.0-30 of 37 Inhalation, Submersion, and Ingestion" (ORNL, September 1988)

12. EPA 402-R-93-081, Federal Guidance Report No. 12, "External Exposure to Radionuclides in Air, Water, and Soil" (ORNL, September 1993)

13. US NRC Regulatory Issue Summary (RIS) 2006-04, "Experience with Implementation of Alternative Source Terms", March 2006.

14. G. Burley, "Evaluation of Fission Product Release and Transport for a Fuel Handling Accident," U.S. NRC Technical Paper (October 1971, Accession Number: 8402080322 in ADAMS or PARS).

15. General Electric Standard Application for Reactor Fuel, GESTAR II (Supplement for the Unite States) Licensing Topical Report, pages US-25

through US-28, NEDE-24011-P-A-14-US, Class III14, June 2000.

16. ENTERGY Calculation VYC-2299, "Radiological AST Fuel Handling Accident Analysis [PSAT 3019CF.QA.05, Rev. 0]" (Jun. 2003)

17. ENTERGY Calculation VYC-2260, "Bounding Core Inventories of Actinides and Fission Products for Design-Basis Applications at 1950 MWt" (Rev. 0, Feb.

2003) 18. ENTERGY Calculation VYC-2206, "Determination of Number of Damaged Fuel Rods due to Refueling Accident", Rev. 0

19. ENTERGY Calculation VYC-3187, "Fuel Handling Accident Supplemental Analysis (Specific to the Spent Fuel Pool", Rev. 0

20. ENTERGY Calculation VYC-2275, "Control Room Air Intake X/Q Due to Release from Reactor Building Blowout Panel Using Arcon96 Methodology" (Rev. 0, April 2003)

21. ENTERGY Calculation VYC-2302, "Radiological AST LOCA Analysis" [PSAT 3019CF.QA.08, Rev. 2]

22. ASME Steam Tables, Sixth Edition

23. NUREG/CR-1918 (ORNL/NUREG-79), Dose Rate Conversion Factors for External Exposure to Photons and Electrons (August 1981) 24. Regulatory Guide 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance

with 10 CFR Part 50, Appendix I (Revision 1, October 1977), Table E-7, Inhalation Dose Factors for Adults, Thyroid and Lung VYNPS DSAR Revision 0 6.0-31 of 37

25. ICRP-30, Limits for Intake of Radionuclides by Workers , Supplement 1, pg 202 26. Regulatory Guide 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance

with 10 CFR Part 50, Appendix I (Revision 1, October 1977), Table E-6, External Dose Factors for Standing on Contaminated Ground, Total Body VYNPS DSAR Revision 0 6.0-32 of 37 6.6 Appendices Appendix A - Fuel Damage from Assembly Drop onto SFP Fuel Racks BACKGROUND

Damage due to a drop of a fuel assembly within the SFP was evaluated using the GESTAR II method (Reference 6.4-15). In GESTAR II Section S.2.2.3.5, GE presents

a methodology for estimating the number of fuel rods that fail as a result of a

fuel handling accident where one fuel bundle is dropped onto a grouping of

additional fuel bundles. This methodology was used to assess the fuel rod damage

impacts of a FHA in the spent fuel pool where one fuel bundle is dropped onto a

second fuel bundle contained within the SPF racks. Figure A-1 shows the pertinent dimensions for the SFP. Critical to the analysis is the maximum drop distance for an assembly, 22.74 inches (1.9 feet), increased to 3 ft to account for the difference in elevations between the top of the racks and

the top of assembly within the racks, as well as any other dimensional uncertainties VYNPS DSAR Revision 0 6.0-33 of 37 Figure A-1 Dimensions of SFP, Fuel Rack and Fuel Handling Equipment METHODOLGY

Key inputs to the analysis are as shown Table 6.3.2.

The basic steps involved in the GESTAR II evaluation methodology are described in

the following paragraphs. Because this methodology is an approximation of the

behavior of an extremely complex series of mechanical interactions, a commentary

describing the explicit and implicit assumptions involved in each step is provided

with description of how each step is implemented.

VYNPS DSAR Revision 0 6.0-34 of 37

1. Calculate the available impact energy as a result of the initial drop.

The GESTAR II methodology is a simplified energy balance approach which examines the kinetic energy of the falling fuel bundle and associated fuel

handling equipment, and from that calculates the number of failed fuel rods

based on an established fuel rod failure energy threshold. The kinetic

impact energy is taken to be equal to the initial potential energy of the

dropped bundle system, that is, the weight of the fuel bundle (with

associated fuel handling equipment weights) multiplied by the drop height.

This assumption avoids the complexities in establishing an expected impact

velocity for the fuel bundle falling through a fluid exerting a significant

drag force. This assumption will result in a conservatively high estimated

impact energy.

2. Calculate the energy available to fail rods.

The calculated impact energy is apportioned amongst the various components

involved in the impact event. One half of the energy is assumed to be

absorbed by the dropping fuel bundle and one half is assumed to be absorbed

by the stationary fuel bundle. Within the stationary fuel bundle, the impact

energy is further divided between the cladding and the remaining structural (non-fuel pellet) material of the fuel bundle. This division is made by

using the ratio of the mass of the fuel cladding to the mass of the

remaining fuel bundle structure. For the GE fuel designs evaluated in the

GESTAR II analysis, the maximum ratio is 0.510.

3. Calculate the energy required to fail one fuel rod.

In the GESTAR II evaluation, detailed for GE-13 fuel, each fuel rod in the

stationary fuel bundles is expected to fail upon absorbing 200 ft-lb of

energy based on a failure criteria of 1% uniform plastic deformation. In

that evaluation, every rod in the dropped bundle is assumed to fail due to

excessive bending moments imposed on the lower tie plate. In VYC-2206, the

failure threshold for the stationary bundles was calculated to be 167 ft-lb

for the GE14 fuel design using a simple scaling of clad cross-sectional

area. Though no details of the derivation of this threshold are provided in

the GESTAR II documentation, it is anticipated that sufficient conservatism

is built in to use this value, particularly given the assumption of complete

fuel rod failure in the dropped bundle and the conservative apportionment of

impact energies amongst the fuel rods described below.

VYNPS DSAR Revision 0 6.0-35 of 37

4. Calculate the number of failed fuel rods as a result of the initial drop.

The number of fuel rods expected to fail as a result of the initial drop are

calculated simply as the energy available to fail fuel rods divided by the

energy required to fail one fuel rod. This method is conservative in that it

assumes that no deformation energy is "wasted" by deforming additional fuel

rods, though not to the point of failure. In other words, no other fuel rods

provide any support for the fuel rods that fail.

5. Calculate the Number of additional fuel rod failures due to the secondary, tip-over impact.

Added to the number of fuel rods failed as a result of the initial impact is

the number of rods expected to fail as the dropped fuel bundle tips over and

impacts additional fuel bundles in the core. Since the analysis considers a

drop within the spent fuel pool, and the tops of the fuel bundles being

stored are below the tops of the spent fuel racks, no additional bundle

impacts result from the tip-over portion of the accident.

ASSUMPTIONS

• All rods in the dropped bundle are conservatively assumed to fail. The number of failed rods for impacted bundles are established from an estimate

of "energy required to fail a fuel rod," given the size (thickness and

diameter) and thus the strength of the GE14 rod cladding, which is slightly

thinner than that for the GE-13 fuel (Reference 6.4-18, VYC-2206, Attachment

2). A strength for the GE14 10x10 using a ratio of cross-sectional areas

for GE-9 8x8 and GE-13 9x9 fuel arrays is determined (Reference 6.4-18, VYC-

2206. Attachment 2). Strength (necessary compression failure energy) for

the GE14 10x10 bundles was estimated to be 170.3 ft-lbs. As a conservative

estimate, the value of 167 ft-lb was used.

• Fuel rods in an assembly are assumed to fail by 1% strain in compression (Reference 6.4-15). It is expected that a GE14 fuel rod will absorb 167 ft-

lbs of energy in this failure mode. Since the partial length rods are not

attached to, and do not contact the upper tie plate, less energy should be

transferred to them if the fuel assembly is impacted. Therefore, assuming

that the part length rods fail at these force levels (if they do fail) is

conservative.

• All fuel assemblies employed in the analysis were assumed to be discharged at the same time so as to maximize the released radioactivity, as the noble gases

and halogens in older assemblies have already decayed to insignificant levels.

VYNPS DSAR Revision 0 6.0-36 of 37

• The GESTAR II (Reference 6.4-15) analysis considers dropping a fuel bundle drop, including a grapple mast and head. The added weight of the grapple

hook assembly and the fuel handling mast are added to the weight of the

dropped fuel bundle.

• All rods in the dropped assembly are conservatively assumed to fail. The GESTAR II (Reference 6.4-15) makes this assumption.
• Toppling of the dropped assembly is assumed to result in no further damage to assemblies in the rack due to the configuration of the rack, which

expends above the top of the assemblies themselves. Conservatism relative to the results of impact is provided by assembly features that

should absorb some impact energy generated by a dropped assembly:

• The raised rectangular lifting-handle assembly at the top of the assembly, which may bend or fracture, dissipating impact energy,
• Expansion springs within the assembly
• The semi-circular insertion guide (assembly base extension) that stands out at the bottom of the assembly and that are used to guide it into the centering

socket at the base of the fuel rack slot. RESULTS Results are summarized below for the failure of fuel rods in the impacted GE14 and

GNF2 10x10 assemblies. Most entries in this table are from Table 6.3.2;

calculated values include their derivation basis, in parentheses. All 92 rods in

the dropped assembly are assumed to fail. Impact of the dropped assembly in the

fuel rack results in an additional 5 GE14 failed rods (for a total of 97 failed

rods due to the accident), and in 6 GNF2 failed rods (for a total of 98). The

number of failed fuel rods for the drop in the SFP will be conservatively based on

98 assemblies, thus representative for both fuel assembly types. VYNPS DSAR Revision 0 6.0-37 of 37 Description Assembly Type GE14 GNF2Maximum drop height (ft) 33 Weight of 10x10 fuel assembly and channel submerged in water (lb) 569.5 580 Dry weight mast & grapple (lb) (a)707.6 707.6 Weight of fuel assembly/channel submerged in water + mast/grapple dry weight (lb) 1277.1 (569.5 + 707.6) 1287.6 (580.0 +707.6) Available Impact Energy (wt x height, ft-lbs) 3831.3 (3

• 1277.1) 3862.8 (3
• 1287.6)

Energy absorbed by dropped fuel assembly (ft-

lbs) 1915.7 (3831.3 / 2) 1931.4 (3862.8 / 2) Energy absorbed by stationary fuel bundle (ft-

lbs) 1915.7 1931.4 Percent energy for stationary clad deformation (GESTAR II) 0.51 0.51 Energy to stationary fuel bundle for clad deformation (ft-lbs) 977.0 (1915.7

• 0.51) 985.0 (1931.4
• 0.51) Failed rods in dropped assembly (assumed all)92 92Energy required to damage stationary fuel rods167 157 1st impact damaged stationary fuel rods (analytical value) 5.9 (977.0 / 167) 6.3 (985.0 / 157)

Total number failed rods (Note: Damaged

stationary rod numbers were rounded down to

whole rods since there can be no partial rod

damage.) 97 (92+5) 98 (92+6) (a) The mast and grapple were conservatively assigned their dry weight since the SFP geometry permits only their partial submersion during the assembly drop.

The dry weight of 707.6 lbs. was calculated based on a wet weight of 619

lbs. (Table 6.3-2, Item #B4), a density of 7.85 g/cc for the mast and

grapple (i.e., that of iron) and a density of 0.983 g/cc for water at 140 o F from ASME Steam Tables, Sixth Edition (Reference 6.4-22): 619

• 7.85 /

(7.85 - 0.983) = 707.6 lb.

VYNPS DSAR Revision 0 7.0-1 of 16 AGING MANAGEMENT TABLE OF CONTENTS

Section Title Page 7.1 SUPPLEMENT FOR RENEWED OPERATING LICENSE ............................. 2 7.2 AGING MANAGEMENT PROGRAMS AND ACTIVITIES .............................. 3 7.2.1 Buried Piping Inspection Program ............................ 3 7.2.2 Diesel Fuel Monitoring Program .............................. 4 7.2.3 Fire Protection Program ..................................... 4 7.2.4 Fire Water System Program ................................... 4 7.2.5 Instrument Air Quality Program .............................. 5 7.2.6 Non-EQ Inaccessible Medium-Voltage Cable Program ............................................................ 5 7.2.7 Oil Analysis Program ........................................ 6 7.2.8 Periodic Surveillance and Preventive Maintenance Program ..................................................... 6 7.2.9 Service Water Integrity Program ............................. 7 7.2.10 Structures Monitoring - Masonry Wall Program. ................ 7 7.2.11 Structures Monitoring - Structures Monitoring Program ..................................................... 7 7.2.12 System Walkdown Program ..................................... 7 7.2.13 Water Chemistry Control - Auxiliary Systems Program ..................................................... 8 7.2.14 Water Chemistry Control - BWR Program ....................... 8 7.2.15 Water Chemistry Control - Closed Cooling Program

............................................................ 8 7.2.16  Bolting Integrity Program ................................... 8 7.2.17  Metal Enclosed Bus Inspection Program ....................... 8 7.2.18  Bolted Cable Connections Program ............................ 9 7.2.19  Neutron Absorber Monitoring Program ......................... 9

7.3 REFERENCES

.......................................................... 10 7.4 LIST OF LICENSE RENEWAL COMMITMENTS ................................. 11

VYNPS DSAR Revision 0 7.0-2 of 16

7.1 SUPPLEMENT FOR RENEWED OPERATING LICENSE

The Vermont Yankee Nuclear Power Station (VYNPS) license renewal application (LRA) (Reference 7.3.1) and information in subsequent related correspondence provided sufficient basis for the NRC to make the findings required by 10 CFR 54.29 (Final Safety Evaluation Report) (References 2, 3 and 4). As required by 10 CFR 54.21(d), this UFSAR supplement contains a summary description of the programs and activities for managing the effects of aging. VYNPS DSAR Revision 0 7.0-3 of 16 7.2 AGING MANAGEMENT PROGRAMS AND ACTIVITIES The integrated plant assessment for license renewal identified aging management programs necessary to provide reasonable assurance that components within the scope of license renewal will continue to perform their intended functions consistent with the current licensing basis (CLB). This section describes the aging management programs and activities that will be required during the period of wet fuel storage.

VYNPS quality assurance (QA) procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR 50, Appendix B. The Quality Assurance Program applies to safety-related structures and components. Corrective actions and administrative (document) control for both safety-related and non-safety related structures and components are accomplished per the existing VYNPS corrective action program and document control program and are applicable to all aging management programs and activities that will be required during the period of wet fuel storage. The confirmation process is part of the corrective action program and includes reviews to assure that proposed actions are adequate, tracking and reporting of open corrective actions, and review of corrective action effectiveness. Any follow-up inspection required by the confirmation process is documented in accordance with the corrective action program. The corrective action, confirmatory process, and administrative controls of the (10 CFR Part 50, Appendix B) Quality Assurance Program are applicable to all aging management programs and activities that will be required during the period of wet fuel storage. 7.2.1 Buried Piping Inspection Program The Buried Piping Inspection Program includes (a) preventive measures to mitigate corrosion and (b) inspections to manage the effects of corrosion on the pressure retaining capability of buried carbon steel, stainless steel, and gray cast iron piping components. Preventive measures are in accordance with standard industry practice for maintaining external coatings and wrappings. Buried components are inspected when excavated during maintenance. Inspections of carbon steel piping segments of standby gas treatment and service water systems will be performed every 10 years. Each of these direct visual

inspections following excavation will cover the entire circumference of at least ten linear feet of piping. Two inspections will cover at least 8% of the total length of in-scope buried fuel oil piping (-40 feet) at least once every 10 years. If trending within the corrective action program identifies susceptible locations, the areas with a history of corrosion problems are evaluated for the need for additional inspection, alternate coating, or replacement. The number of inspections will be increased if the results of soil samples indicate aggressive soil conditions.

VYNPS DSAR Revision 0 7.0-4 of 16 7.2.2 Diesel Fuel Monitoring Program The Diesel Fuel Monitoring Program entails sampling to ensure that adequate diesel fuel quality is maintained to prevent plugging of filters, fouling of injectors, and corrosion of fuel systems. Exposure to fuel oil contaminants such as water and microbiological organisms is minimized by periodic draining and cleaning of tanks and by verifying the quality of new oil before its introduction into storage tanks.

7.2.3 Fire Protection Program The Fire Protection Program includes a fire barrier inspection and a diesel- driven fire pump inspection. The fire barrier inspection requires periodic visual inspection of fire barrier penetration seals, fire barrier walls, ceilings, and floors, and periodic visual inspection and functional tests of fire rated doors to ensure that their functionality is maintained. The diesel-driven fire pump inspection requires that the pump be periodically tested to ensure that the fuel supply line can perform its intended function. Corrective actions, confirmation process, and administrative controls in accordance with the requirements of 10 CFR Part 50 Appendix B are applied to the Fire Protection Program.

7.2.4 Fire Water System Program The Fire Water System Program applies to water-based fire protection systems that consist of sprinklers, nozzles, fittings, valves, hydrants, hose stations, standpipes, and aboveground and underground piping and components that are tested in accordance with applicable National Fire Protection Association (NFPA) codes and standards. Such testing assures functionality of systems. Also, many of these systems are normally maintained at required operating pressure and monitored such that leakage resulting in loss of system pressure is immediately detected and corrective actions initiated. In addition, wall thickness evaluations of fire protection piping are periodically performed on system components using non-intrusive techniques (e.g. volumetric testing) to identify evidence of loss of material due to

corrosion.

A sample of sprinkler heads will be inspected using the guidance of NFPA 25 (2002 Edition) Section 5.3.1.1.1, which states, "Where sprinklers have been in place for 50 years, they shall be replaced or representative samples from one or more sample areas shall be submitted to a recognized testing laboratory for field service testing." This sampling will be repeated every 10 years after initial field service testing.

VYNPS DSAR Revision 0 7.0-5 of 16 7.2.5 Instrument Air Quality Program The Instrument Air Quality Program ensures that instrument air supplied to components is maintained free of water and significant contaminants, thereby preserving an environment that is not conducive to loss of material. Dew point, particulate contamination, and hydrocarbon concentration are periodically checked to verify the instrument air quality is maintained. 7.2.6 Non-EQ Inaccessible Medium-Voltage Cable Program In the Non-EQ Inaccessible Medium-Voltage Cable Program, medium-voltage cables with a license renewal intended function that are exposed to significant moisture and voltage are tested at least once every six years to provide an indication of the condition of the conductor insulation. The specific test performed is a proven test for detecting deterioration of the insulation system due to wetting, such as power factor, partial discharge, polarization index, or other testing that is state-of-the-art at the time the test is performed. Significant moisture is defined as periodic exposures that last more than a few days.

Inspections for water collection in cable manholes containing inaccessible low-voltage and medium-voltage cables with a license renewal intended function will occur at least once every year. Additional condition-based inspections of these manholes will be performed based on: a) potentially high water table conditions, as indicated by high river level, and b) after periods of heavy rain. The inspection results are expected to indicate whether the inspection frequency should be modified. The manhole inspection will include direct observation that cables are not wetted or submerged, that cables/splices and cable support structures are intact, and that dewatering/drainage systems (i.e. sump pumps), if installed, and associated alarms operate properly. Inaccessible low-voltage cables (cables with operating voltage from 400 V to 2 kV) with a license renewal intended function are included in this program. Inaccessible low-voltage cables will be tested for degradation of the cable insulation prior to the period of extended operation and at least once every six years thereafter. A proven, commercially available test will be used for detecting deterioration of the insulation system for inaccessible low-voltage cables potentially exposed to significant moisture. Failure of the cable test results and manhole inspections to meet the acceptance criteria will require corrective actions. The corrective actions will address modifying the cable test frequency and the manhole inspection frequency.

VYNPS DSAR Revision 0 7.0-6 of 16 7.2.7 Oil Analysis Program The Oil Analysis Program maintains oil systems free of contaminants (primarily water and particulates) thereby preserving an environment that is not conducive to loss of material, cracking, or fouling. Activities include sampling and analysis of lubricating oil for detrimental contaminants, water, and particulates.

Sampling frequencies are based on vendor recommendations, accessibility during facility operation, equipment importance to facility operation, and previous

test results. 7.2.8 Periodic Surveillance and Preventive Maintenance Program The Periodic Surveillance and Preventive Maintenance Program includes periodic inspections and tests that manage aging effects not managed by other aging management programs. The preventive maintenance and surveillance testing activities are generally implemented through repetitive tasks or routine monitoring of facility operations. Temperatures are monitored during periodic emergency diesel generator (EDG), John Deere diesel, and control room chilled water condenser surveillance tests to verify that associated heat exchangers are capable of removing the required amount of heat, thereby managing fouling of the heat exchanger tubes.

Periodic surveillance demonstrates the ability of the standby gas treatment system to maintain a test vacuum confirming the absence of aging effects for the reactor building exterior concrete walls.

Periodic leakage testing on the reactor building railroad inner and outer doors verifies the ability of the rubber seals to perform their intended

function.

Periodic inspections using visual or other non-destructive examination techniques verify that the following components are capable of performing their intended function.

• reactor building crane, rails, and girders
• refueling platform carbon steel components
• equipment lock sliding doors
• yard concrete handholes and manholes
• EDG and John Deere diesel intake, air start, and exhaust components
• EDG intake air cooler
• John Deere diesel lube oil coolers and radiators
• housings of control room HVAC package heating and cooling coils, control room chiller, and control room chilled water condensers VYNPS DSAR Revision 0 7.0-7 of 16
• control room ventilation fan duct flexible connections
• nonsafety-related components of the diesel generator air start, and instrument air supply systems
• internal surfaces of carbon steel components in the potable water system containing untreated water
• internal surfaces of carbon steel and copper alloy components in the radwaste system containing untreated water 7.2.9 Service Water Integrity Program The Service Water Integrity Program ensures that the effects of aging on the

service water system (SWS) will be managed for the period of wet fuel storage.

The program includes opportunistic component inspections for erosion, corrosion, and blockage to verify the heat transfer capability of the safety-

related and nonsafety-related heat exchangers cooled by SWS. Chemical

treatment using biocides and periodic cleaning are used to control or prevent

fouling within the SWS heat exchangers

7.2.10 Structures Monitoring - Masonry Wall Program. The objective of the Masonry Wall Program is to manage cracking so that the evaluation basis established for each masonry wall within the scope of license renewal remains valid through the period of wet fuel storage. The program includes all masonry walls identified as performing intended functions in accordance with 10 CFR 54.4. Included walls are the 10 CFR 50.48 required walls and masonry walls in the reactor building, intake structure, control room building, and turbine building. Masonry walls are visually examined at a frequency selected to ensure there is no loss of intended function between inspections. 7.2.11 Structures Monitoring - Structures Monitoring Program Structures monitoring is in accordance with 10 CFR 50.65 (Maintenance Rule) as addressed in Regulatory Guide (RG) 1.160 and NUMARC 93-01. Periodic inspections are used to monitor condition of structures and structural components to ensure there is no loss of structure or structural component

intended function.

7.2.12 System Walkdown Program The System Walkdown Program entails inspections of external surfaces of components subject to aging management review. The program is also credited with managing loss of material from internal surfaces, for situations in which internal and external material and environment combinations are the same such that external surface condition is representative of internal surface

condition. VYNPS DSAR Revision 0 7.0-8 of 16 Surfaces that are not readily accessible, such as piping located in underground vaults, are inspected at least once every 5 years. The inspection frequencies provide reasonable assurance that the effects of aging will be managed such that applicable components will perform their intended function during the period of wet fuel storage.

7.2.13 Water Chemistry Control - Auxiliary Systems Program The purpose of the Water Chemistry Control - Auxiliary Systems Program is to manage aging effects for components exposed to treated water.

7.2.14 Water Chemistry Control - BWR Program

The objective of the Water Chemistry Control - BWR Program is to manage aging effects caused by corrosion and cracking mechanisms. The program relies on monitoring and control of water chemistry based on BWR Water Chemistry Guidelines, 2008 Revision (BWRVIP-190). EPRI guidelines in BWRVIP-190 include recommendations for controlling water chemistry in the torus, condensate storage tank, demineralized water storage tanks, and spent fuel pool.

7.2.15 Water Chemistry Control - Closed Cooling Program The Water Chemistry Control - Closed Cooling Program includes preventive measures that manage loss of material, cracking, and fouling for closed cooling water systems (emergency diesel generator closed cooling water, and chilled water). These chemistry activities provide for monitoring and controlling closed cooling water chemistry using VYNPS procedures and processes based on EPRI guidance for closed cooling water chemistry.

7.2.16 Bolting Integrity Program The Bolting Integrity Program relies on recommendations for a comprehensive bolting integrity program, as delineated in NUREG-1339, and industry recommendations, as delineated in the Electric Power Research Institute (EPRI) NP-5769, with the exceptions noted in NUREG-1339 for safety-related bolting. The program relies on industry recommendations for comprehensive bolting maintenance, as delineated in EPRI TR-104213 for pressure retaining bolting and structural bolting.

7.2.17 Metal Enclosed Bus Inspection Program Under the Metal-Enclosed Bus Inspection Program, internal portions of the isophase bus which runs between the main transformer and the unit auxiliary transformer are inspected for cracks, corrosion, foreign debris, excessive dust buildup, and evidence of water intrusion. Internal bus supports are inspected for structural integrity and cracking. Enclosure assemblies are VYNPS DSAR Revision 0 7.0-9 of 16 visually inspected for evidence of loss of material and, where applicable, enclosure assembly elastomers are inspected to manage cracking and change in

material properties.

7.2.18 Bolted Cable Connections Program The Bolted Cable Connections Program will focus on the metallic parts of the cable connections. This sampling program provides a one-time inspection to verify that the loosening of bolted connections due to thermal cycling, ohmic heating, electrical transients, vibration, chemical contamination, corrosion, and oxidation is not an aging issue that requires a periodic AMP. A representative sample of the electrical cable connection population subject to aging management review will be inspected or tested. Connections covered under the EQ program, or connections inspected or tested as part of a preventive program are excluded from aging management review. The factors considered for sample selection will be application (medium and low voltage), circuit loading (high loading), and location (high temperature, high humidity, vibration, etc.). The technical basis for the sample selected is to be documented.

This program will be implemented prior to the period of extended operation.

7.2.19 Neutron Absorber Monitoring Program The Neutron Absorber Monitoring Program is a new program that will manage loss of material and reduction of neutron absorption capacity of Boral neutron absorption panels in the spent fuel racks. The loss of material and the reduction of the neutron-absorbing capacity will be determined through coupon testing, direct in situ testing or both. Such testing will include periodic verification of boron loss through areal density measurement of coupons or through direct in situ techniques, such as measurement of boron areal density, measurement of geometric changes in the material (blistering, pitting and bulging), and detection of gaps through blackness testing.

VYNPS DSAR Revision 0 7.0-10 of 16

7.3 REFERENCES

1. VYNPS License Renewal Application

2. NUREG-1907, "Safety Evaluation Report Relating to the License Renewal of Vermont Yankee Nuclear Power Station," May 2003.

3. NUREG-1907, Supplement 1, "Safety Evaluation Report Related to the License Renewal of Vermont Yankee Nuclear Power Station," September 2009.

4. NUREG-1907, Supplement 2, "Safety Evaluation Report Related to the License Renewal of Vermont Yankee Nuclear Power Station," April 2011.

VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL COMMITMENT LIST VYNPS DSAR Revision 0 7.0-11 of 16 7.4 LIST OF LICENSE RENEWAL COMMITMENTS During the review of the VYNPS LRA by the staff of the US Nuclear Regulatory Commission (NRC), Entergy Nuclear Operations, Inc. made commitments related to aging management programs (AMPs) to manage the aging effects of

structures and components prior to the period of extended operation. The following table lists these commitments

remaining applicable following permanent cessation of operations and certification of permanent defueling. The

implementation schedules and the sources for each commitment are also provided.

ITEM COMMITMENT IMPLEMENTATION SCHEDULE LRA Section SOURCE 1 Guidance for performing examinations of buried piping will be enhanced to specify that coating degradation and corrosion are

attributes to be evaluated. March 21, 2012 B.1.1 BVY 06-009 3 The Diesel Fuel Monitoring Program will be enhanced to ensure ultrasonic thickness measurement of the fuel oil storage tank

bottom surface will be performed every 10 years during tank cleaning and inspection. Ultrasonic thickness measurement of

the fire pump diesel storage (day) tank bottom will be

performed every 10 years. March 21, 2012 B.1.9 BVY 06-009 BVY 07-018 LBDCR# FCR

26/009 4 The Diesel Fuel Monitoring Program will be enhanced to specify

that UT measurements of the fuel oil storage tank bottom

surface will have acceptance criterion in accordance with

American Petroleum Institute standard API 653 and UT

measurements of the fire pump diesel storage (day) tank bottom

surface will have acceptance criterion in accordance with Steel

Tank Institute standard STI SP001. March 21, 2012 B.1.9 BVY 06-009 BVY 07-018 BVY 10-069 BVY 11-007 8 Procedures will be enhanced to specify that fire damper frames

in fire barriers will be inspected for corrosion. Acceptance

criteria will be enhanced to verify no significant corrosion. March 21, 2012 B.1.12.1 BVY 06-009 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL COMMITMENT LIST VYNPS DSAR Revision 0 7.0-12 of 16 ITEM COMMITMENT IMPLEMENTATION SCHEDULE LRA Section SOURCE 9 Procedures will be enhanced to state that the diesel engine sub-systems (including the fuel supply line) will be observed while the pump is running. Acceptance criteria will be enhanced to verify that the diesel engine did not exhibit signs

of degradation while it was running; such as fuel oil, lube oil, coolant, or exhaust gas leakage. March 21, 2012 B.1.12.1 BVY 06-009 10 Fire Water System Program procedures will be enhanced to

specify that in accordance with NFPA 25 (2002 edition), Section 5.3.1.1.1, when sprinklers have been in place for 50 years a representative sample of sprinkler heads will be submitted to a recognized testing laboratory for field service testing. This

sampling will be repeated every 10 years. March 21, 2012 B.1.12.2 BVY 06-009 11 The Fire Water System Program will be enhanced to specify that

wall thickness evaluations of fire protection piping will be performed on system components using non-intrusive techniques (e.g., volumetric testing) to identify evidence of loss of material due to corrosion. These inspections will be performed

before the end of the current operating term and during the

period of extended operation. Results of the initial

evaluations will be used to determine the appropriate

inspection interval to ensure aging effects are identified

prior to loss of intended function. March 21, 2012 B.1.12.2 BVY 06-009 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL COMMITMENT LIST VYNPS DSAR Revision 0 7.0-13 of 16 ITEM COMMITMENT IMPLEMENTATION SCHEDULE LRA Section SOURCE 13 Implement the Non-EQ Inaccessible Medium-Voltage Cable Program as described in LRA Section B.1.17. Inspections for water accumulation in manholes containing

inaccessible low-voltage and medium-voltage cables with a

license renewal intended function will be performed at least

once every year. Additional condition-based inspections of

these manholes will be performed based on: a) potentially high

water table conditions, as indicated by high river level, and

b) after periods of heavy rain. The inspection results are

expected to indicate whether the inspection frequency should be modified. Inaccessible low-voltage cables (400 V to 2 kV) with a license renewal intended function are included in this program.

Inaccessible low-voltage cables will be tested for degradation

of the cable insulation prior to the period of extended

operation and at least once every six years thereafter. A

proven, commercially available test will be used for detecting

deterioration due to wetting of the insulation system for

inaccessible low-voltage cables. March 21, 2012 B.1.17 BVY 06-009 BVY 10-050 BVY 10-058 17 Enhance the Periodic Surveillance and Preventive Maintenance Program to assure that the effects of aging will be managed as

described in LRA Section B.1.22. March 21, 2012 B.1.22 BVY 06-009 20 Enhance the Structures Monitoring Program to specify that

process facility crane rails and girders, condensate storage

tank (CST) enclosure, CO 2 tank enclosure, N 2 tank enclosure and restraining wall, CST pipe trench, diesel generator cable trench, fuel oil pump house, service water pipe trench, man-way

seals and gaskets, and hatch seals and gaskets are included in

the program. March 21, 2012 B.1.27.2 BVY 06-009 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL COMMITMENT LIST VYNPS DSAR Revision 0 7.0-14 of 16 ITEM COMMITMENT IMPLEMENTATION SCHEDULE LRA Section SOURCE 22 Guidance for performing structural examinations of elastomers (seals and gaskets) to identify cracking and change in material

properties (cracking when manually flexed) will be enhanced in

the Structures Monitoring Program procedure. March 21, 2012 B.1.27.2 BVY 06-009

24 System walkdown guidance documents will be enhanced to perform

periodic system engineer inspections of systems in scope and

subject to aging management review for license renewal in accordance with 10 CFR 54.4 (a)(1) and (a)(3). Inspections

shall include areas surrounding the subject systems to identify hazards to those systems. Inspections of nearby systems that

could impact the subject system will include SSCs that are in

scope and subject to aging management review for license

renewal in accordance with 10 CFR 54.4 (a)(2). March 21, 2012 B.1.28 BVY 06-009 26 Procedures will be enhanced to flush the John Deere Diesel Generator cooling water system and replace the coolant and

coolant conditioner every three years. March 21, 2012 B.1.30.1 BVY 06-009 28 Revise program procedures to indicate that the Instrument Air

Program will maintain instrument air quality in accordance with

ISA S7.3 March 21, 2012 B.1.16 BVY 06-009 30 Revise System Walkdown Program to specify CO2 system inspections every 6 months. March 21, 2012 B.1.28 BVY 06-009 31 Revise Fire Water System Program to specify annual fire hydrant gasket inspections and flow tests. March 21, 2012 B.1.12.2 BVY 06-009 32 Implement the Metal Enclosed Bus Program. Details are provided in a LRA Amendment 16, Attachment 3 and

LRA Amendment 23, Attachment 7. March 21, 2012 B.1.32 BVY 06-058 BVY 07-003 BVY 06-091 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL COMMITMENT LIST VYNPS DSAR Revision 0 7.0-15 of 16 ITEM COMMITMENT IMPLEMENTATION SCHEDULE LRA Section SOURCE 33 Include within the Structures Monitoring Program provisions that will ensure an engineering evaluation is made on a

periodic basis (at least once every five years) of groundwater

samples to assess aggressiveness of groundwater to concrete.

