Text

Defueled Safety Analysis Report, Revision 2
	ML20276A250
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	09/23/2020
From:	Reid B; NorthStar Nuclear Decommissioning Company
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	BVY 20-026
	Download: ML20276A250 (126)
	v • d • e

~

North Star BVY 20-026 September 23, 2020 ATTN: Document Control Desk, U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Northstar Nuclear Decommissioning Co., LLC Vermont Yankee Nuclear Power Station 320 Governor Hunt Rd.

Vernon, VT 05354 802-451-5354 Billy E. Reid, Jr.

Site Vice president 10 CFR 50.71(e) 10 CFR 50.4(b)(6) 10 CFR 50.59(d)(2)

SUBJECT:

Defueled Safety Analysis Report, Revision 2 Vermont Yankee Nuclear Power Station Docket No. 50-271 License No. DPR-28 D$ar Sir or Madam:

Pursuant to 10 CFR 50.71(e) and 10 CFR 50.4(b)(6), NorthStar Nuclear Decommissioning Co.,

LLC, hereby submits Revision 2 of the Vermont Yankee Nuclear Power Station (VY) Defueled Safety Analysis Report (DSAR). The DSAR is maintained considering the guidance contained within NRC Regulatory Guide 1.184, "Decommissioning of Nuclear Power Reactors,"

Revision 1, and serves the same function during decommissioning that the Updated Final Safety Analysis Report (UFSAR) served during operation of the facility.

The Attachment identifies the changes that were incorporated into Revision 2 of the DSAR and lists the DSAR sections affected by each change.

In accordance with 10 CFR 50.71(e) and 10 CFR 50.4(b)(6), please find Enclosed a copy of Revision 2 of the DSAR. Changes to the DSAR are indicated by revision bars. This is a complete revision of the DSAR, and all pages have been converted to Revision 2.

Pursuant to 10 CFR 50.59(d)(2), Northstar Nuclear Decommissioning Co., LLC is reporting that for the interval of September 28, 2017 through September 23, 2020, there were no 10 CFR 50.59 evaluations performed for any changes, tests, or experiments made to the Vermont Yankee Nuclear Power Station.

This letter contains no new regulatory commitments. Should you have any questions concerning this letter, please contact Mr. Thomas B. Silko at (802) 451-5354, Ext 2506.

I declare under penalty of perjury that the foregoing is true and correct. Executed on September 23, 2020.

BVY 20-026 / Page 2 of 2 Sincerely,

Attachment:

Defueled Safety Analysis Report (DSAR) Revision 2 Changes.

Enclosure:

Vermont Yankee Nuclear Power Station Defueled Safety Analysis Report (DSAR), Revision 2.

cc:

Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 Mr. Jack D. Parrott, Senior Project Manager Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Mail Stop T-5A10 Washington, DC 20555 Ms. June Tierney, Commissioner Vermont Department of Public Service 112 State Street - Drawer 20 Montpelier, Vermont 05602-2601

Attachment Vermont Yankee Nuclear Power Station Defueled Safety Analysis Report (DSAR) Revision 2 Changes BVY 20-026 Docket No. 50-271

BVY 20-026 / Attachment / Page 1 of 4 Vermont Yankee Defueled Safety Analysis Report (DSAR) Revision 1 Changes Change#

Sections Affected Brief Description of Change DCR 02/02 4.4.1 Incorporate DSAR changes as a result of Revision 7 to 4.4.2 QAPM. Includes Training related Changes.

5.2.1 5.2.4 5.2.5 5.2.8 5.5.3 5.4.4 Table 7.4 (Item 35)

DCR 02/03 1.1 Incorporate License Amendment #270 "Dry Fuel 1.2 Storage Only Tech Specs" into DSAR. Includes EC 1.3.2 75772 "Final Abandonment of SFPC," and cancellation 1.3.3 of Technical Requirements Manual (TRM).

1.4.1 1.4.2 2.4.3 2.6.1 2.6.3 3.1.2 3.1.3 3.2.1 3.3.1 Figures 3.3.1 - 5 3.3.2 3.3.12 4.4.2 4.5.1 4.5.2 4.6.1 4.7.1 5.6 6.1 6.2.1 6.3 6.3.1 Tables 6.3.1 - 7 Figures 6.3.1 & 2 Figure A-1 6.4.1 6.5 6.6 7.2.19 Table 7.4 (Item 52)

BVY 20-026 / Attachment I Page 2 of 4 Change#

Sections Affected Brief Description of Change DCR 02/04 1.3.4 Abandonment of Area Radiation Monitors (ARM) 2.6.1 Systems (EC 75773) & Process Rad Monitors 3.1.1 (EC 75818).

3.3.3 4.2.2 4.3.1 4.7.1.

4.7.2 Tables 4.7.1.1 - 3 Tables 4.7.2.1 - 2 DCR 02/05 1.3.5 Electrical Power System (Section 3.3.3).

3.3.3.

3.3.9 DCR 02/08 7.2 Incorporate EC 76710 to change safety classification of SSC's {QAPM change request-BVY 17-023).

DCR 02/10 1.3.5 EC 75809 "Abandonment of Service Water."

2.4.3 3.2.5 3.3 3.3.2 DCR 02/13 1.3.5 EC 77 400 "SBO Diesel and Vernon Tie 2.4.3 Abandonment."

3.3.3 DCR 02/14 1.3.5 EC 77498 "Abandon Diesel Fire Pump and B5B Pump."

3.2.5 3.3.4 7.2.3 Table 7.4 (Items 3, 4, 9 &

49)

DCR 02/18 3.1.3 EC 79028 "Abandonment of Seismic Monitor SM-117-1."

DCR 02/19 3.3.4 EC 77793 "Install Fire Water Jockey Pump."

DCR 02/20 7.2.11 Cancellation of Maintenance Rule Program and Procedures.

DCR 02/23 7.2.7 Cancellation of Lube Oil Analysis ProQram.

DCR 02/24 7.2.9 Cancellation of Service Water and Heat Exchanger ProQram.

DCR 02/25 7.2.12 System Engineering Business Core and Support 7.2.16 procedure.

Table 7.4 (Items 24, 30, 34 & 35}

DCR 02/26 1.3.5 EC 78769 "Abandon Cable Vault CO2, Computer room 3.3.3 and SU Transformer systems. Discontinue SR's on 3.3.4 fire barriers."

Table 7.4 {Item 8}

DCR 02/27 2.6.2 DSAR Changes to reflect Revision 40 to the ODCM.

4.5.2 DCR 02/28 1.1 Incorporate License Transfer to Northstar (License 1.3.1 Amendment 271) into DSAR.

2.2.1 2.2.4 DCR 02/29 7.2.6 Deletion of Cable Reliability Program.

7.2.8 Table 7.4 (Item 13}

BVY 20-026 /Attachment/ Page 3 of 4 Change#

Sections Affected Brief Description of Change OCR 02/30 3.1.1 Implementation of PISA Fire Protection Program 3.3.4 (DDFS-FP-10184) and cancellation of site engineering 7.2.3 procedure SEP-FP-VTY-003.

7.2.4 Table 7.4 (Items 10, 11 &

31}

DCR 02/31 2.3.1 Deletion of Met Tower and Meteorological Monitoring 2.3.2 Program.

2.3.3 2.3.4 2.3.5 2.3.6 2.3.7 2.3.8 Tables 2.3.1, 2.3.4 &

2.3.5 Figures 2.3-1 thru 2.3-3 Appendix G.2 DCR 02/32 1.2 Admin Changes for Rev 2.

1.3.1 1.3.6 1.3.8 3.1.2 3.1.3 3.2.1 3.3.5 3.3.10 4.2.2 4.2.3 4.5.1 4.5.2 4.6.1 5.4.2 6.2.2 6.4.1 DCR 02/33 2.4 Major revision of the Hydrology and Biology (2.4) and 2.4.2 Geology and Seismology (2.5) portions of the DSAR.

2.4.3 Includes related content such as Table of Content.

2.4.4 2.4.5 2.4.6 2.4.7 2.4.8 2.4.9 Tables 2.4.1 thru 2.4.11 Figures 2.4-1 thru 2.4-6 Figures 2.4-8 thru 2.4-11 2.5.1 2.5.2 2.5.3 2.5.4 Tables 2.5.1 and 2.5.2 FiQures 2.5-1 thru 2.5-15

BVY 20-026 / Attachment / Page 4 of 4 Change#

Sections Affected Brief Description of Change OCR 02/34 2.2.1 Revise DSAR Section 2.2 (Site Description).

2.2.3 2.2.4 Tables 2.2.4 and 2.2.7 FiQures 2.2-1 & 2.2-5 OCR 02/35 7.1 Delete remainder of DSAR Section 7 (Aging 7.2.1 Management-Renewed Operating License).

7.2.2 7.2.5 7.2.8 7.2.10 7.2.13 7.2.14 7.3 7.4 OCR 02/36 1.3.5 Abandon Switchgear Room CO2 and Heating Boiler 3.3.4 Room detection.

OCR 02/37 1.3.5 Abandon House Heating Boiler, Instrument and 3.3.5 Service Air.

