Text

Defueled Safety Analysis Report, Revision 4
	ML24330A206
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	11/26/2024
From:	Reid B; NorthStar Nuclear Decommissioning Company
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	BVY 24-029
	Download: ML24330A206 (1)
	v • d • e

Billy E. Reid, Jr.

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

BVY 24-029 November 25, 2024 ATTN: Document Control Desk, U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Defueled Safety Analysis Report, Revision 4 Vermont Yankee Nuclear Power Station Docket No. 50-271 License No. DPR-28

Dear 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 4 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 4 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 4 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 4.

Pursuant to 10 CFR 50.59(d)(2), NorthStar Nuclear Decommissioning Co., LLC is reporting that for the interval of September 21, 2022 through September 22, 2024, 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 25, 2024.

NorthStar Nuclear Decommissioning Co., LLC Vermont Yankee Nuclear Power Station 320 Governor Hunt Rd.

Vernon, VT 05354 802-451-5354

BVY 24-029 / Page 2 of 2 Sincerely, BER/tbs

Attachment:

Defueled Safety Analysis Report (DSAR) Revision 4 Changes.

Enclosure:

Vermont Yankee Nuclear Power Station Defueled Safety Analysis Report (DSAR), Revision 4.

cc: Commissioner Vermont Department of Public Service 112 State Street - Drawer 20 Montpelier, Vermont 05602-2601 Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Rd, Suite 102 King of Prussia, PA 19406

BVY 24-029 Docket No. 50-271 Attachment Vermont Yankee Nuclear Power Station Defueled Safety Analysis Report (DSAR) Revision 3 Changes

BVY 24-029 / Attachment / Page 1 of 2 Vermont Yankee Defueled Safety Analysis Report (DSAR) Revision 3 Changes Change #

Sections Affected Brief Description of Change DCR 03/04 1.4.1 3.2.7.1 3.2.8 Revise DSAR in support of GTCC Cask placement on ISFSI Pad.

DCR 04/01 1.1 1.2 1.3.1.2 1.3.3.2 1.4.1 2.2.4 Section 3 TOC 3.2.1 3.2.2 3.2.8 3.3 3.3.1.2 4.3.2 4.4.1.1 Admin Changes for Rev 4 (from Rev 3) consistent with building demolition.

DCR 04/02 Section 3 TOC 3.2.5 3.3 Update DSAR to reflect abandonment of Fire Suppression Water System.

DCR 04/03 TOC Section 4 TOC 4.5 4.5.1 4.5.1.1 4.5.1.2 4.5.1.3 Update DSAR Section 4.5 (4.5.1) to reflect the abandonment of Equipment and Floor Drainage Systems.

DCR 04/04 Section 3 TOC 3.1.1 3.1.2 3.1.3 3.1.3.1 Delete discussion of the stations conformance to the 10 CFR 50 Appendix A General Design Criteria; Section 3.1.1, 3.1.2 and 3.1.3

BVY 24-029 Docket No. 50-271 Enclosure Vermont Yankee Nuclear Power Station Defueled Safety Analysis Report (DSAR), Revision 4 (61 pages excluding this cover sheet)

VYNPS DSAR Revision 4 DEFUELED SAFETY ANALYSIS REPORT VERMONT YANKEE NUCLEAR POWER STATION BVY 24-029 / Enclosure / Page 1 of 61

DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS VYNPS DSAR Revision 4 TOC-1 of 2 SECTION 1 UFSAR, REV 17, TOC

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 4.6 SOLID WASTE MANAGEMENT 4.7 EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING BVY 24-029 / Enclosure / Page 2 of 61

DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS (Continued)

VYNPS DSAR Revision 4 TOC-2 of 2 SECTION 5 5.0 CONDUCT OF OPERATIONS 5.1 ORGANIZATION AND RESPONSIBILITY 5.2 TRAINING 5.3 EMERGENCY PLAN 5.4 QUALITY ASSURANCE PROGRAM 5.5 REVIEW AND AUDIT OF OPERATIONS SECTION 6 6.0 SAFETY ANALYSIS

6.1 INTRODUCTION

6.2 ACCEPTANCE CRITERIA 6.3 DELETED 6.4 SITE EVENTS EVALUATED

6.5 REFERENCES

BVY 24-029 / Enclosure / Page 3 of 61

VYNPS DSAR Revision 4 1.0-1 of 10 INTRODUCTION AND

SUMMARY

TABLE OF CONTENTS Section Title Page

1.1 INTRODUCTION

.......................................................... 2 1.2 DESIGN CRITERIA....................................................... 4 1.3 FACILITY DESCRIPTION.................................................. 6 1.3.1 General...................................................... 6 1.3.1.1 Site and Environs............................... 6 1.3.1.2 Facility Arrangement............................ 8 1.3.2 Fuel Storage and Handling.................................... 8 1.3.2.1 Nuclear Fuel.................................... 8 1.3.3 Radioactive Waste Management................................. 8 1.3.3.1 Equipment and Floor Drainage Systems............ 8 1.3.3.2 Liquid Radwaste System.......................... 8 1.3.3.3 Solid Radwaste System........................... 9 1.3.4 Deleted...................................................... 9 1.3.5 Auxiliary Systems............................................ 9 1.3.5.1 Deleted......................................... 9 1.3.5.2 Deleted......................................... 9 1.3.5.3 Fire Protection System.......................... 9 1.3.5.4 Heating, Ventilating, and Air Conditioning Systems............................ 9 1.3.6 Deleted...................................................... 9 1.3.7 Deleted...................................................... 9 1.3.8 Shielding, Access Control, and Radiation Protection Procedures........................................ 9 1.4