Samples will be monitored for sulfates, pH and chlorides. March 21, 2012 B.1.27 BVY 06-009 34 Implement the Bolting Integrity Program. Details are provided in a LRA Amendment 16, Attachment 2 and

LRA Amendment 23, Attachment 5. March 21, 2012 B.1.31 BVY 06-058 BVY 07-003 BVY 06-091 35 Provide within the System Walkdown Training Program a process

to document biennial refresher training of Engineers to

demonstrate inclusion of the methodology for aging management

of plant equipment as described in EPRI Aging Assessment Field

Guide or comparable instructional guide. March 21, 2012 B.1.28 BVY 06-058 42 Implement the Bolted Cable Connections Program. Details are provided in LRA Amendment 23, attachment 7. March 21, 2012 B.1.33 BVY 07-003 BVY 07-018 46 Enhance the Diesel Fuel Monitoring Program to specify that fuel oil in the fire pump diesel storage (day) tank will be analyzed

according to ASTM D975 and for particulates per ASTM D2276.

Also, fuel oil in the John Deere diesel storage tank will be

analyzed for particulates per ASTM D2276 March 21, 2012 B.1.9 BVY 07-018 BVY 10-069

BVY 11-007 47 Enhance the Diesel Fuel Monitoring Program to specify that fuel oil in the common portable fuel oil storage tank will be

analyzed according to ASTM D975, per ASTM D2276 for

particulates, and per ASTM D2709 for water and sediment. March 21, 2012 B.1.9 BVY 07-018 BVY 10-069 BVY 11-007 49 Revise station procedures to specify fire hydrant hose testing, inspection, and replacement, if necessary, in accordance with

NFPA code specifications for fire hydrant hoses. March 21, 2012 B.1.12 BVY 07-009 VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL COMMITMENT LIST VYNPS DSAR Revision 0 7.0-16 of 16 ITEM COMMITMENT IMPLEMENTATION SCHEDULE LRA Section SOURCE 52 Implement the Neutron Absorber Monitoring Program as described in LRA Section B.1.31. Test one coupon prior to the PEO to measure B-10 areal density

and assess the geometric and physical condition of the tested

coupon. If coupons are not able to be retrieved and tested or

if coupons cannot be demonstrated representative of the Boral

in the Holtec racks, then perform neutron attenuation testing

using in-situ methods, as described in BVY 11-010, (BADGER or

blackness testing method) prior to the end of 2014. March 21, 2012 B.1.31 BVY 10-052 BVY 10-058 BVY 11-013 54 Prior to the PEO, VYNPS will inspect portions of the standby

gas treatment system buried piping. The inspections will

consist of direct visual examination of a minimum of two

sections of piping and cover the entire circumference of at

least ten linear feet of piping in each section. During the PEO, inspections of two carbon steel piping segments

in the standby gas treatment system and four carbon steel piping segments in the service water system will be performed

every 10 years if measured soil resistivity is > 20,000 ohm-cm

and the soil corrosivity index is 10 or less using AWWA C105. If the soil resistivity is < 20,000 ohm-cm or the soil corrosivity index is higher than 10 points using AWWA C105, the

number of inspections of the standby gas treatment system

buried piping will be increased to three and the number of

inspections of the service water system buried piping will be

increased to six. Each of these direct visual inspections

following excavation will cover the entire circumference of at

least ten linear feet of piping. During the PEO, two inspections covering at least 8% of the

total length of in-scope buried fuel oil piping (~40 feet) will

be performed at least once every 10 years. If the soil

resistivity is < 20,000 ohm-cm or the soil corrosivity index is

higher than 10 points using AWWA C105, the percentage of fuel

oil buried piping inspected will be increased to 12%. Soil samples will be taken prior to the period of extended operation and at least once every 10 years thereafter to

confirm the initial sample results. March 21, 2012 B.1.1 BVY 10-052 BVY 10-058 BVY 11-010 BVY 11-013

VYNPS DSAR Revision 0 A-1of216 SEISMIC ANALYSIS TABLE OF CONTENTSSectionTitlePage A.1

SUMMARY

DESCRIPTION...................................................7 A.2 SPECIFIC DESIGN ANALYSIS FOR THE DRYWELL, SUPPRESSION CHAMBER AND REACTOR BUILDING..........................................8 A.3 TURBINE BUILDING.....................................................96 A.4 PLANT STACK..........................................................99 A.4.1 Introduction................................................99 A.4.2 Description of Stack........................................99 A.4.3 Mathematical Model of Plant Stack...........................99 A.4.4 Analytical Procedures -Periods and Mode Shapes.............99 A.4.5 Modal Analysis.............................................100 A.4.6 Results....................................................102 A.4.7 Conclusion.................................................103 A.5 EARTHQUAKE ANALYSIS OF THE CONTROL BUILDING.........................108 A.6 INTAKE STRUCTURE....................................................127 A.7 Deleted.............................................................160 A.8 Deleted.............................................................160 A.9 DESCRIPTION, SCOPE, AND DESIGN METHODOLOGY USED FOR THE

REANALYSIS OF SEISMIC CLASS I PIPING SUBSEQUENT TO INITIAL OPERATION...........................................................160 A.9.1 Introduction...............................................160 A.9.2 Description of Analysis Methodology........................160 A.9.2.1 General........................................160 A.9.2.2 Method 1.......................................161

VYNPS DSAR Revision 0 A-2of216 A.9.2.3 Method 2.......................................161 A.9.2.4 Method 3.......................................161 A.9.2.5 Method 4.......................................161 A.9.2.6 Method 5.......................................161 A.9.2.7 Method6.......................................162 A.9.2.8 Deleted........................................162 A.9.2.9 Method 8.......................................162 A.9.2.10 Method 9.......................................162 A.9.3 Scope of Seismic Class I Piping Systems Modified and Methodology Used.......................................162 A.9.4 Description of Floor Amplified Response Spectra Used for SeismicReanalysis of Piping......................166 A.9.4.1 General........................................166 A.9.4.2 Floor Response Spectra Development.............166 A.9.5 Summary and Conclusions....................................167 A.10 PRIMARY STRUCTURE SEISMIC ANALYSIS..................................168 A.

10.1 INTRODUCTION

...............................................168 A.10.2 MATHEMATICAL MODEL DESCRIPTION.............................168 A.10.2.1 General Description............................168 A.10.2.2 Reactor Pressure Vessel and Internals..........169 A.10.2.2.1 Model Description...................169 A.10.2.2.2 Model Properties....................170 A.10.2.3 Shield Wall/Pedestal Model.....................170 A.10.2.3.1 Model Description...................171 A.10.2.3.2 Model Properties....................171 A.10.2.4 Drywell Model..................................172 A.10.2.4.1 Model Description...................172 A.10.2.4.2 Model Properties....................173 A.10.2.5 Reactor Building Model.........................173 A.10.2.5.1 Model Description...................173 A.10.2.5.2 Model Properties....................174 A.10.2.6 Natural Modes of Combined Model................174 A.10.2.7 Damping........................................175

VYNPS DSAR Revision 0 A-3of216 A.10.3 SEISMIC INPUT DEFINITION...................................175 A.10.3.1 Design Ground Motion...........................175 A.10.3.2 Design Time Histories..........................176 A.10.3.3 Input Location.................................176 A.10.4 METHOD OF ANALYSIS.........................................177 A.

10.5 REFERENCES

.................................................178 A.10.6a DRAWINGS...................................................180 A.10.6b DRAWINGS...................................................181

VYNPS DSAR Revision 0 A-4of216 SAFETY ANALYSIS LIST OF TABLESTable No.TitleA.10.2-1aSummary of Nodal Hydrodynamic Masses for Horizontal DirectionA.10.2-2aHorizontal Model-Nodal Masses/Elevations A.10.2-2bHorizontal Model-Nodal Masses/Elevations A.10.2-2cHorizontal Model-Nodal Masses/Elevations A.10.2-2dHorizontal Model-Nodal Masses/Elevations A.10.2-3aVertical Model-Nodal Masses/Elevations A.10.2-3bVertical Model-Nodal Masses/Elevations A.10.2-4aHorizontal Model-Beam Element Properties North South Direction A.10.2-4bHorizontal Model-Beam Element Properties North South Direction A.10.2-4cHorizontal Model-Beam Element Properties North South Direction A.10.2-5aHorizontal Model-Beam Element Properties East West Direction A.10.2-5bHorizontal Model-Beam Element Properties East West Direction A.10.2-5cHorizontal Model-Beam Element Properties East West Direction A.10.2-5dHorizontal Model-Beam Element Properties East West Direction A.10.2-6aVertical Model-Beam Element Properties A.10.2-6bVertical Model-Beam Element Properties A.10.2-6cVertical Model-Beam Element Properties A.10.2-7aSpring Element Stiffnesses-Horizontal Model A.10.2-8aSpring Element Stiffnesses-Vertical Model A.10.2-9aNatural Frequencies of the N-S Model A.10.2-10aNatural Frequencies of the E-W Model A.10.2-11aNatural Frequencies of the Vertical Model VYNPS DSAR Revision 0 A-5of216 SEISMIC ANALYSIS LIST OF FIGURES ReferenceFigure No.Drawing No.TitleA.3-1Turbine Building Bent No. 4A.3-2Turbine Building Bent No. 5A.4-1Ventilation Stack Mathematical ModelA.4-2Ventilation Stack Maximum ShearA.4-3Ventilation Stack Maximum MovementA.4-4Ventilation Stack Maximum Deflection DiagramA.6-1Plan at EL. 230.00'A.6-2Section A-A and Section B-BA.6-3Intake Structure Service Water Bay Area Mathematical Models (N-S)A.6-4Intake Structure -Service Water Bay Area -Maximum Accelerations (N-S)A.6-5Service Water Bay Area, Maximum Displacements (N-S)A.6-6Intake Structure -Service Water Bay Area, Maximum Shears (N-S)A.6-7Intake Structure -Service Water Bay Area, Maximum Moments (N-S)A.10.2-1Reactor Building Complex Longitudinal SectionA.10.2-2Reactor Building Complex Seismic Horizontal ModelA.10.2-3Reactor Building Complex Seismic Vertical ModelA.10.2-4Stabilizer SystemA.10.2-5Drywell Geometric FigureA.10.3-1AccelT.H.(0.14G), H1-Motion, DT = .01 Sec. NPT = 2049 VYNPS DSAR Revision 0 A-6of216 SEISMIC ANALYSIS LIST OF FIGURES(Cont'd) ReferenceFigure No.DrawingNo.TitleA.10.3-2AccelT.H.(0.14G),H2-Motion, DT = .01 Sec.NPT = 2201A.10.3-3AccelT.H.(0.14G),V-Motion, DT = .01 SLC. NPT = 2201A.10.3-4GESSAR II vs. NRC ARSA.10.3-5GESSAR II vs. NRC ARSA.10.3-6GESSAR II vs. NRC ARS VYNPS DSAR Revision 0 A-7of216A.1

SUMMARY

DESCRIPTIONThis appendix provides the seismic design analyses for the facilityClass I structures. Where formal reports have been made, these are incorporated

unabridged in this appendix. Where formal reports have not been made, details of the seismic design analysis are provided. VYNPS DSAR Revision 0 A-8of216A.2SPECIFICDESIGNANALYSISFORTHEDRYWELL,SUPPRESSIONCHAMBERANDREACTORBUILDING --.. -..._ H.]. SEXTOl\ & ASSOCIATES, ENGINEERS .; A :>J f R A N C I !l (' 0 * .\t P.: L 0 P *" R J( , C A L I F 0 R N lA 10/23/70 A!rLY at*aa TO t U l Ml'aiOI'f ST aDT iAii **10:1 ,.,,, 7.l.lf1 4 August 10, 1967 General El.ec:trie CCIIIp8lly Atcmio Powr Department 175 Curtoer AveJlUII SaD Jose, C&l.itornia 95125 Mr. R. B. Gile 'l'l'a.nalitted be-revitb 11 oar report on tbe aub.)ect* analya1S. 'lbe cSrywll \&1 eoupled vith the re&c'tor buil41ng &Dd &D&l.yaed tor both the empty and tloocle4 CODditiODa. B. J. Very truly your*, s.xt.m /1. B. J. Sexton A.2-l EDgi.neera REPORT ON THE DYNAMIC EARTHQUAKE ANALYSIS OF THE DtmlELL FOR THE VERK>NT YANKEE l'<<X:LEAR POWER STATION TAil report, prepe.J"ed tor the General Electric Cc:mpa.ny, preaezrta the results ot the dytlamic earthquake &Jl&l.ysia tor tbe Ve1'111011t Yankee Dryvell. The dryvell wa.s coupled with the reactor building and analysed for both the empty and flooded ccmdittcma. The maximum envelopes of-shears, DlCIIIents, placements am accelerations are presented for use by the aigaer in the determination of seismic induced stresses. DESCRIPI'IOH OF DRYWELL The drywell is a bulb shaped, velded steel plate assembly fixed to tbe building at Elevation 235.48 and sup* ported at El.evation 309.96. There are tvo appurtenances ed to the drywell; tbe vent pipes which extend into the preaal,ll'e suppresstcm ehlm'ber, &lid. the air lock. '!be vent pipes AN peDdent ot the pressure suppression cbember since they are ated fra:a it by means o£ a bellows cozmection vhich provides no support. The relationship of the dry\lell to the building and pre!leure 81,lppreaaton chamber is illustrated in Figure 1. Figure 2 shows the seometric properties and plate th1ckrless of the well. METHOD OF ANAL'YSIS The dryvell, including the tvo appurtenances, was idealized u a model coasiatillg of twenty-tour lumped masses supported by &n elastic colUIIDl. The properties ot the column were determined by computing the IIKlll!ent of inertia and the etfecti ve shear area ot tbe dryvell between mass points. The drywell \las connected to the reactor bUilding (Referenee

1) to form the coupled system in Figure 3. The properties of the model are listed in Figures 8 and 9* 10/23/70 H. J. SEXTON & ASSOCIATES, ENGINEERS A ?-?. . _ ... 1 10/23/70 A cc::aplete

&D&lyaia was pertormed tor loading coadi tiona; (l) vith the drywell empty, and (2) Vith the drywell :tlooc!ed with vater to Elevation 296.33. The mode shapes &D4 aatural pertode tor both ccm41 tiona were obtained tor the c011pled eyatem u4 :tor 1fbe dryvell alone. co led. a te.m waa aub ected to the ound. 1110t10DI o:t the design ea qualb! vi th a ma:JC1mum aecelera ion v1 y, x e1 ve:re c:OilS red 1D the &D&lyl a. 1J2i ornve per-cent V&! used tor the concrete buil.dips &D4 gp! per-cent tor the dm!ll*. A time history of the *hears, mcaenta, diaplace.enta and accelerations was obtained tor 1850 incrementa or time using &D interval of 0.0055 teeonds in the solution ot the Duhamel A time history ot the toree in the eonaec:t.ion at Elevation 309.90 vu also obtained. All or the above time histories vere enveloped to termine III&XimUm values tor both the empty and flooded coDditiona. For the t'looded cond.i tion, the bydrod;ynam1c effect ot the enclosed water vas conaervati vel.y neglected. As mentioned in Reference l, the stn.ins v1 t.hUl the t0Wldat1on material tor thiS s1 te vere very small IIDd the etteet ot rocking was found to be negliaible. DISCUSSION OF RESULTS Periods 'The *natural periods or vibration tor the drywell were tOWld to be 0.029 seconds for the empty condition and 0.101 aec:onda tor tbe flooded condition. The natural periods or vibr&tion tor tbe coupled system are liSted in the following table: MODE 1 2 3 4 5 6 EMPT'! FLOODED CONDmON CONDITION (seconds) (seconds) 0.321 0.321 0.18o 0.180 0.077 0.101 0.057 0.077 o.o44 0.058 0.038 0.045 H. J. SEXTON & ASSOCIATES , ENGINEERS A.2-3 j 10/23/70 'lbe enwlopel ot *xlww aDd ... nu aro ahow 1D ncure* a. aD4 5* '!bele ftluu illeorporate tbe ettect ot all 1e11a1e i.D4uee4 force* &cti.D& an tbe deywU 1nclu4iq 1Dtrt1& torae* aD4 lN1WDc 41aplac.-ut. !be** c\lr'Yea 1houlc1 be uaed tor t.M H1aa1c cS.e11p ot tbe dr)vell. ))1!Jilac-Dtl 'l'h* envelope or JIIII.Xi.JIIua diapl.aeementa tor the drywell 11 lhovD 1.P rtpre 6. .

• Aooelerat10G8

!be aftlope ot *r1** aceel.er&tiODI tor tbe d.eyveU 11 abow 1D r11\lft 7. Tbele n..l.uea eboul.d be uaed tor tbe d.elip ot el a r Dtl that are 1"1114J.)' &ttaehe4 to the c!rywll. rem:. 111 aM&r tue !be war lup vhieh provide lateral 1upport tor tbe dr7Wll &t JleW.tiCID )09.96 ahoul4 be tor & *z1** Mtaic tor<< ot l,JO.a. ld.pe tor tlM RP't7 cCDUtion aD1 115.9 k1p1 tor t.be flood* ed OCIDIU tiOD. H. J. S.EXTON Q ASS0ClATa8, BN01NEI!RS A.?..-4 3 10/23/70 Deacriptioc nCR.tm 1: Loositu4tn-J Sect1oo * * * * * * * * * * . * * * *

• 5 J'IOUR.e 2: Dr.Ywll -Gecme trie J'1 sure . . . . . . * . . . . . . 6 nGURE 3:

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• 12 PiatBI 9: Bu1141ns &lid DrJwll Seet1 oa Propertiea
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• Aaaoeiatea, J:D.ataeera, 4a ted Febr1&ry 26, 1967. 2. Cbioa12 Bridge &Dd Iron Calculation Sheets Sheet Dated Vei"'DD.Dt Y&Dkee lC-1 reb. 1967 VeniODt Yankee 2C*1 Feb. 1967 V*niiOD't Yazlllze 11<<1pu 1 May 1967 For Reference Only Montiee llo lC-4 s.pt. 1.966 MaDtieello lC-5 Sept. 1966 Noat i eello Sept. 1956 Mootieel..lo lC-7 Sept. 1966 3* Geoer&l. Electric Ccmpe:z:ay APED Dravings S b ee t Dated VenDODt Y&Z:Ikee 729 E 1 0 I>ec. 1966 Vermont Yankee 129 E 252 J&n. 1967 14 K. J , IEXT O N St A SS O C IATES, ENG1N"!£R S 1\. 'l.-l 2/5/71 -. H. J. SEXTON & ASSOCIATES, ENGINEERS IAN P1lANClSCO
• MENLO P.<\1lK , CALtFOI\NtA IN &aPLT &IFia TO* Ul :ICIIIION STJlaiT SAN n.ANCIICC, f4 l OJ (41S) 781*8f14 February 27, 1968 General Electric Company Atomic Power Equipment Deparanent 175 Curtner Avenue *
• r San Jose, California 95125 ATTENTION:

Mr. R. B. Glle Gentlemen: S U BJECT: Vermont Yankee Pressure Suppression Ch<!.mber, Earthquake Analysis Our report on the subject analysis, dated May 15, 1967, shows a c:alculated deflection of 4. 24 mils in the seismic anc h ors for the .flooded condition of the Because of i:s tendency :o*oval und e r seismic loads, the body ot the torus w Ul a l so deflec t relative *to its a n chorages. We calculate this ovallng defiectlon to be 15 mils. Therefore. the maximum calculated seismic defle c tion betwee n the torus and the !oUlldation elab is about 19 mlls. Very truly yours, H. J. Sexton & Associates. Engineers !IF H. J. Se xt on RJF/pb -. H.). SEXTON&. ASSOCIATES, ENGINEERS SAN PRANCISCO

• PARX. CA.LlFORNIA IN AIPt.Y aiPD TO, U.l WL.UlOM n'1111T

.... ., , ..... , ...... .. Gener&l Company Atomic Power Department 175 Avenue San Jose, California 95125 Mr. R. -B. M/C 750 May 15. 1967 SUBJBCT: Vermont Yankee Suppression Chamber. Anal.yeie Gentl.emen: transmitted herewith is our report ot subject analysis. pertinent information and are inc1uGed, aether with discussions ot the analysis and results. HJS/p'b Bnc1oeure H. J

• lt Z-t7 Very yours, 4e ociatea, fi.+"-H.
• Sexton Engineers

.-2/S/71 R2PORT ON tH!: IARtHQVA!E ANALYSIS OF THE PRESSVRE nUPPRESSION CHAMBER P'OR lHE VERMONT YANKEE NUCLEAR l'OWEB STATIQli 'rh.ie report, prepared. the Gcenora.l El.e ctric Compan;r, preeente the reeul.ts a aeiemic anal7*1* the Vermont Taukee Preeeure Suppression Chamber. I)ISORIP':lON euppreeaion chamber ie a torua-ahaped eteel an inaide diameter 27'-a*, and a =ajor diameter 98'-0". It ie supported

• • rtical.ly by '2 columna:

ie provided by pinned anchoraaoe which tranamdt eeie*do loade the chamber'* aoft'it to the concreto dation. the torue ie a in itael.t; the vente, headers, and downoommera are the ob&mber by vhich provide no eupport. Figure 2 ahowe the relationship of the torus and the reactor buildina. ANALYTICAL APpaOACH '.the tol"\le ie ideaJ.ieed ae a dear**

• Tat** vhoee aprina conotant ia dotermined of the pi.na and pl.atee the ancbor-c*

aeoea:abl.iea. the ooluane contribute a amount reeiatance because ot' the creat the tiea. "--*e* both oparatinc and conditione are 4etermined, conservatively ne&leotinc the h.ydrod.yna.mic the water contained i.n the The tot&1 eeiamic load each ie dotermined uaina one-percent provided on Pace H-2-4 of the Plant Deeicn and Ana1ysia Report. Reeulta are eummarieed in the tab1e. 1 .* . 2/517l **-

SUMMARY

OF RESULTS Oon4ition Period STiem;Lc (seconds) Operating 0.037 1016. Fl.ooded 0.051 2188. 2 .. ,. .... .,. "' J1J

• c:: c: n r f A 'T" F. s . E N G t N £ E R S j 1 l I ] * ; .J l ; 1 ' , 2/5/71 --LIST Of PIGY:YS PetoriRtion . b&!. .. PIGURE 1: Reeponee Acce1eration Spectra -----------

--4 Loncitudin&l Beotion -----------

---

5 7IGURB ): P1&D SuppreeeioD Chamber ---------------

6 7IGURE 4: Section Through Suppression Chamber -------7 PIGURE St Detail* Anchorage Aseemb1iee

---

6 fl.2-ZO ' N I f'l! -, ( ) !'; ,.. 'II :n 0 n -> "" "' en . (11 z (') -z ('!\ l"ii "' *.n ****** __ .. -"-*-......... _.. --.,...,.; ...... i ...... ..... ..... ....... '1) ....... c;) 0') ....... J\0.4 t-z ::J (J'l ,.. IP r ZO.'l. 2 cl. o.t;' Ul J _. .-v \U 4 o.l Ul oo v -Z.08 0 0..01 If} -------1-" ?_P,ECiQA VER.MC?NT JOUN t..cn:X:tATEc;,, ENOINEf.RC:, I i I . . I ---- 1.0 _..,. _,. j v .. -----/I-"..--.....-'1..0 v ,........,H . v .......... s.o .J .. . --i I . ---! , __ --------UJ ct.O (Q .01 .ot .04-.o<O .o i .o a .cqo.t 0.'\. \ ....._I-I ..... ........ I I ......_. ...... ---:--..... ...... .......... I'--"J ---'" ;-. ...... i""o \ " "' ......... ! ' ""' ..... ...... r-..... *T ""\ . I 0.? 0.4 OS O,CC, 0.1 Ob 1.0 NA. TU R.AL PE:RIOO , P4G£ .JI*E'*¢ o,r V*RMONT

1. UCl.*.4R POWeR STATION Pt..dNT

,.dNO 4N.dt-YS/S ,'<SPORT. = I .* N l. 1\) ....

/ it 1 1 .f 1 .. E;t/.!:...

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• \ i.o** .. *.* ... --*:.:< . ' .... tl. 28().00' .. :: ::: :*: .. :* .. t:J ;:*:* :: 'l _T *.-. . i ; E/. 252.50' O :,:* c_-:n. :*. I **. i .. .. ::.* ---'-:. .. .-... :**:. : * :. ..! ' .

• ' I /'

*r-* cr 1 .-.-: :-. .. 1 , .. --:(' *. n 1 n * :--, :. i ./ . : * -:*. *. .". .' :." * * : : . .': * ../I : I* --..,---. * . * ** *f .. . . . . 1---..,........ . \. * * * * * * * * .. * ... * .. * .. 'I> * .. * * .. .. ** ** *=' * -* * * * * * * . * * -I: * . . . . . .*. . * * .. * .... ' : 5 L.ONG/TUOINAL SeCTION j='/GURe 2 -* .. " ,.., "-' t: P R S I I J ] i ) ] *t J " l J .. ** ' J j ,

• l l l
• il. J. SEXTON & ASSOCIATES, ENGINEERS 2/5/71 SAN
• Wl!'<ii.O CALIFORNIA

t .... _ _L_ I I tJ. COL TYP. /G G XTE/2101<

COI.l./MNS SSG 5 PLAN PRGSSVRG SUPPReSSION Chi.4MBGR CI/: II OC ILL.-23 ' \g '.A' .... ..... 6 I I j I f J *-H. J. SEXTON & ASSOCIATES, ENGINEERS 2/5/71 IAN PR.ANCISCO

• r.t!NLO CALIFORNIA . 98'-0" M4../0R OIAMGTEI<
• ' ' . -'..--...... ---

........ , ---/ ,--' /\ I / I ', \ ---1 \ I I __).)..--* \ **-t+-*-----.\_ ----* I \A -A l GXT*RIOI<. COLUMN \ \ } ' ; / v --\ \ I ' ---1 /... I .). \, v \\ \ I . f--..:t -r-* \.-,---.---* I I /! I I I I ' I 1 I I . I I I I l l L_J L_.J p i E----1 7 COL U,W.'\' . TYPICAL SGCT/ON PReSSURE" SUPPRESSION CPAMt3ER ,t::'JGURG 4 A.2-2.'f .... ._.,-* H. J. SEXTON & ASSOCIATES, ENGJN!EllS IAN

• J:U!fLO PARl(. C:AI.IFOaNIA 4 2'TYP.

t* 2'7'1"P. SeCTION ScCT/0/'+../ .* l A-A. 2./5/11 8 H.*J. SEXTON & ASSOCIATES, ENGJNEEI\S 2/5/71 SAN fRANCISCO

• MENl.O P.\RK. CALUOaMIA 1e02 JO* VERMONT
• ..,MJ o ..

I!NT .G. G. *u..,KeT SUJ'PAGSS/ON I. STIFFNESS Or CJ.IAM£3ER slil'rness or cht!Jmh:r .svppori provicl60' hy ,; sel or lour plafes pin. Only /Ormdf/on ,t:ui? dS bend1n9 de!/ecflon is negligible.

• ELEVATION dy y t fhlckness o! ,i X
• b-2(?. 72)y A -x6 c t(h-5.44-y) .A . '-fh Vdy _t..Y_ dy k. V _t (:---k-) .u GA
• G t J . (b*S.44y) S.#Gt 'n b*5.44h . o D IJ# \.J

-.X--.1/_ b A -k..v1(b*.S .#(,) For t ..1:.= b=;os.o: fJ .. c;.s* * !(; -

• 7. 20 }( ;oS

' * -J( e.s) fJ. 2-2(., 9 H. ). SEXTON & ASSOCIATES, ENGINEERS SAM PRANCJSCO

• Mf:NI.O *. ,IU(, CALifORNIA 2/S/n JM

.10* . VERMONr YANk'EG St.*cw,. G E. .5UPPI<GSSION ,,. MJ ry.n.4*16.r-., CNC"! &/F OA1'.-M9)< 5r/FrNG5S 0,&" .SUPPOJ<r (C.ONT.) _?.,.r.n . h. Sr ;, ' ness of" oin t . . r-I i ......... ' j ...... ... . . I" ., /" -... ' . c.;' ,_ PLAN r _ k v,.{ o AG r (7"¢l'ln ..."7 1-;..r ........ 1'rY.olcCJ/ (;. For A- /n, 2 G-//.1<> A ¥o* 1:. * .t;' ,t. /. o* . (!Js.s)(//.<i-"10

8) _ .c .s v/. *** Kp = (fi)(I.O)

-4.04 10

c. s/-1/"J:,ess or each Cht:!Mrnber 5f.{aB!!2_rf . . I I I f.= + Rp = 7.20x;o5 + 4.04 tto5 = 2.58 A/a:s .... Ks * )( ;o5 A.2-Z 7 10

11. J.

& ASSOCIATES, ENGINEERS FRANCISCO .. MF.SLO CALIFORNIA

2. TORUS .ST/r,&-N555 Kr 2/5/71 F= KS f-= Ks Llr KrLlr =

.:. ltr -== all VcP/ues Oft(;; Kr-£K's = 2)( z.ssx..as = 5./'"' JC d K,!/'n .3. MASS Flooded Condillon Torf./S Shell t!Je/loJit/5

1. Rein/: Tee Pl/)tn9 .4 tlltn9-s Supporf Rin9.s Mit; C. 1-/eader Misc./Or i'OD or Colvrnr7 Aflacl>menls 1 co!uml?s Total== fl. z-s; 20, k iJ::;-::. 704. k.ipS 70{). 8

• 0 11 !09.0 52.0 107.0 2/0.0 60.0 60.0 ;3?.0

.IOe NO. /20 2 .* .,., G. E. H. ) . SEXTON & ASSOCIATES, ENGINEERS SAN PI\ANCISCO

• MF.NLO P.\RK. CALIFORNIA 2/5/71 . 4:. TORUS PGI<.'OO ¢; RGSPONS* ./* lie .71 s (Jn ,: Js

,.t:"'looder::l C ortdl I /0/') Cono'lf /on Wofer Weiqhl kips //,704-.0 Slrvc!ure wei9nl II 1,555.6 1,555. 6 ro!ol Wei9hf , W , Torus S!l/"iness Kr /IJ* 'I tn 5./GJ(/05 S./6 Per,Oa' T

• e .,. -{;-Fr 0.051 Sa g a l so 0./65 Selsmic Force' ,c'
• f ;c E!!:!Ja. . 7oru s = f /.97

-3 in. 4. Z4 x/0 .. F:s: K!; Ar Cos B

• Ks L1r /Or 12 c9. OperO',.t;n q Cond ifl on Fm.:;x. cKs-L1r lD (2.58J(IOS

b. Flooded Co£)d;:-'lon

,&;, 8)(. =Ks Llr =-(2.58 Jt./0 5 ¥tn)( 4.24 ,( k.. ft.Z-29 REFBRFNC?.S Y*rmopt Yankee Nuclesr lower Stttion Plant Design and Analyeia Report 2. Gtntr&l Electric Company. AliD Arran&ement Dravinc 729E24) Sheet 10 11 '* Qhicaco !r i dge and Iron Compa n y Anohor Bol. t Pl.a.n Drawing F1. Dated '/15/6 7 Earthquake 'l'iee Drawing 208, Reviaion 1, Da t ed 3/13/67 Section and Weight Sheet 2A1, 2, Datod 2/13/67 Sheet 2Bl, Revision 2, 2/16/67 Sheet 2B2. Revis i on 1, 2/16/67 2/Snl " * * ** ,._,, ,..,., t:Nf:T'Nti'!RS 1 3

.. H. J. SEXTON &. ASSOCIATES, ENGINEERS IAN FRANCISCO

• MENLO PAJtK , CALIFORNIA General Electric:

Company Atomic Power Equipment Department 175 Curtner Avenue San Jose, California 95125 Attention: Mr. R. B. Gile, M/C 750 IN aJPL'I auq TO* su vtanoN IAH fltAHCUC:o ..... (411) , ******* May Z.f, 1967

Subject:

Vermont Yankee Reactor Buildtng Reviaed Earthquake Analyaia Gentlemen: herewith ia the subject report which ruu been reviled in accordance with your inltructione, Since the iuu&nce of the oriainal repnr t dated February 1967, a 33'-3" x 56'-9" appenda1e baa been added to th. 1outhwe1terly corner of the building. Fi&W'U 1, A-1, A-9, appropriate: calculatiotu have been altered to include the appendaae and are duianated aa Revision 1; The buildin& wa1 reanalyzed incorporatina the change* in aection propertie* cauaed by the appenda1e. Theae chan1e1 e!!ected a reduction of from one to two percent in the ehear, moment, deflection, and acceleration cliaar&ma 1hown in the ori&inal report. Bec:au*e theae change* are 1ma.U and OD the c:onaer,.ttve aide. it h recommended that they be nealec:ted. The deaian acceleration a.t the Ooor level of the appendaae ahould be the eame u ahown for the building foundation ala.b. RJF/e Encloaure Very truly yo..ar*, A.Z-31 ENGINEERS REPORT Of! THE DYNAMIC EARTHQUAKE Ai1ALYSIS OF THE REACTOR BUILDING FOR THE VERM:>IIT YAI-JKEE POWER STATION This report presents the results of a dynamic analysis of the Rea c tor Building. In the analysis, e mathel!l&tical model of the building is subjected to the design earthquake and the time-history of 1 ts response determined with the aid of a computer system. ima data are provided for use in the aseismatic by others. BUILDING DESCRIPTION The Reactor Building, located at t he Vermont Yankee Nuclear Power Station near Vernon, Vermont, will be founded directly on competent gneiss rock at approximately Elevation 2o6 feet, Mean See. Levei datum. The building is of reinforced concrete construction . except that portion above the operating level, which is of steel construction. Figures A-1 through A-10, Appendix, show the building's general dimensions and arrangement in accordance with the Re!erenc!! Drawings. ANALYTIC CRITERIA The analysis was based on the following data included in the Station's Plant Design & Analysis Report: 1. Design Earthquake (page H-2-2): North 69° West component of the 1952 Taft, California Earthquake normalized to 0.07 gravity. Safe at twice the design earthquake.

