Text

Technical Specifications Proposed Change No. 313: Revision to License and Permanently Defueled Technical Specifications to Reflect Permanent Removal of Spent Fuel from the Spent Fuel Pool
	ML17206A200
Person / Time
Site:	Vermont Yankee
Issue date:	07/20/2017
From:	Boyle J; Entergy Nuclear Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	BVY 17-004
	Download: ML17206A200 (105)
	v • d • e

~ Entergy Entergy Nuclear Operations, Inc.

Vermont Yankee 320 Governor Hunt Rd.

Vernon, VT 05354 802-257-7711 John W. Boyle Director, Nuclear Decommissioning 10 CFR 50.90 BVY 17-004 July 20, 2017 Attn: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Technical Specifications Proposed Change No. 313 Revision to License and Permanently Defueled Technical Specifications to Reflect Permanent Removal of Spent Fuel from the Spent Fuel Pool Vermont Yankee Nuclear Power Station Docket Nos. 50-271 License No. DPR-28

REFERENCES:

1. Letter, Entergy Nuclear Operations, Inc. to USNRC, "Certifications of Permanent Cessation of Power Operations and Permanent Removal of Fuel from the Reactor Vessel ," BVY 15-001 , dated January 12, 2015 (ML15013A426)

2. Letter, Entergy Nuclear Operations, Inc. to USNRC, "Notification of Schedule Change for Dry Fuel Loading Campaign, " BVY 17-013, dated April 12, 2017 (ML17104A050)

3. Letter, Entergy Nuclear Operations, Inc. to USNRC , "Application for Order Consenting to Direct and Indirect Transfers of Control of Licenses and Approving Conforming License Amendment and Notification of Amendment to Decommissioning Trust Agreement ," BVY 17-005, dated February 9, 2017 (ML17045A140)

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Nuclear Operations, Inc. (ENO) hereby requests approval of a proposed amendment to Renewed Facility Operating License DPR-28 for Vermont Yankee Nuclear Power Station (VY). The proposed amendment would revise the Operating License and the Permanently Defueled Technical Specifications (POTS) to reflect removal of all spent nuclear fuel from the spent fuel pool (SFP) and its transfer to dry cask storage within an Independent Spent Fuel Storage Installation (ISFSI). The proposed changes include the

BVY 17-004 I Page 2 of 3 relocation of administrative controls from the POTS to the VY Quality Assurance Program Manual (QAPM) .

By letter dated January 12, 2015 (Reference 1), ENO submitted certifications for permanent cessation of reactor operations at VY and permanent removal of fuel from the reactor vessel pursuant to 10 CFR 50.82{a)(1 ). Therefore as specified in 10 CFR 50.82(a){2) , the 10 CFR Part 50 license for VY no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel. By letter dated April 12, 2017 (Reference 2) , ENO provided notification that off-load of the spent fuel pool (SFP) and transfer of the spent fuel to the ISFSI is expected to be completed in late 2018, provided the criteria for transfer are met, including certain regulatory approvals. In support of this condition , a revision to the VY Operating License and associated POTS is proposed to comport with the requirements for a facility configuration with all spent nuclear fuel in dry storage within an ISFSI. to this letter provides a description and evaluation of the proposed change which includes the no significant hazards consideration determination and environmental considerations. Attachment 2 contains a markup of the Operating License and POTS pages, including changes to the POTS Bases which are provided for information only. Attachment 3 contains the retyped Operating License and POTS pages. Attachment 4 provides the administrative controls that are proposed to be relocated from the POTS to the QAPM .

ENO requests approval of the proposed license amendment by May 31 , 2018. Once approved, the amendment will be implemented within 60 days following ENO notification to the NRC that all spent nuclear fuel has been transferred out of the SFP and placed within the ISFSI.

By letter dated February 9, 2017 (Reference 3) , ENO submitted an application for transfer of the VY licenses, requesting issuance of an Order by December 31 , 2017, for approval of amendments to the VY facility license with authorization for the transfers to take place at any time up to December 31, 2018. NRC review of the license transfer application is anticipated to occur concurrently with review of this amendment request. However, this request to amend the license and POTS to reflect permanent removal of spent fuel from the SFP is not intended to rely on approval of Reference 3, and is being submitted for approval and implementation independent of NRC approval of Reference 3.