SUMMARY

OF RADIATION EFFECTS......................................... 10 1.4.1 Fuel Storage and Handling and Waste Management.............. 10 1.4.2 Accidents and Events........................................ 10 1.5 GENERAL CONCLUSIONS.................................................. 10 BVY 24-029 / Enclosure / Page 4 of 61

VYNPS DSAR Revision 4 1.0-2 of 10

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, fabricated, 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, BVY 24-029 / Enclosure / Page 5 of 61

VYNPS DSAR Revision 4 1.0-3 of 10 2014 which was the date on which operation ceased at VYNPS. ENO also certified that the fuel has been permanently removed from the VYNPS reactor vessel and placed in the spent fuel pool. ENO acknowledged that, following docketing, the VYNPS license no longer authorized operation of the reactor or emplacement or retention of fuel into the reactor vessel.

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 NDC).

The Defueled Safety Analysis Report (DSAR) Revision 0 was derived from Revision 26 of the VYNPS Updated Final Safety Analysis Report (UFSAR) and was developed 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 0 was issued). DSAR Revision 4 continues to reflect the transfer of all spent nuclear fuel into dry cask storage within an Independent Spent Fuel Storage Installation (ISFSI) and removes descriptions of SSCs relied upon for the storage of spent nuclear fuel in the SFP. The DSAR serves the same function during 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.

Per 10 CFR 50.51(b), the Vermont Yankee operating license continues to remain in effect until the Nuclear Regulatory Commission terminates the license.

BVY 24-029 / Enclosure / Page 6 of 61

VYNPS DSAR Revision 4 1.0-4 of 10 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.

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.

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.

BVY 24-029 / Enclosure / Page 7 of 61

VYNPS DSAR Revision 4 1.0-5 of 10 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.

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.

BVY 24-029 / Enclosure / Page 8 of 61

VYNPS DSAR Revision 4 1.0-6 of 10 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 4247' north, 7231' 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.

BVY 24-029 / Enclosure / Page 9 of 61

VYNPS DSAR Revision 4 1.0-7 of 10 1.3.1.1.6 Geology The major structures at the site are supported by bedrock. Compression tests indicated minimum failure of the bedrock to be 16,000 psi (1,152 tons per square foot). An allowable bearing pressure has been established at 50 tons per square foot; however, actual loadings do not exceed 20 tons per square foot.

1.3.1.1.7 Seismology Based on a three-fold seismic evaluation, the site was found to be relatively quiescent from a seismic standpoint. From these studies the design earthquake has been established at 0.07g horizontal ground acceleration and the maximum hypothetical earthquake at 0.14g horizontal ground acceleration. The seismic evaluation consisted of a review of historical data from the New England area, an analysis of instrument and historical records for the Vermont area, and a study of earthquake intensity attenuation with distance for the northeast United States.

1.3.1.1.8 Hydrology The facility is on the Connecticut River in Vernon, Vermont, some 138.3 miles from the river mouth. The river in the vicinity of the facility is comprised of a series of ponds formed by dams constructed for the generation of hydroelectric power. All local surface streams drain to the Connecticut River, and the site is in the direct path of natural drainage to the east of the local watershed. In the vicinity of the site there is also a considerable amount of groundwater which several municipalities utilize as one source of water supply.

1.3.1.1.9 Regional and Site Meteorology The general climatic regime is that of a continental type with some modification from the maritime climate which prevails nearer the coast. For the one-year period between August 1967 and July 1968, temperature inversions occurred 39% of the total time. Seasonal inversion frequencies ranged between 36% and 42%. Wind distribution is biased in the direction of the river due to the channeling effect of the valley.

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

BVY 24-029 / Enclosure / Page 10 of 61

VYNPS DSAR Revision 4 1.0-8 of 10 1.3.1.2 Facility Arrangement The facility arrangement is shown on Drawing 5920-6245. The principal remaining structure of the station is Spent Fuel Storage Installation (ISFSI) storage pad.

1.3.2 Fuel Storage and Handling 1.3.2.1 Nuclear Fuel Nuclear fuel previously used for power generation consists of slightly enriched uranium dioxide pellets contained in sealed Zircaloy tubes. These fuel rods are assembled into individual fuel assemblies. On January 12, 2015, VYNPS certified to the NRC that all nuclear fuel had been permanently removed from the reactor vessel and placed in the spent fuel pool. 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 As part of the decommissioning of the facility, the equipment and floor drain systems have been abandoned.

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

1.3.3.2 Liquid Radwaste System The Liquid Radwaste System is no longer in service. Liquids may be disposed of offsite or discharged to the environs in accordance with applicable permits and regulatory approvals.

BVY 24-029 / Enclosure / Page 11 of 61

VYNPS DSAR Revision 4 1.0-9 of 10 1.3.3.3 Solid Radwaste System Solid radioactive wastes are collected, processed, and packaged for off-site burial. 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 sent to a disposal site for burial.