2. Damping Factors (Table XII-2-1, page XII-2-7):

Reinforcea Concrete Structures ---Steel Frame Structures

---

H.]. SEXTON & ASSOCIATES, ENGIN'EERS It 2-32

3. Petrologic Data (page H-3-25):

Wave Velocity ---13,500 tt/sec Shear Wav& Velocity ----------- 6,500 tt/sec Young's Modul.U3 --------------- 4,l6o ksi Shear or Rigidity Modulus -----1,530 kai Pohson' a Ratio ---------

0. 347 ME'l'HOD OF ANALYSIS An equivalent mass system was selected to represent the buildln..

having nine masses equal to that of the building's operating tion. Included were dead loads, equipment weights, tuel and water weights as provided in the Referenc e Drawings. Effects 01 moment -lhea.r -and axial deformations or the building's structural syste"" are included in the stiffness matrix. Foundation springs and the elastic springs or the building form a coupled system. Tne matical model thus formulated is shown in Figure l. The model was then subjected to the design earthquake motions with the aid of an IBM 7094 digital computer and a program cally developed tor the solution of resporuse ot structures sub.1ected to random motions. This program has a capacity for modeJ.l: having up to 60 masses; its excursion c&pa.city allows 600 incremenn of time. The respoMe of each mass tor each 1110de at every increl!M!nt of time was integrated to give the exact combinati on ot mode pation, including rocking. A total of six runs were made to envel.:Jl)t the et'!ect of shear detomation, which vas tbe governing factor in .. stiffness matrix. Differences in the vari ous runs were negliginle. DISCUSSION OF RESULTS Three modes were chosen for the analysia since the effect of higher modea waa found to be of no consequence. The buildinp was rw in both the East-West and North-South directions. Vibrations The periods of vibration are: Direction E-W N-S First Mode 0.321 sec 0.319 sec Second Mode O.l8o sec 0.179 aec Third Mode 0.077 sec 0.076 sec H. J. SEXTON Q ASSOCIATES. I!NOINBERS Shears, Moments and Displacements The envelopes of maximum seismic shears, moments and in the East-West directions are presented in Figures 2, 3 and 4 re:>pectively, and in the North-South direction in Figures 6, 7 and -?. respectively. These curves are provided for the building design. A c celerations The curves in Figures 5 and 9 are maxim& va.lues for the lute accelerations in the :Eut-West and North-South directions respectively , and are provided for the aseismatic design of ment dements rigidly attached to the building. The term "rigidl.y attached" as 'used here limits the element's period to less than 0.0 5 second. Critical elements of greater periods should be dynamical..ly investigated with interaction of the equi]illlent and the building considered. Combinations I n additio , n to the horizontal values presented herein, the building and its critical ccmponenta must be designed for a. vertica.l eration 2/3 the ground acceleration. These are to be sidered to occur simultaneously and stresses added directly where applicable. A.2-34 3 LIST OF FIGURES Deac:l"iption Mathematical Model E*W Direction Muimm Shear D1agrlll Maximum Moment Diagram Max:imum D1aplac:ement Diagram Maxill1ml Acceleration Diagram N-8 Direction MaxiiiNIIl Shear Diacram MaxillllliD Di spl.acement Diagram Maximum Acceleration Diagram Plan At Elevation: 213.75' At Elevation: 252.50' At Elevation: 280.00' At Elevation: 303.00' At Elevation: 3l8.67' At Elevation: 345 At Elevation: 391.16' N-s & E-W Elevations (Steel BraeiD<<) Longitudinal Seetion Transverse Section Note: The letter "A" denotes appendix. 1 2 3 4 5 6 1 8 9 Al " A4 A5 A7 A9 Al.O 4 ** * .............. &. .. **nl"t ... T1UI. aNGlNB!RS

1. EBASCO Services Inc. R e actor Buildinc General IA.yout SK 5920

• CH -5, Datt4 l/4/6 7 Reactor Buildin<< Above Elevation 345.16' SX 5920 -NJlOl 2. General Electric CC!IR!&, APID Ar'raqement Dr&vina* 7291243 Sheets l throQab 13, Date4 12/29/67 lquip:D!nt Load & noor Loa4a 1291290 D&ted 1/31/67 3. Vtti"IDlt Yank ee Nucle&r Pover Station Plant Design and Anal.yli 1 Report H. J , SEXTON
• AS80C lA T'8 S, U ol 0 l N8ER S /1 , 2-3(,

H. ]

• SEXTON&. ASSOCIATES , ENGINeERS SAN FRANCISCO
• W tS&.O CALifOaNIA WE/G#T (KIPS) SG8 Ic dRACciJ AR£.4 11,697 4,8044itii 2.175, 0) ... 4268,550 N-5 928. t:628. MATH'EM4TIC4L MO.OEL.

/b.. /1.2--37 3, 11.3 . 2.41/ I AAASI H. J. SEXTON&. ASSOCIATES, ENGINEERS IAN F"RAH C liCO

• WEHLO I'A&X, CALIFO&HIA V*RMONT 'YANKEE 4/.IJG. 400 280 MAXIMUM 5/-leAR L)/AGRAA4 c -W .DIReCT/ON 20 A. 2

'c: 0 lu ..J H. J. SEXTON&. ENGINEERS IAN FRANCISCO

• WENLO PAJIK , CALIPOJINIA VeRMONT YANK**

CJL.t:JG. MAXIA4UJ14 MOM*NT E-W 0/,QeCT/ON --t ***** . . 320 280 240 **-200 60 /00 eo MOM*NT 10,000 KIP -FT. F IGI.IRG .! H.]. SEXTON & ASSOCIATES , ENGINEERS 8AN PltANCJIICO

• MI!NLO PAitJC, CALIFOitNIA VeRMONT IJLI)6. /vf,4X/Mf./M

/)t4(5R,4M

• -W LJIRE:CT/ON 240 200 0 FI G U RE 4 H. J. SF.XTON & ASSOCIATES, ENGINEERS IAN FRANCISCO
• W!NLO PARK, CALIFORNIA VeRMONT Y"ANA"'"ee ReACTOR MAX/MUA-1' LJ/AGRAM E-W OIRGCTION

.400 .360 280 240 0 H. J. SEXTON & ASSOCIATES, ENGINEERS IAN FRAN C ISCO

• WENLO P A aX , CALIFO'ItNIA *veA'IWONT Y.4NKEE R*ACTO/i' 8L.DG. .320 280 M.4X/I'WUM S.J.IE.A,Q N-S .DIReCTION 0 4 Sh'G'.4R I N 1 000 K/PS 6 /}.2-LJ 2 H. J. SEXTON & ASSOCIATES.

ENGINEERS IAN PR.ANCISCO

• NENLO PARK , C.ALJFOaNIA VER}40NT YANKe*

M4XIA..?'VM N-10/WE/VT /)/AGRAfrf N-S DI,J<ECT/ON 240 240 200 0 20 40 60 80 MOMC"NT IN 1(/,000 K/P-F?: H. J. SEXTON&. ASSOCIATES, ENGINEERS IAN '"-ANCISCO

• Wl!.NLO PAaK. CALifOtNIA VeRMONT YANK&£ J;'C'ACTOR L'L..OG. MAX/MUM .LJ/.SPL...4CEMGNT LJ/,4(3/i#AM N -S DII<GCT/ON

... 190 -; -'-* -* .. ... r* gt;() . * . . . " . -;:: ' J20 0 i:: 2cSO \(J -..J 240 0 /00 200 30() f<JO OISPLAC*MENT IN M/t..S F/GUR£ d H. J. SEXTON & ASSOCIATES, ENGINEERS SAN FRANCISCO

• MENLO PARK. CALIFORNIA VeRMONT Y'ANK'cc ReACTOR l3L. DG. M4X/,.1/IUM ACC*L£RAT/ON 0/.AGRAM N-.5 OIR*CTION 400 2&0 240 200 o aco o.4o aGo o.do ;,oo ACCELG'RAT/ON

/N ft. z-45 FIGURE 9 APPENDIX H. J. saXTON

• ENGINEE R

,,_, ...... . ... * *.. . ... * .. -.... I .. . *' * * '.* *. #. .. ... . .. .. . * ** .. ;

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..... . . .. *. .. . . . . . .. ; .. . .... ..... .. . . . . ...... .. * ... *.* ... :* .. **;. . . . .. :* : . .. . ...... * * , * # * . . .... * ... , .*. . . . ()I ._.__ __ .-** * ........ : t .......... -ll I 1111 .... .. .. . . .... . *. ... I .* , * & * **.\ .. * . ' . *. . .... ) , ... -.*.**-. .... , Reocfor PLAN AT E'L EV. FIGURG ,d1 ReV/S/ON _& H. J. SEXTON & A SSOC IATES. ENGIN E ERS 'N . ' 144'-0' ...... *. PLA/V AT C:L=V. 252.50 1 FIGURe .42 .... j i i R. J. SEXTON &. ASSOCIATES,

• N t m D ' \. I tl. .. : -'*I **: : . *. : D I t Pt.. 4 N 4 r E 1.. E v. 2 8 a oo 1 H. ]. SEXTON & AS SOCI ATE S, BNGINEERS

) N r-l ----' : ,* . ,. ' PLAN 4T £t.eV. 30.3. 00 1 H. ]. SEXTON 4k ASSOCIATES , l!NGINEBRS A.Z-5o 144! O" ... -.... . : ... * ........ . ; --F/GURE A.:5 (* -*---io-.---*-. >-. . 0 I ' l ..... ' H. J, SEXTON &. ASSOCIATES, ENGINEERS A.7.-'3J j ...... ,, -*N I l ' : DDDD i I H -"""" 1 I \ ' I -10 i i 1 I T 0 0 0 0 I I 1 I j ... --.. ,. ' .. 1 PLAN AT EL.tE V. 345 .. /6 H. J. SEXTON & ASSOClATES, ENGINEERS ft. 2-sz F 0"' -------* -** --I I ROOF AT ELBV. .391.16' FIGU,q£ A7 H. ]. S!XTON Ci A SSOC lA TBS, ENGINE1U\I Typ. NORTI-I # SOUTJ.I e t. E VAT IONS 141'-t:J' eAST c L c VATIONS FIGVRE H. }. SEXTON &. ASSOCIATES, ENGINEERS . . I r I I i ! I . * ,..., .. ., , . .. -... . . . : --.. . -.. .. *: . ... . . . : , . . . . . . .,._ : . . .. " . . . . -.. . . *-, .. *. * ......... . , .,.. . . -: . .. . ... . . . 61.101.()()* . . . . . . . -. . . : . cr : .. . .... .. . . . . . . :**\ : : : .. t r : : : . . . . . FI.Z4a()() ' .. . .. . .. .... * .... I I I .. L.:l ;.*.-* '* . i r; ... ' .. *. I I I *', . :: .. . : . I . ... . . ' . \:: . * . .. . P.?M-7. .. . .-., .. .. . : . -.. 7':* . . *: .. . . . . I'* .. . . .. .. .. .. . . . . . . .. .: . -. -. . . . ... 1.; . . . . . . . .. . . . . -. . . . . . . . . . . . {1.1141$' . . . . . : :: . .. . . ( .. . . : .. *: .. *. . . . . . . * .. : * ... :.: ::* .:*, *:* .. ' ..... ..* .... : . : . . . *. *: _ .. : . . . * . . .* . . . . .* . I LO#G/TUO//V4!.. SECT/OA/ A*9 H. J. I!XTON a ASSO C IATES, ENGINEERS l I .. *a ... II:ZCT 9-vr===; .. .. **. '*. .. .. I 'o**l--___ ...., i: ":'l--....

• r.* :...,:. 1.] i\. ,. _/:.;, Y:_ :. \ . * -*!': * .***
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..... _:_. .. :.** * * .:: * .. =._. *.** *. *:.*. **,*., :.. '* ... *:: _ ... :: :* :. . . . . .. *. . . . . . .. . 1: . -..... -. . , .. TRANSVeRSe SECTION A/0 R. ]. SEXTON &. ASSOCIATES, ENOINB81lS A. 2-6" H. J. SEXTON&. ASSOCIATES. ENGINEERS S A N F R A N C 1 5 C 0

• N l L 0 P .\ R K
• C A L I f 0 a N I A * .. o. /20/

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• Sr;,8. KIPS INIIrT-AI H. J. SEXTON & ASSOCIATES.

ENGINEERS

• MENLO P.\RK. CALIFORNIA M)eHO .* /201 .1o*

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• A2 H. J. SEXTON &. ASSOCIATES, ENGlNIIaS SAJII PRANCISCO
• P\Ril , CALIPOaMIA

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24

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• X 282:

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• Wf::-fLO PARK , CAllfORif1A m ... ItO/ .,.. VGNMONT a*m G.£ ,/;'£ACTOR 4t./I,{)/NG fV* /v,.L/T 4 (fl GI.EV. .:1-IS./6 1) WAt..t.S:

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• ao12 x 7. 4.! '* 500' COLUM,VS: /4S: 260:

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• 200

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• S9 JC ./50 K/;trT. CONCNGTe W.4l..L 5: VOL..
• 291 C(J. 'r. WT.
• 291 Jt./50 K/,r:r.1 COLUMNS: Yt?L. * (525 C(J. rr. WT. * /,525 rKJ*

"" 2*11-(i

• 47. KIPS .
• 29. ..
• 2/S. -

- -229. ... . .,.. ., H. J. SEXTON&. ASSOCIATES, ENGINEERS !AN FRANCISCO

• MENLO P."RK , CALIFOaNIA W*/GI-IT 4 (12 ELeV. .34S.IG2 CONTtJ. S/-/1£/.0 I NG Pt.UG -940. 4 TANKS -12. L..IP"T.tNG SL I NGS -1 4. S,I./IPtOI/\IG CASK -150.

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H. J. SEXTON&. ASSOCIATES , ENGINEERS

• PRANCI S CO
• N!:fLO P.,RIIL CALlfOINIA ... /201 VG.CMONr YANk"GG S\'P' G.6. *u*,*c' I('GACrOI<

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• MENLO P.<\RK , CALIFORNIA Jotl ..a /201 J o* VERMONT YANKE4 c:"*a .. T G. G. *u*..tcCT 8V/I.D/NG FLOO/< St.Ad'S: d, 714 S,C x ./25 KS..C 48" 1,25'0 .,oo GO,. S-!..4 JJS : I. 000 Jt
• 750 .'2 St..AdS: I, 104 X .150 FLOOR VOl..
• cu. F.": a;r.
• 6 55 .150 K/,rr. # CONCRG:r'G WALLS vol. -

cu.r-r. w:: -4t844 J( .150 K/rr.B COL.'-/MN.S vaL.

• 2, 9 20 cu. WT.
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C.MK'O/P../,1:' Dolt.TC - K/PS -750. -760. -. -98. - - 8HCIT NO . .A.R H. J. SEXTON&. ASSOCIATES, ENGINEERS FRASCUCO

• MESLO P."RI(. CAllfOJtNtA Joe NO 1201 JOe Y.4NKt"G'

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• J-1/*/GH'T G (<1 EL.£V.

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(45SUMGO) .. 84. J's Gt...6"CT.'<ICAL GQUIP. (M/SC.)

• SO. }5 PJ , D//'116 (MISC.) 50. /{f. WATG',Q IN rUeL. STO,qAG4 ,Do()t. fDIV,, 6QU I PMGNT IN

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• 8HCCT-A9 H. J. SEXTON & ASSOCIATES, ENGINEERS FRA.NCI'iCO
• ME.SLO PAIUC , CALIFORNIA W*/GI-IT EI..EV. .st..Aes .5L.A8S: 8,948 SF K ./SO KSF 10"

5,G80 k ./2S 48" Z700 A rL.OO,q IJ6AMS r-r. wr.

x .1so lf/;:-r.3 CONCReTE WAt.LS VOL.

• Cl./.rT. WT. -

X./50 N/rr,l C0/1-'C""t!!T.* COLUMNS VOL.

• 5;004 cv. FT. Wi. -.5,004 x. 150 l)I<YWGI.I.

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• /,.442. k" O_c 7/0. -l,te20. -?95'9.
• 7SI. -

-4. -20. -1,0()0. -84. -80. -200. --50. 64. -WEIG,LIT 7 KIPS A. H. J. SEXTON & ASSOCIATES , ENGINEERS P R AN\: IS C O

• N£SLO P\ltK CAllfOaNtA toe /20/ JOt VE,t:lMONT Y"{4NKGI '""-T u. £. I U I J CCT RGAcroR WGIG).IT a */..*V. 252.50'} r'oo,q .SLABS 2-1 1

.:SF JC,.;,!OO CCJNC.i<G/G WALLS VO£. -r-r. rrrK'./£ !!H 2*Z* OA Y! 2*/4**, WT. * = COIVCRGTG COLUMNS VOl.

• _a, Cl..l. ,&-T. W?: -Jl , /50 .lf/tl7. .OI<YWGLL.

c. f2 SUP;O/l£$5

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• Jt T/ON P/PING ¢ W.4TI,q -RECIRCUt.ATION VALVGS
• ReCI.IPCULATION

, ouMPS -f/5 J./0/S'T ¢ SV.CPO.t<T STGGt. {4551./MGI)) -ISOL.AT/01\1

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• MONITORING
• ANO..J/R L I NGS
• f's GLEVATOI<

(A5Sl../Mc0)

• J5 *LECTI<I CAL (MISC.J
• Jj P I PING {M I X.) : CONTJ'./0(.

ROO ORIV£ J./Y O/<Af./1./C VN/T.S -UPPGI< PORT I ON t;I../1 ,0/N G APPEtVa4G£ -WE I GI-lT

• 59S. 400. 850. 2,3G7. 60. 127. 200. 4. 20. t!JO. 1 0. 25. 25. 84. 50. 84-. 89. 84/3. K/PS I INIIM' NO * .4 II r i I \). ( ! .*/ H. J. SEXTON&. ASSOCIATES, ENGINEERS FRANCISCO
• P."RK. CALIFORNIA 5ECT/ON PROP*RTI*5 H. J. SEXTON&. A S SOCIATES, ENGJNEERS

!I A N 'F R A N C I S C 0

• foC F. N L 0 P .\ R K . C A l I F 0 1l N I A 1oa

=Jo=-* __ ____ CL t iiNT G. E. sueJIICT 0/NG cNw*oM../ ?*. . . I. Et.EV. 213.75 1 TO 252.50'@ N-S t E-W AXeS b d t.. .4 J. * -f2 I:Jtl 4 AI. t 1 44.00 0 4G4. 14-92992 , 0 :3.00 it'. 50 t828. G2/, 41/I IS367. /NCt..INcO wAL.t..s (e b d L. A I At. z L 1 2 7F 4)(3. 00 G/.SO 47.2S 654. 748 OCTAGON4L .secr/QJ\1' ,4 T.4N e2.s*

• 3.550 Fr 2 L -

[12 (6-3.60)2 +(52.G9)z]- 895 r-rl tJ . 4(4&) IA f 335CJ

• 5780 E. (.I;, t-At.z)
• f.¢;

r ll fi?-581 .,t. 358 - A l 2 H.]. SEXTON & ASSOCIATES , ENGINEERS "FR.ANClSCO

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4. 25 2JC4.00

/f;J.50 2.50 /7.00 L 0 71.50 :;;t;>.CO 47.00 2B.50 0 0 .38.50 70.00 69.75 7o.oo* 47.00 47.CO 52.00 49.25 f48 284 20 so 10/ 3'2 20 40 30 296 41

24. 5. 2/. 27.

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• N E N L 0 P A K , C 4 L I f 0 a N I A 10* tfo I?QI Jo* Vq-RMOIVT YANK** Sk""' (j. C. eueJ&CT //?EACT"R 15(//(.j)//VG

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• 60?185 r-r# !.4 * .30 .f 1060
• FT2 I (AL z" lo). 5SG,837,.

1; I-185

• 387, A/4 H. J. SEXTON & ASSOCIATES, ENGINEERS fRANCISCO
• MENLO P.,RJC, CAI.lfOJlNIA ... o./20/

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• 2Jf/.OO 14/4.00 0 28d ay o ... Tc CM**oM./ OATC ALz 0 2A/42.00 71.50 2d4 24, 1,451, 879. 4 Jf 2.00 2.00 70.00 16 s.

4 14 2.00 47.00 42 746148. 2 *2.SO 4.00 to 27. lt045. 2.00 20.()0 0 40 1.!35. 0 1(4. '25 47.00 St/ 82. I 5.50 5.50 47.00 .30 7tA. 1/.JCtJJ!/..00 2.00 70.00 //, 2Jt4.00 2.50 G9.75 20 /0. 25.00 2.00 if:'.()O 50 /7. 245,000. so.50 '2.00 70.00 /01 2Jt37.00 . 4.00 14.00 2% 395. 58,0/fO. 2.50 /6.50 28.00 41 94<4. /7.00 2.50 35.25 4.3 22, 63 .. 1547 INCLINE£) W.4L.t. (S*30° COS 2 8*.750 S/N 2 B* .. 250) h d L bd 9 d¢1 A AL 1 I. 72 12 /2.CKJ 2.50 32.25 /6 102. 30 3/, 202* 102. aNIIT NO ,415 A.2-lZ Joe teO /20/ !;L I INT G. c . H. J. SEXTON &. ASSOCIATES, ENGINEERS !!IAN FRANCISCO

• W!NLO CALifORNIA .JI. *LeV. 252-SO'J't) e-WAXI.5{CONT.)

C/RC<Jt..AR .54'Cr/ON A * -1(72.5:-6?. 5 2) --IOtila ,e-r2 z LA= f50 2.437 r-rt r(Al 2 flo)

• 50Q6 79 r 2, 935,159 1-:3 t 202 /02

/85 -I[I['T NO. A/6 H. J. SEXTON & ASSO*CIATES, ENGINEERS N F R. A N C 1 S C 0

• M E N L 0 P *' R K , C A L 1 f 0 ll N 1 A Joe Ho./20 1 VeRMONT V'ANKGG £1.15NT G. G. 8UIIJCCT

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• M!NLO PAilK. CALlfOaMtA 1201 e*

YANKEE et.*w G.E. l<'eACra..r< tJU/t.OIIV'G

• ., .1;../J:' NJ ,.,., 2-,4 , .Ill. £LEV. 28QOO' TO .304.00'f§N-S AXIS (CONTZJ.)

INCt..INcO W4L.L.5 b ci L bel! -12 a'h' -12 e .4 4.00 17.00 2/.50 9/ 45. Ga. 7.00 10-SO 86 23. 2'1. 4.00 /9.00 20.50 69 4s* : !/-1/. CIRCUt.AR SECTION A. +(46:5 2-3:3.5 2)

• 74S F"T.2 ..r.
• !44,563rT.-1

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• 152£ -f 141 + 745 *

-r.z AL 1 55,601. I(l

t 717,061f

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• P..Z-15 i I. 76. 400. /841. .A/<8 H. J. SEXTON & ASSOCIATES, ENGlNEE*ItS SAN FRANCISCO
• MfNLO P.'\ftK ,

M)tl No. 120/ :;.;::Jo-..* __ _ _..___ * .., ,./ r-o.,.. 2*5*CJ c"K'D H./ o.n c 2*H* .lli. E LEV. 280.00 1 TO 303.00 e *-W AXIS l h o' t. A .Lbt:J4 4*/f" .4L2 . /44.00 0 24a. a 2Jt//O /.00 71.5Q 220. 18. 4X2.00 2.00 7aoo 5. 4M2 , 00 47.00 24. /8. .:1.00 2.3.50 2 11 2.00 0 I 9. 0 I I ' 4Jt3.00 z>.oo; 70.oo 1 24. 8. 2S.OO ?.00 "-00; so. 17. 245,000. ! sa so 2.00 it'-00 /01. 4.00 4.oo I 47.00 /6. 2/.

4.00* 4.00 i 2/. I 4.00 t/.00 1 0 21. 0 I 25.50 II. 9. 25 47.00 I/. 9.

4.00; 47.00 94. 1 25. 4.00 18.00 .J7.00 /944. 3.00 24.00 72. <55,698. 7.50 4.00 47.00 40. /1.50 4.00 47.00 61. 3.(/0 22.00 40.50 66

oa '2s7. 4.00 19.00 .35.25 76 . 228(;,. 94,435. 2"1aoo 1 2.00 2/.50 240. 2880.

51 t 3'2G. ... CI'I' HO A/9 .. H. J. SEXTON &. ASSOCIATES, ENGINEERS JAN fRANCISCO

• WE:-fLO CALIJOaNIA
• .. o 12o/ ...,. Vel< MONT Y"ANKGG "'" G. E. .UaJI(CT ReAcroR
• * .,..,./ 9&!1 ..!//: *L*V. 2tBO.OO' TO .303.00'8£-W 4XIS(c0/VTZJ.)

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• f (45.5 1--,e-r.2 ..c -%148, St;3 I.4 * /5"2S .c: 241 I FT.2 AL2 1'25.

6q411. ! fAL t)

• Sit $211 .,. 8471--t-(292 f/44 SG:J c .a,so2 1 4.59 rl:" A. z -77 I. I 21. 400. I, 292. .... ..,_ A20 H. J. SEXTON&. ASSOCIATES, ENGINEERS FRANCISCO
• NF.NLO P.,RK. CALifOaNtA Cl.lt:NT G. e. N. Gt.GV. .303.00 1 TO N-5 AXIS 1; ! d L A Io* /44.00 /.00 7/.50 144. /2.

i /44.00 /.00 .39.50 12. 11aoo .?Po. 221,853. 2x?.OO 2.00 ?aoo 8. :3. .3i200. 2.00 .38.00 4. s .. g.oo '2.00 70.00 10. 147,000. 5Jt3.00 2.00 38.00 /0 .. 2-'2.00 5.00 47.00 12. 9. 2112.00 23.50 12. 9. 2.00 2S.CXJ 0 50. 2,604. 0 . 2.00 0 /01. 21; 4(;}5. 0 2xE.OO 47.00 18. .. /4. $.00 .5.00 9 . 7, 20.00 3.00 13.25 6a 45. .3.00 /5.00 . 24.50 4.00 24.00 98. /31. 5G .. 448. 16.50 56. (310 . 2.SO 5.75 42.0() 14. 40. 3.50 47.00 4J. #. 94,987. 28.00 3.-50 47.00 98. 100. 23.75 .34.00 9S . IO'i.820. 3.00 22.00 1 4. 2.3. 2A5/.00 6.00 22.00 612. s.oo 88.50 0 198. 23,778. 0 2J(8.00 6.00 /7.00 %. 288. 27.744. 27d 120. 2 I A21 H. J. SEXTON & ASSOCIATES, ENGINEERS "AN PI'ANCISCO

• WINLO PAaiiC. CAt.lfOatUA 1201 '"'&J!Y G.£. 8lolaJICT I<GACTO/<

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.. auro .,. ** .,..a. VYNPS DSAR Revision 0A-96of216A.3TURBINEBUILDING Earthquake analysis of the turbine building was done in accordance with the provisions of the Uniform Building Code. To confirm the structural adequacy

of the bents at the diesel generators, dynamic analysis was performed for the

transverse direction on the braced end wall bent at Column Line 4 and the

rigid frame bent on Column Line 5. Because the corrugated structure of the roof permits diaphragm action in the longitudinal direction only, each bent was analyzed separately. Natural

frequencies and mode shapes were obtained by computer analysis. The response

spectra applied were taken from the report by John A. Blume and Associates. A

damping factor for steel of 2 percent was used in accordance with the criteria for the project. The loadings used for the equivalent mass systems consisted of dead (permanent) loads, 50 percent of the design snow load, and the unloaded

turbine building bridge crane in its normal parked position south of Column

Line 5. Two analyses were performed for each bent: basic design seismic

loading was applied to determine if resulting stresses remained within code

allowable, and two times the design seismic loading was applied to determine

if resulting stresses remained within the yield point of the material. The analyses established that the controlling loading condition for the turbine building bents is not due to earthquake, but to the fully loaded crane

with associated horizontal forces and design live loads. No failure of the

bents at the Class I equipment can occur under the loading applied for the dynamic analysis. Schematics of the bents and the mathematical models used in the analysis of the bents appear in Figures A.3-1 and A.3-2. VYNPS DSAR Revision 0A-97of216' -_.__ TURBINE: BlJILDU.JG -Bf:NT NO.4 VERMONT YANKEE NUCLEAR POWER STATION FINAL SAFETY ANALYSIS REPORT TURBINE BUILDING BENT FIGURE A.3-J. VYNPS DSAR Revision 0A-98of216 TOP CllAN6 RAIL FlOOit ':IL' ..:::; ,..... eX ol w ! 2 '-" -<( Ill

• QC l: < l: ----------------0 (!\] ;:( 10 0 -Q -------------------TVR5lNE E>VILDING -BEt-iT NO.5 MATHEMATICAL MDPEl-VERMONT YANKEE NUCLEAR POWER STATION FINAL SAFETY ANALYSIS REPORT TURBINE BUILDING BENT FIGURE VYNPS DSAR Revision 0A-99of216A.4PLANTSTACKA.4.1Introduction This report summarizes the procedures employed and the results obtained from the seismic analysis of the plant (or ventilation) stack for the Vermont