In accordance with 10 CFR 50.91 , a copy of this application , with attachments , is being provided to the State of Vermont , Department of Public Service .

A new regulatory commitment is described in Attachment 5 to this letter.

Should you have any questions concerning this letter, please contact Mr. Coley Chappell at

{802) 451-3374.

BVY 17-004 I Page 3 of 3 I declare under penalty of perjury that the foregoing is true and correct. Executed on July 20, 2017.

JWB/ccc Attachments :

1. Description and Evaluation of the Proposed Changes

2. Markup of the Existing Operating License and Technical Specification Pages

3. Retyped Operating License and Technical Specification Pages

4. Administrative Controls Relocated to the QAPM

5. Regulatory Commitment cc: Mr. Daniel H. Dorman Regional Adm inistrator, Region 1 U.S. Nuclear Regulatory Commission 2100 Renaissance Blvd , Suite 100 King of Prussia, PA 19406-2713 Mr. Jack D. Parrott, Sr. Project Manager Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Mail Stop T-8F5 Washington , DC 20555 Ms . June Tierney , Commissioner Vermont Department of Public Service 112 State Street - Drawer 20 Montpelier, Vermont 05602-2601

BVY 17-004 Docket No. 50-271 Attachment 1 Vermont Yankee Nuclear Power Station Proposed Change 313 Description and Evaluation of the Proposed Changes

BVY 17-004 I Attachment 1 I Page 1 of 46 Technical Specifications Proposed Change No. 313, Revision to License and Permanently Defueled Technical Specifications to Reflect Permanent Removal of Spent Fuel from the Spent Fuel Pool Discussion of Change, Technical Evaluation, No Significant Hazards Consideration Determination, and Environmental Considerations

1.

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, Entergy Nuclear Operations, Inc. (ENO) is proposing an amendment to Renewed Facility Operating License (Operating License) DPR-28 for Vermont Yankee Nuclear Power Station (VY). The proposed amendment would revise the Operating License and revise the Permanently Defueled Technical Specifications (POTS) to reflect removal of all spent nuclear fuel from the spent fuel pool (SFP) and completion of the transfer of spent fuel to dry cask storage within an Independent Spent Fuel Storage Installation (ISFSI) .

The proposed changes include the relocation of administrative controls from the POTS to the VY Quality Assurance Program Manual (QAPM).

By letter dated January 12, 2015 (Reference 1), ENO submitted certifications for permanent cessation of reactor operations at VY and permanent removal of fuel from the reactor vessel pursuant to 10 CFR 50.82(a)(1 ). Therefore as specified in 10 CFR 50.82(a)(2), the 10 CFR Part 50 license for VY no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel. By letter dated April 12, 2017 (Reference 2) , ENO provided notification that off-load of the spent fuel pool (SFP) and transfer of the spent fuel to the ISFSI is expected to be completed in late 2018, provided the criteria for transfer are met, including certain regulatory approvals. In support of this condition , a revision to the VY Operating License and associated POTS is proposed to comport with the requirements for a facility configuration with all spent nuclear fuel in dry storage within an ISFSI.

The existing VY POTS contain Limiting Conditions for Operation (LCOs) that provide for appropriate functional capability of equipment required for safe storage and management of irradiated fuel with fuel stored in the SFP. As such, the existing POTS provide a level of control in excess of that needed for safe storage and management of irradiated fuel with all fuel stored in an ISFSI. The majority of the existing POTS are only applicable when irradiated fuel assemblies are within the SFP. Once all spent fuel assemblies have been transferred to the ISFSI , all remaining LCOs (and associated Surveillance Requirements (SRs)) will no longer be applicable and are being proposed for deletion. The POTS being proposed reflect the removal of all spent fuel from the SFP. The proposed changes will result in a POTS that will be applicable to VY after the last spent fuel assembly has been removed from the SFP and placed within the ISFSI .