1.3.4 Deleted 1.3.5 Auxiliary Systems 1.3.5.1 Deleted 1.3.5.2 Deleted 1.3.5.3 Fire Protection System The Fire Protection Program meets the objectives of 10 CFR 50.48(f). For the latest revision please reference the Plant Fire Protection Program and the ISFSI Fire Protection Program.

1.3.5.4 Heating, Ventilating, and Air Conditioning Systems The original Heating, Ventilating, and Air Conditioning (HVAC) Systems have been abandoned. Monitored and filtered temporary exhaust from the reactor building remain such that the gaseous or particulate contaminants are effectively prevented from entering the cleaner zones.

1.3.6 Deleted 1.3.7 Deleted 1.3.8 Shielding, Access Control, and Radiation Protection Procedures 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 BVY 24-029 / Enclosure / Page 12 of 61

VYNPS DSAR Revision 4 1.0-10 of 10 limit the potential spread of contamination from various areas, particularly areas where maintenance occurs.

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 permitted by 10 CFR 20. Both effects are only a small fraction of the effect of natural background radiation.

For an ISFSI controlled area, 10 CFR 72.104(a) defines the dose that any individual located beyond the controlled area may receive from normal operations and anticipated occurrences associated with the ISFSI.

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.

BVY 24-029 / Enclosure / Page 13 of 61

VYNPS DSAR Revision 4 2.0-1 of 15 SITE AND ENVIRONS TABLE OF CONTENTS Section Title Page 2.1

SUMMARY

DESCRIPTION.................................................... 3 2.2 SITE DESCRIPTION....................................................... 3 2.2.1 Location and Area............................................... 3 2.2.2 Population..................................................... 4 2.2.3 Land Use....................................................... 4 2.2.4 Site Area Boundaries, Exclusion Area, and Low Population Zone............................................. 5 2.2.5 Conclusions.................................................... 8 2.3 METEOROLOGY.......................................................... 11 2.3.1 General....................................................... 11 2.4 HYDROLOGY............................................................ 11 2.4.1 General....................................................... 11 2.4.2 Site Area..................................................... 11 2.4.3 Floods........................................................ 12 2.4.4 Conclusions................................................... 14 2.5 GEOLOGY AND SEISMOLOGY................................................ 14 2.5.1 General....................................................... 14 2.6 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM.......................... 15 2.6.1 General....................................................... 15 2.6.2 Off-Site Dose Calculation Manual............................... 15 2.6.3 Summary of Estimated Doses.................................... 15 BVY 24-029 / Enclosure / Page 14 of 61

VYNPS DSAR Revision 4 2.0-2 of 15 STATION SITE AND ENVIRONS LIST OF FIGURES Reference Figure No.

Drawing No.

Title 2.2-2 Location Map Mile Radius 2.2-3 Location Map Mile Radius BVY 24-029 / Enclosure / Page 15 of 61

VYNPS DSAR Revision 4 2.0-3 of 15 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 42o 47' north latitude and 72o 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.

BVY 24-029 / Enclosure / Page 16 of 61

VYNPS DSAR Revision 4 2.0-4 of 15 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 50-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.

BVY 24-029 / Enclosure / Page 17 of 61

VYNPS DSAR Revision 4 2.0-5 of 15 2.2.4 Site Area Boundaries, Exclusion Area, and Low Population Zone As defined in 10 CFR 20 and 10 CFR 100, the terms "unrestricted area,"

"controlled area," "restricted area," "exclusion area," and "low population zone" each refer to a specific area about the site as a result of applying different radiological health constraints. The "unrestricted area" refers to all areas beyond the site's outer security fence access to which is neither limited nor controlled by the licensee. The "controlled area" refers to all plant areas inside the site boundary, but outside of any restricted area, access to which is limited by the licensee for any reason. Access to the controlled area can be limited to minimize exposures to members of the public from routine radioactive releases from the plant and fixed radiation sources.

"Restricted area" refers to the inner most areas of the plant site and facilities, access to which is limited by the licensee for the purpose of protecting occupationally exposed individuals against undue risks from radiation and radioactive materials. Exclusion area means that area surrounding the reactor, as measured from the reactor center line, in which the reactor licensee has the authority to determine all activities including exclusion or removal of personnel and property from the area. This area may be traversed by a highway, railroad, or waterway, provided those are not so close to the facility as to interfere with normal operations of the facility and provided appropriate and effective arrangements are made to control traffic on the highway, railroad, or waterway, in case of an emergency, to protect the public health and safety. The exclusion area also includes part of the adjacent waterway (Connecticut River) extending across to the opposite shoreline.

Finally, the low population zone is delineated by an area about the plant which includes residential, farming, industrial, etc., activities to some extent, but is not so large or populated to prevent orderly, effective radiological control or evacuation in the event of an accident of an environmentally significant nature.

Thus, these areas and zones are delineated for different purposes and vary in the degree of control that the licensee can exercise from a radiation protection standpoint. The following discussion presents an analysis of each area in relation to the plant and its operations.

BVY 24-029 / Enclosure / Page 18 of 61

VYNPS DSAR Revision 4 2.0-6 of 15

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, including liquid effluents, to levels as-low-as-reasonably achievable. 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.