Yankee Nuclear Station. Envelopes of maximum shears, overturning moments and displacements versus height have been developed and are presented herein. The earthquake criteria used in the development of the above is based on the

recommended earthquake criteria for the project prepared by J. A. Blume &

Associates, Engineers.A.4.2DescriptionofStackThe stack is 318'-0 high, tapered reinforced concrete structure with 7'-0inside diameter at top linearly tapering to 22'-0 inside diameter at the

bottom. The variation of thickness with height and the stack arrangement are shown in Figure 3.2-18.A.4.3MathematicalModelofPlantStack The ventilation stack was treated as a multimass system represented as a flexible cantilever with the base fixed at the top of foundation. Masses were lumped at fifteen mass points, and were considered to be supported by weightless elastic columns. The moments of inertia and effective areas of the

elastic columns as well as the corresponding masses were calculated by

computer. The mathematical model of the ventilation stack is shown in Figure A.4-1.A.4.4AnalyticalProcedures-PeriodsandModeShapes Once the mathematical model is established, the motion of each lumped mass under any external excitation may be written in the matrix form as follows: [M]}{+ [K] {} = F where:[M]=Square mass matrix.[K]=Square matrix of stiffness coefficients including the shear and bending deformations. }{=Column matrix of acceleration vectors. VYNPS DSAR Revision 0A-100of216 {}=Column matrix of lateral displacement and joint rotation vectors.{F}=Column matrix of external vectors. The stiffness matrix [K] is formulated by computing the stiffness coefficients for each joint of the structure and assembling them in the proper sequence to

form the complete square matrix. In the computation of the stiffness matrix, both bending and shear effects are considered. In the above equations of motion, the damping terms are left out intentionally. This is due to the fact that the damped natural frequency is almost the same as the undamped natural frequency for the system with reasonable structural damping factors. In calculating the natural frequencies and the mode shapes, the external load matrix in eq. (1) is set to zero and the displacement vector {} is assumed to take the form of simple harmonic motion or {} = {} Sin t(2)where:{}=Relative amplitude of mode shape vector.=Natural frequency of vibration. After substituting and simplifying, the equations of motion are reduced to the following form:}{}{2 1 M 1 K(3)Solution to this eigenvalue problem exists only for particular values of which correspond to the natural frequencies of vibration of the structure. Eq. (3) is solved by iteration techniques to obtain values of and their corresponding mode shape vectors {}.A.4.5ModalAnalysis VYNPS DSAR Revision 0A-101of216 After all natural frequencies and their mode shapes are determined, the method of modal analysis is employed to calculatethe structural responses. This

method actually simplifies the analysis of a multidegree of freedom system to

the analysis of several equivalent single degree systems, one corresponding to

each normal mode. The governing equation of motion is shown in the following

form: xn x 2 xn x a so n 2 n n n n M 1=x N M 1=x N (t)f Y=A+A B 2+A(4)in which A n=Displacement of any one arbitrarily selected mass.(Usually the topmost mass.) of the nth mode. B n=Damping coefficient = n n. n=Percentage of critical damping of the nth mode. n=Natural frequency of the nth mode. Yso=Maximum ground acceleration. f a(t)=Time function of ground motion. M x=Mass at the xth level. xn=Normalized displacement of the mass. M x of the nth mode.If the two summations on the right-hand side of eq. (4) are denoted by P n , which is defined as the modal participation factor of the nth mode, then (t)f Y P-=A+A B 2+A a so n n n n n n (5)Since the values of B n , n and P n are already known for each normal mode, eqs. (5) which are actually "n" independent equations, can be solved separately and VYNPS DSAR Revision 0A-102of216 their solutions are: d)(t-Sin e (t)f Y P-=(t)A n)(t--a t o n so n n n n(6)The maximum values of eq. (6) are: M n)(t--a t o n so n max n d)(t-Sin e (t)f Y P-=)(t A n n(7)since Y xn (t) = xn A n (t), therefore S P-=d)(t-Sin e (t)f Y P-=)(t Y dn xn n M n)(t--a t o n so xn n max xn n n(8)Where s dnindicates thequantity in the bracket. With the time history of the 1952 Taft Earthquake N 69°W component an excursion was made through the earthquake determining the displacement time history. The inertia forces time-history, and the moment time history records werethen obtained. In addition, the shear and moment time-history data were scanned to determine the maximum values. These values were then used for design.A.4.6Results The results of the seismic analysis in the form of design shears, design moments and relative displacement envelopes are presented in Figures A.4-2 through A.4-4. The computations previously described were performed on a computer. A damping value of four was assigned to all modes.The first ten natural periodsof vibration are listed below:First Mode -T1-1.5262 secondsSecond Mode -T2-0.4245 secondsThird Mode -T3-0.1829 secondsFourth Mode -T4-0.1041 secondsFifth Mode -T5-0.0669 seconds VYNPS DSAR Revision 0A-103of216Sixth Mode -T6-0.0493 secondsSeventh Mode -T7-0.0365 secondsEighth Mode -T8-0.0201 secondsNinth Mode -T9-0.0243 secondsTenth Mode -T10-0.0206 secondsA.4.7Conclusion The subject structure is designed to resist the seismic shears and moments presented herein without the increase in allowable stress for short-term loadings. In addition, the structure has been reviewed to assure that it can resist twice the seismic shears and moments presented herein, without

structural failure. VYNPS DSAR Revision 0A-104of216 560 -5110 2 520 J II 1180 5 1160 6 11110 7 1120 8 ...... LU UJ 1100 9 ..._ )18. z I 0 )80 -...... < > w 10 _. uJ )60 )110 1l )20 )00 12 I 13 280 111 260 15 2110 ,, -, ... -VERMONT YANKEE NUCLEAR POWER STATION FINAL SAFETY ANALYSIS REPORT VENTILATION STACK MATHEMATICAL MODEL FIGURE A.4*l VYNPS DSAR Revision 0A-105of216 S60 ----, 1 -"' 5110 2 ' 520 ---500 II 1180 5 1160 -6 11110 7 .... 1120 .... 8 ' .... ... z -1100 9 "-z 0 -.... "" ,.. \ .... ..J w 10 360 '\ 31!0 11 '\ 320 \ \ 12 300 1) 280 '" u 260 15 2110 0 20 110 60 80 100 120 1'10 160 180 200 270 MAXIMUM SHEAR (KIPS) VERMONT YANKEE NUCLEAR POWER STATION FINAL SAFETY ANALYSIS REPORT VENTILATION STACK SHEAR FIGlJR E A .4-2 VYNPS DSAR Revision 0A-106of216 560 1 ' '\ 520 3\ 500 * \ 80 5 \ -6 \ 1 ..... .., .., 8 .... z -* 9 "" 400 0 \ -..... < > 380 .., \ ..... .., 10 360 \ 340 11 \ 320 300 12 "' -"""' 1) --1'1 """' 260

• 15 240 0 I 2 3 II 5 6 7 8 9 I 0 II 12 13 Ill 15 16 17 18 19 20 21 22 MOMENT ( F T-K I P X I 03 ) VERMONT YANKEE NUCLEAR POWER STATION FINAL SAFETY ANALYSIS REPORT VENTILATION STACK MOMENT FIGURE A.4-3 VYNPS DSAR Revision 0A-107of216 560 1 I 5110 2 v 520 I ' 5 00
• I 1180 s 1160 6 , 11 11 0 ' 7 I t-..... .., 11 20 8 u.. "" -z: 1100 9 Cl ""' > ..... _. 380 .., 360 3110 11 32 0 1 2 3 00 13 280 111 260 15 211 0 0 0.1 0.2 0.3 MA X. OEF L E C T I ON ( FT) V E RM ON T Y ANKE E N UC L EA R P O W E R S T A TI ON FI N AL S AFET Y A NA L YSIS R E P O R T VENTILATION STACK MAX I\1UM DEFLECTION DIAGRAM FIGUR E A.4-4 REPORT ON EARTHQUAKE ANALYSIS OF CONTROL BUILDING AT THE VERMONT YANKEE NUCLEAR POWER STATION VERNON, VERMONT VERMONT YANKEE NUCLEAR POWER CORPORATION A. s -t CONTENTS I -!NT RODUCT ION II-PURPOSE Ill-SCOPE . IV -BASIC THEORY AND METHOD OF ANALYSES A -Mathemat i cal Model B -Equations of Motion c -Natural Frequenc y and Mode Shapes D -Model A n alyses . E -Structural Responses . Building Plans and El e vation* Mathematical Model . Per io ds and Mode Shapea Max im um Accelerations Maximum Shears DRAWINGS Reap*onae Acceleration Spectra . 1 1 2 3 3 5 12 13 14 15 1 6 I -INTRODUCTION As a requirement of design !or the Vermont Yankee Nuclear f#**wer Station Control Building, a dynamie a.nalysia waa performed upon this structure to determine the accelerations and deform*tions to be expectf"d at various levels within the structure.

The dynamic analysis used was developed by Ebaaco Servicers Incorporated and performed using a Burroughs 5500 digital computer. The Vermont Yankee structures are designed to resist. within n**l'mal delign limih, an earthquake with a ground acceleration correspon d mg to the response spectrum of the N 69° W component of the 1952 Taft Earth quake, normalized to 0.07 e:ravity. They are further designed to s urvivt an earthquake witb a similar response apectrum and an acceleration of 0.14 gravity loss of !unction. 11-PURPOSE This report aumm.ariz.es the theory and approach used for the dynam1 c analysis of such structures and presentl the model used for the C**ntrtll Building along with the results of the dynamic analysis. I 11 -SCOPE This report sununa r izes the basic theory used to develop the dyna..mic a.na.lyah, the mathematical model used, the acceleration

• , the maximum deformation-, shear -, and moment -curves obtained.

I V -BASIC THEORY. AND METHOD OF ANALYSIS A -MATHEMATICAL MODEL 1n determining the responses of a complex structure aubJect to the ground excitation from an earthquake, it h often required to find a silnpliliecl structural model to represent the original structure so that the mathematical operations may be c arried out without too much <'OTT\* plexity. This simplified structural model is usually known as the mathr* matical model. The mathematical model chosen to represent the .,rlgt.n&i complex structure ia a. luznped mas a system. To use this system the dead load at each floor level and the weights of the walls or columns at the adjacent floors are considered lumped at a point and then connected by the weightleu elastic bars which represent the sti!fncsses of the columna or walls between floors. The base of this lumped mass systf'n was assumed fixed. For this structurt-such an assumption was more coneervative than a less rigid foundation. B -EQUATIONS OF MOTION Once the mathematical model is established, the motion of each lumped mass under any external excitation may be written in the ma.tri.X form as follows: [K)ltd= IF} (1) Where [MJ = Sc,uare mass matrix [K] =Square matrix of stiffness coeff i cients including the shear and bending deformations. = Column matrix of acceleration vectors. = Column matrix of lateral displacement and jo i nt rotation vectors. IFf = Column matrix of external load vectors. The sti!fneas matrix [ K] ia formulated by computing the stif!neu ficients !or each joint o! the original structure and assembling them in the proper sequence to form the complete square matrix. In the tation of the stiffness matrix, it is assumed that all joints at the same level have the same displacements (i.e., translations and rotations). ln the above equations of motion, the damp i ng terms are left out tentionally. This is due to the fact that the damped natural frequency is abnost the same as the undamped natural frequency for system with reasonable structural damping factors. (For 10 percent critical damp i ng w d = 0.995w .) C-NATURAL FREQUENCY AND MODE SHAPES In calculating the !requencie1 and the mode ahapes, the ternal load matrix in eq. (1) ia aet to zero and the cliapacement vector lll l is assumed to take the form of simple harmonic motion or f Ld = f " f Sin w t (2) Where l f l = Relative amplitude of mode shape vector. w = Natural frequency of vibration. Alter aubatituting and simplifying, the equations of motion are reduced to the following form: = 1 ;:;2 -----(3) Solution to thit eiienvalue problem exists only for particular values of w which correspond to the natural frequencies of vibration of the Eq. (3) ia aolved by iteration techniques to obtain values of w and their reaponding mode shape vectors l 0 I . D-MODAL ANALYSIS After all natural frequencies and their mode shapes are determined, the method of modal analysis is employed to calculate the structural sponses. This method actuall y simplifies the analysis of a multidegree of freedom system to the analysis of several equivalent single degree systema, one corresponding to each normal mode. The governing tion of motion is shown in the following: N A + ZB A + z A = n n *n w n n y so fa (t) x 1 Mx f xn N M " 2 ____ (4) in which A n w n ).. n = ;: ::: ! x xn X= 1 Displacement o£ any one arbitrarily mass * (Usually the topmost mass.) of the nth mode. Damping coeffi ci ent = A n wn. Percentage of critical damping of the nth rnode. Natural frequency of the nth mode. ft , 6-'5 3 = Maximum ground acceleration. = Time function of ground motion. Man at the xth level. c N or malized d i splacement o{ the maaa. M of the nth mode. x l! the two summations on the right-hand tide of eq. (4) are denoted by P

• n which b defined as the modal participation

!actor of the nth mode, then . . . An+ ZBnAn + wnAn = -Pn Y 10 £a(t)-----(S) Since the values o f B , w and P are already known for each normal n n n mode, eqa. (S) which aTe actually " n" i ndependent equationa, can be aolved aeparately and their solutions are: t p y A (t) = _ n ao J n w n o ! (t) e a The maximum values of eq. (6) are: *Xn w n Sin w (t-'t)d t---(6) D = -P {y*o J t. l w n(t-<) Sin w --(7) n "'n o a n JM tinee Y (t) = f A (t), therefore xn xn n Y (t) = -P e v *o f {.. t xn max n xn w n 0 = -P ' sd ____ (8) n xn n Where Sdn in d ic:at et the quantity in the bracket. Finally the total ment is the eummatio n of the displa c em e nt o£ each no rm al mode, that i a

• N y x (t}m.a.x = -n!'l p n e )CD S drt ----(9) Eq. (9) g ive a &he upper limit of the di*pla c ement.* ol any ma*** However, a* we ean rea*on.ably auume that all the maximum displacements o! all normal mode* do not neceaaarUy occur at the eame time, therefore, for the purpose of design, the root-mean square-method is adopted from the statistical point of view, thus: Y (t) = [r (P 0 S )2] 1 /Z (1 0) x max n xn dn -------E -STRUCTURAL RESPONSES Knowing the function Y 50 fa (t), or the record of the design earthquake.

the value of Sd can be calculated and plotted into a curve using the natural period of vibration as abscissa and the maximum displacement as thf' ordinate. This curve is known as the displacement response spectrum. To construct the velocity response spectrum, it is found that the spectral displacement Sd is directly related to the spectral velocity Sv by nat1. ral frequency w , or Sv = Similarly, the spectral acceleration Sa is equal to the product of w and the spectral velocity, that is: Sa = w Sv = .. 2 sd. ln the design o! the control building the response epectra employed were those corresponding to the N 69° W component of the l95Z Taft Earthquake. !'he structural responses are then calculated through the following operations

Where

• Mn M X 0xn vn 6xn Fxn Svn wn Mn vn F xn = = = N -M \2 {x=l xn x N 2 -Mx x;l *'* M ... n 5 vn v n 0 (xn M X = Effective mass. i n nth mode. = Mas s concentrat e d at xth level. = Normalized d i spla c ement of Mx in the nth mode. = Base shear in nth mode. = Deflection of the xth l e vel in nth mode. = Inertia force at the xth le v el in nth mode. = Spectral velocity in nth mode. "' Natural frequency of nth mode. A.s-7 s The modal re1pon*e values !or each mode computed above are combmt-d by the root-mean-square-method to determine the total responses of the structure, thus N z 1/Z (F ) = o: F xn) X n=l ----(11) N 2 1/2. (6 ) = (! ll xn) X n=l ----(lZ) b z J J} ) ' ex ' ><J . I -..I
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performed. The results of such an analysis on the intake structure are summarized in this report, and they include the maximum accelerations, deformations, shears, and external moments at all levels of this appurtenance. The Vermont Yankee Class I structures must be designed to resist, within the normal design limits, an earthquake having a ground acceleration corresponding to the response spectrum of the N69 o W component of the Taft Earthquake of 1952, normalized to 0.07 gravity. They should also be designed to survive the same

earthquake at a higher spectral acceleration, namely 0.14 gravity, without any loss of function. In both cases 5% of critical damping was assumed over the entire model.A.6.2PurposeandScope This report summarizes the theory and approach used for the dynamic analysis, and presents the mathematical model and results obtained therefrom.A.6.3DescriptionThe intake structure is a partially buried reinforced concrete, box-like structure, with one side open (the intake end) and with internal walls separating the interior compartments (see Figures A.6-1 and A.6-2). The floor slab rests on bedrock at elevation 187.00 feet, and the top of the deck slab is

at an elevation 237.00 feet. The enclosures for the main pumps and the service

pumps are located above the deck slab. Those for the service pumps are reinforced concrete units which are monolithically integrated with the main part of the structure and are classified as Class I but the main pump housing

is a steel frame with metal siding and is not a Class I superstructure. The

material around the buried portion is a dense sand fill.A.6.4MathematicalModel In the E-W direction the stiffness of the structure below deck level is, for all practical purposes, infinite in value. Using a one-degree of freedom

model in this direction, the period was found to be 0.039 second, which would produce no significant magnification of the effects of the accelerations through the structure. VYNPS DSAR Revision 0A-128of216In the N-S direction, two models were used. One was a two-mass vertical cantilever, fixed at its base andsupported at deck level by a spring, which

represented the stiffness of the sand fill at that level. The second model was identical except for the absence of the restraining sand-fill stiffness or

spring. In both models the mass of the water, which was assumed to be at

elevation 235.00 feet, was lumped at the corresponding mass points, together

with the gravity loads on the structure. The intake structure was analyzed

under conditions of maximum submergence in order to achieve a condition of

maximum structural flexibility. An empty structure would act as a stiffer one

and its period would correspond to a lower point on the spectral curve, well

to the left of that of maximum response. Of the two models, the one without

the lateral outer spring was more conservative, but the responses did not

differ very much. Inclusion of the spring, however, is more realistic. To validate the actual presence of a lateral spring due to the "passive" effect of the sand-fill, a column of dense sand of the same height as the actual fill was first analyzed dynamically. Its natural period was found to be about 0.77 second, which was much higher than that of the substructure

taken by itself (0.023 second flooded; and 0.08 second unflooded). This meant that the structure moved the sand-fill and not vice-versa.The spring constant of this sand-fill was determined with: k/ft 10 x 1.72=)v-Am(1 EA m=K 5 2 s where:A=1850 ft 2 (the upper third of wall contact surface)m=0.65m'=1.25E=1740 k/ft 2v=0.2A.6.5EquationsofMotion Once a mathematical model is established, the motion of each lumped mass under any external excitation may be written in the matrix form as follows: [M]{} + [K] {} = {F}(1) VYNPS DSAR Revision 0A-129of216 where:M=Square mass matrix.K=Square matrix of stiffness coefficients including the shear and bending deformations. }{=Column matrix of acceleration vectors. {}=Column matrix of lateral displacement and joint rotation vectors.{F}=Column matrix of external load vectors. The stiffness matrix [K] is formulated by computing the stiffness coefficients for each joint of the original structure and assembling them in the proper

sequence to form the complete square matrix. In the computation of the stiffness matrix, it is assumed that all joints at the same level have the

same displacements (i.e., translations and rotations). In the above equations of motion, the damping terms are left out intentionally. This is due to the fact that the damped natural frequency is

almost the same as the undamped natural frequency for the system with

reasonable structural damping factors. (For 10 percent critical damping d=0.995.)A.6.6NaturalFrequencyandModeShapes In calculating the natural frequencies and the mode shapes, the external load matrix in eq. (1) is set to zero and the displacement vector{}is assumed to take the form of simple harmonic motion or {} = {} sin t(2)where:{} = Relative amplitude of mode shape vector.= Natural frequency of vibration. VYNPS DSAR Revision 0A-130of216 After substituting and simplifying, the equations of motion are reduced to the following form: }{1=}]{M[]K[2 1(3)Solution to this eigenvalue problem exists only for particular values of which correspond to the natural frequencies of vibration of the structure. Eq. (3) is solved by iteration techniques to obtain values of and their corresponding mode shape vectors {}.A.6.7ModalAnalysis After all natural frequencies and their mode shapes are determined, the method of modal analysis is employed to calculate the structural responses. This method actually simplifies the analysis of a multidegree of freedom system to theanalysis of several equivalent single degree systems, one corresponding to

each normal mode. The governing equation of motion is shown in the following: 2 xn x N 1=x xn x N 1=x a so n n n n M M (t)f Y=A n 2+A B 2+A(4)in which A n=Displacement of any one arbitrarily selected mass.(Usually the topmost mass.) of the n th mode.B n=Damping coefficient = n n. n=Percentage of critical damping of the n th mode. n=Natural frequency of the n th mode.Y so=Maximum ground acceleration. f a(t)=Time function of ground motion. M x=Mass at the X th level. xn=Normalized displacement of the mass. M x of the n th mode. VYNPS DSAR Revision 0A-131of216If the two summations on the right-hand side of eq. (4) are denoted by P n , which is defined as the modal participation factor of the n th mode, then An+ 2 B n An+ n A n=-P n Yso f a (t)(5)Since the values of B n , n and P n are already known for each normal mode, eqs. (5) which are actually "n" independent equations, can be solved separately and their solutions are: d)(t-sin)(f o t Y P-=(t)A)(t-a n so n n n n n e t(6)The maximum values of eq. (6) are: M e t d)(t-sin)(f o t Y P-=)(t A n n n)(t--a n so n max n(7)since Y xn (t) = xn A n (t), therefore (8)where s dn indicates the quantity in the bracket. Finally the total displacement is the summation of the displacement of each normal mode, that is: S P 1=n N-=)(t Y dn xn n max x(9)Eq. (9) gives the upper limit of the displacements of any mass. However, as we can reasonably assume that all the maximum displacements of all normal

modes do not necessarily occur at the same time, therefore, for the purpose of design, the root-mean square-method is adopted from the statistical point of view; thus, (10)S P-=d)(t-Sin e (t)f Y P-=)(t Y dn xn n M n)(t--a t o n so xn n max xn n nS P 2=)(t Y dn xn n max x 2/1 VYNPS DSAR Revision 0A-132of216A.6.8StructuralResponses Knowing the function Y so f a (t), or the record of the design earthquake, the value of S d can be calculated and plotted into a curve using the natural period of vibration as abscissa and the maximum displacement as the ordinate. This curve is known as the displacement response spectrum. To construct the velocity response spectrum, it is found that the spectral displacement S d is directly related to the spectral velocity S v by the natural frequency , or S v= S d. Similarly, the spectral acceleration S a is equal to the product of and the spectral velocity, that is S a= S v= 2 S d. In the design of the control building the response spectra employed were those corresponding to the N69°W component of the 1952 Taft Earthquake. The structural responses are

then calculated through the following operations: x xn N x x xn N x n M M M 2 1 2 1*S M=V vn n*n nx xn N x M 1 M V n=F xn x xn (x n) = [K]-1 (F xn)where: M*n=Effective mass in n th mode.M x=Mass concentrated at x th level. xn=Normalized displacement of M x in the n th mode.V n=Base shear in n th mode.xn=Deflection of the x th level in n th mode.F xn=Inertia force at the x th level in n th mode.S vn=Spectral velocity in n th mode. n=Natural frequency of n th mode. VYNPS DSAR Revision 0A-133of216 The modal response values for each mode computed above are combined by the root-mean-square-method to determine the total responses of the structure,

thus,

F=)F (2 xn 1=n N 1/2 x (11)

xn 1=n N 1/2 x 2=)((12) VYNPS DSAR Revision 0A-134of216 N . H.o* , &. 1ft I 113 s* I' ' .... -I I !1 I Cli . ' I II _) lrir ) I'

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-'n J *:ar I --*-_] u I . b f*s' t u c6, -.r -1 ' r __, . -.!U !§.. ,--' I . I 3./:J' h ... ...., --l'liJ ' . r l.s II 4t.o* ?:J.o* ..a.a 1 , lEY. 10. DATE EBASCO SERVICES INCORPOfP.ATED PLAN AT EL. 230.00' A-lmf* co* 1-l ... k.:t._ KA'P(*2r/ -FIGURE A.6-l DATP I VYNPS DSAR Revision 0A-135of216 fl. '2?'2.1;' EL.237.0' b N.W.l. SECTION A* A EL.231.0' EL.190.0' SECT\ON B .. B REV. 110. DATE EDASCO SERVICES INCORPORATED NEW YORK SECTIONS A-Dav.C*H A ..... OYD SCAIIEI'*ro' CH.-FIGURE A.6-2 DATE I VYNPS DSAR Revision 0A-136of216 El. 265.5 EL. 235.5 E L. I 88. 5 L A PT. (FT.) (FT .2) I 30 10.8 2 Ll7 S2LI.O 3 (a) I w (FT. ij) ( K) 72.0 1.511 8850.0 257,300.0 3 2 (b) 3 damping: 5% VERMONT YANKEE NUCLEAR POWER STATION FINAL SAFETY ANALYSIS REPORT INTAKE STRUCTURE SERVICE WATER BAY AREA MATHEMATICAL MODELS (N*S) FIGURE A.6*3 VYNPS DSAR Revision 0A-137of216 80 EL. 265.5 70 60 50 EL. 235.5 .... ..... ...., ... 110 I-::c "' <:. 30 ::c 20 10 EL. 188.5 0 0.01 .02 .03 .07 . 101 ' ' ' ' ' ' ' ' ' MODEL (b) ' ' ' MODEL (a) ' ' . ' . ' ' . ' ' ' ' ' . 107 . 10 J I I I I I I I I ' ' ' ' I I .011 .05 .06 .07 .08 .09 .10 .II .12 MAXIMUM ACCELERATIONS (N-S) (G) VERMONT YANKEE NUCLEAR POWER STATION FINAL SAFETY ANALYSIS REPORT INTAKE STRUCTURE

• SERVICE WATER BAY AREA
• MAXIMUM ACCELERATIONS (N*S) FIGURE A.6-4 VYNPS DSAR Revision 0A-138of216 80 ,. 0 70 N 0 "' 0 "' 0 60 0 0 0 t= 50 w w .... ...... 1-:c <..0 w 30 :c 20 10 0 -N (") ,. ..., 0.01 0.02 0.03 0 oo 0 8 oo . . MAXIHUM 01 SP L ACEMENTS (N-S) (FT) VERMONT YANKEE NUCLEAR POWER S TATION FINAL SAFETY ANALYS I S REPORT INTAKE STRUCTURE
• SERVICE WATER BAY AREA
• MAXIMUM DISPLACEMENTS (N*S) FIGUR E A.6-5 VYNPS DSAR Revision 0A-139of216 80 -"" ,.._ --EL. 265.5 70 60 -"" ""' 0 ;;;; .-. 50 "' ..., 1-co a> .., --EL. 235.5 .., co I ... 40 ,; I 1-I ::c I C> -I .., 30 ::c .. _.I ...J 20 :;;I .... 0 0 X I 10 I I EL. 188.5 0 100 200 300 1100 500 600 700 800 900 1.000 MAXIMUM SHEARS (N-S) (KIPS) VERMONT YANKEE NUCLEAR POWER STATION FINAL SAFETY ANALYSIS REPORT INTAKE STRUCTURE SERVICE WATER BAY AREA MAXIMUM SHEARS (N-S) FIGURE A.6-6 VYNPS DSAR Revision 0A-140of216 80 70 60 -"' 50 ,. ..... ,. w UJ .... qo ..... :z:: UJ "' 30 20 10 EL. 0 0 0 N 265.5 EL. 23S. S MODEL 188.5 8 0 0 0 0 0 0 0 .; .; .,; 0 0 0 0 MODEL (a) 20.000 30,000 40.000 50,000 MAXIMUM MOMENTS (FT-KIPS)

VERMONT YANKEE NUCLEAR POWER STATION FINAL SAFETY ANALYSIS REPORT INTAKE STRUCTURE -SERVICE WATER BAY AREA -MAXIMUM MOMENTS (N-S) FIGURE A.6-7 REPORT ON EARTHQUAKE STRUCTURAL ANALYSIS OF INTAKE STRUCTURE AND EQUIPMENT LOCK BUILDING AND FREQUENCY ANALYSIS OF DIESEL GENERA TOR FOUNDATION VERMONT YANKEE NUCLEAR POWER STATION VERNON , VERMONT VERMONT YANKEE NUCLEAR POWER CORPORATION C ONTENTS r -INTRODUCTION . 1 I -PURPOSE AND SCOPE. III -DESCRIPTION . IV -MATHEMATICAL MODELS A -I ntake Structure B -Equipment Lock Building V -THEORY A -Equations of Motion B -Natural FrC'quencies and Mode Shapes C -Structural Responst>s V 1 -RESPONSES VI I -DIESEL GENERATOR FOUNDA T tON FiGURES 2 3 4 5 5 6 7 r -INTRODUCTION As a requiremt>nt for Lht: design of the Vermont Yankee Nuclear Power Station* s Class 1 Structures, a dynami c analysis of each of these structures was performed. The of the analyses on the ff\take StrPc:h*re and the Equipment Lock Building are summadzed in this report, and they elude the com maximum values of the accelerations, deformations, shears and moments at all levels of these appurtenances. The Vermont Yankee Class 1 Structures must be design*ed to resist, within the specified design limits, an earthquake having a ground acceleration corresponding to the response spectrum of the N 69° W component of t.he Tnft Earthquake of 1952, normalized to 0.07 gravity (Figure 1 0). They should also be designed to survive a proportioncilely more intense earthquake with a higher spectral acceleration, namely 0.14 gravity, without any loss of !unction. In both cases the entire structure was Aiven a damping factor of (5 percent of critical damping). In the last part of this report the natural frequencies of the Diesel Generator Foundation for the rockin$!. sioesway, and vertical modes are compared with the frequency of .the aenerator. I I -PURPOSE AND SCOPE This report summarizes the theory and approach used for the frequency and seiomic analyses and presents the mathematical models and results obtained therefrom. I I l -DESCRIPTION The Intake Structure is a partially buried reinforced concrete box-like structure with the intake end open and with internal walls separating the i nterior compartments. See Figures l and 2.. The floor slab rests on bedr oc k, whi c h i s at El. 1 8 7.0, and th e top o f theo de c k slab is at EL 237.0. A. r.o-II The enclosures for the main pul'Tlps and service pumps *r e located above deck level. Those for the service pumps are of reinforced concrete lithically integrated into the main part of the structure and are classified as Class 1, while the main pump housing is a light steel frame with metal siding and is not considered as a Class l structure. The i ntake structure is backfilled with a dense sand fill up to El. 236.5 on the north and south faces, and up to El. Z54.0 on the west face. The Equipment Lock Build is a one -story reinforced concrete Class l supported on a compacted sandfill. It consists of two outer walls with pilasters every 18 ft-3 in. on center s in the long direction. The short sides are open, with sliding steel doors suspended from overhead rails. The roof is a 13-1/2 in. slab and the floor is a mat foundat ion. See Figure 3. IV -MATHEMATICAL MODELS A -INTAKE STRUCTURE In the east-west direction the stiffness of the structure is for all tical purposes infinite in value. Using a one-degree of freedom model in this direction the period was found to be 0.039 second, which would produce no significant magnification of the gr ound acceleration throughout the structure. In the north -sou th direction two models were used; see Figure 4. One was a vertical ca ntilever fixed at its base and supported at deck level by a spring, which represented the stiffness of the sandfill at that level. The second model was identical, except for the absence of the restraining sandfill stiffness. In both models the mass of the water, which was assumed to be up to El. 235.0. was lumped at the two mass levels together with the gravity loads of the structure and equipment. The Intake Structure was analyzed under condit i ons of maximum water c.>levalion in o rder to achieve maximum structural flexibility. An empt y structure would act as a stiffer one and it s A ,{,-P-2 ( period would correspond to a lower point on the Spectral Curve, well to the left of the peak response. Of the two models, the one without the outer spring was more conservative, but the responses of both models did not differ by more than 10 percent. Inclusion of the Spring was more realistic and the results are shown on Figure 5. To validate the elast i c restraining action of the sandfill, in the form of a lateral spring, an isolated column of dense sand, of the same height as the actual !ill, was first analyzed dynamically. [ts natural period was found to be 0.77 sec., which was much higher than that of the substructure taken by itself {period of substructure = 0.023 sec., flooded; and 0.08 sec., empt y). This led to the conclusion that the st r ucture impinged upon the s:1ndfill, and not vice -versa, during an earthquake. The spring constant o! this spring was determined as follows: Ks = m* EA = 1.12 x tos k/rt. A/'A m (1-vZ) where: A :: 1850 nZ (the upper third of wall contact surface) m = 0.65 m* = l.ZS !or the sandfill: E = 1740 k/ ftZ v = o.z B -EQUIPMENT LOCK BUILDING In ilie north -south direction the structure can be co nsidered as a rigid structure, without magnification of the ground a ccele rat ion anywhere i n the structure. In the east-west direction the structure ts modeled as a two-mass cantilever (one mass at roo! leve l and the other at base level) supported on rocking and translational springs. See Figure 6. The Spring Constants arc found as follows>:*: l) KH = 2 (1 + v) G Bx vBL Where: G = E/2 (l + v) E: 36 000 KSF v = o.zs Bx = 1. B = 78 ft L = 31 ft :. KH = 1. 773 x 106 K/ ft. 2) = G ( 1-v) Where: B(> = 0.43 K(> = 6.185 x 108 ft_K/ RAD. *"D esign Procedures for Dynami cally Loaded Foundations" by Whitman and Richart -Journal of the Soil Mechanics and Foundations Division-November 1967. V -THEORY A -EQUATIONS OF MOTION Once a mathematical model is established, the motion of each lumped mass under any external excitation may be written in rnatrix form as follows: [ M] ltd Where: [M] :: [K) = I 61 = :: = + ( K] IL\ I = I F l ( 1 ) Square mass matrix -including base mass and ba.se mass moment of inertia. Square matrix of stiffness coefficients including the shear and bending de formations, and the base translation and rocking stif fnesses. Column matrix of acceleration vectors. Column matrix of lateral displacement and joint rotation vectors. Column matrix of external load vector s. The stiffness matrix ( K I is formulated by computing the stiffness ficients for each joint of the original structure and assembling them in the proper sequence to form the complete square matrix. In the tation of the stiffness matrix, it is assumed that all joints at the same level have the same displacements (i.e., translations and rotations). In the above equations of motion, the damping terms are left out intentionally. This is due to the fact that the damped natural frequency is almost the same as the undamped natural frequency for the system with reasonable structural damping factors. (For 10 percent critical damping .. : d = 0.995 t*'.) R -NATURAL FREQUENCIES AND MODE SHAPES In calculating the natural frequencies and the mode shapes, the ternal load matrix in eq. (1) is set to zero and the displacement vector I ';! is assumed lo take the form of simple hartnonic motion or I :\ I = I 4 I Sin " t (2) Where 11> J = Relative amplitude of mode :shape vector. "' = Natural frequency of vibration. After substituting and simplifying, the equations of motion are reduced to the following form: r K 1 -1 = 1 W7 I (]) I (3) Solution to this eigenvalue problem exists only for particular values of *' which correspond to the natural frequencies of vibration of the structure. Eq. (3) is solved by iteration techniques to obtain values of w and their corresponding mode shape vectors I. C -STRUCTURAL RESPONSES After all natural frequencies and their mode shapes have been determined, the Response Spectrum is used to read off the spectral velocities corresponding to the natural frequencies. The method of Modal Analysis is then employed to calculate the structural responses. This meth o d actually simplifies the analysis of a multidegree of freedom system by using seve l*al equivalent single-degree systems. one to each Normal Mode. S i n c e the spectral displacement Sd i a directly related to th e:-spe c t r al velocity Sv by the natural W, or Sv a W Sd and. sim i larly. t h<' spectr<ll accele ra tion Sa is equal t o W Sv = wZ Sd, the structural rc:.ponses are then calculated through the £ollowine opeT"aUons

INERT IA FORCE AT XTH LEVEL tN NTH MODE: where: : -.1 n = mass at level x normaliz ed displa c ement o£ Mx in the nth mode pa rtic ipation factor o{ nth mode fi Mx(/ncn X : 1 (where: N =number of levels) spectral acceleration

= .., nSvnt !or nth mode natural frequency in nth mode The modal respon se values for each mode computed above are bined by the root*mean -square method to determine the total responses o{ t he structu re , thus N Fk-. )1/z Force per level: Fx = ( ); (5) n = 1 N "* z ) 1/z Displacement

\x = l. (6) xn per level: n = 1 where: 1\xn = f K J -l (Fxn) and [Kl = stiffness matrix of entiT"e structure N = num.ber o! modes vr -RESPONSES Th e mode shapes. periods , and responses of the Intake Structure are shown in Figure 5. and tbe mode shapes , periods. and responses of t he Equlpment Lock Building are shown in 6 and 7

• i V It -DIESEL GENERATOR FOUNDATION The diesel generator weighs 108 kp1 and operates at 900 rpm. lts f o undation is shown in Figure 8. Three indepe ndent modes of vibration were Investigated, namely, rocking, sidesway, and vertical.