Pending Licensing Actions under NRC Review There are several other pending licensing actions currently under NRC review that involve proposed changes to the Operating License. For clarity, and to aid in the review of this submittal, the marked-up and retyped Operating License pages included in Attachments 2

BVY 17-004 I Attachment 1 I Page 2 of 46 and 3, respectively, indicate where changes are proposed by these separate licensing actions.

A general description of these separate licensing actions is provided below.

By letter dated February 9, 2017 (Reference 3), ENO submitted an application for transfer of the VY licenses, requesting issuance of an Order by December 31, 2017, for approval of amendments to the VY facility license with authorization for the transfers to take place at any time up to December 31 , 2018. NRG review of the license transfer application (LTA) is anticipated to occur concurrently with review of this amendment request. However, this request to amend the license and POTS to reflect permanent removal of spent fuel from the SFP is not intended to rely on approval of Reference 3, and is being submitted for approval and implementation independent of NRG approval of Reference 3.

ENO has also submitted separate license amendment requests for proposed changes to License Condition 3.G "Security Plan" for the VY ISFSI Physical Security Plan (PSP) and changes to cyber security plan (CSP) requirements.

As noted above, for completeness and to aid in NRG review, all pages of the Operating License and POTS are included in the marked-up pages (Attachment 2) and retyped pages (Attachment 3), with references to the separately proposed changes associated with the LTA, ISFSI PSP, and CSP, as appropriate, provided for information only.

2. DETAILED DESCRIPTION AND BASIS FOR THE PROPOSED CHANGES The proposed amendment would modify the VY Operating License and POTS to comport to the condition of all irradiated fuel in dry storage within the onsite Independent Spent Fuel Storage Installation (ISFSI) at VY using casks certified for use under a general 10 CFR 72 license. The amendment would also revise the VY POTS to eliminate operational requirements and certain design requirements involving storage of spent fuel that will no longer be applicable following the transfer of the last spent fuel assembly from the SFP to the ISFSI.

A new POTS design requirement is proposed to prohibit storage of spent fuel in the SFP to comport with the proposed VY ISFSI Emergency Plan (Reference 4) and the proposed VY ISFSI Physical Security Plan (Reference 5) , which are predicated on completion of the offload of spent fuel from the SFP and transfer to the ISFSI.

The proposed changes to the POTS also involve relocating administrative controls from Section 5, Administrative Controls, to the VY QAPM , and subsequently controlling them in accordance with 10 CFR 50.54(a) . This relocation is being proposed pursuant to the criteria contained in 10 CFR 50 .36 and in accordance with the recommendations and guidance contained in NRG Administrative Letter 95-06 (Reference 6).

Proposed changes to the Operating License and a brief description are as follows :

License Finding h regarding the previous NRG findi ng related to actions for management of aging on the functionality of structures and components, as identified during the VY license renewal process, will be deleted as no longer applicable;

BVY 17-004 I Attachment 1 I Page 3 of 46

License Condition 3.C, Reports, regarding requirements in the Technical Specifications administrative controls will be deleted since these are proposed to be relocated to the VY QAPM;

License Condition 3.N, Mitigation Strategy License Conditions, regarding mitigation strategies for large fires and explosions, will be deleted as no longer applicable;

Licensing Condition 3.P regarding license renewal commitments included as information in the VY Updated Final Safety Analysis Report (UFSAR) , will be deleted as no longer required ; and

Operating License pages will be re-formatted to remove references to License Condition paragraphs deleted by previously issued Amendments in order to consolidate the remaining paragraphs.