BVY 24-029 / Enclosure / Page 19 of 61

VYNPS DSAR Revision 4 2.0-7 of 15

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.

BVY 24-029 / Enclosure / Page 20 of 61

VYNPS DSAR Revision 4 2.0-8 of 15

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.

BVY 24-029 / Enclosure / Page 21 of 61

Radius Location Map Mile VYNPS DSAR Revision 4 2.0-9 of 15 Figure 2.2-2 Vermont Yankee Defueled Safety Analysis Report

'ltW 0

5 I

I I

~~----,

10 15 Mile J.

i I

BVY 24-029 / Enclosure / Page 22 of 61

Vermont Yankee Defueled Safety Analysis Report Location Map Mile Radius Figure 2.2-3 VYNPS DSAR Revision 4 2.0-10 of 15 0

5 10 15 Mile BVY 24-029 / Enclosure / Page 23 of 61

VYNPS DSAR Revision 4 2.0-11 of 15 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 HYDROLOGY HISTORICAL INFORMATION BELOW NOT REQUIRED TO BE REVISED.

2.4.1 General The site is at mile 138.3 above the mouth of the Connecticut River, located on the west bank of the river, on the pond formed by the Vernon Dam and Hydroelectric Station, licensed by the Federal Energy Regulatory Commission as Project No. 1094. The site is about 3,500 feet upstream from the Vernon Hydroelectric Station, on the same side of the river. The Vernon Hydroelectric Station is the furthest downstream of a series of six hydroelectric projects totaling over 456,000 kW on the river. Storage reservoirs, whose contents total over 330,000 acre-feet, are also usable for power generation.

Three of the dams, at 32, 75, and 132 miles above the site, are relatively low structures developing heads of from 29 to 62 feet, with small amounts of pondage. The large storage reservoirs are from 150 to 260 miles upstream from Vernon.

2.4.2 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.

BVY 24-029 / Enclosure / Page 24 of 61

VYNPS DSAR Revision 4 2.0-12 of 15 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.

BVY 24-029 / Enclosure / Page 25 of 61

VYNPS DSAR Revision 4 2.0-13 of 15 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.

BVY 24-029 / Enclosure / Page 26 of 61

VYNPS DSAR Revision 4 2.0-14 of 15 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.

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. 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.

BVY 24-029 / Enclosure / Page 27 of 61

VYNPS DSAR Revision 4 2.0-15 of 15 2.6 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 2.6.1 General A program is required to monitor the radiation and radionuclides in the environs of the plant. The program provides:

2.6.1.1 Representative measurements of radioactivity in the highest potential exposure pathways, and 2.6.1.2 Verification of the accuracy of the effluent monitoring program and modeling of environmental pathways.

2.6.1.3 The program is contained in the ODCM, conforms to the guidance of 10 CFR 50, Appendix I, and includes the following:

a. Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM.

b. A Land Use Census to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the monitoring are made if required by the results of this census, and

c. Participation in an Inter-laboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

2.6.2 Off-Site Dose Calculation Manual The ODCM provides the information and methodologies used to evaluate the impact of radiological liquid and gaseous effluent discharged from the plant.

The ODCM is used to demonstrate that the plant complies with the requirements of 40 CFR 190, 10 CFR 20, and the dose guidelines of 10 CFR 50, Appendix I.

2.6.3 Summary of Estimated Doses Estimates of the maximum whole body dose and the maximum organ dose to an individual member of the public at or beyond the site boundary (i.e., in the unrestricted area) that would be received as a result of the release of liquid and gaseous effluents, and direct radiation (e.g., ISFSI) dose estimates, can be found in the Annual Radioactive Effluent Release Reports submitted to the NRC.

BVY 24-029 / Enclosure / Page 28 of 61

VYNPS DSAR Revision 4 3.0-1 of 10 FACILITY DESIGN AND OPERATION TABLE OF CONTENTS Section Title Page 3.1 DESIGN CRITERIA....................................................... 2 3.1.1 Deleted................................................ 2 3.1.2 Deleted.....................................................2 3.1.3 Loading Considerations for Structures, Foundations, Equipment and Systems...................................... 2 3.1.3.1.

Seismic Classification......................... 2 3.1.4 References................................................. 3 3.2 FACILITY STRUCTURES.................................................. 4 3.2.1 Deleted.................................................... 4 3.2.2 Deleted.................................................... 4 3.2.3 Deleted................................................... 4 3.2.4 Deleted................................................... 4 3.2.5 Deleted................................................... 4 3.2.6 Deleted.................................................... 4 3.2.7 Independent Spent Fuel Storage Installation................ 4 3.2.7.1 Description.................................... 4 3.2.7.2 Seismic Analysis............................... 4 3.2.8 References

................................................. 6 3.3 SYSTEMS 7

3.3.1 Fuel Storage and Handling.................................. 7 3.3.1.1 Nuclear Fuel................................... 7 3.3.1.2 Spent Fuel Storage............................. 10 BVY 24-029 / Enclosure / Page 29 of 61

VYNPS DSAR Revision 4 3.0-2 of 10 3.1 DESIGN CRITERIA 3.1.1 Deleted 3.1.2 Deleted 3.1.3 Loading Considerations for Structures, Foundations, Equipment and Systems The Reactor Building and all other Class I structures except the ISFSI storage pad were founded on firm bedrock. The ISFSI storage pad is founded on engineered fill placed on existing soil.