Their re1pective natural frequencies are shown in the table below. Mode Natural (rpm) Frequency Rocking 595 Side sway 53 Vertical 687 The respective models and their spring constants and masse!f are shown in Flgure 9. The vertical natural frequency is the most c ritica.l. and its value is 76 percent o! the operating frequency, which is acceptable. N ' .., tr3 s* . 62.0' 42.5' J *o m.. I' I I i I . I . .I I r 1 J -i ... -f1 J ] r !' h I .., ---I I ;a ,LJ -""' . !!.. -'n *::t __ :*r-1 f '1.5 0 .. u _, --r-'* ----,..... I I . 11-, ---.. r ll 41.0' PI..A"' AT EBA8CO ARYICU INCORPORATED N.W 'YO-T 21.o* 41.C1 !4.5

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):::, *""' I N -.1 lillJ !i :I CJ I o ... r:s "'z l.Sf ... < a I ... t:g: ::3 '8 l!J " I ,.. "' -< " ' ,.. "' f .. "'"' ,... -... " -t:>* "',.m 0 j .. 1\ () "'o ., it li' :EO! .. 0.3 I I i I I I I I I I I I I I I I e o.z z = .. t= c CIC UJ irl (.) 0 cc 0.1 UJ z: "" CIC 0 0.51 DAMPING 0 0.2 0.4 0.6 0.8 RESPONSE ACCELERATION SPECTRA VERMONT YANKEE NUCLEAR POWER , JOHN A. BLUL1E & ASSOCIATES , ENGINEERS 1.0 .1.2 1.4 1.6 1.8 2J 21 2J NATURAL PERIOD (seconds) 2.8 3. VYNPS DSAR Revision 0A-160of216A.7DeletedA.8DeletedA.9DESCRIPTION,SCOPE,ANDDESIGNMETHODOLOGYUSEDFORTHEREANALYSISOFSEISMICCLASSIPIPINGSUBSEQUENTTOINITIAL OPERATIONA.9.1Introduction Since the Vermont Yankee Nuclear Power Plant became operational in 1972, there have been many modifications and upgrades to Seismic Class I piping systems.

Although several different seismic design methods were used, all methods were conservative with respect to original design criteria. The purpose of this section is to describe these methodologies, identify which methodology was used for each Seismic Class I piping system, and define the seismic design

criteria to be used for future analyses.A.9.2DescriptionofAnalysisMethodologyA.9.2.1General The purpose of this section is to briefly describe the seismic analysis methodologies used to qualify Seismic Class I piping at Vermont Yankee. Piping system design methodology consists of piping stress analysis and design qualification of pipe supports. Pipe support design methodology has been consistent throughout the life of the plant.The remainder of this section summarizes analysis methods used to analyze Seismic ClassI piping at Vermont Yankee.Piping analysis methodologies described below as Methods1, 2, and 3 refer to original design of Vermont Yankee; Methods4, 5, 6, 8 and 9 refer to designs subsequent to original startup. The chart in SectionA.7.3 correlates each piping system (or portion of piping system) to the analysis methodology

described in the remainder of this section. VYNPS DSAR Revision 0A-161of216A.9.2.2Method1 Original design of the recirculation piping was performed by General Electric Company. The Recirculation System was replaced in 1986. Method1 is no longer applicable to Vermont Yankee.A.9.2.3Method2Original design of all other Seismic ClassI piping (except branch lines and some small bore piping) was performed by the architect-engineer, Ebasco Services.A.9.2.4Method3Original design of remaining Seismic ClassI small bore piping and branch lines were designed by Ebasco Services using a chart method of analysis.A.9.2.5Method4 Torus Attached Piping (TAP) systems modified in support of the MARK I Containment Program were analyzed considering all required loading conditions, such as deadweight, thermal, seismic and MARK I loads. Dynamic seismic

analyses were performed using FSAR criteria for both Operating Basis

Earthquake (OBE) and Safe Shutdown Earthquake (SSE) loading. Load combinations and stress levels were evaluated in accordance with the MARK I

Containment Program Structural Acceptance Criteria Plant Unique Analysis Application Guide. The piping was evaluated to ASME Section III -1977 code

allowables.A.9.2.6Method5 Piping systems modified in support of various licensing commitments were performed considering all required loading conditions, such as deadweight, thermal,and seismic loads. Dynamic analyses based upon FSAR criteria and/or Regulatory Guide1.60, ground spectra, and Regulatory Guide1.61, damping values, were used for seismic analyses. Seismic analyses were performed for the SSE scenario only. The piping was evaluated to ANSIB31.1-1977 code allowables. VYNPS DSAR Revision 0A-162of216A.9.2.7Method6 This method was utilized in the Seismic Reanalysis Program (SRP). The SRP pertains to all Seismic ClassI piping not modified as a result of previous licensing commitments (Method6 does not apply to the Reactor Recirculation System). The piping was computer analyzed, with dynamic analyses used to evaluate seismic loadings. Ground spectra, based upon Regulatory Guide1.60,and floor spectra with ASME Code Case N-411 damping (conforming to Regulatory Guide1.84) defined the seismic loading for these piping systems. Seismic analyses were performed for the SSE scenario only. The piping was evaluated to ANSIB31.1-1977 code allowables.A.9.2.8DeletedA.9.2.9Method8 Piping systems modified in support of various licensing commitments were analyzed considering all required loading conditions, such as deadweight, thermal and seismic loads. Dynamic seismic analyses were performedusing the FSAR criteria. These seismic analyses considered the OBE and SSE scenarios. The piping was evaluated to ANSI B31.1-1967, 1980 code allowables.A.9.2.10Method9 This method applies to piping that is computer analyzed and considers all required loading conditions, such as deadweight, thermal, and seismic loads. Dynamic seismic analysis was used. The piping was evaluated to ANSI B31.1-1977 code allowables.A.9.3ScopeofSeismicClassIPipingSystemsModifiedandMethodology Used Piping System Part No.Description Flow Diagram Method Service Water SystemSW/1RHR Service WaterG19115961ARRUs 5 and 7 OutletG19115951BUPS OutletG1911596 VYNPS DSAR Revision 0A-163of216A.9.3ScopeofSeismicClassIPipingSystemsModifiedandMethodology Used Piping System Part No.Description Flow Diagram Method2RHR Service Water Supply Discharge, Suction, Diesel

Supply, and Cross Tie to SW

Part 9G19115962ARRUs 5 and 7 InletG19115952BRRUs 17A and 17B InletG19115952CFrom RRUs 17A and 17B to SW Part 2DG19115962D8-Inch SW 34 Ground Flow Reactor Building WestG19115963Service Water at Intake StructureG19115964Reactor Building to SWG19115964ARRUs 6 and 8 SWG19115964BRRU 9 OutletG19115966SW to RBCCW Heat ExchangerG19115966AUPS InletG19115967RBCCW Heat Exchanger Discharge and MGLO Cooler DischargeG19115967AFrom SW Part 7 to Valve SW-302 and SW-23CG19115968SW Supply in Reactor Building to Turbine BuildingG19115968ASW Supply Header in Reactor BuildingG19115969SW Supply to RBCCW Heat ExchangerG19115969AUPS InletG191159610Main SW Supply to Reactor Building and Cooling Tower ReturnG191159610ARRUs 6 and 8 InletG191159610BRRU 9 InletG1911596 VYNPS DSAR Revision 0A-164of216A.9.3ScopeofSeismicClassIPipingSystemsModifiedandMethodology Used Piping System Part No.Description Flow Diagram Method11Service Water to Alternate CoolingG191159612SW Supply to DG-AG191159613SW Supply to DG-BG191159514SW Diesel OutletG191159615SW Discharge in Turbine BuildingG191159616Supply to Seismic Cell of Cooling TowerG1911592P8-1AService Water Pump Moter CoolantG1911595P8-1BService Water Pump Motor CoolantG1911595P8-1CService Water Pump Motor CoolantG1911595P8-1DService Water Pump Motor CoolantG1911595DGSKID/1Diesel A & B -Air Jacket CoolantG1911595DGSKID/2Expansion Tank Vent to Atmosphere-DG "A"G1911596EDGV/PI 1&2From the After Cooler to the Expansion TankG1911596SBSupply Line to Circulating Water Pump Cooling SystemG1911599SBSupply Line to Chlorination SystemG1911599SBMiscellaneous Small Bore PipeG1911593,5 Condensate Storage Tank and Condensate Transfer System (CST)CST/3CST Feed from CST Storage Tank to Core Spray PumpG19117664CST Feed from CST Storage Tank to HPCI PumpG19117667RCIC Suction LineG1911766 Fuel Oil Transfer System (FO) VYNPS DSAR Revision 0A-165of216A.9.3ScopeofSeismicClassIPipingSystemsModifiedandMethodology Used Piping System Part No.Description Flow Diagram MethodFO/3Diesel Fuel OilG1911626SBMiscellaneous Small Bore PipeG1911623,5 Radwaste SystemRW/1Drywell Floor Sump Pump DischargeG19117762Drywell Equipment Drain Sump Pump DischargeG1911776SBMiscellaneous Small Bore PipeG1911775 Diesel Generator Starting Air (SA) and Exhaust (DG) SystemsDGPart 1:Diesel Generator "B" ExhaustG1911606Part 2:Diesel Generator "A" ExhaustG1911606SAPart 1:Diesel Generator "A" Starting AirG1911606Part 2:Diesel Generator "B" Starting AirG1911606SBMiscellaneous Small Bore PipeG1911603,5 Service AirSBMiscellaneous Small Bore Pipe G1911603,5 VYNPS DSAR Revision 0A-166of216A.9.4DescriptionofFloorAmplifiedResponseSpectraUsedforSeismicReanalysisofPipingA.9.4.1General In response to several I&E Bulletins issued 1979, Vermont Yankee decided to upgrade Seismic Class I piping systems outside of the drywell. The analysis

methodology was chosen to be consistent with current practices, including

Systematic Evaluation Program (SEP) acceptance criteria. For example, Amplified Response Spectra (ARS) were to be developed using Regulatory Guides1.60 and 1.61, and pipingand pipe supports were evaluated for seismic

loading using the Safe Shutdown Earthquake (SSE or MHE) condition only. ARS

so developed were used in this Seismic Reanalysis Program. However, prior to

completion of plant modifications, NRC approval was obtained to use increased damping values per Code CaseN-411 (now approved by Regulatory Guide1.84) for load reconciliation purposes. Final analyses of piping in the as-built condition were performed using these increased (ASMEN-411) damping values, whichwill constitute the damping design criteria for all future piping

analyses at Vermont Yankee.A.9.4.2FloorResponseSpectraDevelopment Floor response spectra were developed as follows:1.Synthetic time histories, whose response spectra closely simulate Regulatory Guide1.60 ground spectra, were generated for each of two

horizontal directions and for the vertical direction.2.The modal superposition time history technique of dynamic analysis was used to compute response time histories at building floor elevations, using the above synthetic time histories as input and structural damping values from Regulatory Guide1.61.3.Floor time histories were developed into floor response spectra using Regulatory Guide1.61 and ASME Code CaseN-411 damping. VYNPS DSAR Revision 0A-167of216A.9.5SummaryandConclusions Seismic design methodologies described in this section meet or exceed original licensing requirements. In particular, conservatism of piping analysis methods used subsequent to initial plant startup are evident since:1.All reanalyses were performed using modern methods.2.Branch piping was included in the piping models.3.All pipe supports were included in the piping models.4.Dynamic analysis was used to evaluate both seismic and other transient loadings.5.ARS developed from Regulatory Guide1.60 provided conservative ground motion.Reanalysis of piping systems using the methods described in SectionsA.9.2.5 to A.9.2.10 was limited to complete systems/structural boundaries. (For example, if a 12-inch diameter header was analyzed using Method5, then a 4-inch diameter branch line was also analyzed using Method5, and no mixing of analysis methodologies was permitted.)Seismic Design Method6(SectionA.9.2.7) will be used to qualify all Seismic ClassI piping systems which may be added or modified in the future.In summary, it is concluded that the Seismic ClassI piping systems at Vermont Yankee have been designed to withstand all seismic (and other) loadings and to

remain operable. VYNPS DSAR Revision 0A-168of216A.10PRIMARYSTRUCTURESEISMICANALYSISA.

10.1INTRODUCTION

This Appendix describes the mathematical model and methodology used to develop amplified response spectra for Vermont Yankee Nuclear Power Station. The actual amplified response spectra are contained in Reference A.10.5.15.A.10.2MATHEMATICALMODELDESCRIPTIONA.10.2.1GeneralDescription The structures to be analyzed for seismic loads, in this study termed as primary structures, include the Reactor Building and its internal structures. The Reactor Building complex shown in Figure A.10.2-1 consists of the Reactor

Building, drywell, shield wall, pedestal, and Reactor Pressure Vessel (RPV)

and internals. These structures are mathematically idealized by centerline beam element models as shown in Figures A.10.2-2 and A.10.2-3. Each stick is made of a series of connected massless beam elements. Masses are lumped to nodal points. The couplings between individual structures are represented by

either linear springs or rigid links. The Reactor Building, including the suppression chamber, is a complex three-dimensional structure. In this study, it is simplified into plane symmetric structures in the North-South (NS) andEast-West (EW) directions.

Uncoupled NS and EW models were developed. The effect of coupling was

accounted for by correction factors. The correction factors of 1.05 in the NS

direction and 1.15 in the EW direction are considered to encompass the overall

torsional effects. The Reactor Building internal structures, e.g., drywell, shield wall, pedestal, RPV, and internals are considered axisymmetric structures. Their mass and stiffness properties are identical in NS and EW

directions. The primary structure seismic analysis model has been constructed

based on operating conditions. Since the Reactor Building internal structures

are either directly or indirectly coupled to the Reactor Building, any effect due to asymmetry of the Reactor Building would be transmitted to its internal

structures. Therefore, the torsional correction factors defined for the

Reactor Building are considered as global correction factors and have been incorporated in the analyses as input scale factors. VYNPS DSAR Revision 0A-169of216 With the exception of the RPV and its internals, the structures are nodalized identically in the horizontal and vertical models. The RPV and internals, being stiffer in the vertical direction, are represented by different horizontal and vertical models; a 64-node horizontal model and a simplified

15-node vertical model. Because of the nature of centerline modeling, the

horizontal and vertical models can be uncoupled. The stiffness properties in

the horizontal models are defined in terms of effective shear areas and

moments of inertia. The stiffness properties in the vertical model are defined in terms of axial cross-sectional areas. The model properties are shown in TableA.10.2-1 through Table A.10.2-8. The soil is not modeled since the structures are founded on rock with a shearwave velocity of 6500 ft/sec. The effect of soil-structure interaction for plants founded on rock can be neglected. Therefore, the seismic analyses are

performed using the fixed base model. The drawings used in the construction of the seismic analysis model (see Section A.10.6) were reviewed against drawings available at the site. The site drawings reviewed also included "as-built" drawings. It was determined

that with relevance to seismic analysis work, there were insignificant

variations between site drawings and those used in the generation of the primary structure seismic analysis model. A repair to the core shroud has been installed. As a result, the structural models have been revised. The models described in this Appendix are no longer current. See AppendixK and Reference A.10.5.16 for information concerning the structural models.A.10.2.2ReactorPressureVesselandInternalsA.10.2.2.1ModelDescription The reactor pressure vessel is a 205" nominal ID cylindrical shell capped by hemispherical heads at both ends. It is supported by the RPV pedestal near the bottom. In the horizontal direction, the RPV is restrained through the

RPV stabilizers and refueling bellows by the shield wall and drywell, respectively. The RPV internals considered in the model are the fuel assemblies, guide tubes, CRD housings, steam separators, standpipes, shroud, top guide beam, and core plate. Stiffnesses of light components, such as jet

pumps, in-core assemblies, feedwater sparger, and steam dryer assembly, have not been considered, but their masses are included. The multiple-unit internal components are modeled as single-segmented beams based on the

assumptions that they move in phase under dynamic loads. VYNPS DSAR Revision 0A-170of216 The RPV and internals are modeled in detail in the horizontal direction (Figure A.10.2-2). The core plate and top guide beam are modeled as rigid links due to their large in-plant stiffnesses. The shroud support plate is

rigidly linked to the vessel wall in the translational direction and has a finiterotational stiffness. The hydrodynamic mass effect is considered to

account for the coupling of the RPV and internal components by the water in

between.In the vertical direction, the RPV and internals are relatively stiff. A simplified model is thus developed for this direction, and is shown in FigureA.10.2-3. The core plate and top guide beam do not provide linkage

between the shroud and fuel assemblies/guide tubes; thus, only their masses

are considered. The CRD housings below the vessel are not explicitly included

in the model due to their high vertical stiffnesses. The hydrodynamic mass coupling effect is neglected, but the water masses are added to the

appropriate nodes. The RPV is considered uncoupled from the shield wall and drywell in the vertical direction due to the absence of the RPV stabilizer

action and very flexible refueling bellows.A.10.2.2.2ModelProperties For the RPV and internals horizontal and vertical models, the nodal elevations and masses are presented in Tables A.10.2-2and A.10.2-3, respectively. The

horizontal RPV and internals model is represented by Nodes 1 through 64 in

Table A.10.2-2. The vertical RPV and internals model is represented by Nodes1 through 15 in Table A.10.2-3. The summary of the hydrodynamic massesin the horizontal direction is shown in Table A.10.2-1.The element section properties are presented in Tables A.10.2-4 through A.10.2-6. Spring element stiffnesses are presented in Tables A.10.2-7 and

A.10.2-8. The horizontal model is represented by Elements 1 through 62 in Tables A.10.2-4 and A.10.2-5, and the vertical model by Elements 1 through 14 in Table A.10.2-6. Due to the axisymmetric nature of the model, the same nodal masses and element properties are applicable in both the NS and EW directions.A.10.2.3ShieldWall/PedestalModel VYNPS DSAR Revision 0A-171of216A.10.2.3.1ModelDescriptionThe shield wall consists of either 1/4" or 1-1/2" thick outer steel plates and 1/4" thick inner steel plates connected together by wide flanged steel columns placed symmetrically around the circumference. The outer and inner steel liner plates enclose a concrete fill whose structural properties were not considered, but whose weight was considered for calculation of nodal masses. Inner radius of the shield wall is 10' x 3/16" and the outer radius is 12'x5-3/16". The top of the shield wall is horizontally connected to the drywell by the star truss and to the RPV by the RPV stabilizers. This Stabilizer System, shown in Figure A.10.2-4, provides coupling of the shield wall to the drywell and the RPV in the horizontal direction only. The pedestal is made up of reinforced concrete with 1/4" thick outer steel ring plates and 1-1/4" thick inner steel ring plates. The concrete is assumed uncracked for section property calculations. Theinner radius of the pedestal

is 8' x 7" and the outer radius is 12' x 7". The bottom of the pedestal is embedded in the Reactor Building concrete which anchors the drywell, as well. There are two platforms spanning from the pedestal/shield wall to the drywell at Elevations 268.98' and 250.98'. The structural elements of the platforms

are radial beams and intermediate lateral restraint beams. The connections at

the drywell end are slotted joints in the horizontal direction. In the vertical direction, the end connections of the radial beams are essentially pin-pin joints. Therefore, no stiffness couplings were considered between the

shield wall/pedestal and the drywell through the platform beams. The models were developed using Drawings A.10.6.47 through A.10.6.54. The effects of rotational inertias on the natural frequencies of the uncoupled

shield wall/pedestal model were studied. The results showed that the rotary

inertias have minimal effects on the frequencies of interest and are thus neglectedin the model. The shield wall/pedestal is an axisymmetric structure. The nodal masses, element, and material properties in both NS and EW directions are identical.A.10.2.3.2ModelProperties The shield wall/pedestal is represented by Nodes 65 to 77 for the horizontal and vertical models in Figures A.10.2-2 and A.10.2-3.The nodal elevations, locations, and masses are indicated on Figures A.10.2-2and A.10.2-3 and in Tables A.10.2-2 and A.10.2-3. VYNPS DSAR Revision 0A-172of216 The shield wall/pedestal has numerous penetrations. The section properties were calculated with due consideration given to the openings. The section

properties of the pedestal concrete were calculated conservatively without

considering the steel reinforcement. The beam element properties are given in TablesA.10.2-4 through A.10.2-6.A.10.2.4DrywellModelA.10.2.4.1ModelDescriptionThe drywell essentially is a thin steel shell structure. It is a bulb-shaped, welded steel plate assembly embedded in the Reactor Building concrete at Elevation 235.48' andlaterally connected at Elevation 309.96' to the Reactor Building by the shear lugs and to the shield wall by the star truss. Furthermore, the drywell head is coupled to the reactor pressure vessel through the refueling bellows. The geometry and major dimensions of the drywell are shown in Figures A.10.2-1 and A.10.2-5. The drywell is considered

uncoupled from the suppression chamber because of the flexible bellows used in the connecting vent pipes. The Stabilizer System is shown in FigureA.10.2-4. There are two platforms spanning from the drywell to the shield wall/pedestal at Elevations 268.98' and 250.98'. The structural elements of these platforms

are radial beams and intermediate lateral restraint beams. The connections to

the drywell are slotted joints in the horizontal direction. The connections to both the drywell and the shield wall/pedestal ends are pin-pin joints in

the vertical direction. Therefore, no stiffness couplings were considered

between the drywell and shield wall/pedestal through these platforms.The drywell was originally modeled by Sexton & Associates (ReferenceA.10.5.1) in the form of a centerline beam model consisting of 24 lumped masses. For this study, the Sexton model was reviewed and spot-checked using the data in

Reference A.10.5.1, and Drawings A.10.6.55 and A.10.6.56. The following modifications to the model were made as a result of the review of the Sexton

model:1.The weights of the two platforms, spray headers, and liners were added. This results in approximately 15% more weight than that included in the Sexton model.2.The section properties at lower elevations were recalculated to account for the nonuniform plate thickness around the circumference. VYNPS DSAR Revision 0A-173of2163.The equivalent cylinder beam element modeling approach used for the noncylindrical portions of the drywell was not adequate for obtaining the

natural frequencies of the drywell as a whole, due to the large spherical

portion (about 50% of the drywell in height) and its direct connection

with the support. The uncoupled drywell, fixed at Elevation 235.48' and

simply supported at Elevation 309.96' was first analyzed for natural

frequencies for both the beam and shell models in the horizontal and the vertical directions. The vertical and horizontal beam-type frequencies of

the shell model were extracted from the zero and first Fourier harmonic

terms, respectively, using ASHSDO2 code (Reference A.10.6.10). The

results showed that the beam models overestimate the fundamental frequencies of the drywell in both directions byroughly a factor of 2.

Accordingly, the Young's modulus was adjusted for the beam models such

that their horizontal and vertical fundamental frequencies match with

those calculated for the shell model. The results from the modal frequency analyses of the uncoupled drywell by considering shear lugs connecting the drywell to the Reactor Building as a

spring element and also as a rigid connection revealed negligible change

in the natural frequencies. Modeling shear lugs as a rigid link is

therefore adequate. The drywell is considered as an axisymmetric structure and its properties in both the NS and EW directions are identical.A.10.2.4.2ModelProperties The drywell is represented by Nodes 78 to 103 for the horizontal and vertical models in Figures A.10.2-2 and A.10.2-3.The nodal elevations, locations, and masses are given in Tables A.10.2-2 and A.10.2-3 and in Figures A.10.2-2 and A.10.2-3.The beam element properties are presented in Tables A.10.2-4 through A.10.2-6.A.10.2.5ReactorBuildingModelA.10.2.5.1ModelDescriptionThe Reactor Building, Figure A.10.2-1, is a reinforced concrete structure rising from a 6' thick basemat at Elevation 213.75' to the refueling floor at

Elevation 345.16'. The portion of the building above the refueling floor is of braced structural steel construction. VYNPS DSAR Revision 0A-174of216 These models are centerline beam models developed from a spring element model which expresses the building properties in terms of direct mass and stiffness

matrices. Structural masses were lumped to each floor level. The weights of

the spent fuel pool water and torus in operating condition, not included in

the Vermont Yankee supplied models, were added and distributed to appropriate

nodes. Two massless nodes at Elevations 309.96' and 235.48' were introduced

to provide structural coupling with the drywell. Section properties for the

steel portion of the building between Elevations 345.16' to 387.42' have been

adjusted to correspond to concrete material properties. As mentioned above, the Reactor Building models used in this analysis were made equivalent to a spring element model. To ensure the validity of equivalence, the bending rotation is not allowed in the horizontal models. Consequently, each node has only a translational degree of freedom

unrestrained. Three uncoupled (NS, EW, and vertical) models were used in the analysis. Due to asymmetry of the Reactor Building, coupled translational and torsional

responses would result in structures excited under horizontal ground motion.

The extent of coupling would depend on the magnitude of the eccentricities and the relation between the uncoupled translational frequencies and uncoupled torsional frequencies. In lieu of using a coupled horizontal model, torsional

correction factors of 1.05 and 1.15 were used to increase the results of the

uncoupled NS and EW analyses, respectively, to account for the overall

torsional effects.A.10.2.5.2ModelProperties The Reactor Building is represented by Nodes 105 to 113 in the horizontal direction (Figure A.10.2-2) and by Nodes 104 to 113 in the vertical direction (Figure A.10.2-3). The top node number 104 is inactive in the horizontal

model.The nodal elevations, locations, and masses are presented in Tables A.10.2-2and A.10.2-3 and in Figures A.10.2-2 and A.10.2-3.The beam element properties are given in Tables A.10.2-4 through A.10.2-6. A.10.2.6NaturalModesofCombinedModelThe eigenvalues for the combined models are presented in Tables A.10.2-9through A.10.2-11. Table A.10.2-9 presents results for the NS direction. Table A.10.2-10 shows results for the EW direction. The vertical direction results are presented in Table A.10.2-11. Frequencies up to 33 Hz were

extracted. VYNPS DSAR Revision 0A-175of216A.10.2.7Damping For the Vermont Yankee plant, the SSE analysis used 2% of the critical damping for Control Rod Drive (CRD) guide tubes and housings, 6% for the fuel assemblies, and 4% for the RPV vessel, refueling bellows, RPV stabilizers, star truss, and other internal components. For the drywell, 4% damping was used, and 7% for Reactor Building, biological shield wall, and pedestal. For the Operating Basis Earthquake (OBE) analysis, 1% of the critical damping was used for CRD guide tubes and housings, 6% for the fuel assemblies (horizontal), 2% for the RPV, refueling bellows, RPV stabilizers, star truss, and other internal components. Two percent (2%) for the drywell and 4% for the Reactor Building, shield wall, and pedestal were used. For the vertical analysis, damping values are the same except 4% damping for the fuel assemblies was used for the OBE analysis. Damping values used in this analysis are based on Reference A.10.5.3, except fuel assemblies for which damping under horizontal excitations is based on

test results.A.10.3SEISMICINPUTDEFINITION This section describes the input time histories for the Vermont Yankee primary structure seismic analysis.A.10.3.1DesignGroundMotion The maximum ground accelerations at the site are specified as a 0.14g Safe Shutdown Earthquake (SSE) and 0.07g Operating Basis Earthquake (OBE). The above-defined maximum ground accelerations are used for all three components of the design ground motion -NS, EW, and vertical. It should be noted that

the maximum vertical ground acceleration specified in Reference A.10.5.6 is

two-thirds of the maximum horizontal ground acceleration. According to

Reference A.10.5.14, the spectral acceleration of the vertical design response

spectra reaches the maximum vertical ground acceleration at 50 Hz, but maintains the maximum horizontal ground acceleration at33 Hz. Since the frequencies of interest in seismic analysis are up to 33 Hz, the maximum horizontal ground acceleration was specified as Zero Period Acceleration (ZPA) for the vertical synthetic time history to ensure its spectra will envelop the NRC Regulatory Guide 1.60 vertical design spectra. The design response spectra, are broad-band spectra constructed according to the NRC Regulatory

Guide 1.60 requirements with the scale factors equal to the maximum ground accelerations specified for the site. VYNPS DSAR Revision 0A-176of216A.10.3.2DesignTimeHistories For subsequent seismic analysis, the input motions in the form of acceleration time histories are required. Three synthetic time histories were generated and calibrated to the NRC Regulatory Guide 1.60 design response spectra. They

are mutually orthogonal and statistically independent. Two of them are

horizontal components designated at H1 and H2. The other is the vertical

component. The H1 component is more conservative than the H2 component.

Also, the EW torsional correction factor is larger than the NS component. To

minimize the degree of conservatism, the H1 component was used for the NS time history analysis, and H2 for the EW analysis. These synthetic time histories are digitized at a constant time step of 0.01second and have a duration of 22 seconds, except the H1 component which has a duration of 20.48 seconds. The plots of the time histories having a maximum acceleration of 0.14g are shown in Figures A.10.3-1, A.10.3-2, and

A.10.3-3.The criteria for the adequacy of the synthetic time histories are based on Article II.1.b of Reference A.10.5.6. In addition, consideration is given to

the frequency/amplification content of the synthetic time histories in

comparison with the eigenvalues of the primary structure. The percent critical damping for the elements of the primary structure range from 1.0% to 7.0%. Consequently, comparisons between the synthetic time

history spectra and Regulatory Guide 1.60 spectra are made for plots corresponding to 1/2%, 2%, 5%, and7% oscillator damping which bound the

primary structure model damping range. Plots of the H1, H2, and vertical response spectra comparison between the synthetic time history spectra and Regulatory Guide 1.60 spectra for 2% damping are provided in FiguresA.10.3-4 through A.10.3-6. The vertical time history has been scaled up by a factor of 1.1 in the analysis. The scaling

was required to satisfy the criteria of Reference A.10.5.6.A.10.3.3InputLocation The synthetic acceleration time histories are applied at the basemat of the primary structure model. The free-field input motion can be applied directly

to the fixed base model because the shear wave velocity for the site subgrade is 6500 ft/sec, which can be regarded as rock formation. VYNPS DSAR Revision 0A-177of216A.10.4METHODOFANALYSIS The spe cif ic met hod of ana lys is, inc lud ing the ide ntity of computer programs used is contained in Reference A.10.5.15. A

general summary is provided below:SeismicMethodology for Primary Structure Seismic Analysis Seismic Analysis Item Seismic Analysis MethodDesign Basis Input MotionRegulatory Guide 1.60 for Free-Field (Reference A.10.5.2)Dynamic AnalysisModal Superposition Using Base Support Acceleration Time History AnalysisDampingPrimary Structure Damping According to Regulatory Guide 1.61 (ReferenceA.10.5.3) and SectionA.10.2.7Dynamic ModelMathematical Centerline Beam Element Model VYNPS DSAR Revision 0A-178of216A.