Proposed changes to the POTS and a brief description are as follows:

Section 1 .0, "Definitions," defines terms that are proposed to be eliminated or relocated from the POTS, therefore will be deleted in its entirety as no longer needed;

Sections 3.0/4.0, "Limiting Conditions of Operation and Surveillance Requirement (SR)

Applicability," provides general requirements applicable to all SRs, will be deleted in their entirety since all existing POTS to which they apply are being proposed for deletion ;

Sections 3.1/4.1 , "Radioactive Effluents," applying to the release of radioactive effluents from the facility, were determined to not meet the criteria specified in § 50.36(c)(2)(ii) ,

and therefore proposed to be relocated to a licensee-controlled procedure subject to the requirements in§ 50.59 and other applicable regulations ;

Sections 3.2/4.2, "Spent Fuel Storage," apply to spent fuel stored in the SFP, and therefore will be deleted in their entirety;

Section 5.0, "Design Features, describes the site locations and requirements for storage of spent nuclear fuel, will be modified to reflect the condition of a permanently defueled reactor and permanent removal of spent fuel from the SFP; and

Section 6.0, "Administrative Controls, describes the organization and responsibilities for the site, as well as programs and procedures, certain modifications and deletions are proposed to conform with other proposed changes and the condition of all spent fuel stored within the ISFSI, and the majority of this section is proposed to be relocated from the POTS to the VY QAPM, with the exception of Section 6.5,"High Radiation Area,

which will be retained in the POTS.

General Analysis Applicable to the Proposed Changes The proposed amendment would modify the VY Operating License and POTS by deleting requirements that are no longer applicable to a facility with no spent nuclear fuel stored in the SFP, while modifying the remaining portions to comport with the condition of all spent nuclear fuel stored within an ISFSI . The proposed changes involve the relocation of certain administrative controls from the POTS to the VY QAPM. This amendment is proposed to be implemented within 60 days following ENO's notification to the NRC that all spent fuel assemblies have been transferred out of the SFP and placed in dry storage within the ISFSI ,

therefore implementation of the approved amendment is predicated on this condition.

BVY 17-004 I Attachment 1 I Page 4 of 46 VY plans to use a decommissioning method in which most fluid systems are drained and the plant is left in a stable condition until final dismantlement. Administrative controls that are required to be in place when decontamination or dismantling activities of radioactive systems, structures, and components are being performed are designed to minimize the likelihood of an off-normal or accident event, and thereby the consequences of such an event. The proposed changes to the Operating License and POTS do not have an adverse impact on the remaining decommissioning activities or any of their postulated consequences .

The spent fuel will be stored in the ISFSI until it is shipped off site consistent with the schedules described in the Post-shutdown Decommissioning Activities Report (PSDAR) (Reference 7) and the Irradiated Fuel Management Program (Reference 8).

During decommissioning with all spent fuel in dry storage within an ISFSI , there are no previously installed plan systems relied upon for the safe storage of spent fuel. In this condition there are no credible accidents whose prevention or mitigation would need to be addressed by facility Technical Specifications (TS). The spent fuel storage casks and canisters used in the ISFSI are subject to their own Certificate of Compliance and associated TS.

Section 6 of the VY Defueled Safety Analysis Report (DSAR) describes the design basis accidents (OBAs) related to the SFP. These postulated accidents are predicated on spent fuel being stored in the SFP. With the removal of the spent fuel from the SFP, there are no remaining spent fuel assemblies to be monitored and there are no credible accidents that require the actions of a Certified Fuel Handler, Shift Manager, or a Non-certified Operator to prevent occurrence or mitigate the consequences of an accident.

10 CFR 50.2 defines safety-related structures, systems, and components (SSCs) as those that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition ; or (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in § 50.34(a)(1) or§ 100.11 of this chapter, as applicable.

The first two criteria (integrity of the reactor coolant pressure boundary and safe shutdown of the reactor) are not applicable to a plant in a permanently defueled condition. The third criterion is related to preventing or mitigating the consequences of accidents that could result in potential offsite exposures exceeding limits. However, after all nuclear spent fuel assemblies have been transferred to dry cask storage within an ISFSI , there are no longer any SSCs at VY that are required to be relied upon for accident prevention or mitigation , since the only design basis accident described in the current VY Defueled Safety Analysis Report (DSAR) Section 6 involves a fuel handling accident (FHA) analyzed using Alternate Source Term methodology using the applicable guideline exposures specified in§ 50.67(b)(2) . Since the FHA is predicated on spent fuel stored in the SFP, it is no longer applicable as a design basis accident once all spent fuel is offloaded from the SFP and transferred to dry storage within the ISFSI.