3.3.6 OCR 02/38 3.2.6 Process Computer System, Cooling Tower Deep Basin 3.3.9 and Lighting Systems.

3.3.11 OCR 02/39 4.6.1 Solid Waste Management (4.6) and Area Radiation Table 4.6.1.1 Monitoring System (4.7.2).

4.7.2 Table 4.7.2.3 Figure 4.7.2-2 OCR 02/40 4.3.2 Liquid Waste Management systems (4.5).

4.5.1 4.5.2 Tables 4.5.2.1 thru 4.5.2.5 Figure 4.5.2-8

Enclosure Vermont Yankee Nuclear Power Station Defueled Safety Analysis Report (DSAR), Revision 2

( 118 pages excluding this cover sheet)

BVY 20-026 Docket No. 50-271

BVY 20-025 / Enclosure / Page 1 of 118 VYNPS DEFUELED SAFETY ANALYSIS REPORT VERMONT YANKEE NUCLEAR POWER STATION DSAR Revision 2

BVY 20-025 /Enclosure/ Page 2 of 118 DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS SECTION 1

1.0 INTRODUCTION

AND

SUMMARY

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 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 4.5 LIQUID WASTE MANAGEMENT SYSTEMS 4.6 SOLID WASTE MANAGEMENT 4.7 EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING VYNPS DSAR Revision 2 TOC-1 of 2

BVY 20-025 / Enclosure I Page 3 of 118 SECTION 5 5.0 SECTION 6 6.0 VYNPS DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS (Continued)

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 SAFETY ANALYSIS 6.1 6.2 6.3 6.4

6.5 INTRODUCTION

ACCEPTANCE CRITERIA Deleted SITE EVENTS EVALUATED REFERENCES DSAR Revision 2 TOC-2 of 2

BVY 20-025 / Enclosure / Page 4 of 118 INTRODUCTION AND

SUMMARY

TABLE OF CONTENTS Section Title

1. 1 INTRODUCTION.......................................................... 3

1. 2 DESIGN CRITERIA....................................................... 5

1. 3 FACILITY DESCRIPTION.................................................. 7 VYNPS 1.3.1

1. 3.2

1. 3. 3 1.3. 4

1. 3. 5

1. 3. 6 1.3. 7

1. 3. 8 General...................................................... 7 1.3.1.1 1.3.1.2 Site and Environs............................... 7 Facility Arrangement............................ 9 Fuel Storage and Handling.................................... 9 1.3.2.1 Nuclear Fuel.................................... 9 Radioactive Waste Management................................. 9 1.3.3.1 1.3.3.2 1.3.3.3 Equipment and Floor Drainage Systems............ 9 Liquid Radwaste System......................... 10 Solid Radwaste System.......................... 10 Deleted..................................................... 10 Auxiliary Systems........................................... 10 1.3.5.1 1.3.5.2 1.3.5.3 1.3.5.4 1.3.5.5 1.3.5.6 Electrical Power Systems....................... 10 Deleted........................................ 10 Fire Protection System......................... 10 Heating, Ventilating, and Air Conditioning Systems........................... 11 Deleted........................................ 11 Process Sampling System........................ 11 Deleted..................................................... 11 Station Water Purification, Treatment and Storage........... 11 1.3.7.1 1.3.7.2 Deleted,....................................... 11 Potable and Sanitary Water System.............. 12 Shielding, Access Control, and Radiation Protection Procedures.................................................. 12 DSAR Revision 2

1. 0-1 of 14 7

BVY 20-025 / Enclosure / Page 5 of 118 1.3.8.1 General........................................ 12

1. 3. 9 Structural Loading Criteria................................. 13

1. 4

SUMMARY

OF RADIATION EFFECTS......................................... 14 1.5 VYNPS

1. 4.1

1. 4. 2 Fuel Storage and Handling and Waste Management.............. 14 Accidents and Events...........*.....................*...... 14 GENERAL CONCLUSIONS........................................*......... 14 DSAR Revision 2

1. 0-2 of 14 l

BVY 20-025 / Enclosure/ Page 6 of 118

1.1 INTRODUCTION

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 (10 CFR 50).

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.

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.

The operating license for VY was issued on March 21, 1972, and commercial operation commenced on November 30, 1972.

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.

The operating license for VY was renewed for an additional 20 years on March 21, 2011.

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 VYNPS DSAR Revision 2

1. 0-3 of 14

BW 20-025 / Enclosure / Page 7 of 118 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.

On August 16, 2018, ENO notified the NRC that as of August 1, 2018, all spent nuclear fuel assemblies have been transferred out of the spent fuel pool and have been placed in dry storage within the ISFSI.

On August 15, 2018, ENO implemented License Amendment #270 which reflected the permanent removal of all spent nuclear fuel from the spent fuel pool (SFP) and transfer to the fuel to dry cask storage within an Independent Spent Fuel Storage Installation (ISFSI).

This License Amendment prohibited the placement of fuel within the spent fuel pool and made other conforming revisions to the VY Operation License and Technical Specifications to reflect the permanently-shutdown status of VY, as well as the reduced scope of structures, systems and components necessary to ensure plant safety since all spent fuel has been permanently moved to the ISFSI.

On January 11, 2019, the operating license was transferred from Entergy Nuclear Vermont Yankee, LLC to Northstar Vermont Yankee, LLC and Northstar Nuclear Decommissioning Company, LLC (Northstar NOC).

The Defueled Safety Analysis Report (DSAR) Revision O was derived.from Revision 26 of the VYNPS Updated Final Safety Analysis Report (UFSAR) and was devel*oped as a licensing basis document that reflects the permanently defueled condition of VYNPS.

DSAR, Revision 1 reflected changes to the facility to be consistent with guidance provided in Regulatory Guide 1.184, Revision 1 (e.g.,

deleted or modified section or portions of sections describing systems, structures and components (SSCs) that are repetitive, or not associated with the storage of spent fuel, or have been abandoned, changed, and/or are no longer utilized since Revision O was issued).

DSAR Revision 2 reflects the removal of all spent nuclear fuel from the spent fuel pool (SFP) and the transferral into dry cask storage within an Independent Spent Fuel Storage Installation (ISFSI) and removes descriptions of SSC's relied upon for the storage of spent nuclear fuel in the SFP.

The DSAR serves the same function during SAFSTOR and decommissioning that the UFSAR served during operation of the facility.

The criteria used to evaluate the major SSCs and the conclusions of the evaluations are provided in appropriate station documents.

Northstar Vermont Yankee acknowledges that the 10 CFR 50 operating license continues to remain in effect until the Nuclear Regulatory Commission terminates the license.

VYNPS DSAR Revision 2

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BVY 20-025 / Enclosure / Page 8 of 118 1.2 DESIGN CRITERIA The principal architectural and engineering criteria for the design and construction of the station, no longer apply with the removal of all spent nuclear fuel from the SFP and its transfer to dry cask storage within an ISFSI.

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 2

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BVY 20-025 / Enclosure / Page 9 of 118 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 10 CFR 20.

VYNPS DSAR Revision 2

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--7

BVY 20-025 / Enclosure / Page 1 0 of 118 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 5920-6245.

1.3.1.1.2 Site Ownership Northstar 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 Northstar Vermont Yankee, LLC and Northstar Nuclear Decommissioning Company, LLC. 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 local 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 2

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BVY 20-025 / Enclosure / Page 11 of 118 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 hydr?electric 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 2

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BVY 20-025 /Enclosure/ Page 12 of 118 1.3.1.2 Facility Arrangement The facility arrangement is shown on Drawing 5920-6245.

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 an Independent Spent Fuel Storage Installation (ISFSI) storage pad.

1. 3. 2 Fuel Storage and Handling 1.3.2.l Nuclear Fuel Nuclear fuel preyiously 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. As of August 1, 2018, all nuclear fuel is stored at the Independent Spent Fuel Storage Installation (ISFSI) Facility.

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 10 CFR 20 and within the limits specified in 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 are:

1. The chemical waste sump and equipment and floor sumps in the Radwaste Building and Reactor Building that contain or potentially contain radioactive liquids are routed to the Torus.

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

VYNPS DSAR Revision 2

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BVY 20-025 / Enclosure / Page 13 of 118 1.3.3.2 Liquid Radwaste System The Liquid Radwaste System is no longer in service.

The system has been drained to the extent practical.

The Torus-as-CST System processes water collected from the chemical waste sump and equipment and floor drains in the Radwaste Building and Reactor Building.

This water is stored in the Torus and may be disposed of offsite or discharged to the environs in accordance with applicable permits and regulatory approvals.

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.

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 Deleted

1. 3. 5 Auxiliary Systems 1.3.5.1 Electrical Power Systems Off-site power is supplied to the facility from the 115 kV switchyard via two startup transformers and a local 13.8 kV distribution system along Governor Hunt Road.

1.3.5.2 Deleted 1.3.5.3 Fire Protection System Water for the Fire Protection System is supplied by an electric-motor driven vertical turbine-type pump, located in the intake structure.

This pump supplies water to the facility fire loop with its various hydrants and subsequently to the standpipe connections, sprinklers, and deluge systems throughout portions of the facility.

Supplementing these water systems are portable fire extinguishers are located throughout the facility.

Consideration has been given to the use of noncombustible and fire-resistant materials throughout the facility.