10.5REFERENCES

1.FSAR for VermontYankee Nuclear Power Station, Vernon, Vermont.2.USNRC Regulatory Guide 1.60, "Design Response Spectra for Seismic Design of Nuclear Power Plants," Revision 1, December 1973.3.USNRC Regulatory Guide 1.61, "Damping Values for Seismic Design of Nuclear PowerPlants," October 1973.4.USNRC Regulatory Guide 1.92, "Combining Model Responses and Spatial Components in Seismic Response Analysis," Revision 1, February 1976.5.USNRC Regulatory Guide 1.122, "Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components," Revision 1, February 1978.6.USNRC Standard Review Plan, 3.7.1, "Seismic Design Parameters," Revision1, July 1981.7.USNRC Standard Review Plan, 3.7.2, "Seismic System Analysis," Revision 1, July 1981.8.SAP4G07User's Manual, "Static and Dynamic Analysis of Mechanical and Piping Components by Finite Element Method," G. C. Mok (Editor), NEDO-10909, Revision 7, December 1979.9."SPECA04 User's Manual," J. M. Lee, NEDE-25181, September 1979 (User's Manual for SPECA04 and SPECA05 are the same).10.ASHSD01 User's Manual, G. C. Mok, NEDE-23646, Revision 0, July 1977, and ASHSD02 User's Manual, G. C. Mok, NEDE-23646, Revision 2, January 1983.11."RINEX01 User's Manual," K. K. Fujikawa, NEDE-30824, November 1984.12.NUREG/CR-0098, "Development of Criteria for Seismic Review of Selected Nuclear Power Plants," Newmark and Hall.13.ASME CODE N-411, "Alternative Damping Values for Seismic Analysis of Piping Section III, Divisions 1, 2, and 3."14.Journal of the Power Division, November 1973, "Seismic Design Spectra for Nuclear Power Plants," N. M. Newmark, J. A. Blume, and K. K. Kapur. VYNPS DSAR Revision 0A-179of21615.General Electric Company Report 23A4591, "Primary Structure Seismic Analysis," Revision 1, June 1985.16.EDCR 95-406, "Core Shroud Modifications." VYNPS DSAR Revision 0A-180of216A.10.6aDRAWINGS Reactor Pressure Vessel Drawings Description GE Drawing No. EBASCO Drawing No.*15Reactor104R940, Rev. 105920-3773, Rev. 10 -3774, Rev.716RV Stabilizer922D166, Rev, 05920-5540, Rev. 017CRD Housing Support730E492-Sheet 1 (Rev. 1)-Sheet 2 (Rev. 2) 5920-564, Rev. 1 -565, Rev. 2*18Refueling Bellows919D218, Rev. 55920-17, Rev. 619Reactor Primary System WTS and Volumes 729E267, Rev. 1 9 5920-276, Rev. 1 -277, Rev. 1 -278, Rev. 120Reactor Vessel919D294-Sheet 1 (Rev. 8)-Sheet 2 (Rev. 7) -Sheet 3 (Rev. 1) -Sheet 4 (Rev. 6) -Sheet 5 (Rev. 3)-Sheet 6 (Rev. 3) -Sheet 7 (Rev. 10) 5920-18, Rev. 8 -19, Rev. 7

-20, Rev. 1

-21, Rev. 6

-22, Rev. 3

-23, Rev. 3

-24, Rev. 1021Core Structure729E959, Rev. 15920-532, Rev. 122Steam Separator730E165, Rev. 15--23CRD Housing919D260, Rev. 16--24CRD Guide Tube885D686, Rev. 255920-5002, Rev. 2325Control Rod706E855, Rev. 25--26In-Core Housing117C1419, Rev. 22--*27In-Core Guide Tube129B3588, Rev. 125920-2417, Rev. 728Steam Dryer729E956-Sheet 1 (Rev. 3)-Sheet 2 (Rev. 1) 5920-493, Rev. 2 -494, Rev. 1*29Steam Dryer Unit885D738, Rev. 75920-343, Rev. 830Hold Down (Top Guide)886D688, Rev. 35920-6018, Rev. 3 31Shroud729E959-Sheet 1 (Rev. 3)-Sheet 2 (Rev. 0) 5920-528, Rev. 3 -529, Rev. 032Top Guide729E959, Rev. 25920-531, Rev. 233Core Support729E957, Rev. 35920-539, Rev. 334Shroud Head and Separator730E750, Rev. 45920-576, Rev. 435Shroud Head Bolt920D228, Rev. 0--*36Jet Pump730E438, Rev. 95920-4799, Rev. 737Control Rod (Outline)814E749, Rev. 7-- 38CRD Guide Tube (Outline)919D914, Rev. 6-- 39CRD (Assembly)761E387, Rev. 18--*40CRD (Outline)846D298, Rev. 105920-6015, Rev. 6Note:The drawings identified by an "*" are drawings which had deviation between GE San Jose and site revision levels. The drawings were reviewed, and it was determined that with reference to seismic analysis work, there were insignificant variations

between site drawings and those used in the generation of the primary structure

seismic analysis model. VYNPS DSAR Revision 0A-181of216A.10.6bDRAWINGS Reactor Pressure Vessel Drawings (Continued) Description GE Drawing No. EBASCO Drawing No.*41CRD Housing (Outline)919D274, Rev. 95920-2410, Rev. 7*42In-Core Housing (Outline)117C1684, Rev. 105920-2409, Rev. 643Steam Separator (Outline)920D230, Rev. 45920-567, Rev. 4 44Steam Separator (Inlet and Discharge)161F302, Rev. 13--45Steam Separator (Inlet)920D239, Rev. 16--46Reactor Pressure Vessel Vendor Drawings 213A8464, Rev. 0 Shield Wall / Pedestal Drawings47Vessel Support Arrangement719E450, Rev. 35920-44, Rev. 3*48Shield Wall and Pedestal Arrangements729E525-Sheet 1 (Rev. 2)5920-491, Rev 349Primary Containment Loading729E252, Rev. 45920-46, Rev. 450Recirculation Loop Suspension729E909, Rev. 65920-422, Rev. 6 -423, Rev. 451Piping Restraints729E911, Rev. 55920-424, Rev. 5 -425, Rev. 552Sec Containment Floor Loading729E290, Rev. 15920-49, Rev. 1 53Motor Removal Recirc Loop Pump729E477, Rev. 15920-421, Rev. 1 54Shield Wall / Pedestal Vendor Drawing 213A8463, Rev. 0 Drywell Drawings / Documents*55Primary and Secondary Containment729E243-Sheet 10, (Rev. 1)5920-13, Rev.256Drywell Vendor Drawings213A8466, Rev. 057GE Document, 21A5837, Revision 3, "Drywell and Suppression Chambers Containment Vessels"Note:The drawings identified by an "*" are drawings which had deviation between GE San Jose and site revision levels. The drawings were reviewed, and it was determined that with reference to seismic analysis work, there were insignificant variations

between site drawings and those used in the generation of the primary structure

seismic analysis model. VYNPS DSAR Revision 0A-182of216 TABLE A.10.2-1a Summary of Nodal Hydrodynamic Masses for Horizontal DirectionNODE POINTHYDRODYNAMIC MASS (lb-sec 2/in)IJKDIAGONALOFF-DIAGONALM(I,I)M(J,J)M(K,K)M(I,J)M(J,K)5839104.6337.84-54.275738178.6548.39-69.975637148.9065.19-76.525536309.80226.05-251.635435350.94250.06-275.91533418232.01219.0835.96-170.21-48.87523317513216503115493014232.01219.0835.96-170.21-48.874829185.57129.42-136.13472812224.23200.0734.76-164.50-35.57462711244.90218.5137.96-179.66-38.85452610228.79204.1335.47-167.84-36.29449117.1242.11-43.1843885.8030.39-31.2441741.5312.81-13.7024629.036.35-7.6252.58 VYNPS DSAR Revision 0A-183of216 TABLE A.10.2-2aHorizontal Model -Nodal Masses/ElevationsNODE NO.DESCRIPTIONELEVATION (IN)MASS (LB-SEC 2/IN)1 2 3 4 5 6 CRD HOUSINGS-132.81-99.00 -66.00 -33.00 0.0 21.9 14.166 27.993 27.654 27.655 46.172 17.678 7 8 9 10 11 12 13 CRD GUIDE TUBES 42.19 59.72 98.25 129.00 150.00 186.00 202.19 24.026 22.446 27.740 20.721 22.823 20.897 62.019 14 15 16 17 18 19 FUEL 229.19 256.19 283.19 310.19 337.19 364.19 111.078 111.078 111.078 111.078 111.078 55.539 20 21 22 23 24 CRD HOUSINGS-132.81-99.00 -66.00 -33.00 21.90 14.166 27.993 27.654 36.830 84.849 25 26 27 28 29 SHROUD 98.25 129.0 150.0 186.0 202.19 0.0 18.941 20.862 19.102 62.521 VYNPS DSAR Revision 0A-184of216 TABLE A.10.2-2bNODE NO.DESCRIPTIONELEVATION (IN)MASS (LB-SEC 2/IN)30 31 32 33 34 35 36 37 SHROUD 229.19 256.19 283.19 310.19 337.19 364.19 402.75 426.75 19.764 19.764 19.764 19.764 19.764 50.367 33.890 33.456 38 39 40 STEAM SEPARATORS 466.0 510.0 554.0 58.577 66.842 30.827 41 43 44 BOTTOM HEAD 42.19 59.72 98.25 48.33 112.57 176.785 45 46 47 48 49 50 51 52 53 54 55 56 57 58 RPV 129.0 150.0 186.0 202.19 229.19 256.19 283.19 310.19 337.19 364.19 402.75 426.75 466.0 510.0 79.856 87.958 80.535 66.647 80.533 74.045 74.045 74.045 74.045 90.994 87.76 88.728 116.785 101.0 VYNPS DSAR Revision 0A-185of216 TABLE A.10.2-2cNODE NO.DESCRIPTIONELEVATION (IN)MASS (LB-SEC 2/IN)59 60 61 RPV 537.0 595.0 644.0 208.79 220.15 186.69 62 63 64 RPV TOP HEAD 682.25 719.87 758.88 139.28 37.92 9.4865SHIELD WALL TOP529.92194.10 66 67 68 69 70 71 72 SHIELD WALL 490.44 414.36 351.48 280.92 210.48 126.96 24.96 372.67 434.78 438.66 489.13 631.47 687.11 749.2273TOP OF PEDESTAL-27.00809.13 74 75 76 77 PEDESTAL-84.00-148.08 -215.04 263.04 584.89 654.76 740.17 1120.6078DRYWELL TOP HEAD770.4086.1879JNCT. DRYWELL & DRYWELL TOP HEAD729.9665.73 VYNPS DSAR Revision 0A-186of216 TABLE A.10.2-2dNODE NO.DESCRIPTIONELEVATION (IN)MASS (LB-SEC 2/IN)80 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 100 101 102 103 DRYWELL 654.96 618.96 555.96 516.48 492.96 444.96 396.96 336.96 272.28 237.36 201.24 152.52 97.32 45.36-6.60-58.32-110.04 -152.04 -194.04 -235.80 -275.04 -295.68 -311.04 -347.04 248.45 55.64 111.02 Massless 73.24 29.50 32.87 38.04 75.31 96.01 104.81 78.93 70.39 202.90 185.04 89.54 132.76 92.39 104.30 83.59 54.61 1022.77 45.29 Massless 104 105 106 107 108 109 110 111 112 REACTOR BUILDING 1146.00 1238.88 938.88 621.00 516.48 432.96 156.96-173.28-377.28 Inactive 3034.90 34342.75 32840.17 Massless 31476.66 46318.94 70323.58 Massless113TOP OF BASEMAT-638.04106685.25 VYNPS DSAR Revision 0A-187of216 TABLE A.10.2-3aVertical Model -Nodal Masses/ElevationsNODE NO.DESCRIPTIONELEVATION (IN)MASS (LB-SEC 2/IN)1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 65 66 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 85 86 FUEL GUIDE TUBES RPV BOTTOM HEAD SHROUD RPV SHIELD WALL TOP SHIELD WALL TOP OF PEDESTAL PEDESTAL DRYWELL TOP HEAD JNCT. DRYWELL &DRYWELLTOP HEAD DRYWELL 364.19 202.19 0.0 426.75 402.75 283.19 98.25 595.0 466.0 364.19 186.0 150.0 98.25 59.72 42.19 529.92 490.44 414.36 351.48 280.92 210.48 126.96 24.96-27.00 -84.00-148.08 -215.04 -263.04 770.40 729.96 654.96 618.96 555.96 516.48 492.96 444.96 396.96 100.03 604.56 371.68 208.32 259.12 98.82 513.60 802.08 306.51 400.89 301.76 127.89 301.75 112.57 375.97 194.10 372.67 434.78 438.66 489.13 631.47 687.11 684.52 744.43 584.89 654.76 703.93 1120.60 86.18 65.73 248.45 55.64 111.02 Massless 73.24 29.50 32.87 NOTE: The mathematical properties above are representative only and should not be used for design purposes. For design input properties and the analysis of record, consult EDCR 95-406, "Core Shroud Modifications" and

EDCR 97-422, "Cycle 20 Reload." VYNPS DSAR Revision 0A-188of216 TABLE A.10.2-3bVertical Model -Nodal Masses/ElevationsNODE NO.DESCRIPTIONELEVATION (IN)MASS (LB-SEC 2/IN)87 88 89 90 91 92 93 94 95 96 97 98 99 100 101 102 103 104 105 106 107 108 109 110 111 112 113 DRYWELL REACTOR BUILDING TOP OF BASEMAT 336.96 272.28 237.36 201.24 152.52 97.32 45.36-6.60-58.32-110.04 -152.04 -194.04 -235.80 -275.04-295.68-311.04 -347.04 1146.00 1238.88 938.88 621.00 516.48 432.96 156.96-173.28 -377.28 -638.04 38.04 75.31 96.01 104.81 78.93 70.39 267.34 249.48 89.54 132.76 92.39 129.66 96.27 54.61 1022.77 45.29 Massless 952.78 2082.12 34342.75 32840.17 Massless 31476.66 46318.94 70323.58 Massless 106685.25 NOTES:1.Nodes 16-64 are dummy nodes.2.Elevations are with respect to vessel 0.0 VYNPS DSAR Revision 0A-189of216 TABLE A.10.2-4aHorizontal Model -Beam Element Properties North-South Direction ELEMENT NUMBER I NODE JNODEDESCRIP. LENGTH (IN)SHEAR AREA (IN 2)I (IN 4)E (LB/IN 2)POISSON'S RATIO ()1 2 3 4 5 6 1 2 3 4 5 6 2 3 4 5 6 7 33.81 33.00 33.00 33.00 21.9 20.29 176.44 176.44 176.44 217.61 217.61 435.21 1364.74 1364.74 1364.74 1613.57 1613.57 3227.14 27.6x10 6 27.6x10 6 27.6x10 6 27.6x10 6 25.55x10 6 25.55x10 6 0.3 0.3 0.3 0.3 0.3 0.3 7 8 9 10 11 12 7 8 9 10 11 12 8 9 10 11 12 13 17.53 38.53 30.75 21.0 36.0 16.19 244.17 244.17 6840.5 6840.5 25.55x10 6 25.55x10 6 0.3 0.3 13 14 15 16 17 18 13 14 15 16 17 18 14 15 16 17 18 19 27.0 27.0 27.0 27.0 27.0 27.0 304.89 304.89 2886.96 2886.96 11.03x10 6 11.03x10 6 0.41 0.41 19 20 21 22 20 21 22 23 21 22 23 24 33.81 33.0 33.0 54.9 176.44 176.44 176.44 217.61 1364.74 1364.74 1364.74 1613.57 27.6x10 6 27.6x10 6 27.6x10 6 27.6x10 6 0.3 0.3 0.3 0.3 23 24 25 26 27 28 29 30 31 25 26 27 28 29 30 31 32 33 26 27 28 29 30 31 32 33 34 30.75 21.00 36.00 16.19 27.00 27.00 27.00 27.0 27.0 444.64 444.64 445.5 451.5 2.909x10 6 2.909x10 6 4.305x10 6 3.05x10 6 25.55x10 6 25.55x10 6 0.3 NOTE: The mathematical properties above are representative only and should not be used for design purposes. For design input properties and the analysis of record, consult EDCR 95-406, "Core Shroud Modifications" and

EDCR 97-422, "Cycle 20 Reload." VYNPS DSAR Revision 0A-190of216 TABLE A.10.2-4b ELEMENT NUMBER I NODE J NODE DESCRIP.LENGTH (IN)SHEAR AREA (IN 2)I (IN 4)E (LB/IN 2)POISSON'S RATIO ()32 33 34 34 35 36 35 36 37 27.0 38.56 24.0 465.99 526.81 300.89 1.230x10 7 8.847x10 6 1.227x10 6 25.55x10 6 0.3 35 36 37 37 38 39 38 39 40 39.25 44.0 44.0 492.78 1210.5 375.25 29130.5 30994.0 20960.11 25.55x10 6 0.3 38 39 40 5 24 43 24 41 44 21.9 20.29 38.53 969.87 1455.28 1740.04 2.376x10 6 2.019x10 7 18.49x10 6 27.00x10 6 0.3 41 42 43 44 45 46 47 48 49 50 51 52 53 54 55 56 57 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 30.75 21.00 36.0 16.19 27.00 27.00 27.00 27.00 27.00 27.00 38.56 24.00 39.25 44.0 27.0 58.0 49.0 1769.4 1765.7 1765.7 2108.89 3.356x10 7 1.979x10 7 1.979x10 7 5.914x10 7 27.00x10 6 0.3 58 59 60 61 62 63 62 63 64 38.25 37.62 39.01 3948.0 1030.0 331.1 1.130x10 8 1.043x10 7 3.340x10 6 27.00x10 6 0.3614143B. HEAD17.531781.01.986x10 7 27.00x10 6 0.3 VYNPS DSAR Revision 0A-191of216 TABLE A.10.2-4c ELEMENT NUMBER I NODE J NODE DESCRIP.LENGTH (IN)SHEAR AREA (IN 2)I (IN 4)E (LB/IN 2)POISSON'S RATIO ()627343SKIRT86.72636.261.125x10 7 28.3x10 6 0.3 63 64 65 66 67 68 69 70 65 66 67 68 69 70 71 72 66 67 68 69 70 71 72 73 39.48 76.08 62.88 70.56 70.44 83.52 102.00 51.96 494.95 475.22 494.95 1001.85 1001.85 431.20 494.95 500.45 9.16x10 6 8.98x10 6 9.16x10 6 20.59x10 6 20.59x10 6 7.99x10 6 9.16x10 6 9.26x10 6 29.07x10 6 0.30 71 72 73 74 75 73 74 75 76 77 74 75 76 77 103 57.00 64.08 66.96 48.00 84.00 20752.50 22508.10 338.39x10 6 366.16x10 6 3.45x10 6 0.25 76 77 78 79 79 80 40.44 75.00 650.88 636.48 16.10x10 6 16.80x10 6 9.70x10 6 0.30 78 79 80 81 82 83 84 85 86 87 88 89 90 91 80 81 82 83 84 85 86 87 88 89 90 91 92 93 81 82 83 84 85 86 87 88 89 90 91 92 93 94 36.00 63.00 39.48 23.00 48.00 48.00 60.00 64.68 34.92 36.12 48.72 55.20 51.96 51.96 764.64 707.04 777.60 777.60 396.00 396.00 396.00 396.00 1586.88 1730.88 568.80 659.52 727.20 766.08 20.20x10 6 23.00x10 6 30.70x10 6 30.70x10 6 15.60x10 6 15.60x10 6 15.60x10 6 15.60x10 6 63.80x10 6 82.70x10 6 40.20x10 6 62.50x10 6 82.60x10 6 97.50x10 6 9.70x10 6 0.30 VYNPS DSAR Revision 0A-192of216 TABLE A.10.2-5aHorizontal Model -Beam Element Properties East-West Direction ELEMENT NUMBER I NODE JNODEDESCRIP. LENGTH (IN)SHEAR AREA (IN 2)I (IN 4)E (LB/IN 2)POISSON'S RATIO ()1 2 3 4 5 6 1 2 3 4 5 6 2 3 4 5 6 7 33.81 33.00 33.00 33.00 21.9 20.29 176.44 176.44 176.44 217.61 217.61 435.21 1364.74 1364.74 1364.74 1613.57 1613.57 3227.14 27.6x10 6 27.6x10 6 27.6x10 6 27.6x10 6 25.55x10 6 25.55x10 6 0.3 0.3 0.3 0.3 0.3 0.3 7 8 9 10 11 12 7 8 9 10 11 12 8 9 10 11 12 13 17.53 38.53 30.75 21.0 36.0 16.19 244.17 244.17 6840.5 6840.5 25.55x10 6 25.55x10 6 0.3 0.3 13 14 15 16 17 18 13 14 15 16 17 18 14 15 16 17 18 19 27.0 27.0 27.0 27.0 27.0 27.0 304.89 304.89 2886.96 2886.96 11.03x10 6 11.03x10 6 0.41 0.41 19 20 21 22 20 21 22 23 21 22 23 24 33.81 33.0 33.0 54.9 176.44 176.44 176.44 217.61 1364.74 1364.74 1364.74 1613.57 27.6x10 6 27.6x10 6 27.6x10 6 27.6x10 6 0.3 0.3 0.3 0.3 23 24 25 26 27 28 29 30 31 25 26 27 28 29 30 31 32 33 26 27 28 29 30 31 32 33 34 30.75 21.00 36.00 16.19 27.00 27.00 27.00 27.0 27.0 444.64 444.64 445.5 451.5 2.909x10 6 2.909x10 6 4.305x10 6 3.05x10 6 25.55x10 6 25.55x10 6 0.3 NOTE: The mathematical properties above are representative only and should not be used for design purposes. For design input properties and the analysis of record, consult EDCR 95-406, "Core Shroud Modifications" and

EDCR 97-422, "Cycle 20 Reload." VYNPS DSAR Revision 0A-193of216 TABLE A.10.2-5b ELEMENT NUMBER I NODE JNODEDESCRIP. LENGTH (IN)SHEAR AREA (IN 2)I (IN 4)E (LB/IN 2)POISSON'S RATIO ()32 33 34 34 35 36 35 36 37 27.0 38.56 24.0 465.99 526.81 300.89 1.230x10 7 8.847x10 6 1.227x10 6 25.55x10 6 0.3 35 36 37 37 38 39 38 39 40 39.25 44.0 44.0 492.78 1210.5 375.25 29130.5 30994.0 20960.11 25.55x10 6 0.3 38 39 40 5 24 43 24 41 44 21.9 20.29 38.53 969.87 1455.28 1740.04 2.376x10 6 2.019x10 7 18.49x10 6 27.00x10 6 0.3 41 42 43 44 45 46 47 48 49 50 51 52 53 54 55 56 57 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 30.75 21.00 36.0 16.19 27.00 27.00 27.00 27.00 27.00 27.00 38.56 24.00 39.25 44.0 27.0 58.0 49.0 1769.4 1765.7 1765.7 2108.89 3.356x10 7 1.979x10 7 1.979x10 7 5.914x10 7 27.00x10 6 0.3 58 59 60 61 62 63 62 63 64 38.25 37.62 39.01 3948.0 1030.0 331.1 1.130x10 8 1.043x10 7 3.340x10 6 27.00x10 6 0.3614143B. HEAD17.531781.01.986x10 7 27.00x10 6 0.3 VYNPS DSAR Revision 0A-194of216 TABLE A.10.2-5c ELEMENT NUMBER I NODE JNODEDESCRIP. LENGTH (IN)SHEAR AREA (IN 2)I (IN 4)E (LB/IN 2)POISSON'S RATIO ()627343SKIRT86.72636.261.125x10 7 28.3x10 6 0.3 63 64 65 66 67 68 69 70 65 66 67 68 69 70 71 72 66 67 68 69 70 71 72 73 39.48 76.08 62.88 70.56 70.44 83.52 102.00 51.96 494.95 475.22 494.95 1001.85 10001.85 431.20 494.95 500.45 9.16x10 6 8.98x10 6 9.16x10 6 20.59x10 6 20.59x10 6 7.99x10 6 9.16x10 6 9.26x10 6 29.07x10 6 0.30 71 72 73 74 75 73 74 75 76 77 74 75 76 77 103 57.00 64.08 66.96 48.00 84.00 20752.50 22508.10 338.39x10 6 366.16x10 6 3.45x10 6 0.25 76 77 78 79 79 80 40.44 75.00 650.88 636.48 16.10x10 6 16.80x10 6 9.70x10 6 0.30 78 79 80 81 82 83 84 85 86 87 88 89 90 91 80 81 82 83 84 85 86 87 88 89 90 91 92 93 81 82 83 84 85 86 87 88 89 90 91 92 93 94 36.00 63.00 39.48 23.00 48.00 48.00 60.00 64.68 34.92 36.12 48.72 55.20 51.96 51.96 764.64 707.04 777.60 777.60 396.00 396.00 396.00 396.00 1586.88 1730.88 568.80 659.52 727.20 766.08 20.20x10 6 23.00x10 6 30.70x10 6 30.70x10 6 15.60x10 6 15.60x10 6 15.60x10 6 15.60x10 6 63.80x10 6 82.70x10 6 40.20x10 6 62.50x10 6 82.60x10 6 97.50x10 6 9.70x10 6 0.30 VYNPS DSAR Revision 0A-195of216 TABLE A.10.2-5d ELEMENT NUMBER I NODE J NODE DESCRIP.LENGTH (IN)SHEAR AREA (IN 2)I (IN 4)E (LB/IN 2)POISSON'S RATIO ()92 93 94 95 96 97 98 99 100 101 94 95 96 97 98 99 100 101 102 103 95 96 97 98 99 100 101 102 103 112 51.72 51.72 42.00 42.00 41.76 39.24 20.64 15.36 36.00 30.24 792.00 799.20 944.64 925.92 892.80 1293.12 1226.88 1179.36 1095.12 869.76 108.00x10 6 111.00x10 6 128.00x10 6 121.00x10 6 108.00x10 6 140.00x10 6 120.00x10 6 106.00x10 6 85.20x10 6 52.60x10 6 9.7x10 6 0.30102-------------- 103 104 105 106 107 108 109 110 105 106 107 108 109 110 111 112 106 107 108 109 110 111 112 113 300.00 317.88 104.52 83.52 276.00 330.24 204.00 260.76 263.52 230356.80 178161.12 178161.12 139864.32 275040.00 289152.00 289152.00 3222.80x10 6 40766.00x10 6 41266.00x10 6 41266.00x10 6 41159.00x10 6 113870.0x10 6 169910.0x10 6 169910.0x10 6 3.64x10 6 0.20 VYNPS DSAR Revision 0A-196of216 TABLE A.10.2-6aVertical Model -Beam Element Properties ELEM.NO.NODE DESCRIPTION LENGTH (in)AXIAL AREA (in 2)MODULUS OF ELASTICITY (LB/IN 2)POISSON'S RATIO IJ112FUEL162.040.7311.03x10 6 0.41223CRD GUIDE TUBE & HOUSINGS202.19307.4925.6x10 6 0.334524.035.8925.6x10 6 0.3456SHROUD119.56953.4325.6x10 6 0.3567184.94896.9025.6x10 6 0.3689129.03531.4027.0x10 6 0.37910101.813531.4027.0x10 6 0.381011178.193531.4027.0x10 6 0.39111236.03531.4027.0x10 6 0.3101213RPV51.753535.8027.0x10 6 0.311131438.533480.0727.0x10 6 0.312141517.532628.8527.0x10 6 0.31315342.19437.5327.0x10 6 0.3141473RPV SKIRT86.721272.5228.3x10 6 0.3 NOTE: The mathematical properties above are representative only and should not be used fordesign purposes. For design input properties and the analysis of record, consult EDCR 95-406, "Core Shroud Modifications" and EDCR 97-422,"Cycle 20 Reload." VYNPS DSAR Revision 0A-197of216 TABLE A.10.2-6b ELEM.NO.NODE DESCRIPTION LENGTH (in)AXIAL AREA (in 2)MODULUS OF ELASTICITY (LB/IN 2)POISSON'S RATIO IJ 63 64 65 66 67 68 69 70 65 66 67 68 69 70 71 72 66 67 68 69 70 71 72 73 SHIELD WALL 39.48 76.08 62.88 70.56 70.44 83.52 102.00 51.96 989.90 950.45 989.90 2003.70 2003.70 862.40 989.90 1000.90 29.07x10 6 0.30 71 72 73 74 75 73 74 75 76 77 74 75 76 77 103 PEDESTAL 57.00 64.08 66.96 48.00 84.00 41800.00 45360.00 3.45x10 6 0.25 76 77 78 79 79 80 DYRWELL TOP HEAD 40.44 75.00 1301.76 1272.96 6.678x10 6 0.30 78 79 80 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 80 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 100 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 100 101 DRYWELL 36.00 63.00 39.48 23.52 48.00 48.00 60.00 64.68 34.92 36.12 48.72 55.20 51.96 51.96 51.72 51.72 42.00 42.00 41.76 39.24 20.64 1529.28 1414.08 1555.20 1555.20 792.00 792.00 792.00 792.00 3173.76 3461.76 1137.60 1319.04 1454.40 1532.16 1584.00 1598.40 1889.28 1851.84 1785.60 2586.24 2453.76 6.678x10 6 0.30 VYNPS DSAR Revision 0A-198of216 TABLE A.10.2-6c ELEM.NO.NODE DESCRIPTION LENGTH (in)AXIAL AREA (in 2)MODULUS OF ELASTICITY (LB/IN 2)POISSON'S RATIO IJ 99 100 101 102 103 104 105 106 107 108 109 110 101 102 103 104 105 106 107 108 109 110 111 112 102 103 112 105 106 107 108 109 110 111 112 113 DRYWELL REACTOR BUILDING 15.36 36.00 30.24 207.12 300.00 317.88 104.52 83.52 276.00 330.24 204.00 260.76 2358.72 2190.24 1739.52 388.80 532.80 359596.80 356443.20 356443.20 285652.80 405936.00 417312.00 417312.00 6.678x10 6 3.64x10 6 0.30 0.20 NOTE: ELEMENTS 15-62 ARE DUMMY ELEMENTS. VYNPS DSAR Revision 0A-199of216 TABLE A.10.2-7aSpring Element Stiffnesses -Horizontal Model Element No.LocationDirectionStiffness K 1CRD HousingTranslational0 K 2Shroud SupportRotational3.662 x 10 11 in-lb/rad K 3Refueling BellowsTranslational4.563 x 10 6 lb/in K 4Refueling BellowsRotational189.1 x 10 6 in-lb/rad K 5RPV StabilizersTranslational20.0 x 10 6 lb/in K 6 Star Truss (Shield Wall/Drywell)Translational2.33 x 10 7 lb/in VYNPS DSAR Revision 0A-200of216 TABLE A.10.2-8aSpring Element Stiffnesses -Vertical ModelELEMENT NO.LOCATIONDIRECTIONSTIFFNESS K 1 Shroud Support PlateVertical1.899x10 7 lb/in K 2 Shroud Support StiltVertical9.02x10 7 lb/in VYNPS DSAR Revision 0A-201of216 TABLE A.10.2-9aNatural Frequencies of the N-S Model Mode No.Circular Frequency (Rad/Time) Frequency (Hertz)Period (Time)Node No.11.5293E 012.4339E 004.1086E-0110522.7248E 014.3366E 002.3059E-011633.7610E 015.9857E 001.6706E-012043.9080E 016.2199E 001.6078E-014054.5453E 017.2340E 001.3824E-012066.7225E 011.0699E 019.3465E-02978.6268E 011.3730E 017.2833E-02989.9356E 011.5813E 016.3239E-029291.1215E 021.7849E 015.6024E-0214101.1635E 021.8518E 015.4001E-0291111.2118E 021.9287E 015.1850E-0240121.3430E 022.1375E 014.6783E-0240131.3877E 022.2086E 014.5277E-0222141.4900E 022.3714E 014.2169E-0222151.6454E 022.6187E 013.8187E-0278161.9249E 023.0635E 013.2642E-027171.9483E 023.1009E 013.2249E-027182.0332E 023.2359E 013.0904E-023192.2131E 023.5223E 012.8391E-023 NOTE: The mathematical properties above are representative only and should not be used for design purposes. For design input properties and the analysis of record, consult EDCR 95-406, "Core Shroud Modifications" and

EDCR 97-422, "Cycle 20 Reload." VYNPS DSAR Revision 0A-202of216TABLEA.10.2-10aNatural Frequencies of the E-W Model Mode No.Circular Frequency (Rad/Time) Frequency (Hertz)Period (Time)Node No.12.0869E 013.3214E 003.0108E-0110522.7249E 014.3368E 002.3058E-011633.7611E 015.9860E 001.6706E-012043.9083E 016.2203E 001.6076E-014054.6560E 017.4102E 001.3495E-012066.7391E 011.0726E 019.3235E-02978.6270E 011.3730E 017.2832E-02989.9478E 011.5832E 016.3161E-029291.1215E 021.7850E 015.6023E-0214101.1997E 021.9094E 015.2372E-0240111.2216E 021.9443E 015.1433E-0240121.3432E 022.1377E 014.6778E-0240131.3899E 022.2121E 014.5206E-0222141.4901E 022.3716E 014.2165E-0222151.6462E 022.6200E 013.8168E-0278161.9289E 023.0699E 013.2574E-027172.0089E 023.1972E 013.1277E-023182.0990E 023.3407E 012.9934E-023 NOTE: The mathematical properties above are representative only and should not be used for design purposes. For design input properties and the analysis of record, consult EDCR 95-406, "Core Shroud Modifications" and EDCR 97-422, "Cycle 20 Reload." VYNPS DSAR Revision 0A-203of216 TABLE A.10.2-11a Natural Frequencies of the Vertical Model MODE NUMBER CIRCULAR FREQUENCY (RAD/TIME) FREQUENCY (HERTZ)PERIOD (TIME)NODE NUMBER14.3545E 016.9304E 001.4429E-0110429.1957E 011.4635E 016.8328E-027839.9173E 011.5784E 016.3356E-027841.0804E 021.7195E 015.8158E-0210451.1876E 021.8901E 015.2908E-02161.5689E 022.4969E 014.0049E-02171.9873E 023.1629E 013.1617E-021 NOTE: The mathematical properties above are representative only and should not be used for design purposes. For design input properties and the analysis of record, consult EDCR 95-406, "Core Shroud Modifications" and EDCR 97-422, "Cycle 20 Reload." VYNPS DSAR Revision 0A-204of216 f .Qeaclor t!t.. * ** DRY'W6ll G/..Z.!O.