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. There is no longer any Class I equipment required to be in service for the current facility configuration.

Class II - There is no longer any Class II structures or equipment required to be in service for the current facility configuration.

BVY 24-029 / Enclosure / Page 30 of 61

VYNPS DSAR Revision 4 3.0-3 of 10 3.1.4 References

1. Deleted.

2. Deleted.

3. Deleted.

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 Storage 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."

BVY 24-029 / Enclosure / Page 31 of 61

VYNPS DSAR Revision 4 3.0-4 of 10 3.2 FACILITY STRUCTURES 3.2.1 Deleted 3.2.2 Deleted 3.2.3 Deleted 3.2.4 Deleted 3.2.5 Deleted 3.2.6 Deleted 3.2.7 Independent Spent Fuel Storage Installation 3.2.7.1 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 the 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. There are 58 casks containing spent fuel located on the ISFSI storage pads. 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.

A single cask containing GTCC material removed from the reactor vessel is located on the ISFSI pad. The activity to locate a cask containing GTCC material was performed under DCR-21-01-0 (Reference 17).

The spent fuel cask manufacturers 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.

BVY 24-029 / Enclosure / Page 32 of 61

VYNPS DSAR Revision 4 3.0-5 of 10 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 69º 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 seismic event.

The sliding analysis evaluated coefficients of friction ranging from 0.0 to stimulate icing conditions on the pad up to a maximum of 0.8. The results of the analysis show that the maximum horizontal displacements of the casks for any condition are much smaller than half the free distance between the casks and much less than the distance between the edge of the external casks and the edge of the pad. This analysis also shows that the casks are stable and remain upright. The sliding analysis is provided in Reference 8 for the East pad and 14 for the West BVY 24-029 / Enclosure / Page 33 of 61

VYNPS DSAR Revision 4 3.0-6 of 10 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. Deleted

2. Deleted

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)."

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 BVY 24-029 / Enclosure / Page 34 of 61

VYNPS DSAR Revision 4 3.0-7 of 10 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."

17. Design Change Record DCR-21-01-0, Place a Holtec HI-SAFE Cask Containing GTCC Material in the ISFSI.

3.3 SYSTEMS 3.3.1 Fuel Storage and Handling 3.3.1.1 Nuclear Fuel 3.3.1.1.1 Objective The nuclear fuel provides a high integrity assembly containing fissionable material which could be arranged in a critical array. The assembly efficiently transfers decay heat to the 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.

BVY 24-029 / Enclosure / Page 35 of 61

VYNPS DSAR Revision 4 3.0-8 of 10 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.

BVY 24-029 / Enclosure / Page 36 of 61

VYNPS DSAR Revision 4 3.0-9 of 10 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 10X10 array with two central water rods, as well as both short and long partial length rods. Some of the design features include the following:

Improved part-length rod configuration for improved Cold Shutdown Margin (CSDM) and efficiency.

Modified fuel rod clad thickness to diameter ratio (T/D) with increased uranium mass for increased bundle energy.

Modified channel that interacts with the LTP to control leakage flow while eliminating finger springs for ease of channeling operations.

Improved Inconel X-750 grid type spacer with Flow Wings for increased margin to Boiling Transition and reduced pressure drop.

BVY 24-029 / Enclosure / Page 37 of 61

VYNPS DSAR Revision 4 3.0-10 of 10 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.

BVY 24-029 / Enclosure / Page 38 of 61

VYNPS DSAR Revision 4 4.0-1 of 9 RADIOACTIVE WASTE MANAGEMENT TABLE OF CONTENTS Section Title Page 4.1 SOURCE TERMS................................................... 2 4.2 RADIATION SHIELDING............................................. 2 4.2.1 Objective...................................... 2 4.2.2 Design Basis.................................... 2 4.3 HEALTH PHYSICS INSTRUMENTATION.................................. 3 4.3.1 Objective...................................... 3 4.3.2 Description.................................... 3 4.4 RADIATION PROTECTION............................................ 5 4.4.1 Health Physics.................................. 5 4.4.1.1 Personnel Monitoring Systems............... 5 4.4.1.2 Personnel Protective Equipment............. 5 4.4.1.3 Change Area and Shower Facilities.......... 5 4.4.1.4 Access Control............................ 6 4.4.1.5 Laboratory Facilities...................... 6 4.4.1.6 Bioassay Program.......................... 6 4.4.2 Radioactive Materials Safety Program............ 7 4.4.2.1 Facilities and Equipment................... 8 4.4.2.2 Personnel and Procedures................... 8 4.4.2.3 Required Materials........................ 8 4.5 LIQUID WASTE MANAGEMENT........................................9 4.5.1 Nonradioactive Water Drainage System............9 4.5.2 Liquid Radioactive Wastes...................... 9 4.6 SOLID WASTE MANAGEMENT.......................................... 9 4.7 EFFLUENT RADIOLOGICAL MONITROING AND SAMPLING.................... 9 4.7.1 Process Radiation Monitoring Instrumentation.... 9 BVY 24-029 / Enclosure / Page 39 of 61

VYNPS DSAR Revision 4 4.0-2 of 9 4.1 SOURCE TERMS In the permanently defueled condition VYNPS will no longer produce fission, corrosion, or activation products from operation. The radioactive inventory that remains is primarily attributable to activated reactor components and structural materials and residual radioactivity. The accumulation of small amounts of solid waste may easily be controlled. Any future planned liquid effluent releases will be evaluated prior to release, and appropriate controls will be established. The Offsite Dose Calculation Manual ensures that VYNPS complies with 10 CFR 50, Appendix I.