VYNPS DSAR Revision 2 1.0-10 of 14

BVY 20-025 / Enclosure / Page 14 of 118 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 Deleted 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

1. 3. 7 1.3.7.1 VYNPS Deleted Station Water Purification, Treatment and Storage Deleted DSAR Revision 2 1.0-11 of 14

BVY 20-025 / Enclosure / Page 15 of 118 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.l 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 10 CFR 20 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 10 CFR 20.

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 2 1.0-12 of 14

BVY 20-025 / Enclosure / Page 16 of 118 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 2 1.0-13 of 14

BVY 20-025 / Enclosure / Page 17 of 118 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 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 radwaste transfer cask drop event.

The calculated consequences are substantially below the dose limits given in 10 CFR 100 for the transfer cask drop event.

With the removal of all spent nuclear fuel from the SFP and its transfer into dry cask storage within an ISFSI, a fuel handling accident is no longer possible and is no longer discussed within the DSAR.

A further description of the radwaste transfer cask drop event 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 2 1.0-14 of 14

BVY 20-025 / Enclosure / Page 18 of 118 SITE AND ENVIRONS TABLE OF CONTENTS Section Title 2.1

SUMMARY

DESCRIPTION.................. *................ *.............. 4

2. 2 SITE DESCRIPTION..................................................... 4 2.2.1 2.2.2 2.2.3 2.2.4 2.2.5 Location and Area............*..........*................... 4 Population.................................................. 5 Land Use...................................................* 5 Site Area Boundaries, Exclusion Area, and Low Population Zone......................................... 6 Conclusions................................... *............. 9 2. 3 METEOROLOGY....................................... *................. 12 2.3.1 General.................................................... 12

2. 4 HYDROLOGY..*........................*............................... 12 2.4.1 2.4.2 2.4.3 2.4.4 General................................*................... 12 Site Area.............................. *................... 12 Floods..................................................... 13 Conclusions.........*.......................... *............ 15

2. 5 GEOLOGY AND SEISMOLOGY.............................................. 16 2.5.1 VYNPS General.................................................... 16 DSAR Revision 2 2.0-1 of 21

BVY 20-025 / Enclosure / Page 19 of 118 SITE AND ENVIRONS TABLE OF CONTENTS Section Title 2.6 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM....................... 17 VYNPS 2.6.1 2.6.2 2.6.3 2.6.4 2.6.5 Objectives................................................. 1 7 Monitoring Network......................................... 18 2.6.2.1 2.6.2.2 2.6.2.3 2.6.2.4 Direct Radiation............................... 18 Airborne....................................... 19 Waterborne..................................... 19 Ingestion...................................... 20 Land Use Census............................................ 20 Emergency Surveillance..................................... 20 Reports.................................................... 21 DSAR Revision 2 2.0-2 of 21

BVY 20-025 /Enclosure/ Page 20 of 118 STATION SITE AND ENVIRONS Figure No.

2.2-2 2.2-3 VYNPS Reference Drawing No.

LIST OF FIGURES Title Location Map -

10-Mile Radius Location Map -

25-Mile Radius DSAR Revision 2 2.0-3 of 21

BVY 20-025 /Enclosure/ Page 21 of 118 2.1

SUMMARY

DESCRIPTION HISTORICAL INFORMATION BELOW NOT REQUIRED TO BE REVISED.

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 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.

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 Northstar Vermont Yankee, LLC and a narrow strip of land between the Connecticut River and the east boundary of the VYNPS property to which Northstar 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° 47' north latitude and 72° 31' west longitude.

Figures 2.2-2 and 2.2-3 locate the site.

The site plot plan, exclusion area boundary and site area boundaries for both gaseous and liquid effluents can be found on site on Drawing 5920-6245.

VYNPS DSAR Revision 2 2.0-4 of 21

BVY 20-025 I Enclosure / Page 22 of 118 2.2.2 Population HISTORICAL INFORMATION BELOW NOT REQUIRED TO BE REVISED.

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.

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.

The total resident population within 50 miles for 2000 is estimated to be about 1,467,343.

Based on this region's projected growth rate of 4% over the next 10 years, the estimated SO-mile population for the year 2010 is 1,526,037..

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.

2.2.3 Land Use HISTORICAL INFORMATION BELOW NOT REQUIRED TO BE REVISED.

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 (Governor Hunt Road) 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 and is approximately 0.5 miles west of the plant at its closest approach.

Via this rail line, various hazardous materials are shipped past the site.

No other significant off-site sources of hazardous materials have been identified within five (5) miles of the site.

VYNPS DSAR Revision 2 2.0-5 of 21

BVY 20-025 / Enclosure / Page 23 of 118 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 expos~res 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.

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BVY 20-025 / Enclosure / Page 24 of 118

1. Controlled Area The controlled area for the VYNPS site consists of a significant portion of the 125-acre property area owned by Northstar 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 Northstar Vermont Yankee, LLC and Northstar Nuclear Decommissioning Company, LLC. Normal 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.

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.

Thus, the overall boundary area for the plant may be found 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 2 2.0-7 of 21

BVY 20-025 / Enclosure I Page 25 of 118

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.

VYNPS DSAR Revision 2 2.0-8 of 21

BVY 20-025 / Enclosure / Page 26 of 118

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

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.

Thus, the responsibilities of the licensee are met from both radiological protection and plant security standpoints.

2.2.5 Conclusions HISTORICAL INFORMATION BELOW NOT REQUIRED TO BE REVISED.

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 2 2.0-9 of 21

BVY 20-025 / Enclosure / Page 27 of 118 ltW 0

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25-Mile Radius Figure 2.2-3 DSAR Revision 2 2.0-11 of 21

BVY 20-025 / Enclosure / Page 29 of 118 2.3 METEOROLOGY 2.3.1 General Vermont Yankee has the capability to receive meteorological data from established local weather services, such as local news and Weather television and radio stations, including the Internet.

VY has reduced the risk of a credible accident now that it has entered decommissioning.

There is no postulated design basis accident or reasonably conceivable beyond design basis event that can result in a radiological release that exceeds EPA Protective Action Guideline Limits beyond the exclusion area boundary.

Therefore, methods for assessing an off-site release are no longer warranted.

Meteorological methods that provide local wind direction and wind speed data are adequate to protect on-site workers and members of the general public that may be on site.

2.4 2.4.1 HYDROLOGY HISTORICAL INFORMATION BELOW NOT REQUIRED TO BE REVISED.

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 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.

VYNPS DSAR Revision 2 2.0-12 of 21

BVY 20-025 / Enclosure / Page 30 of 118 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 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.

The Connecticut River Basin above Vernon, Vermont includes a drainage area of 6,266 square miles.

Based on the Probable Maximum Flood (PMF) studies performed for this site, the PMF stillwater level at the site has been determined to be at an elevation of 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 Standard Project Flood (SPF).

For conservatism, the maximum inflow of 71,000 cfs for this reservoir, which is located about 22 miles upstream from Vernon, was considered to be translated downstream and directly added onto the SPF peak discharge.

This coincident dam failure concurrent with the assumed SPF discharge results in a maximum stillwater elevation at the site of 240.8 feet MSL.

VYNPS DSAR Revision 2 2.0-13 of 21

BVY 20-025 / Enclosure/ Page 31 of 118 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.

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 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months after the postulated failure of the dam.

The time of the peak flood at the site is about 47 hours5.439815e-4 days 0.0131 hours 7.771164e-5 weeks 1.78835e-5 months 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 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

The time from the start of dam failure until the initial rise at the site is 5.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

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 hours0 days 0 hours 0 weeks 0 months .

The time from the start of dam failure until the initial rise at the site is 7.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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

VYNPS DSAR Revision 2 2.0-14 of 21

BVY 20-025 / Enclosure / Page 32 of 118 In summary, the flood levels at the VYNPS site due to upstream dam failures are well below the PMF level at the site.

2.4.4 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.

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.

VYNPS DSAR Revision 2 2.0-15 of 21

BVY 20-025 / Enclosure / Page 33 of 118 2.5 GEOLOGY AND SEISMOLOGY HISTORICAL.INFORMATION BELOW NOT REQUIRED TO BE REVISED.

2.5.1 General 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.

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 Independent 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.

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BVY 20-025 / Enclosure / Page 34 of 118 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 station 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 and decommissioning of the station.

VYNPS DSAR Revision 2 2.0-17 of 21

BVY 20-025 / Enclosure I Page 35 of 118 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), 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.

With the cessation of operations and reduced risk of radioactive releases from the facility, the direct radiation monitoring network is reduced to representative meteorological sectors around the facility with a land border with the state of Vermont.

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 10 locations.

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.

VYNPS DSAR Revision 2 2.0-18 of 21

BVY 20-025 /Enclosure/ Page 36 of 118 2.6.2.2 Airborne Air is sampled for particulates at offsite locations as described in the ODCM (including one control).

The samples are collected by passing the air through a glass fiber filter.

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 monthly for gross beta radioactivity.

These filters are composited for each sampling station and are analyzed quarterly for gamma-emitting radionuclides.

Increased sampling frequency or additional analyses may be required on air particulate filters 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.l 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.