• • .. . . .. ; :.:** :*-: *. SPUR UIG : .. *:. *. ::::!:* .. * .* < '* .\ VERMONT YANKEE NUCLEAR POWER STATION REACTOR BUILDING COMPLEX LONGITUDINAL SECTION FIGURE A.l0.2-l VYNPS DSAR Revision 0A-205of216 b4 7 58.88 1\!'V TOP H!:AD 719.87 0:.3 Refueling 682.25 RP\' H!:hD l'1.A..'IGE

6ellcws 64t..OO BELLO\.IS 595.00 STEA.'-1 OUTLET !.U'V Sub:!.lize::

55t..OO TOP OF STEAM 510.00 466.00 TOP OF P!PES 1.26.75 TOP OF SHROUD H!;.!l 1.02. 75 BOTTOM OF 364.19 .337.19

• MASS NODES 310.19 0 BEAK ELE!-!ENTS 283.19 --vvv-SPRING ROTATIONAL SPRING 256.19 =z= RIGID LINKS 229.19 HINGE CONNECTION 202.19 186.00 INLET 150.00 'RECIRC. OU'TLET 129.00 98.25 SHROUD SUPPORT 59.72 TOP OF SXIRT 42.19 TOP OF CRD P.OUSINC 21.90 o.oo TOP OF -33.00 1
• ELEVAl'IONS ARE P!D!STAI IN

-66.00 AND ARE WITH TO VESSEL 0.0. 2. VESSEL ELEVATION 0.0 2 -99.00 CORRESPONDS TO Pl.A)."T G) ELEVATION 266.92 FT. -132.81 'l1 1 IPV /INTEIU:ALS BORIZORTAL MOD!L VERMONT YANKEE NUCLEAR POWER STATION (.0 > UACTOB. BUILDING COMPLEX w SEISMIC BOI.IZO!ITAL MODEL ex: 'II'TI"..rmS"

• _, n. 2-2 VYNPS DSAR Revision 0A-206of216 REACTOR EL.370.16 105 EL. 345 .16 __ _ 106 El.31B.57 l 07 SHEAR 309. 96 108 LUGS El. 303.00 109 EL.280.00 110 El.252.50 111 EL.235.48 112 DP.YWEll HISEOMENT EL.213. 75 113 TOP OF BASEMAT 308.00 304.00-85 300.00_ 86 295.00-87 289.61-88 286.70-89 283.69-90 279.63_ 91 92 93 94 262.06-254.25-250.75-100 101 241.00-102 103 LEGEl\D: e MASS NODES 0 MASSLESS NODES -'\JV\r TRANSLATIONAL SPRINGS .....-o ROTATIONAL SPRING ++-++-RIGID LINKS 61 REFUELING BELLQ'I.!S 65 59 66 307.79 67 301 .45 68 296.21 69 290.33 284.46 277 .so __ 269.00 264.67 Top of 74 __ 259_92 Pedestal __ 254.58 __ 249.00 245.00 238.00 --PEDESTAL EHBEOI-!ENT ELEVATIONS ARE PLANT ELEVATIONS AND tN FEET. u: ::::::. VERMONT YANKEE NUCLEAR POWER STATION REACTOR BUILDING COMPLEX SEISMIC HORIZONTAL MODEL

______ _J VYNPS DSAR Revision 0A-207of216 ....... ..... ' ' \. 8 426.75 364.19 202.19 0 .00. ' I I I I o MASS NODES 0 BEA."l -f\/\1'-TRANSLA!!ONAL S?RINGS 595.00 466.CO 402.75 19 2Bs.19 186.00 150.00 98.25 59.72 42.19 -27.00 1. ELEVA!!ONS ARE EXPRESSED IN INCRES AND WITH RESPECT TO VESSEL 0. 0. 2. VESsn*o.o COUES?ONDS TO-'PLAii"r E!.EVAnCN 266.92 rt. VERTICAL RPV AND INTERNALS MODEL VERMONT YANKEE NUCLEAR POWER STATION REACTOR BUILDING COMPLEX SEISMIC VERTICAL HODEL FIGURE A.l0.2-3 VYNPS DSAR Revision 0A-208of216 REACTOR BU!LDING EL.3&7.42 104 £L.370.16 EL .345.16 EL.318.67 SHEAR 30 LUGS EL.303. uO El.280.00 El.252.50 9.96 El.235.48 .. DRYWEll EHSEOMENT El.213 .75 *TOP OF--BASEMAT 105 106 107 108 109 110 111 112 rY-3 y DR':,.ELL [l.331 .12_. 78 327.75_. 79 321 .50-*1 80 318.50-81 313.25-62 >a3 308.00 84 85 300.00_ 86 295.00-87 289.61-88 286.70-89 283.69-90 279.63_ 91 92 270.70_ 93 266.37-94 262.06-95 257.75_ 96 254.25-97 250.75-98 247.27-99 244.00-100 242*.28-101 241.00-102 103 !Qg: X LEGEND: o !-tASS NOD::S 0 !AS SLESS NODES M 65 6& 67 66 il 69 70 711 72 -311.08 3C7. 79 301 .45 296.21 290.33 284.46 277. so __269.00 73. 74 264.67 iop of __ 259_92 Pedestal 75 77 __ 254.58 249.00 __ 245.00 238.00 PEDESTAL EMBEDMENT ELEVATIONS PLANT !LE\'A! IONS IN FEET. VERMONT YANKEE NUCLEAR POWER STATION REACTOR BUILDING COMPLEX SEISMIC VERTICAL MODEL FIGURE A.10.2-3 (CONT.) ____ .,.J VYNPS DSAR Revision 0A-209of216 Base 5upport skirt I Concrete .. *;.*.:,.;.;.*.

• . . . . . pedestal
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• ac .us nauu A.l0.3-4 VYNPS DSAR Revision 0A-215of216 (f) :: c f-0 z C[ (f) l l VS. NRC RRS NnvrHilER

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__ L_LLLLLLJ I I II II I q I I I I I I I! I 10 -I 10 ° 10 I FREQUENCY !HZl 10 2 T 1.6WIO -lt,ojllS/1111 VERMONT YANKEE NUCLEAR POWER STATION G ESSAR II vs. NRC ARS I F IGURE A,l0,3-S VYNPS DSAR Revision 0A-216of216 (f) I C) z GESSRR II VS. NRC ARS NnvrHnrn 12G. t!lnll I I L CA ldN 0 HI' N( GESSAn J I IVT FREE F lEU " 2. 0 PEnCHIT 800 -1 1-1 II VT ZPA,0.!4G B¢.+<jPFCJAA ZPA D.._l!lfi_j 1 1-+--l-H-1-600 Q z 0 ._. 1-CI rc ll.J _) ILl u u ([ 400-200 1-l-1-l-l--1-1-l-I I I i i I i I I : i ' l I I i i I l i / : : 1 ' tf :;Jr..-:: i . : y¥ '; o -1 i I 10 -1 lU 0 10 I FREQUENCY rHZl 10 2 SCIUS US£0 T 1.6110 VERMONT YANKEE NUCLEAR POWER STAT ION G E S SAR II vs. NRC ARS FIGURE A.l 0.3-6 VYNPS DSAR Revision 0 G.2-1 of 40 APPENDIX G.2 CURRENT ON-SITE METEOROLOGICAL PROGRAM TABLE OF CONTENTS Section Title Page G.2.1Introduction.........................................................4G.2.2Description of the Monitoring Program................................ 4G.2.3Results..............................................................5G.2.3.1Wind Data .................................................5G.2.3.2Inversion Data ............................................6 VYNPS DSAR Revision 0 G.2-2 of 40 CURRENT ON-SITE METEOROLOGICAL PROGRAM LIST OF TABLES Table No. Title G.2.1 Meteorological Data Recovery Rates for 1980 G.2.2 Joint Frequency Distribution of Wind Speed, Wind Direction, and Stability Class (Stability Based on 295-33 Foot Delta-

T)(35.0 FT Wind Data) G.2.3 Joint Frequency Distribution of Wind Speed, Wind Direction, and Stability Class (Stability Based on 295-33 Foot Delta-T)

(297 foot level FT Wind Data) G.2.4 Wind Direction Persistence Summary (35 foot level) G.2.5 Wind Direction Persistence Summary (297 foot level) G.2.6 Inversion Persistence Summary (198-33 foot Delta T) G.2.7 Inversion Persistence Summary (295-33 foot Delta T) VYNPS DSAR Revision 0 G.2-3 of 40 CURRENT ON-SITE METEOROLOGICAL PROGRAM LIST OF FIGURES Reference Figure No. Drawing No. Title G.2-1 Location of Primary and Backup Meteorological Towers G.2-2 "Spring Wind Rose (35 foot level) March 1980 May 1980" G.2-3 Summer Wind Rose (35 foot level) June 1980 August 1980 G.2-4 Autumn Wind Rose (35 foot level) September 1980 November 1980 G.2-5 Winter Wind Rose (35 foot level) January 1980 February 1980; December 1980 G.2-6 Annual Wind Rose (35 foot level) January 1980 December 1980 G.2-7 Spring Wind Rose (297 foot level) March 1980 May 1980G.2-8 Summer Wind Rose (297 foot level) June 1980 August 1980 G.2-9 Autumn Wind Rose (297 foot level) September 1980 November 1980 G.2-10 Winter Wind Rose (297 foot level) January 1980 February 1980; December 1980 G.2-11 Annual Wind Rose (297 foot level) January 1980 December 1980 VYNPS DSAR Revision 0 G.2-4 of 40 G.2 CURRENT ON-SITE METEOROLOGICAL PROGRAM G.2.1 Introduction The On-Site Meteorological Data Collection Program was upgraded in early 1976 to meet the intent of Revision 0 of Regulatory Guide 1.23. This report

describes the current on-site monitoring program and presents wind and

stability data summaries for one full year of operation; January 1, 1980

through December 31, 1980. A discussion of the data summaries is included, and a comparison is made between data collected by the initial monitoring

program (August 1967 - July 1968) and data collected by the current monitoring

program (January 1980 - December 1980). It is concluded that results from

both monitoring programs are compatible, and that both programs produced data

bases which are representative of site meteorology. G.2.2 Description of the Monitoring Program The current Meteorological Monitoring System includes both a primary and a backup system. The primary system utilizes a guyed 305-foot tower located

on-site as shown in Figure G.2-1. The parameters measured on the tower

include the following: Wind speed at the 35-foot and 297-foot levels (3-cup anemometer sensors) Wind direction at the 35-foot and 297-foot levels (airfoil vane sensors) Temperature at the 33-foot level (RTD located in a radiation-shielded aspirator) Delta-temperature between the 198-33 foot and between the 295-33 foot levels (RTDs located in radiation-shielded aspirators) In addition, both precipitation and barometric pressure are measured on the ground.The translator cards for the tower sensors are located in an instrument shed near the base of the tower. The analog output is then digitally transmitted

to one of the plant process computer's remote data acquisition terminals. The

plant process computer periodically scans each parameter and then digitally

compiles and records the data as 15-minute averages. The 15-minute averages

are available for display in the Control Room. A digital recorder, located in

the relay house is also utilized as an auxiliary data logger. The entire

system is currently supplied by redundant power sources.

A 140-foot guyed tower used previously for meteorological monitoring was VYNPS DSAR Revision 0 G.2-5 of 40 reinstrumented during 1980 to serve as a backup tower. This tower's location is also shown in Figure G.2-1. The parameters measured on the backup tower

include: Wind speed at the 100-foot level (3-cup anemometer sensor) Wind direction at the 100-foot level (airfoil vane sensors) Delta-temperature between the 135-33 foot levels (RTDs located in radiation-shielded aspirators) The translator cards for the tower sensors are located in an instrument shed near the base of the tower. The analog is then transmitted digitally to the

Control Room and sampled by the plant process computer. The signals are also

captured by a digital recorder, which is also utilized as an auxiliary data

logger.G.2.3 Results The primary meteorological system was the data source for the 1980 data summaries which follow. The digital recording system was the principal data

collection mechanism. The data base consists of hourly data where the first

15-minute average collected each hour is used to represent the hour. The

analog strip chart recorders were utilized as backup data loggers for quality

control analysis, and data from the strip charts were used to fill in gaps in

the digital data base. The resulting data recovery rates, which are well

above the Regulatory Guide 1.23 goal of 90%, are presented in Table G.2.1. G.2.3.1 Wind Data Seasonal and annual wind roses for all stabilities combined from each tower level are presented in Figures G.2-2 through G.2-11. Annual three-way joint

frequency summaries of wind speed, wind direction, and stability (stability

defined as a function of delta-temperature per Rev. 0 of Regulatory Guide

1.23) for both tower levels are also presented in Tables G.2.2 and G.2.3.

Comparison with the initial monitoring program results shows good agreement.

The seasonal and annual wind roses from each tower level continue to

illustrate the channelling effect of the Connecticut River Valley upon the

winds. Winds generally blow with the highest frequency from the NNW and SSE.

The annual average wind speeds for the current monitoring program were 6.2 mph

for the 35-foot level and 9.2 mph for the 297-foot level. These average wind

speeds compare well with the 140-foot level annual average wind speed of 7.5

mph for the initial monitoring program, if one considers the expected

variation of wind speed with height from the surface. VYNPS DSAR Revision 0 G.2-6 of 40 Wind direction persistence summaries for 1980 are presented in Tables G.2.4 and G.2.5. Summarizing the persistence information, 74.9% of the 35-foot and

69.5% of the 297-foot cases during 1980 were one-hour events, and only one of

the 35-foot and nine of the 297-foot persistence cases were more than 15 hours

long. This compares well with the initial monitoring program which found that

69.0% of its persistence cases were one-hour events and only five persistence

cases were more than 15 hours long. G.2.3.2 Inversion Data The annual stability class frequency distribution for the initial and current monitoring programs compare as follows: Stability Case Initial Program Current Program 140-5 Foot Delta-T 198-33 Foot Delta-T (Ref. Table G.2.2) 295-33 Foot Delta-T (Ref. Table G.2.3) Unstable (A,B,C) 25.4% 14.9% 8.1% Neutral (D) 25.1% 37.1% 43.0% Stable (E,F,G) 49.5% 48.0% 48.9% Differences in the above frequency distributions can be expected due to the differences in measurement heights. Measurement heights are important because

the largest temperature gradients occur near the ground due to surface heating

during the day and radiative cooling at night. Inversions, defined as a positive value of delta-temperature, occurred 33.6% and 32.7% of the time during 1980 for the 198-33 foot and 295-33 foot

delta-temperature measurements, respectively. Inversions occurred at the site

nearly 39% of the time during the initial data collection period.