4.2 RADIATION SHIELDING 4.2.1 Objective Radiation shielding is utilized as appropriate to limit 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.

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.

BVY 24-029 / Enclosure / Page 40 of 61

VYNPS DSAR Revision 4 4.0-3 of 9 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.

4.

Beta-Gamma dose rate meters are used for determining stay times for radiation workers and for posting radiation area warning signs.

High range beta-gamma meters provide dose rate information during any event involving high levels of radiation. Neutron dose rate meters respond to and provide an indication of the entire spectrum of neutrons encountered around the ISFSI.

BVY 24-029 / Enclosure / Page 41 of 61

VYNPS DSAR Revision 4 4.0-4 of 9

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 facility, and are used by facility personnel to detect surface contamination on clothing, skin, etc.

3.

Dosimeter readers, which contain the equipment for measuring the dose received by personal dosimeters. These instruments are located in an off-site dosimeter processing facility under contract with Vermont Yankee.

4.

Multi-channel gamma spectrometer, which consists of a NaI, or HpGe crystal, and analyzer circuits necessary for the identification of individual isotopes by gamma ray energy.

5.

Laboratory alpha and beta-gamma counters, which are used for measuring low levels of radioactivity in specially prepared samples such as smears, air particulate sample filters, etc.

6.

Body-burden counters, which are used to assess internal contamination from both natural sources and from inhaled/absorbed radioactive gases or particulates.

BVY 24-029 / Enclosure / Page 42 of 61

VYNPS DSAR Revision 4 4.0-5 of 9 4.4 RADIATION PROTECTION 4.4.1 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 for each individual issued personnel monitoring equipment.

4.4.1.2 Personnel Protective Equipment Special protective clothing and respiratory equipment are furnished and worn as necessary to protect personnel from radioactive contamination.

4.4.1.3 Change Area and Shower Facilities A change area is provided where personnel may obtain clean protective clothing required for station work. Temporary change areas are provided when required.

Decontamination shower facilities are maintained on-site to assist in timely personnel decontamination. Monitoring equipment is used to assess the effectiveness of personnel decontamination efforts.

BVY 24-029 / Enclosure / Page 43 of 61

VYNPS DSAR Revision 4 4.0-6 of 9 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.

BVY 24-029 / Enclosure / Page 44 of 61

VYNPS DSAR Revision 4 4.0-7 of 9 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 QAPM, 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.

BVY 24-029 / Enclosure / Page 45 of 61

VYNPS DSAR Revision 4 4.0-8 of 9 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 Sealed sources for calibration of radioactive monitoring equipment 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.

BVY 24-029 / Enclosure / Page 46 of 61

VYNPS DSAR Revision 4 4.0-9 of 9 4.5 LIQUID WASTE MANAGEMENT 4.5.1 Nonradioactive Water Drainage System Surface run-off from within the Protected Area may carry 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.

4.5.2 Liquid Radioactive Wastes Liquid radioactive wastes 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.

4.6 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 Reactor Building ventilation exhaust.

BVY 24-029 / Enclosure / Page 47 of 61

VYNPS DSAR Revision 4 5.0-1 of 5 CONDUCT OF OPERATIONS TABLE OF CONTENTS Section Title Page 5.1 ORGANIZATION AND RESPONSIBILITY....................................... 2 5.2 TRAINING............................................................. 2 5.2.1 Program Description (General)............................... 2 5.2.2 General Employee Training................................... 2 5.2.2.1 Access to Plant................................. 2 5.2.3 Deleted.....................................................2 5.2.4 Deleted..................................................... 2 5.2.5 Deleted..................................................... 3 5.2.6 Training Records............................................ 3 5.2.7 Training Program Approval and Evaluation.................... 3 5.2.8 Responsibility.............................................. 3 5.3 EMERGENCY PLAN........................................................ 4 5.4 QUALITY ASSURANCE PROGRAM............................................. 4 5.4.1 Scope....................................................... 4 5.4.2 Responsibilities............................................ 4 5.4.3 Implementation.............................................. 4 5.4.4 Management Evaluation.......................................

5 5.5 REVIEW AND AUDIT OF OPERATIONS........................................ 5 5.5.1 General..................................................... 5 5.5.2 Independent Safety Review................................... 5 5.5.3 Independent Management Assessment............................ 5 BVY 24-029 / Enclosure / Page 48 of 61

VYNPS DSAR Revision 4 5.0-2 of 5 5.1 ORGANIZATION AND RESPONSIBILITY The Vermont Yankee Nuclear Power Station organization, including the responsibilities and duties of staff personnel, are detailed in the Vermont Yankee Quality Assurance Program Manual.