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BVY 20-025 / Enclosure/ Page 37 of 118 2.6.2.4 Ingestion 2.6.2.4.1 Deleted 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 former milk sampling station quarterly, 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 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 2 2.0-20 of 21

BVY 20-025 / Enclosure / Page 38 of 118 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 2 2.0-21 of 21

BVY 20-025 / Enclosure / Page 39 of 118 FACILITY DESIGN AND OPERATION TABLE OF CONTENTS Section Title 3; 1 DESIGN CRITERIA....................................................... 6 3.1.1 3.1.2 3.1. 3 3.1. 4 Conformance with 10 CFR 50 Appendix A General Design Criteria..................................................... 6 Classification of Structures, Systems and Components........ 8 Loading Considerations for Structures, Foundations, Equipment and Systems....................................... 9 3.1.3.1.

3.1.3.2 Seismic Classification......................... 14 Seismic Design................................. 15 References................................................. 18

3. 2 FACILITY STRUCTURES.................................................. 21 3.2.1 3.2.2 3.2.3 3.2.4 3.2.5 3.2.6 VYNPS Reactor Building........................................... 21 3.2.1.1 3.2.1.2 3.2.1.3 Function....................................... 21 Description.................................... 21 Seismic Analysis............................... 23 Turbine Building........................................... 25 3.2.2.1 3.2.2.2 Function....................................... 25 Description.................................... 2 5 Plant Stack................................................ 25 3.2.3.1 3.2.3.2 Description.................................... 25 Seismic Analysis............................... 2 6 Control Room Building...................................... 26 3.2.4.1 3.2.4.2 Description.................................... 2 6 Seismic Analysis............................... 2 6 Circulating Water Intake and Discharge Structures.......... 27 3.2.5.1 3.2.5.2 Intake Structure............................... 27 Discharge and Aerating Structure............... 27 Deleted.................................................... 28 DSAR Revision 2 3.0-1 of 48

BVY 20-025 / Enclosure / Page 40 of 118 3.2.7 3.2.8 Independent Spent Fuel Storage Installation................ 28 3.2.7.1 3.2.7.2 Description.................................... 28 Seismic Analysis............................... 29 References................................................. 30

3. 3 SYSTEMS.............................................................. 33 VYNPS 3.3.1 3.3.2 3.3.3 3.3.4 3.3.5 3.3.6 3.3.7 3.3.8 3.3.9 Fuel Storage and Handling.................................. 34 3.3.1.1 3.3.1.2 Nuclear Fuel................................... 34 Spent Fuel Storage............................. 37 Deleted.................................................... 37 Electrical Power Systems................................... 37 3.3.3.1 3.3.3.2 Deleted........................................ 37 Auxiliary Power System......................... 37 Fire Protection System..................................... 38 3.3.4.1 3.3.4.2 3.3.4.3 3.3.4.4 3.3.4.5 Objective...................................... 38 Design Basis................................... 38 Description.................................... 39 Inspection and Testing......................... 40 References..................................... 41 Heating, Ventilating and Air Conditioning Systems.......... 41 3.3.5.1 3.3.5.2 3.3.5.3 3.3.5.4 Objective...................................... 41 Design Bases................................... 41 Description.................................... 42 Inspection and Testing......................... 45 Deleted.................................................... 45 Process Sampling........................................... 45 3.3.7.1 3.3.7.2 3.3.7.3 Objective...................................... 45 Design Basis................................... 45 Description.................................... 4 6 Deleted.................................................... 47 Deleted.................................................... 48 DSAR Revision 2 3.0-2 of 48

BVY 20-025 / Enclosure / Page 41 of 118 VYNPS 3.3.10 3.3.11 3.3.12 Deleted.................................*.......*.......... 48 Deleted.................................................... 48 Torus-as-CST System........................................ 4 8 3.3.12.1 3.3.12.2 3.3.12.3 Objective...................................... 48 Design Basis................................... 4 8 Description.................................... 4 8 DSAR Revision 2 3.0-3 of 48

BVY 20-025 / Enclosure / Page 42 of 118 Table No.

3.1.1 3.1-2 VYNPS FACILITY DESIGN AND OPERATION LIST OF TABLES Title Allowable Stresses for Class I Structures Safety Margins for Several Critical Portions of Major Class I Structures DSAR Revision 2 3.0-4 of 48

BVY 20-025 / Enclosure / Page 43 of 118 Figure No.

3.2-18 VYNPS Reference Drawing No.

FACILITY DESIGN AND OPERATION LIST OF FIGURES Title Main Stack Geometry DSAR Revision 2 3.0-5 of 48

BVY 20-025 / Enclosure / Page 44 of 118 3.1 DESIGN CRITERIA 3.1.1 Conformance with 10 CFR ~ 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 2 3.0-6 of 48

BVY 20-025 / Enclosure / Page 45 of 11 B 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 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 radiation monitoring system provides 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 accidental criticality for spent fuel.

VYNPS DSAR Revision 2 3.0-7 of 48

BVY 20-025 I Enclosure / Page 46 of 118 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 10 CFR 50.82(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 2 3.0-8 of 48

BVY 20-025 / Enclosure/ Page 47 of 118 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.

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.

VYNPS DSAR Revision 2 3.0-9 of 48

BVY 20-025 I Enclosure I Page 48 of 118 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 H

E E'

w W'

VYNPS is the jet force or pressure on the structure due to rupture of any one pipe.

is the force on the structure due to thermal expansion of pipes.

is the design earthquake load.

is the maximum hypothetical earthquake load.

is the load due to wind.

is the load due to tornado.

DSAR Revision 2 3.0-10 of 48

BVY 20-025 / Enclosure / Page 49 of 118 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:

Loading Consideration A.

Reactor Building and All Other Class I Structures, excluding the primary containment

1.

D+L+R+E

2.

D+L+R+E'

3.

D+L+W'

4.

D+L+W B.

Class I Tanks

1.

2.

3.

D+L+H+E D+L+H+W D+L+H+E' Allowable Stress Normal allowable code stresses are used.

The customary increase in design stresses for the loading combinations considered is not permitted.

Stresses are allowed to approach the yield point for ductile materials, and 0.85 times the ultimate strength for concrete.

Normal allowable code stresses and customary increases in stresses are used for these load combinations.

Normal allowable code stresses are used.

The customary increases in design stresses for the load combinations considered are not permitted.

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 2 3.0-11 of 48

BVY 20-025 / Enclosure / Page 50 of 118 To assure the required properties of concrete poured during cold weather, placing of concrete with ambient temperatures around 15°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 ~ere encouraged to comment on all phases of the program including cold weather concreting.

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:

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.

VYNPS DSAR Revision 2 3.0-12 of 48

BVY 20-025 /Enclosure/ Page 51 of 118 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 refuel 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.

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 tornado loading, metals are allowed to approach their yield point, and concrete,.

its ultimate strength.

VYNPS DSAR Revision 2 3.0-13 of 48

BVY 20-025 / Enclosure / Page 52 of 118 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 10 CFR 100 for a low probability event.

The ISFSI storage pad (comprised of an East and West pad) is classified as Important to Safety Class C (ITS-C) as defined in 10 CFR 72.3.

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 10 CFR 20 for unrestricted areas.

It should be recognized, however, that this failure constitutes an accident and that 10 CFR 50.67 rather than 10 CFR 20 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 10 CFR 50.67 at the site boundary.

3.1.3.1.1 Class I Structures The following is a listing of the Class I structures associated with the storage of irradiated fuel:

Independent Spent Fuel Storage Installation (ISFSI) pad 3.1.3.1.2 Class I Equipment There is no longer any Class I equipment required to be in service for the current facility configuration.

3.1.3.1.3 Class II Structures Administration Building Intake and Discharge Structures All Other Structures, not listed in Paragraph 3.1.3.1.2, that have seismic design requirements.

VYNPS DSAR Revision 2 3.0-14 of 48

BVY 20-025 /Enclosure/ Page 53 of 118 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)

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 refuel 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.

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.

It is noted that Appendix A "Seismic Analysis" of the DSAR Revision 0, contained historical information regarding the seismic design analysis for various SSCs.

VYNPS DSAR Revision 2 3.0-15 of 48

BVY 20-025 / Enclosure / Page 54 of 118 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:

Item Reinforced Concrete Structures Steel Frame Structure Bolted or Riveted Assembly Welded Assembly (equipment and supports)

Vital Piping System Wood Structures with Bolted Joints Percent of Critical Damping 5.0 2.0 2.0 1.0 Various 5.0 Summaries of the seismic analyses for the Reactor Building, Control Room Building, plant stack, intake structure, and deep basin are given in the Facility Structures section under the respective structure.

Detailed analyses for the ISFSI storage pad is contained in References 4 through 8 for the East Storage Pad and References 9 through 14 for the West Storage Pad.

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.

VYNPS DSAR Revision 2 3.0-16 of 48

BVY 20-025 I Enclosure / Page 55 of 118 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.

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.

VYNPS DSAR Revision 2 3.0-17 of 48

BVY 20-025 / Enclosure / Page 56 of 118 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."

9. Calculation VYC-3175, "Determination of Soil Parameters for ISFSI Expansion Concrete Storage Pad."

10.Calculation VYC-3176, "Development of Response Spectra Consistent Time Histories for ISFSI Expansion Concrete Storage Pad."

11.Calculation VYC-3177, "Development of Strain Dependent Soil Properties for ISFSI Expansion Concrete Storage Pad."

12.Calculation VYC-3178, "Soil Structure Interaction Analysis and Cask Stability/Sliding of ISFSI Expansion Concrete Storag~ Pad."

13.Calculation VYC-3179, "Liquefaction Potential for ISFSI Expansion Concrete Storage Pad."