Tables G.2.6 and G.2.7 present the annual inversion persistence summaries for

1980. The longest inversion measured during the year lasted 38 hours. VYNPS DSAR Revision 0 G.2-7 of 40 TABLE G.2.1 Meteorological Data Recovery Rates for 1980 Parameter Possible Hours Usable Hours Recovery Rate 35-Foot Wind Speed 8784 8723 99.3% 297-Foot Wind Speed 8784 8721 99.3% 35-Foot Wind Direction 8784 8763 99.8% 297-Foot Wind Direction 8784 8573 97.6% 33-Foot Temperature 8784 8734 99.4% 198-33 Foot Delta-T 8784 8703 99.1% 295-33 Foot Delta-T 8784 8710 99.2% Precipitation 8784 8474 96.5% Solar Radiation 8784 8768 99.8% Composite (35' WS, 35' WD, 198-33' DT) 8784865998.6% Composite (297' WS, 297' WD, 295-33' DT) 8784847496.5% VYNPS DSAR Revision 0 G.2-8 of 40 Table G.2.2 Joint Frequency Distribution of Wind Speed Wind Direction, and Stability Class (Stability Based on 198-33 Foot Delta-T) 35.0 FT WIND DATA STABILITY CLASS A CLASS FREQUENCY (PERCENT) = 6.14 WIND DIRECTION FROM SPEED (MPH) N NNE NE ENE EESESESSESSSWSW WSWWWNWNWNNWVRBLTOTAL CALM (1) (2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 C-3 (1) (2)2.38 .02 7 1.32.08 6 1.13.07 7 1.32.08 7 1.32.08 10 1.88.12 9 1.69.10 5.94 .06 4.75 .05 0 0.00 0.00 2.38 .02 2.38 .02 1.19 .01 1.19 .01 1.19 .01 3.56 .03 0 0.00 0.00 67 12.59.77 4-7 (1) (2)14 2.63.16 10 1.88.12 13 2.44.15 19 3.57.22 28 5.26.32 17 3.20.20 14 2.63.16 20 3.76.23 7 1.32.08 4.75 .05 2.38 .02 1.19 .01 1.19 .01 3.56 .03 17 3.20.20 36 6.77.42 0 0.00 0.00 206 38.72 2.38 8-12 (1) (2)16 3.01.18 3.56 .03 2.38 .02 0 0.00 0.00 5.94 .06 4.75 .05 14 2.63.16 45 8.46.52 19 3.57.22 5.94 .06 1.19 .01 2.38 .02 5.94 .06 4.75 .05 22 4.14.25 62 11.65.72 0 0.00 0.00 209 39.29 2.41 13-18 (1) (2)4.75 .05 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.19 .01 0 0.00 0.00 0 0.00 0.00 1.19 .01 1.19 .01 3.56 .03 1.19 .01 11 2.07.13 27 5.08.31 0 0.00 0.00 49 9.21.57 19-24 (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.19.01 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.19.01 GT 24 (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 ALL SPEEDS (1)(2)36 6.77.42 20 3.76.23 21 3.95.24 26 4.89.30 40 7.52.46 31 5.83.36 37 6.95.43 71 13.35.82 30 5.64.35 9 1.69.10 6 1.13.07 6 1.13.07 10 1.88.12 10 1.88.12 51 9.59.59 128 24.06 1.48 0 0.00 0.00 532 100.00 6.14 (1) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C=CALM (WIND SPEED LESS THAN OR EQUAL TO .60 MPH) VYNPS DSAR Revision 0 G.2-9 of 40 TABLE G.2.2 35.0 FT WIND DATA STABILITY CLASS B CLASS FREQUENCY (PERCENT) = 4.09 WIND DIRECTION FROM SPEED (MPH) N NNE NE ENE EESESESSESSSWSW WSWWWNWNWNNWVRBLTOTAL CALM (1) (2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 C-3 (1) (2)0 0.00 0.00 7 1.98.08 2.56 .02 4 1.13.05 5 1.41.06 5 1.41.06 5 1.41.06 4 1.13.05 3.85 .03 2.56 .02 1.28 .01 1.28 .01 1.28 .01 1.28 .01 1.28 .01 3.85 .03 0 0.00 0.00 45 12.71.52 4-7 (1) (2)14 3.95.16 5 1.41.06 3.85 .03 8 2.26.09 19 5.37.22 8 2.26.09 11 3.11.13 9 2.54.10 5 1.41.06 1.28 .01 3.85 .03 2.56 .02 1.28 .01 6 1.69.07 12 3.39.14 25 7.06.29 0 0.00 0.00 132 37.29 1.52 8-12 (1)(2)9 2.54.10 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.28.01 1.28.01 2.56.02 7 1.98.08 10 2.82.12 0 0.00 0.00 4 1.13.05 4 1.13.05 5 1.41.06 11 3.11.13 12 3.39.14 34 9.60.39 0 0.00 0.00 100 28.25 1.15 13-18 (1) (2)2.56 .02 1.28 .01 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 2.56 .02 3.85 .03 0 0.00 0.00 2.56 .02 1.28 .01 11 3.11.13 9 2.54.10 12 3.39.14 28 7.91.32 0 0.00 0.00 71 20.06.82 19-24 (1) (2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 2.56 .02 1.28 .01 1.28 .01 2.56 .02 0 0.00 0.00 6 1.69.07 GT 24 (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 ALL SPEEDS (1)(2)25 7.06.29 13 3.67.15 5 1.41.06 12 3.39.14 25 7.06.29 14 3.95.16 18 5.08.21 22 6.21.25 21 5.93.24 3.85.03 10 2.82.12 8 2.26.09 20 5.65.23 28 7.91.32 38 10.73.44 92 25.99 1.06 0 0.00 0.00 354 100.0 4.09 (1) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C=CALM (WIND SPEED LESS THAN OR EQUAL TO .60 MPH) VYNPS DSAR Revision 0 G.2-10 of 40 TABLE G.2.2 35.0 FT WIND DATA STABILITY CLASS C CLASS FREQUENCY (PERCENT) = 4.64 WIND DIRECTION FROM SPEED (MPH) N NNE NE ENE EESESESSESSSWSW WSWWWNWNWNNWVRBLTOTAL CALM (1) (2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 C-3 (1) (2)5 1.24.06 5 1.24.06 2.50 .02 3.75 .03 3.75 .03 6 1.49.07 7 1.74.08 6 1.49.07 2.50 .02 1.25 .01 2.50 .02 3.75 .03 1.25 .01 0 0.00 0.00 5 1.24.06 4 1.00.05 0 0.00 0.00 55 13.68.64 4-7 (1) (2)18 4.48.21 5 1.24.06 6 1.49.07 6 1.49.07 12 2.99.14 10 2.49.12 6 1.49.07 2.50 .02 3.75 .03 3.75 .03 3.75 .03 4 1.00.05 1.25 .01 7 1.74.08 12 2.99.14 24 5.97.28 0 0.00 0.00 122 30.35 1.41 8-12 (1) (2)9 2.24.10 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 2.50 .02 0 0.00 0.00 3.75 .03 20 4.98.23 14 3.48.16 4 1.00.05 9 2.24.10 6 1.49.07 10 2.49.12 15 3.73.17 10 2.49.12 23 5.72.27 0 0.00 0.00 125 31.09 1.44 13-18 (1)(2)7 1.74.08 1.25 .01 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.25 .01 1.25 .01 1.25 .01 0 0.00 0.00 1.25 .01 4 1.00.05 11 2.74.13 21 5.22.24 16 3.98.18 22 5.47.25 0 0.00 0.00 86 21.39.99 19-24 (1) (2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.25 .01 1.25 .01 2.50 .02 3.75 .03 6 1.49.07 0 0.00 0.00 13 3.23.15 GT 24 (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.25.01 0 0.00 0.00 0 0.00 0.00 1.25.01 ALL SPEEDS (1)(2)39 9.70.45 11 2.74.13 8 1.99.09 9 2.24.10 17 4.23.20 16 3.98.18 17 4.23.20 29 7.21.33 20 4.98.23 8 1.99.09 15 3.73.17 18 4.48.21 24 5.97.28 45 11.19.52 47 11.69.54 79 19.65.91 0 0.00 0.00 402 100.00 4.64 (1) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C=CALM (WIND SPEED LESS THAN OR EQUAL TO .60 MPH) VYNPS DSAR Revision 0 G.2-11 of 40 TABLE G.2.2 35.0 FT WIND DATA STABILITY CLASS D CLASS FREQUENCY (PERCENT) = 37.14 WIND DIRECTION FROM SPEED (MPH) N NNE NE ENE EESESESSESSSWSW WSWWWNWNWNNWVRBLTOTAL CALM (1) (2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 C-3 (1) (2)39 1.21.45 15.47 .17 22.68 .25 26.81 .30 40 1.24.46 23.72 .27 41 1.27.47 39 1.21.45 33 1.03.38 32 1.00.37 28.87 .32 42 1.31.49 24.75 .28 42 1.31.49 52 1.62.60 60 1.87.69 0 0.00 0.00 558 17.35 6.44 4-7 (1) (2)106 3.30 1.22 42 1.31.49 28.87 .32 20.62 .23 35 1.09.40 57 1.77.66 101 3.14 1.17 145 4.51 1.67 65 2.02.75 15.47 .17 22.68 .25 28.87 .32 44 1.37.51 54 1.68.62 112 3.48 1.29 231 7.18 2.67 0 0.00 0.00 1105 34.36 12.76 8-12 (1) (2)74 2.30.85 16.50 .18 14.44 .16 9.28 .10 12.37 .14 25.78 .29 21.65 .24 92 2.86 1.06 80 2.49.92 16.50 .18 20.62 .23 29.90 .33 130 4.04 1.50 130 4.04 1.50 119 3.70 1.37 170 5.29 1.96 0 0.00 0.00 957 29.76 11.05 13-18 (1)(2)34 1.06.39 1.03 .01 3.09 .03 0 0.00 0.00 0 0.00 0.00 1.03 .01 1.03 .01 7.22 .08 14.44 .16 1.03 .01 6.19 .07 8.25 .09 89 2.77 1.03 109 3.39 1.26 107 3.33 1.24 126 3.92 1.46 0 0.00 0.00 507 15.76 5.86 19-24 (1) (2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.03 .01 2.06 .02 0 0.00 0.00 0 0.00 0.00 3.09 .03 12.37 .14 13.40 .15 28.87 .32 24.75 .28 0 0.00 0.00 83 2.58.96 GT 24 (1) (2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 4.12 .05 2.06 .02 0 0.00 0.00 0 0.00 0.00 6.19 .07 ALL SPEEDS (1)(2)253 7.87 2.92 74 2.30.85 67 2.08.77 55 1.71.64 87 2.71 1.00 106 3.30 1.22 164 5.10 1.89 284 8.83 3.28 194 6.03 2.24 64 1.99.74 76 2.36.88 110 3.42 1.27 299 9.30 3.45 352 10.95 4.07 420 13.06 4.85 611 19.00 7.06 0 0.00 0.00 3216 100.00 37.14 (1) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C=CALM (WIND SPEED LESS THAN OR EQUAL TO .60 MPH) VYNPS DSAR Revision 0 G.2-12 of 40 TABLE G.2.2 35.0 FT WIND DATA STABILITY CLASS E CLASS FREQUENCY (PERCENT) = 30.70 WIND DIRECTION FROM SPEED (MPH) N NNE NE ENE EESESESSESSSWSW WSWWWNWNWNNWVRBLTOTAL CALM (1) (2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 C-3 (1) (2)50 1.88.58 26.98 .30 30 1.13.35 22.83 .25 31 1.17.36 35 1.32.40 42 1.58.49 59 2.22.68 104 3.91 1.20 130 4.89 1.50 143 5.38 1.65 122 4.59 1.41 150 5.64 1.73 127 4.78 1.47 132 4.97 1.52 99 3.72 1.14 0 0.00 0.00 1302 48.98 15.04 4-7 (1) (2)38 1.43.44 16.60 .18 1.04 .01 7.26 .08 16.60 .18 32 1.20.37 60 2.26.69 87 3.27 1.00 85 3.20.98 28 1.05.32 29 1.09.33 42 1.58.49 79 2.97.91 91 3.42 1.05 163 6.13 1.88 166 6.25 1.92 0 0.00 0.00 940 35.36 10.86 8-12 (1) (2)16.60 .18 1.04 .01 0 0.00 0.00 0 0.00 0.00 2.08 .02 6.23 .07 18.68 .21 37 1.39.43 39 1.47.45 3.11 .03 5.19 .06 8.30 .09 52 1.96.60 44 1.66.51 50 1.88.58 67 2.52.77 0 0.00 0.00 348 13.09 4.02 13-18 (1)(2)3.11 .03 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.04 .01 6.23 .07 7.26 .08 0 0.00 0.00 0 0.00 0.00 1.04 .01 4.15 .05 7.26 .08 17.64 .20 17.64 .20 0 0.00 0.00 63 2.37.73 19-24 (1) (2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 2.08 .02 1.04 .01 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.04 .01 0 0.00 0.00 1.04 .01 0 0.00 0.00 0 0.00 0.00 5.19 .06 GT 24 (1) (2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 ALL SPEEDS (1)(2)107 4.03 1.24 43 1.62.50 31 1.17.36 29 1.09.33 49 1.84.57 73 2.75.84 121 4.55 1.40 191 7.19 2.21 236 8.88 2.73 161 6.06 1.86 177 6.66 2.04 173 6.51 2.00 286 10.76 3.30 269 10.12 3.11 363 13.66 4.19 349 13.13 4.03 0 0.00 0.00 2658 100.00 30.70 (1) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C=CALM (WIND SPEED LESS THAN OR EQUAL TO .60 MPH) VYNPS DSAR Revision 0 G.2-13 of 40 TABLE G.2.2 35.0 FT WIND DATA STABILITY CLASS F CLASS FREQUENCY (PERCENT) = 13.87 WIND DIRECTION FROM SPEED (MPH) N NNE NE ENE EESESESSESSSWSW WSWWWNWNWNNWVRBLTOTAL CALM (1) (2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 C-3 (1) (2)14 1.17.16 10.83 .12 8.67 .09 10.83 .12 10.83 .12 9.75 .10 18 1.50.21 43 3.58.50 51 4.25.59 116 9.66 1.34 197 16.40 2.28 157 13.07 1.81 109 9.08 1.26 68 5.66.79 71 5.91.82 39 3.25.45 0 0.00 0.00 930 77.44 10.74 4-7 (1) (2)6.50 .07 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 4.33 .05 3.25 .03 3.25 .03 8.67 .09 15 1.25.17 18 1.50.21 38 3.16.44 32 2.66.37 19 1.58.22 20 1.67.23 47 3.91.54 37 3.08.43 0 0.00 0.00 250 20.82 2.89 8-12 (1) (2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 2.17 .02 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 2.17 .02 2.17 .02 1.08 .01 5.42 .06 8.67 .09 0 0.00 0.00 20 1.67.23 13-18 (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.08 .01 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.08 .01 19-24 (1) (2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 GT 24 (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 ALL SPEEDS (1)(2)20 1.67.23 10.83.12 8.67.09 10.83.12 14 1.17.16 14 1.17.16 21 1.75.24 51 4.25.59 66 5.50.76 134 11.16 1.55 235 19.57 2.71 192 15.99 2.22 130 10.82 1.50 89 7.41 1.03 123 10.24 1.42 84 6.99.97 0 0.00 0.00 1201 100.00 13.87 (1) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C=CALM (WIND SPEED LESS THAN OR EQUAL TO .60 MPH) VYNPS DSAR Revision 0 G.2-14 of 40 TABLE G.2.2 35.0 FT WIND DATA STABILITY CLASS G CLASS FREQUENCY (PERCENT) = 3.42 WIND DIRECTION FROM SPEED (MPH) N NNE NE ENE EESESESSESSSWSW WSWWWNWNWNNWVRBLTOTAL CALM (1) (2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 C-3 (1) (2)5 1.69.06 1.34 .01 7 2.36.08 5 1.69.06 5 1.69.06 2.68 .02 11 3.72.13 5 1.69.06 26 8.78.30 33 11.15.38 39 13.18.45 25 8.45.29 26 8.78.30 21 7.09.24 12 4.05.14 13 4.39.15 0 0.00 0.00 236 79.73 2.73 4-7 (1) (2)1.34 .01 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.34 .01 0 0.00 0.00 2.68 .02 0 0.00 0.00 2.68 .02 11 3.72.13 14 4.73.16 7 2.36.08 4 1.35.05 4 1.35.05 4 1.35.05 3 1.01.03 0 0.00 0.00 53 17.91.61 8-12 (1) (2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 2.68 .02 5 1.69.06 0 0.00 0.00 7 2.36.08 13-18 (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 19-24 (1) (2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 GT 24 (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 ALL SPEEDS (1)(2)6 2.03.07 1.34.01 7 2.36.08 5 1.69.06 6 2.03.07 2.68.02 13 4.39.15 5 1.69.06 28 9.46.32 44 14.86.51 53 17.91.61 32 10.81.37 30 10.14.35 25 8.45.29 18 6.08.21 21 7.09.24 0 0.00 0.00 296 100.00 3.42 (1) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C=CALM (WIND SPEED LESS THAN OR EQUAL TO .60 MPH) VYNPS DSAR Revision 0 G.2-15 of 40 TABLE G.2.2 35.0 FT WIND DATA STABILITY CLASS ALL CLASS FREQUENCY (PERCENT) = 100.00 WIND DIRECTION FROM SPEED (MPH) N NNE NE ENE EESESESSESSSWSW WSWWWNWNWNNWVRBLTOTAL CALM (1) (2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 C-3 (1) (2)115 1.33 1.33 71.82 .82 77.89 .89 77.89 .89 101 1.17 1.17 90 1.04 1.04 133 1.54 1.54 161 1.86 1.86 223 2.58 2.58 314 3.63 3.63 412 4.76 4.76 352 4.07 4.07 312 3.60 3.60 260 3.00 3.00 274 3.16 3.16 221 2.55 2.55 0 0.00 0.00 3193 36.87 36.87 4-7 (1) (2)197 2.28 2.28 78.90 .90 51.59 .59 60.69 .69 115 1.33 1.33 227 1.47 1.47 197 2.28 2.28 271 3.13 3.13 182 2.10 2.10 80.92 .92 111 1.28 1.28 116 1.34 1.34 149 1.72 1.72 185 2.14 2.14 367 4.24 4.24 522 6.03 6.03 0 0.00 0.00 2808 32.43 32.43 8-12 (1) (2)124 1.43 1.43 20.23 .23 16.18 .18 9.10 .10 22.25 .25 38.44 .44 58.67 .67 201 2.32 2.32 162 1.87 1.87 28.32 .32 39.45 .45 51.59 .59 204 2.36 2.36 205 2.37 2.37 220 2.54 2.54 369 4.26 4.26 0 0.00 0.00 1766 20.39 20.39 13-18 (1)(2)50.58.58 3.03.03 3.03.03 0 0.00 0.00 0 0.00 0.00 1.01.01 3.03.03 17.20.20 25.29.29 1.01.01 10.12.12 16.18.18 118 1.36 1.36 147 1.70 1.70 163 1.88 1.88 220 2.54 2.54 0 0.00 0.00 777 8.97 8.97 19-24 (1) (2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 3.03 .03 3.03 .03 0 0.00 0.00 0 0.00 0.00 4.05 .05 16.18 .18 17.20 .20 33.38 .38 32.37 .37 0 0.00 0.00 108 1.25 1.25 GT 24 (1) (2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 4.05 .05 3.03 .03 0 0.00 0.00 0 0.00 0.00 7.08 .08 ALL SPEEDS (1)(2)486 5.61 5.61 172 1.99 1.99 147 1.70 1.70 146 1.69 1.69 238 2.75 2.75 256 2.96 2.96 391 4.52 4.52 653 7.54 7.54 595 6.87 6.87 423 4.89 4.89 572 6.61 6.61 539 6.22 6.22 799 9.23 9.23 818 9.45 9.45 1060 12.24 12.24 1364 15.75 15.75 0 0.00 0.00 8659 100.00 100.00 (1) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C=CALM (WIND SPEED LESS THAN OR EQUAL TO .60 MPH) VYNPS DSAR Revision 0 G.2-16 of 40 TABLE G.2.3 Joint Frequency Distribution of Wind Speed, Wind Direction, and Stability Class (Stability Based on 295-33 Foot Delta-T) 297.0 FT WIND DATA STABILITY CLASS A CLASS FREQUENCY (PERCENT) = .91 WIND DIRECTION FROM SPEED (MPH) N NNE NE ENE EESESESSESSSWSW WSWWWNWNWNNWVRBLTOTAL CALM (1) (2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 C-3 (1) (2)0 0.00 0.00 0 0.00 0.00 2 2.60.02 2 2.60.02 0 0.00 0.00 0 0.00 0.00 3 3.90.04 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1 1.30.01 0 0.00 0.00 1 1.30.01 0 0.00 0.00 9 11.69.11 4-7 (1) (2)2 2.60.02 2 2.60.02 0 0.00 0.00 0 0.00 0.00 1 1.30.01 1 1.30.01 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 2 2.60.02 5 6.49.06 0 0.00 0.00 13 16.88.15 8-12 (1) (2)1 1.30.01 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 4 5.19.05 1 1.30.01 12 15.58.14 6 7.79.07 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1 1.30.01 10 12.99.12 0 0.00 0.00 35 45.45.41 13-18 (1)(2)1 1.30.01 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1 1.30.01 6 7.79.07 3 3.90.04 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 2 2.60.02 1 1.30.01 4 5.19.05 0 0.00 0.00 18 23.38.21 19-24 (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 2 2.60.02 0 0.00 0.00 2 2.60.02 GT 24 (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 ALL SPEEDS (1) (2)4 5.19.05 2 2.60.02 2 2.60.02 2 2.60.02 1 1.30.01 5 6.49.06 5 6.49.06 18 23.38.21 9 11.69.11 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 3 3.90.04 4 5.19.05 22 28.57.26 0 0.00 0.00 77 100.00.91 (1) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C=CALM (WIND SPEED LESS THAN OR EQUAL TO .60 MPH) VYNPS DSAR Revision 0 G.2-17 of 40 TABLE G.2.3 297.0 FT WIND DATA STABILITY CLASS B CLASS FREQUENCY (PERCENT) = 2.62 WIND DIRECTION FROM SPEED (MPH) N NNE NE ENE EESESESSESSSWSW WSWWWNWNWNNWVRBLTOTAL CALM (1) (2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 C-3 (1) (2)1.45 .01 1.45 .01 1.45 .01 2.90 .02 1.45 .01 1.45 .01 2.90 .02 2.90 .02 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.45 .01 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 12 5.41.14 4-7 (1) (2)12 5.41.14 3 1.35.04 1.45 .01 2.90 .02 8 3.60.09 8 3.60.09 4 1.80.05 2.90 .02 1.45 .01 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 9 4.05.11 0 0.00 0.00 50 22.52.59 8-12 (1)(2)8 3.60.09 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 6 2.70.07 5 2.25.06 11 4.95.13 12 5.41.14 2.90 .02 0 0.00 0.00 1.45 .01 1.45 .01 1.45 .01 5 2.25.06 23 10.36.27 0 0.00 0.00 75 33.78.89 13-18 (1) (2)5 2.25.06 3 1.35.04 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.45 .01 1.45 .01 4 1.80.05 5 2.25.06 0 0.00 0.00 1.45 .01 0 0.00 0.00 6 2.70.07 2.90 .02 8 3.60.09 30 13.51.35 0 0.00 0.00 66 29.73.78 19-24 (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.45.01 4.180.05 12 5.41.14 0 0.00 0.00 17 7.66.20 GT 24 (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 2.90.02 0 0.00 0.00 2.90.02 ALL SPEEDS (1)(2)26 11.71.31 7 3.15.08 2.90.02 4 1.80.05 9 4.05.11 16 7.21.19 12 5.41.14 19 8.56.22 18 8.11.21 2.90.02 1.45.01 1.45.01 7 3.15.08 5 2.25.06 17 7.66.20 76 34.23.90 0 0.00 0.00 222 100.00 2.62 (1) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C=CALM (WIND SPEED LESS THAN OR EQUAL TO .60 MPH) VYNPS DSAR Revision 0 G.2-18 of 40 TABLE G.2.3 297.0 FT WIND DATA STABILITY CLASS C CLASS FREQUENCY (PERCENT) = 4.60 WIND DIRECTION FROM SPEED (MPH) N NNE NE ENE EESESESSESSSWSW WSWWWNWNWNNWVRBLTOTAL CALM (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 C-3 (1)(2)8 2.05.09 4 1.03.05 1.26 .05 2.51.02 3.77 .04 4 1.03.05 4 1.03.05 4 1.03.05 0 0.00 0.00 1.26.01 0 0.00 0.00 1.26.01 2.51 .02 0 0.00 0.00 2.51.02 3.77 .04 0 0.00 0.00 39 10.00.46 4-7 (1)(2)13 3.33.15 4 1.03.05 6 1.54.07 3.77.04 9 2.31.11 18 4.62.21 8 2.05.09 8 2.05.09 2.51.02 1.26.01 1.26 .01 0 0.00 0.00 1.26.01 2.51 .02 10 2.56.12 18 4.62.21 0 0.00 0.00 104 26.67 1.23 8-12 (1)(2)9 2.31.11 5 1.28.06 0 0.00 0.00 0 0.00 0.00 1.26.01 5 1.28.06 4 1.03.05 21 5.38.25 4 1.03.05 4 1.03.05 0 0.00 0.00 0 0.00 0.00 1.26 .01 9 2.31.11 12 3.08.14 28 7.18.33 0 0.00 0.00 103 26.41 1.22 13-18 (1)(2)8 2.05.09 1.26.01 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.26 .01 0 0.00 0.00 5 1.28.06 4 1.03.05 1.26.01 0 0.00 0.00 1.26 .01 8 2.05.09 11 2.82.13 9 2.31.11 37 9.49.44 0 0.00 0.00 86 22.05 1.01 19-24 (1)(2)2.51.02 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 3.77.04 5 1.28.06 8 2.05.09 26 6.67.31 0 0.00 0.00 44 11.28.52 GT 24 (1)(2)2.51 .02 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.26 .01 1.26.01 2.51.02 8 2.05.09 0 0.00 0.00 14 3.59.17 ALL SPEEDS (1) (2)42 10.77.50 14 3.59.17 7 1.79.08 5 1.28.06 13 3.33.15 28 7.18.33 16 4.10.19 38 9.74.45 10 2.56.12 7 1.79.08 1.26 .01 2.51.02 16 4.10.19 28 7.18.33 43 11.03.51 120 30.77 1.42 0 0.00 0.00 390 100.00 4.60 (1) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C=CALM (WIND SPEED LESS THAN OR EQUAL TO .60 MPH) VYNPS DSAR Revision 0 G.2-19 of 40 TABLE G.2.3 297.0 FT WIND DATA STABILITY CLASS D CLASS FREQUENCY (PERCENT) = 43.03 WIND DIRECTION FROM SPEED (MPH) N NNE NE ENE EESESESSESSSWSW WSWWWNWNWNNWVRBLTOTAL CALM (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 C-3 (1)(2)34.93.40 24.66 .28 23.63 .27 26.71.31 23.63 .27 40 1.10.47 48 1.32.57 22.60.26 21.58.25 12.33.14 5.14 .06 10.27.12 11.30 .13 5.14.06 16.44.19 42 1.15.50 0 0.00 0.00 362 9.93 4.27 4-7 (1)(2)59 1.62.70 29.80.34 21.58 .25 19.52.22 28.77 .33 38 1.04.45 90 2.47 1.06 95 2.61 1.12 50 1.37.59 15.41.18 11.30 .13 4.11.05 12.33.14 20.55 .24 48 1.32.57 154 4.22 1.82 0 0.00 0.00 693 19.01 8.18 8-12 (1)(2)105 2.88 1.24 44 1.21.52 24.66.28 13.36.15 15.41.18 18.49 .21 41 1.12.48 170 4.66 2.01 93 2.55 1.10 22.60.26 29.80.34 32.88 .38 81 2.22.96 119 3.26 1.40 68 1.87.80 203 5.57 2.40 0 0.00 0.00 1077 29.54 12.71 13-18 (1)(2)82 2.25.97 11.30.13 9.25.11 10.27 .12 8.22 .09 12.33 .14 8.22.09 36.99.42 80 2.19.94 14.38.17 17.47.20 16.44 .19 77 2.11.91 192 5.27 2.27 125 3.43 1.48 217 5.95 2.56 0 0.00 0.00 914 25.07 10.79 19-24 (1)(2)46 1.26.54 3.08.04 3.08 .04 0 0.00 0.00 0 0.00 0.00 1.03 .01 1.03.01 3.08 .04 15.41.18 1.03.01 5.14.06 3.08 .04 26.71.31 99 2.72 1.17 97 2.66 1.14 148 4.06 1.75 0 0.00 0.00 451 12.37 5.32 GT 24 (1)(2)25.69 .30 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.03.01 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 5.14.06 9.25 .11 18.49.21 34.93.40 57 1.56.67 0 0.00 0.00 149 4.09 1.76 ALL SPEEDS (1) (2)351 9.63 4.14 111 3.04 1.31 80 2.19.94 68 1.87.80 74 2.03.87 109 2.99 1.29 188 5.16 2.22 327 8.97 3.86 259 7.10 3.06 64 1.76.76 67 1.84.79 70 1.92.83 216 5.92 2.55 453 12.42 5.35 388 10.64 4.58 821 22.52 9.69 0 0.00 0.00 3646 100.00 43.03 (1) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C=CALM (WIND SPEED LESS THAN OR EQUAL TO .60 MPH) VYNPS DSAR Revision 0 G.2-20 of 40 TABLE G.2.3 297.0 FT WIND DATA STABILITY CLASS E CLASS FREQUENCY (PERCENT) = 34.06 WIND DIRECTION FROM SPEED (MPH) N NNE NE ENE EESESESSESSSWSW WSWWWNWNWNNWVRBLTOTAL CALM (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 C-3 (1)(2)88 3.05 1.04 41 1.42.48 48 1.66.57 49 1.70.58 68 2.36.80 90 3.12 1.06 110 3.81 1.30 54 1.87.64 19.66.22 18.62.21 15.52 .18 7.24.08 18.62 .21 19.66.22 39 1.35.46 72 2.49.85 0 0.00 0.00 755 26.16 8.91 4-7 (1)(2)103 3.57 1.22 15.52.18 5.17 .06 16.55.19 17.59 .20 52 1.80.61 131 4.54 1.55 125 4.33 1.48 41 1.42.48 23.80.27 15.52 .18 17.59.20 25.87.30 37 1.28.44 77 2.67.91 267 9.25 3.15 0 0.00 0.00 966 33.47 11.40 8-12 (1)(2)70 2.43.83 6.21.07 0 0.00 0.00 2.07.02 7.24.08 14.49 .17 41 1.42.48 87 3.01 1.03 70 2.43.83 15.52.18 15.52.18 17.59 .20 53 1.84.63 60 2.08.71 62 2.15.73 258 8.94 3.04 0 0.00 0.00 777 26.92 9.17 13-18 (1)(2)31 1.07.37 4.14.05 0 0.00 0.00 0 0.00 0.00 3.10.04 9.31 .11 10.35.12 26.90 .31 44 1.52.52 4.14.05 3.10.04 3.10 .04 26.90.31 44 1.52.52 30 1.04.35 75 2.60.89 0 0.00 0.00 312 10.81 3.68 19-24 (1)(2)4.14.05 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 2.07 .02 14.49.17 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 4.14.05 7.24 .08 9.31 .11 22.76.26 0 0.00 0.00 62 2.15.73 GT 24 (1)(2)1.03 .01 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.03.01 6.21.07 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.03 .01 1.03.01 1.03.01 3.10 .04 0 0.00 0.00 14.49.17 ALL SPEEDS (1) (2)297 10.29 3.50 66 2.29.78 53 1.84.63 67 2.32.79 95 3.29 1.12 165 5.72 1.95 292 10.12 3.45 295 10.22 3.48 194 6.72 2.29 60 2.08.71 48 1.66.57 44 1.52.52 127 4.40 1.50 168 5.82 1.98 218 7.55 2.57 697 24.15 8.23 0 0.00 0.00 2886 100.00 34.06 (1) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C=CALM (WIND SPEED LESS THAN OR EQUAL TO .60 MPH) VYNPS DSAR Revision 0 G.2-21 of 40 TABLE G.2.3 297.0 FT WIND DATA STABILITY CLASS F CLASS FREQUENCY (PERCENT) = 13.03 WIND DIRECTION FROM SPEED (MPH) N NNE NE ENE EESESESSESSSWSW WSWWWNWNWNNWVRBLTOTAL CALM (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 C-3 (1)(2)48 4.35.57 42 3.80.50 22 1.99.26 21 1.90.25 30 2.72.35 40 3.62.47 40 3.62.47 30 2.72.35 22 1.99.26 13 1.18.15 13 1.18.15 4.36.05 17 1.54.20 11 1.00.13 23 2.08.27 47 4.26.55 0 0.00 0.00 423 38.32 4.99 4-7 (1)(2)41 3.71.48 8.72.09 2.18 .02 5.45.06 13 1.18.15 34 3.08.40 66 5.98.78 39 3.53.46 25 2.26.30 12 1.09.14 16 1.45.19 19 1.72.22 16 1.45.19 29 2.63.34 37 3.35.44 92 8.33 1.09 0 0.00 0.00 454 41.12 5.36 8-12 (1)(2)12 1.09.14 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.09.01 2.18.02 10.91.12 15 1.36.18 12 1.09.14 5.45.06 5.45 .06 6.54 .07 22 1.99.26 11 1.00.13 29 2.63.34 75 6.79.89 0 0.00 0.00 205 18.57 2.42 13-18 (1)(2)1.09.01 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 4.36.05 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 2.18.02 4.36 .05 0 0.00 0.00 8.72.09 0 0.00 0.00 19 1.72.22 19-24 (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.09.01 0 0.00 0.00 0 0.00 0.00 2.18.02 0 0.00 0.00 3.27 .04 GT 24 (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 ALL SPEEDS (1) (2)102 9.24 1.20 50 4.53.59 24 2.17.28 26 2.36.31 44 3.99.52 76 6.88.90 116 10.51 1.37 84 7.61.99 63 5.71.74 30 2.72.35 34 3.08.40 29 2.63.34 58 5.25.68 55 4.98.65 89 8.06 1.05 224 20.29 2.64 0 0.00 0.00 1104 100.00 13.03 (1) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C=CALM (WIND SPEED LESS THAN OR EQUAL TO .60 MPH) VYNPS DSAR Revision 0 G.2-22 of 40 TABLE G.2.3 297.0 FT WIND DATA STABILITY CLASS G CLASS FREQUENCY (PERCENT) = 1.76 WIND DIRECTION FROM SPEED (MPH) N NNE NE ENE EESESESSESSSWSW WSWWWNWNWNNWVRBLTOTAL CALM (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 C-3 (1)(2)2 1.34.02 1.67.01 3 2.01.04 1.67.01 1.67 .01 4 2.68.05 2 1.34.02 3 2.01.04 4 2.68.05 3 2.01.04 3 2.01.04 0 0.00 0.00 2 1.34.02 2 1.34.02 1.67.01 2 1.34.02 0 0.00 0.00 34 22.82.40 4-7 (1)(2)3 2.01.04 1.67.01 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 2 1.34.02 7 4.70.08 5 3.36.06 7 4.70.08 3 2.01.04 3 2.01.04 8 5.37.09 6 4.03.07 8 5.37.09 12 8.05.14 10 6.71.12 0 0.00 0.00 75 50.34.89 8-12 (1)(2)1.67.01 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 2 1.34.02 2 1.34.02 3 2.01.04 0 0.00 0.00 3 2.01.04 3 2.01.04 11 7.38.13 3 2.01.04 2 1.34.02 7 4.70.08 0 0.00 0.00 37 24.83.44 13-18 (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.67 .01 0 0.00 0.00 0 0.00 0.00 1.67.01 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 1.67.01 0 0.00 0.00 3 2.01.04 19-24 (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 GT 24 (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 ALL SPEEDS (1) (2)6 4.03.07 2 1.34.02 3 2.01.04 1.67.01 1.67 .01 6 4.03.07 11 7.38.13 11 7.38.13 14 9.40.17 6 4.03.07 10 6.71.12 11 7.38.13 19 12.75.22 13 8.72.15 15 10.07.18 20 13.42.24 0 0.00 0.00 149 100.00 1.76 (1) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C=CALM (WIND SPEED LESS THAN OR EQUAL TO .60 MPH) VYNPS DSAR Revision 0 G.2-23 of 40 TABLE G.2.3 297.0 FT WIND DATA STABILITY CLASS ALL CLASS FREQUENCY (PERCENT) = 100.00 WIND DIRECTION FROM SPEED (MPH) N NNE NE ENE EESESESSESSSWSW WSWWWNWNWNNWVRBLTOTAL CALM (1)(2)0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 C-3 (1)(2)181 2.14 2.14 113 1.33 1.33 100 1.18 1.18 103 1.22 1.22 126 1.49 1.49 179 2.11 2.11 209 2.47 2.47 115 1.36 1.36 66.78.78 47.55.55 36.42 .42 22.26.26 50.59.59 39.46 .46 81.96.96 167 1.97 1.97 0 0.00 0.00 1634 19.28 19.28 4-7 (1)(2)233 2.75 2.75 62.73.73 35.41 .41 45.53.53 76.90 .90 153 1.81 1.81 306 3.61 3.61 274 3.23 3.23 126 1.49 1.49 54.64.64 46.54 .54 48.57.57 60.71.71 96 1.13 1.13 186 2.19 2.19 555 6.55 6.55 0 0.00 0.00 2355 27.79 27.79 8-12 (1)(2)206 2.43 2.43 55.65 .65 24.28.28 15.18.18 24.28 .28 49.58.58 104 1.23 1.23 318 3.75 3.75 200 2.36 2.36 48.57.57 52.61.61 59.70 .70 169 1.99 1.99 203 2.40 2.40 179 2.11 2.11 604 7.13 7.13 0 0.00 0.00 2309 27.25 27.25 13-18 (1)(2)128 1.51 1.51 19.22.22 9.11 .11 10.12 .12 11.13.13 23.27 .27 20.24.24 78.92 .92 140 1.65 1.65 19.22.22 22.26.26 20.24 .24 119 1.40 1.40 255 3.01 3.01 173 2.04 2.04 372 4.39 4.39 0 0.00 0.00 1418 16.73 16.73 19-24 (1)(2)52.61 .61 3.04.04 3.04.04 0 0.00 0.00 0 0.00 0.00 1.01 .01 1.01 .01 5.06.06 29.34.34 1.01 .01 5.06.06 3.04.04 34.40 .40 112 1.32 1.32 118 1.39 1.39 212 2.50 2.50 0 0.00 0.00 579 6.83 6.83 GT 24 (1)(2)28.33.33 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 0 0.00 0.00 2.02.02 6.07.07 0 0.00 0.00 0 0.00 0.00 5.06.06 11.13.13 20.24.24 37.44 .44 70.83 .83 0 0.00 0.00 179 2.11 2.11 ALL SPEEDS (1) (2)828 9.77 9.77 252 2.97 2.97 171 2.02 2.02 173 2.04 2.04 237 2.80 2.80 405 4.78 4.78 640 7.55 7.55 792 9.35 9.35 567 6.69 6.69 169 1.99 1.99 181 1.90 1.90 157 1.85 1.85 443 5.23 5.23 725 8.56 8.56 774 9.13 9.13 1980 23.37 23.37 0 0.00 0.00 8474 100.00 100.00 (1) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2) = PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C=CALM (WIND SPEED LESS THAN OR EQUAL TO .60 MPH) VYNPS DSAR Revision 0 G.2-24 of 40 TABLE G.2.4 Wind Direction Persistence Summary (35-foot level) WIND DIRECTION PERSISTENCE

SUMMARY

= NUMBER OF OBSERVATIONS AND PERCENT PROBABILITY DIRECTION PERSISTENCE (HOURS) DIRECTION 1 2 3 4 5 6 7891011121314151617 18192021222324GT.24TOTAL N 254 73 61 91 18 96 8 99 2 99 1 99 2 100 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 0 346 NNE 120 82 22 97 3 99 1 100 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 0 146 NE 113 90 8 96 2 98 1 98 1 99 0 99 1 100 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 0 126 ENE 104 84 17 98 2 99 1 100 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 0 124 E 153 80 31 96 6 99 0 99 1 99 1 100 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 0 192 ESE 177 83 25 95 6 98 4 100 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 0 212 SE 234 78 43 93 15 98 5 99 2 100 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 0 299 SSE 273 67 83 87 26 94 9 96 6 97 5 99 5 100 0 100 0 100 1 100 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 0 408 S 340 79 46 90 19 95 10 97 7 99 4 100 0 100 0 100 2 100 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 0 428 SSW 304 84 47 97 6 99 3 100 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 0 360 SW 347 78 75 95 14 98 4 99 1 100 1 100 0 100 1 100 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 0 443 WSW 357 82 68 97 9 99 2 100 1 100 1 100 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 0 438 W 439 76 94 92 26 97 12 99 3 99 1 99 3 100 0 100 1 100 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 0 579 WNW 443 75 108 93 16 96 18 99 4 94 2 100 1 100 1 100 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 0 593 NW 500 71 121 89 39 94 21 97 12 99 2 99 2 94 0 99 2 100 1 100 0 100 0 100 0 100 0 100 1 100 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 701 NNW 384 57 133 77 67 87 27 91 17 94 18 97 4 97 3 98 2 98 7 99 1 99 1 99 1 99 1 100 2 100 1 100 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 669TOTAL 4542 982 274 126 57 36 185791111310 00000000 6064 VYNPS DSAR  Revision 0 G.2-25 of 40 TABLE G.2.5 Wind Direction Persistence Summary (297-foot level)  WIND DIRECTION PERSISTENCE 

SUMMARY

= NUMBER OF OBSERVATIONS AND PERCENT PROBABILITY DIRECTION PERSISTENCE (HOURS) DIRECTION 1 2 3 4 5 6 7891011121314151617 18192021222324GT.24TOTAL N 351 69 85 86 33 93 13 95 12 97 5 98 4 99 1 99 1 100 0 100 0 100 0 100 1 100 0 100 0 100 1 100 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 507 NNE 169 83 25 95 8 99 0 99 1 100 1 100 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 0 204 NE 109 82 16 94 6 98 1 99 0 99 0 99 0 99 1 100 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 0 133 ENE 134 88 16 99 0 99 1 99 1 100 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 0 152 E 160 83 24 96 3 97 4 99 1 100 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 0 192 ESE 264 81 52 97 5 98 3 99 2 100 1 100 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 0 327 SE 317 72 79 89 31 96 6 98 7 99 1 100 1 100 0 100 1 100 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 0 443 SSE 301 66 74 83 33 99 19 94 7 96 10 98 4 99 3 99 2 100 0 100 0 100 0 100 1 100 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 454 S 205 65 60 84 17 89 13 93 10 96 2 97 4 98 2 99 1 99 1 99 1 100 0 100 1 100 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 317 SSW 140 91 12 99 1 99 1 100 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 0 154 SW 119 85 17 97 4 100 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 0 140 WSW 98 80 19 95 3 98 2 99 0 99 1 100 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 0 123 W 220 73 49 89 20 96 6 98 2 98 2 99 3 100 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 0 302 WNW 246 62 79 82 31 89 24 95 7 97 3 98 0 98 2 98 3 99 1 99 0 99 1 100 0 100 0 100 0 100 0 100 1 100 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 398 NW 336 69 81 86 33 93 14 96 10 98 2 98 6 100 0 100 1 100 0 100 0 100 0 100 0 100 1 100 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 484 NNW 336 47 147 68 77 78 38 84 35 89 21 91 10 93 10 94 7 95 9 97 5 97 6 98 3 98 4 99 0 99 0 99 1 99 0 99 2 99 0 99 1 100 0 100 0 100 0 100 3*100 715TOTAL 3505 835 305 145 95 49 321916116765012 02010003 5045

• Of these three occurrences, one lasted 26 hours, the second lasted 34 hours, and the third lasted 36 hours.

TABLE G.2.6 Inversion Persistence Summary (198-33 foot Delta-T) THE LONGEST INVERSION LASTED 38 HOURS OF THE LONGEST INVERSIONS, NUMBER 1 STARTED 18 HOURS INTO DAY 327 THIRD COLUMN DEFINES THE PERCENT PROBABILITY THAT IF AN INVERSION OCCURS, ITS DURATION WILL BE LESS THAN THE

NUMBER OF HOURS SPECIFIED VYNPS DSAR Revision 0 G.2-26 of 40 Duration (hours)Number of Observations Percent Probability1 146 27.39 2 60 38.65 3 49 47.84 4 34 54.22 5 31 60.04 6 22 64.17 7 26 69.04 8 20 72.80 9 20 76.55 10 27 81.61 11 24 86.12 12 26 90.99 13 21 94.93 14 12 97.19 15 4 97.94 16 3 98.50 17 2 98.87 18 1 99.06 19 1 99.25 20 2 99.62 21 0 99.62 22 0 99.62 23 1 99.81 24 0 99.81 25 0 99.81 26 0 99.81 TABLE G.2.6 Inversion Persistence Summary (198-33 foot Delta-T) THE LONGEST INVERSION LASTED 38 HOURS OF THE LONGEST INVERSIONS, NUMBER 1 STARTED 18 HOURS INTO DAY 327 THIRD COLUMN DEFINES THE PERCENT PROBABILITY THAT IF AN INVERSION OCCURS, ITS DURATION WILL BE LESS THAN THE

NUMBER OF HOURS SPECIFIED VYNPS DSAR Revision 0 G.2-27 of 40 Duration (hours)Number of Observations Percent Probability27 0 99.81 28 0 99.81 29 0 99.81 30 0 99.81 31 0 99.81 32 0 99.81 33 0 99.81 34 0 99.81 35 0 99.81 36 0 99.81 37 0 99.81 38 1 100.00 THE LONGEST INVERSION LASTED 38 HOURS OF THE LONGEST INVERSIONS, NUMBER 1 STARTED 18 HOURS INTO DAY 327 THIRD COLUMN DEFINES THE PERCENT PROBABILITY THAT IF AN INVERSION OCCURS, ITS DURATION WILL BE LESS THAN THE

NUMBER OF HOURS SPECIFIED VYNPS DSAR Revision 0 G.2-28 of 40 TABLE G.2.7 Inversion Persistence Summary (295-33 foot Delta-T) Duration (hours)Number of Observations Percent Probability1 140 27.94 2 50 37.92 3 49 47.70 4 25 52.69 5 27 58.08 6 19 61.88 7 31 68.06 8 18 71.66 9 17 75.05 10 23 79.64 11 25 84.63 12 29 90.42 13 15 93.41 14 15 96.41 15 4 97.21 16 2 97.60 17 3 98.20 18 4 99.00 19 1 99.20 20 1 99.40 21 0 99.40 22 0 99.40 23 1 99.60 24 0 99.60 25 1 99.80 26 0 99.80 27 0 99.80 THE LONGEST INVERSION LASTED 38 HOURS OF THE LONGEST INVERSIONS, NUMBER 1 STARTED 18 HOURS INTO DAY 327 THIRD COLUMN DEFINES THE PERCENT PROBABILITY THAT IF AN INVERSION OCCURS, ITS DURATION WILL BE LESS THAN THE

NUMBER OF HOURS SPECIFIED VYNPS DSAR Revision 0 G.2-29 of 40 TABLE G.2.7 (Continue) Inversion Persistence Summary (295-33 foot Delta-T) Duration (hours)Number of Observations Percent Probability28 0 99.80 29 0 99.80 30 0 99.80 31 0 99.80 32 0 99.80 33 0 99.80 34 0 99.80 35 0 99.80 36 0 99.80 37 0 99.80 38 1 100.00 VYNPS DSAR Revision 0 G.2-30 of 40 Vermont Yankee Defueled Safety Analysis Report Revision0Location of Primary and Backup Meteorological Towers Figure G.2-1 VYNPS DSAR Revision 0 G.2-31 of 40Vermont Yankee Defueled Safety Analysis Report Revision0 Spring Wind Rose (35-Foot Level) March 1980 - May 1980 Figure G.2-2 VYNPS DSAR Revision 0 G.2-32 of 40Vermont Yankee Defueled Safety Analysis Report Revision0 Summer Wind Rose (35-Foot Level) June 1980 - August 1980 Figure G.2-3 VYNPS DSAR Revision 0 G.2-33 of 40Vermont Yankee Defueled Safety Analysis Report Revision0 Autumn Wind Rose (35-Foot Level) September 1980 - November 1980 Figure G.2-4 VYNPS DSAR Revision 0 G.2-34 of 40Vermont Yankee Defueled Safety Analysis Report Revision0 Winter Wind Rose (35-Foot Level) January 1980 - February 1980; December 1980 Figure G.2-5 VYNPS DSAR Revision 0 G.2-35 of 40Vermont Yankee Defueled Safety Analysis Report Revision0 Annual Wind Rose (35-Foot Level) January 1980 - December 1980 Figure G.2-6 VYNPS DSAR Revision 0 G.2-36 of 40Vermont Yankee Defueled Safety Analysis Report Revision0 Spring Wind Rose (297-Foot Level) March 1980 - May 1980 Figure G.2-7 VYNPS DSAR Revision 0 G.2-37 of 40Vermont Yankee Defueled Safety Analysis Report Revision0 Summer Wind Rose (297-Foot Level) June 1980 - August 1980 Figure G.2-8 VYNPS DSAR Revision 0 G.2-38 of 40Vermont Yankee Defueled Safety Analysis Report Revision0 Autumn Wind Rose (297-Foot Level) September 1980 - November 1980 Figure G.2-9 VYNPS DSAR Revision 0 G.2-39 of 40Vermont Yankee Defueled Safety Analysis Report Revision0 Winter Wind Rose (297-Foot Level) January 1980 - February 1980; December 1980 Figure G.2-10 VYNPS DSAR Revision 0 G.2-40 of 40Vermont Yankee Defueled Safety Analysis Report Revision0 Annual Wind Rose (297-Foot Level) January 1980 - December 1980 Figure G.2-11}}