5.2 TRAINING 5.2.1 Program Description (General)

The objective of the Training Program is to provide qualified personnel to operate and maintain the permanently defueled facility in a safe manner, including the storage and handling of irradiated fuel. Any required operations, craft, technician, engineering staff, and general employee training requirements are described in ISFSI Training and Qualification procedures or in-processing procedure(s).

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 Deleted 5.2.4 Deleted BVY 24-029 / Enclosure / Page 49 of 61

VYNPS DSAR Revision 4 5.0-3 of 5 5.2.5 Deleted 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.

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.

5.2.8 Responsibility The ISFSI training representatives are responsible for the conduct and administration of the specified training activities.

BVY 24-029 / Enclosure / Page 50 of 61

VYNPS DSAR Revision 4 5.0-4 of 5 5.3 EMERGENCY PLAN The ISFSI-Only Emergency Plan was reviewed and approved by the NRC and implemented following the placement of all fuel in dry fuel storage. For the latest revision of the Emergency Plan, contact Licensing or refer to the records management system.

5.4 QUALITY ASSURANCE PROGRAM 5.4.1 Scope This section establishes the criteria to be applied to systems requiring Quality Assurance which prevent or mitigate the consequences of postulated accidents which could cause undue risk to the health and safety of the public.

The structures, systems, components, and other items requiring quality assurance are listed in the Vermont Yankee Quality Assurance Program Manual (QAPM).

5.4.2 Responsibilities

1. Compliance with the requirements of the VY QAPM based on the criteria of Title 10 of the Code of Federal Regulations, Part 50, Appendix B, and as committed to within the QAPM, shall be the responsibility of all personnel involved with activities affecting operational safety. 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 QAPM are delineated in the Organization,"

section of the QAPM.

5.4.3 Implementation Establishment of an effective Quality Assurance Program is assured through consideration of, and conformance with, the Regulatory Position in the Regulatory Guides as listed within the QAPM. Implementation of this program is assured through Quality Assurance procedures, derived from Quality Assurance policies, goals, and objectives.

BVY 24-029 / Enclosure / Page 51 of 61

VYNPS DSAR Revision 4 5.0-5 of 5 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.

BVY 24-029 / Enclosure / Page 52 of 61

VYNPS DSAR Revision 4 6.0-1 of 9 SAFETY ANALYSIS TABLE OF CONTENTS Section Title Page

6.1 INTRODUCTION

........................................................ 3 6.2 ACCEPTANCE CRITERIA.................................................. 4 6.2.1 DBA Acceptance Criteria..................................... 4 6.2.2 Site Event Acceptance Criteria.............................. 4 6.3 DELETED............................................................. 4 6.4 SITE EVENTS EVALUATED................................................ 5 6.4.1 High Integrity Container (HIC) Drop Event................... 5 6.4.1.1 Analytical Methodology.......................... 5 6.4.1.2 Assumptions.................................... 5 6.4.1.3 Inputs......................................... 6 6.4.1.4 Radiological Consequences/Results............... 7

6.5 REFERENCES

.......................................................... 9 BVY 24-029 / Enclosure / Page 53 of 61

VYNPS DSAR Revision 4 6.0-2 of 9 SAFETY ANALYSIS LIST OF TABLES Table No.

Title 6.4.1 HIC Drop Source Term Release Activity BVY 24-029 / Enclosure / Page 54 of 61

VYNPS DSAR Revision 4 6.0-3 of 9 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.

BVY 24-029 / Enclosure / Page 55 of 61

VYNPS DSAR Revision 4 6.0-4 of 9 6.2 Acceptance Criteria 6.2.1 DBA Acceptance Criteria A fuel handling accident (FHA) was previously discussed in this section as the only design basis accident (DBA) remaining 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) 10% of 10 CFR 100 Acceptance Criteria (rem)

EAB and LPZ 25 (whole body) 2.5 (whole body) 300 (thyroid) 30 (thyroid, critical organ)

(1)

EAB and LPZ dose acceptance criteria from 10 CFR 100.11 6.3 Deleted BVY 24-029 / Enclosure / Page 56 of 61

VYNPS DSAR Revision 4 6.0-5 of 9 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 (F) at the time of the postulated incident.

6.4.1.2 Assumptions Sandia National Laboratory has conservatively estimated, for a severity Category 3 transportation accident (which includes 99% of urban and 94% of rural accidents), no more than 1% (0.01) of any package contents would be released. For the purposes of the analysis, it was assumed that 0.5% of the released activity becomes aerosolized as a result of the fire.

BVY 24-029 / Enclosure / Page 57 of 61

VYNPS DSAR Revision 4 6.0-6 of 9 A HIC of 150 feet3 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.