14.Calculation VYC-3181, "Structural Concrete Design for ISFSI Expansion Concrete Storage Pad."

VYNPS DSAR Revision 2 3.0-18 of 48

BVY 20-025 / Enclosure / Page 57 of 118 TABLE 3.1.1 Allowable Stresses for Class I Structures Reinforcing Concrete Concrete Steel Maximum Maximum Maximum Allowable Allowable Loading Conditions Allowable Compressive Shear Stress Stress Stress

1.

Loading as defined f~

1.10 g 0.50 Fy 0.45 without E Wand W'

2.

Loading as defined f~

g 0.667 Fy 0.60 1.467 excluding E, E' and W'

3.

Loading as defined with E' or W' present f~

See Note A 0.85 Seismic load (0.14g)

25% Live Load is considered concurrent with seismic load.

Fy is the minimum yield point of the steel used.

f~ is. the compressive strength of concrete.

Concrete Maximum Structural Allowable Steel Bearing Tension on Stress Net Section 0.25 f~

0.60 Fy 0.333 f~

0.80 Fy See Note A Note A: Stresses permitted to approach but not exceed yield stress of the material.

VYNPS Structural Structural Steel Shear Steel on Gross Section 0.40 Fy 0.53 Fy 0.60 Fy Compression on Gross Section Varies with slenderness ratio Varies with slenderness ratio Varies with slenderness ratio DSAR Revision 2 3.0-19 of 48 Structural Steel Bending 0.66 Fy to 0.60 Fy 0.88 Fy to 0.80 Fy See Note A

BVY 20-025 /Enclosure/ Page 58 of 118 Structure

1. RPV Pedestal

a.

Circumferential Stresses.........

b.

Vertical Stresses................

2. RB Biological Wall Beam

a.

At face..........................

b.

At face..........................

c.

At midspan.......................

d.

At midspan.......................

3. Control Room Footing

a.

Face of column...................

b.

Face of column...................

4. Intake Structure Service Bay

a.

South enclosure..................

NOTE:

Loads as defined in Section 3.1.3.

TABLE 3.1.2 Safety Margins for Several Critical Portions of Major Class I Structures Controlling Loading Condition R

D+L+E+R D+L+E D+L+E' D+L+E D+L+E' D+L+E D+L+E' D+L+W' Allowable Stress in psi Concrete Reinforcing 3400 36,000 1800 20,000 1800 20,000 3400 36,000 1800 20,000 400 36,000 1350 20,000 2550 36,000 2550 36,000 Actual Stresses in psi Concrete Reinforcing 1720 34,000 877 19,644 950 18,100 1080 20,800 1640 16,400 2380 23,600 450 15,600 810 28,200 300 15,000 Safety Margins (Allowable/Actual)

Concrete Reinforcing

1. 97

1. 06 2.06

1. 02

1. 90 1.10 3.14

1. 73 1.16 1.22 1.43 1.53 3.00 1.28 3.15 1.28 8.50 2.40 VYNPS DSAR Revision 2 3.0-20 of 48

BVY 20-025 / Enclosure / Page 59 of 118 3.2 FACILITY STRUCTURES 3.2.1 Reactor Building 3.2.1.1 Function The Reactor Building is no longer required for the safe storage of nuclear fuel.

The following information is historical.

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.

VYNPS DSAR Revision 2 3.0-21 of 48

BVY 20-025 / Enclosure / Page 60 of 118 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.

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:

VYNPS DSAR Revision 2 3.0-22 of 48

BVY 20-025 /Enclosure/ Page 61 of 118

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.

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, VYNPS DSAR Revision 2 3.0-23 of 48

BVY 20-025 /Enclosure/ Page 62 of 118 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.

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.

VYNPS DSAR Revision 2 3.0-24 of 48

BVY 20-025 / Enclosure/ Page 63 of 118 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.

3.2.2 3.2.2.1 Turbine Building Function The Turbine Building SSCs have been abandoned.

3.2.2.2 Description SSCs within the Turbine Building have been abandoned.

Equipment and floor drain sumps are routed to a batch tank.

Tank contents are sampled prior to being transferred, disposed of (via offsite shipments), or discharged to the environs in accordance with applicable permits and regulatory approvals.

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 Inside diameter at top VYNPS 318 ft 7 ft DSAR Revision 2 3.0-25 of 48

BVY 20-025 / Enclosure / Page 64 of 118 Outside diameter at base Thickness at top 27.5 ft 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~.

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.

VYNPS DSAR Revision 2 3.0-26 of 48

BVY 20-025 / Enclosure / Page 65 of 118 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; a fire water pump; 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 a fire water pump, and two radwaste dilution pumps.

Bays are provided with trash rack and stoplog guides, 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.

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 air entrainment, energy dissipation, and warm water dispersion VYNPS structure to provide of discharged water.

DSAR Revision 2 3.0-27 of 48

BVY 20-025 I Enclosure / Page 66 of 118 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 3.2.7 3.2.7.1 Deleted Independent Spent Fuel Storage Installation Description The ISFSI Storage Pad (comprised of an East and West pad) is monolithic reinforced concrete slabs supported by compacted structural fill placed on existing soils.

The two storage pads provide structural support for up to 58 spent fuel storage casks with four extra positions to provide sufficient room to be able to access any individual cask should t_he need arise, and 3 spaces available for storage of Greater-than-Class-C (GTCC) storage casks.

The East Storage Pad can store up to 40 casks arranged in a 5 X 8 array.

The West Storage Pad can store up to 25 casks in a 5 X 5 array.

The spent fuel storage casks are free standing on the pad.

There is temperature monitoring available for each cask if desired.

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.

VYNPS DSAR Revision 2 3.0-28 of 48

BVY 20-025 / Enclosure / Page 67 of 118 3.2.7.2 Seismic Analysis A dynamic analysis of each of the ISFSI storage pads was performed.

This analysis is composed of several parts.

A subsurface investigation was performed to establish bedrock elevations and soil properties beneath each pad (References 4 for the East pad and 10 for the West pad).

The design of the East pad meets the requirements of Revision 3 of Section 3.7.1 of NUREG-0800 which was in effect at the time of its design.

A single set of three artificial time histories for the Design Basis Earthquake was developed for input to the seismic analysis (Reference 5).

The design of the West pad meets the requirements of Revision 4 of Section 3.7.1 of NUREG-0800 which was in affect the time of its design.

Five sets of three artificial time histories for the Design Basis Earthquake were developed for input to the seismic analysis (Reference 12).

These time histories envelope the design response spectra for the site, the North 69° West component of the Taft Earthquake, normalized to 0.14g for the Design Basis Earthquake.

The earthquake(s) 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 profile to the deepest bedrock depth under each pad and one profile to the shallowest bedrock depth under each pad.

This analysis is further described and provided in Reference 6 for the East pad and 13 for the West pad.

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 7 for the East pad and 14 for the West pad.

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 s'eismic event.

VYNPS DSAR Revision 2 3.0-29 of 48

BVY 20-025 / Enclosure / Page 68 of 118 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 for the East pad and 14 for the West pad.

References 9 (East pad) and 16 (West pad) provide 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.

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" 10 Report VY-ROT-14-00005, "Geotechnical Soils Report for DFS-PAD Data Report to Support the Expansion of the Independent Spent Fuel Storage Installation (ISFSI)."

VYNPS DSAR Revision 2 3.0-30 of 48

BVY 20-025 / Enclosure / Page 69 of 118

11. Calculation VYC-3175, "Determination of Soil Parameters for ISFSI Expansion Concrete Storage Pad."

12. Calculation VYC-3176, "Development of Response Spectra Consistent Time Histories for ISFSI Expansion Concrete Storage Pad."

13. Calculation VYC-3177, "Development of Strain Dependent Soil Properties for ISFSI Expansion Concrete Storage Pad."

14. Calculation VYC-3178, "Soil Structure Interaction Analysis and Cask Stability/Sliding of ISFSI Expansion Concrete Storage Pad."

15. Calculation VYC-3179, "Liquefaction Potential for ISFSI Expansion Concrete Storage Pad."

16. Calculation VYC-3181, "Structural Concrete Design for ISFSI Expansion Concrete Storage Pad."

VYNPS DSAR Revision 2 3.0-31 of 48

BW 20-025 / Enclosure / Page 70 of 118

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BVY 20-025 / Enclosure/ Page 71 of 118 3.3 SYSTEMS The following systems have been or are in the process of being abandoned and removed from service.

Abandonment includes, where appropriate, draining piping and tanks, removing electrical power, removal of combustible liquids and placing the abandoned SSC in its lowest energy condition.