Therefore , there are no SSCs at VY that meet the definition of a safety-related SSC as defined

BVY 17-004 I Attachment 1 I Page 5 of 46 in§ 50.2, and the proposed deletion of requirements in the POTS does not involve any SSCs credited in any accident analysis at VY after the complete offload of the SFP and transfer of spent fuel to the ISFSI.

10 CFR 50.36, "Technical Specifications," promulgates the regulatory requirements related to the content of Technical Specifications. As detailed in subsequent sections of this proposed amendment, this regulation lists four criteria to define the scope of equipment and parameters that must be included in a plant's Technical Specifications. A discussion of the applicability of these four criteria in the permanently defueled condition with all fuel removed from the spent fuel pool is provided in Section 3.1 of this submittal. In a permanently defueled condition with all spent fuel in dry storage within an ISFSI , the scope of equipment and parameters that need be included in the VY POTS is limited to a description of the design features and high radiation area adm inistrative controls .

The proposed changes related to the relocation of certain administrative controls do not affect operating procedures or administrative controls that have the function of preventing or mitigating any accidents applicable to the safe management of irradiated fuel or decommissioning of the facility.

Detailed Discussion The following table identifies each section in the Operating License and POTS that is being changed , the proposed changes , and the basis for each change . Changes to the Operating License are listed first, followed by changes to the POTS. Changes to the POTS Bases are provided for information in Attachment 2.

Additional administrative changes to the Operating License and POTS are proposed to remove unnecessary references to paragraphs deleted by the proposed changes or previously issued Amendments , and to reformat and consolidate the remaining sections and paragraphs. These changes are as shown in the markup provided in Attachment 2 and retyped pages provided in .

A change is proposed to the Cover Page for Appendix A to Technical Specifications (TS) to delete reference to "Bases." This is an administrative and conforming change to reflect the proposed changes to the POTS. Deletions of the Bases are shown in Attachment 2, and with there no longer being any Bases remaining in the POTS , and the TS Bases Control Program in TS 6.7.E proposed to be deleted as described below, this reference on the Cover Page is no longer needed .

Changes to the POTS Table of Contents are proposed to reflect the elimination of sections associated with other proposed changes to the POTS and previously issued Amendments, as well as repagination of remaining sections. This is an administrative change and is shown in Attachments 2 and 3.

BVY 17-004 I Attachment 1 I Page 6 of 46 Detailed Discussion of Proposed Changes to the VY Operating License LICENSE FINDING h. AGING EFFECTS Existing license Finding h. Proposed license Finding h.

h. Actions have been identified and have Deleted.

been or will be taken with respect to: ( 1) managing the effects of aging on the functionality of structures and components that have been identified to require review under 10 CFR 54.21 (a)(1) during the period of extended operation, and (2) time-limited aging analyses that have been identified to require review under 10 CFR 54.21 (c), such that there is reasonable assurance that the activities authorized by this license will continue to be conducted in accordance with the current licensing basis, as defined in 10 CFR 54.3 for the facility, and that any changes made to the facility's current licensing basis in order to comply with 10 CFR 54.29(a) are in accordance with the Act and the Commission's regulations.

Basis POTS The certifications for permanent cessation of reactor operations at VY and permanent removal of fuel from the reactor vessel pursuant to 10 CFR 50.82(a)(1) have been docketed ,

therefore by 10 CFR 50.82(a)(2), the 10 CFR Part 50 license for VY no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel. As such, this finding, related to the "period of extended operation" based on the referenced regulations at 10 CFR Part 54, is no longer applicable to the VY facility, and is appropriate to delete.