BVY 24-029 / Enclosure / Page 58 of 61

VYNPS DSAR Revision 4 6.0-7 of 9 6.4.1.4 Radiological Consequences/Results 10%of10CFR100Dose AcceptanceCriteria (rem)

Calculated Dose (rem)

EAB 2.5rem(wholebody) 6.52E03(a)

(2 9.59E03(b) hours) 16.1E03(c) 30rem(thyroid,alsoappliedto 2.03E03(thyroid) thecriticalorgan) 4.58(lung)

(a)Dosefromstandingoncontaminatedground(2hrexposure)

(b)Dosefromcloudpassageoverheadduetoresinfireandaerosol release (c)Sumofgroundplaneexternalplusairbornefromcloud BVY 24-029 / Enclosure / Page 59 of 61

Table 6.4.1 HIC Drop Source Term Release Activity Liner A2 Values 2

LSA Limit3

Total Activity4 Drop Release Activity5 Nuclide1 (Ci)

(mCi/gm)

(Ci)

Cr51 600 0.3 1020 0.051 Mn54 20 0.3 1020 0.051 Fe55 1000 0.3 1020 0.051 Co58 20 0.3 1020 0.051 Co60 7

0.3 1020 0.051 Fe59 10 0.3 1020 0.051 Ni59 900 0.3 1020 0.051 Ni63 100 0.3 1020 0.051 Sb124 5

0.3 1020 0.051 Zn65 30 0.3 1020 0.051 Ag110m 7

0.3 1020 0.051 Sr89 10 0.3 1020 0.051 Sr90 0.4 0.005 17 0.00085 Zr95 20 0.3 1020 0.051 NB95 20 0.3 1020 0.051 Tc99 25 0.3 1020 0.051 I1296 2

NA 0.34 0.000017 Cs134 10 0.3 1020 0.051 Cs137 10 0.3 1020 0.051 Ce141 25 0.3 1020 0.051 Ce144 7

0.3 1020 0.051 Pu238 0.003 0.0001 0.34 0.000017 Pu 239/240 0.002 0.0001 0.34 0.000017 Am241 0.003 0.0001 0.34 0.000017 Cm242 0.2 0.005 17 0.00085 Footnotes:

Cm 243/244 0.01 0.0001 0.34 0.000017 19415.7 0.970785 1

NuclideListing: Alistingofradionuclidesthattypicallyaredeterminedbylaboratoryanalysistobe presentinRWCUresin. Shortlivedgaseousandvolatileradionuclidesarenotdetectedintypical radwastestreams.

2 A2: Forinformationalpurposes,quantitiesofnormalform(notspecialform)radionuclides,expressedin curies,permittedbyDOTtobecontainedinaTypeAdisposalpackage. Referto49CFR173.435forlisting.

3 LSALimit: DOTdeterminedLowSpecificActivityconcentrationlimit,expressedinunitsofmillicuriespergram ofmaterial. UnderregulationsdatedJanuary1989,LSAisafunctionofthetabulatedA2variableabove. Refer to49CFR173.403(n)(4)fortherelationship.

4 TotalActivity: Becauseconcentrationanddistributionofradionuclidesinwasteareexpectedtovaryover time,itisassumedforpurposesofthisradiologicalaccidentanalysisthatallradionuclidesareattheirupper limit. Inreality,asmallnumberofradionuclidesmightbeexpectedtoapproachalimitingconditionwhilethe majoritywouldbeatsomelowerlevel. Totalactivityisbasedonthefollowing: A) 150ft3(4.25m3)liner waste,densityof50lb/ft3=4.248E+06cc@0.8gm/ccgiving3.40E+06gm. B) EachnuclideisattheLSAlimit.

5 ReleaseActivity: Thequantityofeachnuclideassumedtobereleasedfromthewastelinertoformthesource term. Thereleaseactivityisbasedon: A) Linerdropincidentresultsinlinerfailureandreleaseof1%total contents. B) Ofthe1%materialreleased,0.5%isaerosolizedtoforma"releasecloud"sourceterm. The releasefractionis0.01andtheaerosolfractionis0.00005ofthetotalHICactivity).

6 I129islimitedby10CFR61burialrequirementsratherthanDOT. TheclassCdisposallimitforI129,aslistedin 10CFR61.55,Table1,is0.08Ci/m3(orCi/cc).

VYNPS DSAR Revision 4 6.0-8 of 9 BVY 24-029 / Enclosure / Page 60 of 61

6.5 References

1.

BVY 15-001, Certifications of Permanent Cessation of Power Operations and Permanent Removal of Fuel from the Reactor Vessel, Vermont Yankee Nuclear Power Station, January 12, 2015.

2.

Holtec International Final Safety Analysis Report for the Hi-Storm 100 Cask System, (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 4 6.0-9 of 9 BVY 24-029 / Enclosure / Page 61 of 61

BVY 24-029, Defueled Safety Analysis Report, Revision 4

Contents

Text

SUBJECT:

Dear Sir or Madam:

Attachment:

Enclosure:

1.0 INTRODUCTION

SUMMARY

1.1 INTRODUCTION

SUMMARY

SUMMARY

6.1 INTRODUCTION

6.5 REFERENCES

SUMMARY

1.1 INTRODUCTION

SUMMARY

1.1 INTRODUCTION

SUMMARY

SUMMARY

SUMMARY

6.1 INTRODUCTION

6.5 REFERENCES

Navigation menu

BVY 24-029, Defueled Safety Analysis Report, Revision 4

Text

SUBJECT:

Dear Sir or Madam:

Attachment:

Enclosure:

1.0 INTRODUCTION

SUMMARY

1.1 INTRODUCTION

SUMMARY

SUMMARY

6.1 INTRODUCTION

6.5 REFERENCES

SUMMARY

1.1 INTRODUCTION

SUMMARY

1.1 INTRODUCTION

SUMMARY

SUMMARY

SUMMARY

6.1 INTRODUCTION

6.5 REFERENCES

Navigation menu

Search