VYNPS High Pressure Coolant Injection System Main Steam Heater Drains and Vents Automatic Depressurization System Air Evacuation, Auxiliary Steam, Advanced Off Gas Condensate & Condensate Demineralizer System Containment Air Dilution System Circulating Water & Circulating Water Priming System (includes cooling tower equipment)

Feedwater & Feedwater Controls Systems Hydrogen, Hydrogen Water Chemistry, Nitrogen Supply & Oxygen Injection Systems MG Lube Oil System Reactor Protection and Primary Containment Isolation System River Water Temperature and Toxic Gas Monitoring Systems Reactor Core Isolation Cooling System Recirculation Pumps, MG Sets & Flow Control System Main Turbine Generator, TBCCW, Stator Cooling Seal Oil, Lube Oil, Isophase Bus Cooling Standby Liquid Control System 22K and 345K Volts AC Electrical System Control Rod Drive & Hydraulic Control Unit Systems Core Spray System Nuclear Boiler and Nuclear Boiler Vessel Instrumentation Systems Neutron Monitoring System Reactor Building Closed Cooling Water System Residual Heat Removal & RHR Service Water Systems Radwaste System Process Rad Monitor and Turbine Building Area Rad Monitor Reactor Water Clean-Up System Demineralized Water Transfer and Makeup Demineralizer Systems Post Accident Sampling System Primary Containment/Penetration System DSAR Revision 2 3.0-33 of 48

BVY 20-025 / Enclosure / Page 72 of 118 Primary Containment Atmospheric Control Emergency Diesel Generator and Fuel Oil Systems Normal Fuel Pool Cooling System (FPCS)

Standby Fuel Pool Cooling System (SBFPC)

Fire Protection System, Sprinklers/Detectors (Partial Abandonment) 3.3.1 Potable Water Reconfiguration For SAFSTOR Service Water System 400V DC System Removal Of Low Level Radwaste Site and Other Non-SSC 24V DC RPS Neutron Monitoring System Batteries Vital MG Set MG-2-lA Stack Gas III Radiation Monitor (RM-17-155)

Sentry Lights Fuel Storage and Handling 3.3.1.1 Nuclear Fuel 3.3.1.1.1 Objective Buildings 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 Holtec cask system 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.

VYNPS DSAR Revision 2 3.0-34 of 48

BVY 20-025 / Enclosure / Page 73 of 118 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.

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.

VYNPS DSAR Revision 2 3.0-35 of 48

BVY 20-025 / Enclosure / Page 7 4 of 118 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.

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 lOXlO 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.

VYNPS DSAR Revision 2 3.0-36 of 48

BVY 20-025 /Enclosure/ Page 75 of 118 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.

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 Storage of spent fuel in dry casks at the Independent Spent Fuel Storage Installation facility is licensed in accordance with 10 CFR 72 and is not within the scope of the 10 CFR 50 Defueled Final Safety Analysis Report.

Spent fuel shall not be stored in the spent fuel pool as per Technical Specification 5.2.

The spent fuel pool can be utilized for decommissioning activities including refiling with water for radiation shielding.

3.3.2 3.3.3 3.3.3.1 3.3.3.2 VYNPS Deleted Electrical Power Systems Deleted Auxiliary Power System DSAR Revision 2 3.0-37 of 48

BVY 20-025 / Enclosure / Page 76 of 118 3.3.3.2.1 Objective The objective of the Auxiliary Power System is to provide a reliable power supply to all station loads.

3.3.3.2.2 Design Basis The Station Auxiliary Power System shall have the capacity and capability to supply the required facility loads.

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; dry chemicals; detection and alarm systems; and formerly 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.

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.

VYNPS DSAR Revision 2 3.0-38 of 48

BVY 20-025 / Enclosure / Page 77 of 118 3.3.4.3 Description The Vermont Yankee Fire Protection Program makes use of detection and suppression systems, separation criteria, formerly rated fire barriers and seals, fire stops, procedures and fire watches, standpipe hose connections, and training.

The fire protection program for the permanently defueled state has been developed based on the applicable requirements of 10 CFR 50.48.

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 an electric motor-driven vertical turbine-type pump with a capacity of 2,500 gpm at 125 psi discharge pressure.

The pump is located in the intake structure and discharges 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 pressure in the Fire Main System is maintained at approximately 100 psig by a jockey pump designed to keep the system piping filled when the electric motor-driven pump is not operating.

Operation of the fire pump is controlled from pressure switches in the discharge piping.

The motor-driven pump starts at a predesignated system pressure (typically 85 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 yard piping consists of a 12-inch underground piping loop around the entire station, with valved branches serving 10 fire hydrants.

Valved branches from the piping loop supply water for interior 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.

VYNPS DSAR Revision 2 3.0-39 of 48

BVY 20-025 I Enclosure / Page 78 of 118 A heat traced and insulated fire protection header in the Turbine Building supplies an interior Reactor Building loop.

This loop services two standpipes and fifteen standpipe hose connections.

Selected abandoned water suppression deluge valves have retained the connected heat actuated devices (HADs) for early indication of a fire event in areas which they were previously installed.

These valves are alarmed and signal the control room fire alarm panel.

The Cable Vault is protected by ionization detectors.

The Switchgear Room is protected by ionization detectors coincident with thermal detection.

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, CO2 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.

Formerly rated fire barriers have been retained for the purpose of inhibiting fire spread across adjacent structures.

Features that achieve this goal are selected fire resistant wall assemblies, doors and dampers.

These features are explicitly not credited in the Fire Hazards Analysis as meeting a specific barrier rating, therefore surveillances are not required.

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 pump, water suppression systems, 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.

VYNPS DSAR Revision 2 3.0-40 of 48

BVY 20-025 / Enclosure/ Page 79 of 118 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.

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 leakage through systems that exhaust to the plant stack.

VYNPS DSAR Revision 2 3.0-41 of 48

BVY 20-025 / Enclosure / Page 80 of 118 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:

Outdoor Summer:

90°F dry bulb, 75°F wet bulb Winter:

-12°F dry bulb Indoor Reactor Building:

Maximum:

Minimum:

100°F 65°F 55°F (occupied areas)

(refuel floor)

(occupied areas other than refuel floor)

Control Room and Service Building:

3.3.5.3.1 Maximum:

Minimum:

78°F dry bulb, 50% relative humidity 72°F dry bulb 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.

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 100°F, 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.

VYNPS DSAR Revision 2 3.0-42 of 48

BVY 20-025 / Enclosure / Page 81 of 118 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.

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.

VYNPS DSAR Revision 2 3.0-43 of 48

BVY 20-025 / Enclosure / Page 82 of 118 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.

A purge exhaust fan permits exhausting the drywell or the suppression chamber.

The upstream end of the purge exhaust fa*n 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).

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.

3.3.5.3.2 Deleted 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 a chilled water pump, two air-cooled condensers, piping, and controls.

A separate air-cooled chiller unit was installed to provide equivalent or better primary, or backup, cooling for the Control Room.

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.

VYNPS DSAR Revision 2 3.0-44 of 48

BW 20-025 /Enclosure/ Page 83 of 118 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 lA/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.

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.

3.3.6 Deleted 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 2 3.0-45 of 48

BVY 20-025 / Enclosure / Page 84 of 118 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 as appropriate.

VYNPS DSAR Revision 2 3.0-46 of 48

BVY 20-025 / Enclosure / Page 85 of 118 Description 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 Makeup a)

C9ndensate storage tank Location Outlet pipe Pump discharge Pump discharge Pump discharge Pump discharge Pump discharge Inlet pipe Outlet pipe Outlet pipe Outlet pipe Outlet pipe Pump discharge 3.3.7.3.2 Radwaste Building Sample Panel Purpose 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 Water quality 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 Stack sample Ventilation gases a)

Reactor Building b)

Radwaste Building Location Stack Fan discharge Fan discharge Purpose Particulate and gaseous activity 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.

3.3.8 VYNPS Deleted DSAR Revision 2 3.0-47 of 48

I BW 20-025 / Enclosure / Page 86 of 118 3.3.9 Deleted 3.3.10 Deleted 3.3.11 Deleted 3.3.12 Torus-as-CST System 3.3.12.1 Objective The objective of the Torus-as-CST System is to recirculate water in the Torus.

3.3.12.2 Design Basis The Torus-as-CST System utilizes the Torus for water storage.

The system recirculates water from the Torus and processes it through filters and demineralizers.

3.3.12.3 Description The chemical waste sump and sumps (equipment and floor) in the Radwaste Building and Reactor Building are routed to the Torus.

The Torus-as-CST water treatment system, installed after permanent plant shutdown, recirculates and cleans Torus water.

The System contains parallel paths of pumps, filters and demineralizers.

Suction is taken from the Torus and discharge is back to the Torus.

Water stored in the torus may be disposed offsite or discharged to the environs in accordance with applicable permits and regulatory.approvals.