The purpose of aging management programs is, in part, to ensure that aging effects of equipment important to safe operation of the reactor are managed so that the functionality of SSCs is maintained during the facility's period of extended operation. For a permanently shutdown facility, most of the equipment subject to aging management programs is no longer in use and functionality does not need to be maintained. During decommissioning, some equipment functionality, such as for equipment related to the fire protection system to address fire events that could result in radiological hazards per the requirements of 10 CFR 50.48(f) , may be required beyond the permanent cessation of operations and therefore may be subject to an aging management program . Prior to cessation of operations, VY license renewal commitments for aging management were incorporated into Section 15, "Aging Management Programs," of the Updated Final Safety Analysis Report (UFSAR) , and subsequent to permanent defueling of the reactor relocated to Section 7, "Aging Management," of the Defueled Safety Analysis Report (DSAR), updated in accordance with 10 CFR 50.71(e) . Therefore, changes to these license renewal commitments continue to be evaluated pursuant to the criteria in 10 CFR 50.59. On this basis, the NRC staff has previously found removal of license conditions related to aging management commitments to be acceptable (see Reference 9 and Reference 10).

BVY 17-004 I Attachment 1 I Page 7 of 46 LICENSE CONDITION 3.C, REPORTS Existing License Condition 3.C Proposed License Condition 3.C C. Reports C. This paragraph deleted by Amendment No.

XXX .

Entergy Nuclear Operations, Inc. shall make reports in accordance with the requirements of the Technical Specifications.

Basis The elimination of this License Condition is an administrative change , because the proposed changes to the POTS will not contain reporting requirements since the administrative controls for the reporting requirements are proposed to be eliminated or relocated to the QAPM, as described and evaluated in the proposed changes to the POTS.

LICENSE CONDITION 3.E, ENVIRONMENTAL CONDITIONS Existing License Condition 3.E.1 O Proposed License Condition 3.E.1 O

10. A report shall be submitted to MDPH and 10. A report shall be submitted to MDPH and MDC by May 15 of each year of plant MDC by May 15 of each year, specifying operation , specifying the total quantities the total quantities of radioactive materials of radioactive materials released to the released to the Connecticut River during Connecticut River during the previous the previous calendar year. The report calendar year. The report shall contain shall contain the following information:

the following information:

Basis This proposed change is to remove reference to "plant operation" with respect to the requirements to submit an annual report to the Massachusetts Department of Public Health and the Metropolitan District Commission for radioactive materials released to the Connecticut River.

This is an administrative change to clarify that the report will continue to be submitted each year that there are radioactive releases to the Connecticut River.

BVY 17-004 I Attachment 1 I Page 8 of 46 LICENSE CONDITION 3.E, ENVIRONMENTAL CONDITIONS Existing License Condition 3.E.13 Proposed License Condition 3.E.13

13. Entergy Nuclear Operations , Inc. shall 13. This paragraph deleted by Amendment No.

establish and maintain a system of XXX.

emergency notification to the states of Vermont and New Hampshire , and the Commonwealth of Massachusetts, satisfactory to the appropriate public health and public safety officials of those states and the Commonwealth, which provides for:

a. Notice of site emergencies as well as general emergencies.

b. Direct microwave communication with the state police headquarters of the respective states and the Commonwealth when the transmission facilities of the respective states and the Commonwealth so permit, at the expense of Entergy Nuclear Operations, Inc.

c. A verification or coding system for emergency messages between Entergy Nuclear Operations, Inc. and the state police headquarters of the respective states and the Commonwealth.

Basis This License Condition involved the establishment and maintenance of a system of emergency notification to the states of Vermont and New Hampshire and the Commonwealth of Massachusetts. This equipment is not relied upon by the existing VY Permanently Defueled Emergency Plan , or the proposed VY ISFSI Emergency Plan (Reference 4) and VY ISFSI Physical Security Plan (Reference 5). The proposed change is to delete this condition in its entirety in order to comport with the existing Emergency Plan and Physical Security Plan and the proposed plans for the VY ISFSI.

BVY 17-004 I Attachment 1 I Page 9 of 46 LICENSE CONDITION 3.N, MITIGATION STRATEGY LICENSE CONDITION Existing License Condition 3.N Proposed License Condition 3.N N. Mitigation Strategy License Condition Deleted.

Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(a) Fire fighting response strategy with the following elements:

1. Pre-defined coordinated fire response strategy and guidance

2. Assessment of mutual aid fire fighting assets

3. Designated staging areas for equipment and materials

4. Command and control

5. Training of response personnel (b) Operations to mitigate fuel damage considering the following:

1. Protection and use of personnel assets

2. Communications

3. Minimizing fire spread

4. Procedures for implementing integrated fire response strategy

5. Identification of readily-available pre-staged equipment

6. Training on integrated fire response strategy

7. Spent fuel pool mitigation measures (c) Actions to minimize release to include consideration of:

1. Water spray scrubbing

2. Dose to onsite responders Basis This license condition is proposed for deletion in its entirety. After all spent fuel is stored within the ISFSI, the mitigation strategies are no longer required. The license condition incorporated the requirements for Section B.5.b mitigation strategies of the Interim Compensatory Measures (ICM) Order EA-02-026 dated February 25, 2002 (Reference 11). Subsequently, 10 CFR 50.54(hh)(2) became effective on May 26, 2009. This license condition provides mitigation strategies and other requirements for loss of large areas of the plant due to large fires and explosions. However, as stated in§ 50.54(hh)(3), this section of the requlations does not apply

BVY 17-004 I Attachment 1 I Page 1O of 46 to a permanently defueled reactor that has submitted the certifications under§ 50.82(a) for permanent cessation of operations and permanent removal of fuel from the reactor vessel , which have been submitted for VY and docketed (Reference 1). By letter dated November 28 , 2011 (Reference 12), the NRC rescinded Item 8 .5.b of the ICM Order. Therefore , neither the ICM Order nor§ 50.54(hh) continue to apply to VY.

It is noted that SECY-16-0142, Draft Final Rule - Mitigation of Beyond-Design-Basis Events, dated December 15, 2016 (Reference 13), would , in part, relocate requirements existing in

§ 50.54(hh)(2) for mitigation of the effects of a loss of large area of the plant due to explosions or fire to a new section 10 CFR 50.155. As discussed in Enclosure 1 to SECY-16-0142, the NRC concludes that it is inappropriate to apply requirements to implement mitigation measures for large fires and explosions to a permanently shutdown and defueled reactor, where the fuel was moved to an ISFSI or removed from the site. The applicability of the proposed § 50.155 states that holders of operating licenses for which the NRC has docketed the§ 50.82(a)(1) certifications need not meet the requirements of this section once all irradiated fuel has been permanently removed from the SFP. Since this will be the condition of the VY facility prior to implementing this amendment request, VY will not need to meet the requirements as described in SECY 0142.

BVY 17-004 I Attachment 1 I Page 11 of 46 LICENSE CONDITION 3.P, LICENSE RENEWAL COMMITMENTS IN THE UFSAR Existing License Condition 3.P Proposed License Condition 3.P P. The information in the UFSAR supplement, Deleted.

submitted pursuant to 10 CFR 54.21 (d) , as revised during the license renewal application process, and as supplemented by Commitment Nos. 1-5, 6 (as revised by Entergy Nuclear Vermont Yankee, LLC letter dated May 19, 2011), 7-36, 38, 39, 42, 43 ,

and 45-55 of Appendix A of Supplement 2 of NUREG-1907 shall be incorporated as part of the UFSAR which will be updated in accordance with 10 CFR 50.71{e). As such ,

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. may make changes to the programs and activities described in the UFSAR supplement and Commitment Nos. 1-5, 6 (as revised by Entergy Nuclear Vermont Yankee, LLC letter dated May 19, 2011 ), 7-36, 38, 39, 42, 43, and 45-55 of Appendix A of Supplement 2 of NUREG-1907 provided Entergy Nuclear Vermont Yankee , LLC and Entergy Nuclear Operations, Inc. evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