VYNPS DSAR Revision 2 3.0-48 of 48

BVY 20-025 / Enclosure / Page 87 of 118 RADIOACTIVE WASTE MANAGEMENT TABLE OF CONTENTS Section Title Page 4. 1 SOURCE TERMS.......................................................... 3

4. 2 RADIATION SHIELDING...........................*...............*....... 3 4. 2. 1 Objective.......... *................................... 3 4.2.2 Design Basis.......... *................................ 3 4

2. 3 Description........................................... 4 4.2.3.1 Materials Description........................... 4 4.2.3.2 4.2.3.3 Reactor Building................................ 4 Main Control Room.......................*... *..... 4 4.2.4 Surveillance and Testing............................. 5

4. 3 HEALTH PHYSICS INSTRUMENTATION........................................ 5 4. 3. 1 Objective....................*............... :........ 5 4. 3. 2 Description........*.................................. 5

4. 4 RADIATION PROTECTION.................................................. 7 4.4.1 Health Physics..............................*......... 7

4. 4. 1. 1 Personnel Monitoring Systems...........*........ 7 4.4.1.2 4.4.1.3 4.4.1.4 4.4.1.5 Personnel Protective Equipment.................. 7 Change Area and Shower Facilities............... 8 Access Control.................................. 8 Laboratory Facilities........................... 8

4. 4. 1. 6 Bioassay Program.................... *........... 8 4.4.2 Radioactive Materials Safety Program................. 9

4. 4

2. 1 Facilities and Equipment....................... 1 0 4.4.2.2 Personnel and Procedures...............*....... 10 4.4.2.3 Required Materials........................*.... 10 4.5 LIQUID WASTE MANAGEMENT SYSTEMS...........................*.......... 10 VYNPS 4.5.1 Equipment and Floor Drainage Systems.....*.......... 10 4.5.1.1 Objective...................................... 11 4.5.1.2 4.5.1.3 4.5.1.4 Design Basis................................... 11 Description.................................... 11 Inspection and Testing......................... 15 4.5.2 Liquid Radwaste System.............................. 15 4.5.2.1 Deleted........................................ 15 4.5.2.2 4.5.2.3 4.5.2.4 Design Bases.*................................. 15 Deleted......*....................*............ 15 Evaluation.............................*....... 15 DSAR Revision 2 4.0-1 of 18

BVY 20-025 / Enclosure / Page 88 of 118

4. 6 SOLID WASTE MANAGEMENT............................................... 17 4.7 EFFLUENT RADIOLOGICAL MONITROING AND SAMPLING........................ 17 4.7.1 VYNPS Process Radiation Monitoring Instrumentation........ 17 DSAR Revision 2 4.0-2 of 18

BVY 20-025 / Enclosure / Page 89 of 118 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 4.2.1 RADIATION SHIELDING 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 10 CFR 20.

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.

VYNPS DSAR Revision 2 4.0-3 of 18

BVY 20-025 / Enclosure / Page 90 of 118 Within the controlled access boundary are areas, designated Zone III, which will allow up to 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months per week occupancy and are designed for 6 mrem/hr.

Controlled areas that are designed for 100 mrem/hr allowing occupancy up to 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months per week are designated Zone IV.

Section 6.5 of the Vermont Yankee Permanently Defueled Technical Specifications describes the radiation protection controls for all radiation areas with dose rates exceeding 100 mrem/hr.

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 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 Main Control Room is shielded so that no individual exposure will exceed the limits set forth in Criterion 19, Appendix A of 10 CFR Part 50.

VYNPS DSAR Revision 2 4.0-4 of 18

BVY 20-025 / Enclosure/ Page 91 of 118 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.

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.

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 BF3 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.

VYNPS DSAR Revision 2 4.0-5 of 18

BVY 20-025 / Enclosure / Page 92 of 118 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.

5.

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.

6.

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 facilityi 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.

VYNPS DSAR Revision 2 4.0-6 of 18

BVY 20-025 / Enclosure/ Page 93 of 118

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.

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 4.4.1 RADIATION PROTECTION Health Physics All badged employees of Vermont Yankee with plant access are given training in radiological safety and in the requirements for working in the plant commensurate with their job duties.

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.

VYNPS DSAR Revision 2 4.0-7 of 18

BVY 20-025 / Enclosure / Page 94 of 118 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 faciliti*es 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 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.

VYNPS DSAR Revision 2 4.0-8 of 18

BVY 20-025 / Enclosure / Page 95 of 118

4. 4. 2 Radioactive Materials Safety Program All Vermont Yankee personnel who work in radiologically controlled areas are given training in radiological safety.

Training Program content is specified in appropriate training 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.

Accountability of sources that are exempt from leak testing required by the ODCM, 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.

VYNPS DSAR Revision 2 4.0-9 of 18

BVY 20-025 / Enclosure / Page 96 of 118 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 site training program.

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 4.5.1 VYNPS LIQUID WASTE MANAGEMENT SYSTEMS Equipment and Floor Drainage Systems DSAR Revision 2 4.0-10 of 18

BVY 20-025 / Enclosure / Page 97 of 118 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.

Nonradioactive drainage systems shall be arranged to assure that no infiltration of radioactive waste will occur.

Fluids from radioactive and potentially radioactive drains will be collected, stored, and/or analyzed prior to disposal in accordance with 10 CFR 20.

Nonradioactive equipment and floor drains empty into the Storm Sewer System and then discharge into 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 VYNPS DSAR Revision 2 4.0-11 of 18

BW 20-025 /Enclosure/ Page 98 of 118 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 from the Radwaste and Reactor Buildings are pumped to the Torus.

Liquid wastes from the Turbine Building are pumped manually to tankers for offsite disposal.

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.

The chemical waste sump and equipment/floor drain sumps in the Radwaste Building and Reactor Building are routed to the Torus.

Torus water is processed through the Torus-as-CST System.

Water stored in the Torus may be disposed offsite or discharged to the environs in accordance with applicable permits and regulatory approvals.

Equipment/floor drain sumps in the Turbine Building are routed to a common sump.

Common sump contents are sampled prior to being transferred, disposed of (via offsite shipments), or discharged to the environs in accordance with applicable permits and regulatory approvals.

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.

4.5.1.3.2 Deleted 4.5.1.3.3 Deleted 4.5.1.3.4 Radioactive Liquid Chemical Drainage Systems 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 Radwaste Building sumps.

VYNPS DSAR Revision 2 4.0-12 of 18

BVY 20-025 / Enclosure / Page 99 of 118 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 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.

VYNPS DSAR Revision 2 4.0-13 of 18

BVY 20-025 / Enclosure I Page 100 of 118 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 10 CFR 50, Appendix I.

VYNPS DSAR Revision 2 4.0-14 of 18

BVY 20-025 / Enclosure / Page 101 of 118 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.

4.5.2 Liquid Radwaste System 4.5.2.1 Deleted 4.5.2.2 Design Bases Liquid radwaste shall be contained 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 10 CFR 20.1301.

4.5.2.3 Deleted 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.

HISTORICAL INFORMATION BELOW NOT REQUIRED TO BE REVISED.

The outside storage tanks located within approximately 1.5 foot high concrete dikes, have been drained and abandoned.

VYNPS DSAR Revision 2 4.0-15 of 18

BVY 20-025 / Enclosure / Page 102 of 118 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 10 CFR 20 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 10 CFR 20, 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 controls provided by sample tank liquid analysis.

When in use this process radiation monitor would provide a warning to the appropriate facility personnel that the activity of the discharge had reached a pre-set limit.

When appropriate, facility personnel could then take action to reduce processed liquid flow or terminate flow entirely to assure that the releases do not exceed the limits for which the facility is licensed.

This processed radiation monitor has been abandoned and is no longer in use since discharges from the Liquid Radwaste System to the Connecticut River are no longer expected and the setpoint calculation methodology for this process radiation monitor has been removed from the ODCM.

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.

VYNPS DSAR Revision 2 4.0-16 of 18

BVY 20-025 / Enclosure / Page 103 of 118 4.6 Solid Waste Management As Vermont Yankee is in the process of decommissioning, solid radwaste is being shipped off-site for permanent disposal.

The transportation of the waste is in accordance with applicable regulations.

4.7 Effluent Radiological Monitoring and Sampling 4.7.1 Process Radiation Monitoring Instrumentation Due to the decommissioning process of the facility, the remaining functional monitoring system is the Plant Stack Radiation Monitor which is discussed below.

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 The Plant Stack Radiation Monitoring System provides for particulate sampling so that a determination of the total amounts of particulate activity release is possible.

4. 7. 1. 1. 3 Description The primary channel provides filter media which may be manually analyzed in the plant laboratory by gamma spectroscopy to evaluate long-lived isotopic composition of particulates in plant stack effluents.

VYNPS DSAR Revision 2 4.0-17 of 18

BVY 20-025 / Enclosure / Page 104 of 118 The primary stack monitoring channel consists of four (4.) sampling chambers.

Composites of long-lived particulates 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 flow enters the stack.

The sample train is branched prior to the point of measurement.

Branch I consists of one I-131 charcoal cartridge filter and an 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.

Manual controls for pump motors are located in the station stack Victoreen Room.

All 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 and radio-gas grab samples are provided at elevation 462', several feet downstream of the isokinetic sample probe.

VYNPS DSAR Revision 2 4.0-18 of 18

BVY 20-025 / Enclosure / Page 105 of 118 CONDUCT OF OPERATIONS TABLE OF CONTENTS Section Title 5.1 ORGANIZATION AND RESPONSIBILITY..................................... 2 5

2 TRAINING.............................................................. 2 5.2.1 5.2.2 5.2.3 5.2.4 5.2.5 5.2.6 5.2.7 5.2.8 Program Description (General)............................... 2 General Employee Training................................... 2 5.2.2.1 Access to Plant................................. 2 Fire Brigade Training....................................... 2 ISFSI Security Shift Supervisor (ISSS)

Training.................................................... 2 ISFSI Craft, Technician, and Engineering Training.................*.................................. 3 Training Records............................................ 3 Training Program Approval and Evaluation..................*. 3 Responsibility.............................................. 3 5

3 EMERGENCY PLAN........................................................ 4

5. 4 QUALITY ASSURANCE PROGRAM............................................. 4 5.4.1 5.4.2 5.4.3 5.4.4 Scope....................................................... 4 Responsibilities......................................,..... 4 Implementation.............................................. 4 Management Evaluation....................................... 5

5. 5 REVIEW AND AUDIT OF OPERATIONS........................................ 5 VYNPS 5.5.1 5.5.2 5.5.3 General..................................................... 5 Independent Safety Review................................... 5 Independent Management Assessment........................... 5 DSAR Revision 2 5.0-1 of 5

BVY 20-025 / Enclosure / Page 106 of 118 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 5.2.1 TRAINING 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.

Any required operations, craft, technician, engineering staff, and general employee training requirements are described in ISFSI Training and Qualification document(s) or in-processing procedure(s).

Training associated with the ISFSI is implemented and maintained using a Systems Approach to Training (SAT)~ in accordance with 10 CFR 50.120, 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 related to incipient fire fighting.

5.2.4 ISFSI Security Shift Supervisor (ISSS) Training The initial and continuing training programs for the personnel performing ISFSI operator functions are based on a SAT.

VYNPS DSAR Revision 2 5.0-2 of 5

BVY 20-025 / Enclosure/ Page 107 of 118 5.2.5 ISFSI Craft, Technician, and Engineering Training The initial and continuing training programs for the ISFSI personnel performing the duties of 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 ISFSI training and qualification procedure and in-process training procedure are approved by appropriate facility management, as specified in applicable facility procedures. This ensures that the content and the intent of the training procedures provide the necessary training for personnel associated with the safe storage and handling of irradiated fuel and management of radioactive waste.

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 responsible manager, the ISFSI training representatives are responsible for the conduct and administration of the specified training activities, including:

1.

Initial and continuing training programs for the ISSS/ISFSI Security Officers/Operators.

2.

Fire brigade training.

3.

Initial and continuing training programs for personnel performing the duties of instrumentation and control, maintenance, and engineering.

4.

Initial and continuing training programs for personnel performing the duties of chemistry and radiation protection.

5.

General employee training.

VYNPS DSAR Revision 2 5.0-3 of 5

BVY 20-025 / Enclosure / Page 108 of 118 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 5.4.1 QUALITY ASSURANCE PROGRAM 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 10 CFR 50 Appendix Bin 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 as listed within the VYQAPM. Implementation of this program is assured through Quality Assurance procedures, derived from Quality Assurance policies, goals, and objectives.

VYNPS DSAR Revision 2 5.0-4 of 5

BVY 20-025 / Enclosure / Page 109 of 118 5.4.4 Management Evaluation The independent and safety review function and Independent Management Assessments independently review activities to provide additional assurance that VY is maintained in accordance with the Operating License and applicable regulations that address nuclear safety.

These independent safety review functions are performed in accordance with the Quality Assurance Manual and associated implementing procedures.

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 Independent Safety Review The responsibilities and authorities of the Independent Safety Review are described in an approved Quality Assurance Program Manual implementing procedure.

5.5.3 Independent Management Assessment The Independent Management Assessments (IMA) are periodically performed to monitor overall performance and confirm that activities affecting quality comply with the QAPM and that the QAPM is effectively implemented.

The IMA is performed in accordance with the Quality Assurance Manual and associated implementing procedures.

VYNPS DSAR Revision 2 5.0-5 of 5

BVY 20-025 / Enclosure / Page 110 of 118 SAFETY ANALYSIS TABLE OF CONTENTS Section Title

6.1 INTRODUCTION

........................................................... 3

6. 2 ACCEPTANCE CRITERIA................................................... 4 6.2.1 6.2.2 DBA Acceptance Criteria...................................... 4 Site Event Acceptance Criteria............................... 4

6. 3 DELETED............................................................... 4

6. 4 SITE EVENTS EVALUATED................................................. 5 6.5 VYNPS 6.4.1 High Integrity Container (HIC) Drop Event.................... 5 6.4.1.1 6.4.1.2 6.4.1.3 6.4.1.4 Analytical Methodology.......................... 5 Assumptions..................................... 5 Inputs.......................................... 6 Radiological Consequences/Results............... 7 REFERENCES............................................................ 9 DSAR Revision 2 6.0-1 of 9

BVY 20-025 / Enclosure/ Page 111 of 118 SAFETY ANALYSIS LIST OF TABLES Table No.

Title 6.4.1 HIC Drop Source Term Release Activity VYNPS DSAR Revision 2 6.0-2 of 9

BVY 20-025 /Enclosure/ Page 112 of 118 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.

In addition, Vermont Yankee Technical Specification 5.2 has been revised to prohibit the storage of spent fuel in the spent fuel pool.

As such, there is no longer the possibility of experiencing a fuel handling accident.

This chapter discusses the postulated drop of a high integrity container (HIC) containing radioactive resins.

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, as deemed necessary, during the conduct of decommissioning activities.

VYNPS DSAR Revision 2 6.0-3 of 9

BVY 20-025 / Enclosure / Page 113 of 118 6.2 Acceptance Criteria 6.2.1 OBA Acceptance Criteria A fuel handling accident (FHA) was previously discussed in this section as the only design basis accident (OBA) remai"ning at Vermont Yankee.

However, since all fuel has been transferred to the ISFSI pad, a fuel handling-accident is no longer possible and thus is no longer discussed in this document.

6.2.2 Site Event Acceptance Criteria The HIC drop acceptance criteria are based on 10% of the 10 CFR 100 dose acceptance criteria.

10 CFR 100 Acceptance Criteria (1)

(rem)

EAB and 25 (whole body)

LPZ 300 (thyroid)

(1)

EAB and LPZ dose acceptance criteria 6.3 Deleted VYNPS 10% of 10 CFR 100 Acceptance Criteria (rem) 2.5 (whole body) 30 (thyroid, critical from 10 CFR 100.11 organ)

DSAR Revision 2 6.0-4 of 9

BVY 20-025 /Enclosure/ Page 114 of 118 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 6.5-3) for the air submersion pathway, Regulator Guide 1.109 (Reference 6.5-4) for the inhalation pathway and all nuclides except I-129, ICRP-30 (Reference 6.5-5) for the inhalation pathway and I-129, and Regulator Guide 1.109 (Reference 6.5-6) 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 m2 ) 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 (Fl 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 2 6.0-5 of 9

BVY 20-025 / Enclosure I Page 115 of 118 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 10 CFR 61 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/m3

The breathing rate for the organ dose is 8000 m3/yr (2.537E-04 m3/sec), from RG 1.109.

VYNPS DSAR Revision 2 6.0-6 of 9

BVY 20-025 / Enclosure / Page 116 of 118 6.4.1.4 Radiological Consequences/Results 10% of 10 CFR 100 Dose Calculated Acceptance Criteria Dose (rem)

(rem)

EAB 2.5 rem (whole body) 6.52E-03 1"'

(2 9.59E-03 (b) hours) 16.lE-03 (cl 30 rem (thyroid, also applied to 2.03E-03 (thyroid) the critical organ) 4.58 (lung) 1"' Dose from standing on contaminated ground (2-hr exposure) 10' Dose from cloud passage overhead due to resin fire and aerosol release lcJ Sum of ground plane external plus airborne from cloud VYNPS DSAR Revision 2 6.0-7 of 9

BVY 20-025 I Enclosure / Page 117 of 118 Table 6.4.1 HIC Drop Source Term Release Activity Liner A2 Drop Values LSA Total Release 2

Limit 3 Activity 4 Activity 5 Nuclide 1 1Q1 (mCifgm)

.(g)_

1Q1 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-llOm 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 1-129 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-0.002 0.0001 0.34 0.000017 239/240 Am-241 0.003 0.0001 0.34 0.000017 Cm-242 0.2 0.005 17 0.00085 Cm-0.01 0.0001 0.34 0.000017 243/244 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 ft3 (4.25 m3) liner waste, density of 50 lb/ft3 = 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 -

1-129 is limited by 10CFR61 burial requirements rather than DOT. The class C disposal limit for 1-129, as listed in 10CFR61.55, Table 1, is 0.08 Ci/m3 (or µCi/cc).

VYNPS DSAR Revision 2 6.0-8 of 9

BVY 20-025 / Enclosure / Page 118 of 118 6.5 Refe~ences

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, (applicable revision).

3.

NUREG/CR-1918 (ORNL/NUREG-79), Dose Rate Conversion Factors for External Exposure to Photons and Electrons (August 1981)

4.

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

5.

ICRP-30, Limits for Intake of Radionuclides by Workers,Supplement 1, pg 202

6.

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 2 6.0-9 of 9

BVY 20-026, Defueled Safety Analysis Report, Revision 2

Contents

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SUBJECT:

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Enclosure:

1.0 INTRODUCTION

SUMMARY

1.1 INTRODUCTION

SUMMARY

SUMMARY

6.5 INTRODUCTION

SUMMARY

SUMMARY

1.1 INTRODUCTION

SUMMARY

SUMMARY

SUMMARY

6.1 INTRODUCTION

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BVY 20-026, Defueled Safety Analysis Report, Revision 2

Text

SUBJECT:

Attachment:

Enclosure:

1.0 INTRODUCTION

SUMMARY

1.1 INTRODUCTION

SUMMARY

SUMMARY

6.5 INTRODUCTION

SUMMARY

SUMMARY

1.1 INTRODUCTION

SUMMARY

SUMMARY

SUMMARY

6.1 INTRODUCTION

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