Basis This License Condition involved the management of license renewal commitments. The purpose of aging management programs is, in part, to ensure that aging effects of equipment important to safe operation of the reactor are managed so that the functionality of SSCs is maintained during the facility's period of extended operation. For a permanently shutdown facility, most of the equipment subject to aging management programs is no longer in use and functionality does not need to be maintained . During decommissioning , some equipment functionality, such as for equipment related to the fire protection system to address fire events that could result in radiological hazards per the requirements of 10 CFR 50.48{f) , may be required beyond the permanent cessation of operations and therefore may be subject to an aging management program . Prior to cessation of operations, VY license renewal commitments for aging management were incorporated into Section 15, "Aging Management Programs," of the Updated Final Safety Analysis Report (UFSAR} , and subsequently relocated to Section 7, "Aging Management," of the Defueled Safety Analysis Report (DSAR) , updated in accordance with 10 CFR 50.71{e) . Changes to these license renewal commitments continue to be evaluated pursuant to the criteria in 10 CFR 50.59. On this basis, the NRC staff has previously found removal of license conditions for aging management commitments acceptable (see Reference 9 and Reference 10). The spent fuel storage cask systems used in the ISFSI are subject to their own Certificate of Compliance and associated Technical Specifications.

BVY 17-004 I Attachment 1 I Page 12 of 46 Detailed Discussion of Proposed Changes to the VY POTS POTS SECTION 1.0, DEFINITIONS Existing POTS 1.0 Proposed POTS 1.0 1.0 DEFINITIONS Deleted The succeeding frequently used terms are explicitly defined so that a uniform interpretation of the specifications may be achieved.

A. Certified Fuel Handler - A Certified Fuel Handler is an individual who complies with the provisions of the Certified Fuel Handler training program.

B. Immediate - Immediate means that the required action will be initiated as soon as practicable considering the safe operation of the unit and the importance of the required action .

C. Operable - A system , subsystem, train ,

component or device shall be operable or have operability when it is capable of performing its specified function(s).

Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls ,

normal or emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem , train, component or device to perform its function(s) are also capable of performing their related support function(s).

D. Operating - Operating means that a system or component is performing its intended functions in its required manner.

Basis As stated in this section, the purpose of the definitions is to provide uniform interpretation of frequently used terms in the POTS. The proposed changes to other POTS sections either eliminate or relocate the information that references these terms. Since the terms are no longer needed after the spent fuel has been removed from the SFP and transferred to the ISFSI, it is acceptable to delete this section in its entirety with no impact on continued sate storage and maintenance of spent fuel in the ISFSI.

BVY 17-004 I Attachment 1 I Page 13 of 46 POTS SECTIONS 3.0/4.0, LIMITING CONDITIONS OF OPERATION AND SURVEILLANCE REQUIRMEMENT (SR) APPLICABILITY Existing POTS 3.0 and 4.0 Proposed POTS 3.0 and 4.0 3.0 LIMITING CONDITIONS FOR Deleted OPERATION APPLICABILITY 4.0 SURVEILLANCE REQUIREMENT (SR)

APPLICABILITY SR 4.0.1 SRs shall be met during the specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR .

Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failu re to perform a Surveillance within the specified frequency shall be failure to meet the LCO except as provided in SR 4.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.

SR 4.0.2 Unless otherwise stated in these specifications, periodic surveillance tests ,

checks , calibrations , and examinations shall be performed within the specified surveillance intervals . These intervals may be adjusted plus 25%.

SR 4.0.3 If it is discovered that a surveillance was not performed within its specified frequency, declaring applicable Limiting Conditions for Operation (LCOs) not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified frequency, whichever is greater. This delay period is permitted to allow performance of the surveillance . A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the risk impact shall be managed.

BVY 17-004 I Attachment 1 I Page 14 of 46 (Continued)

If the surveillance is not performed within the delay period , applicable LCOs must immediately be declared not met, and applicable LCOs must be entered .

When the surveillance is performed within the delay period and the surveillance is not met (i.e. , acceptance criteria are not satisfied) , applicable LCOs must immediately be declared not met, and applicable LCOs must be entered.

Basis All information in Section 3.0 was previously deleted. SR 4.0.1 , SR 4.0.2, and SR 4.0.3 provide general requirements applicable to all surveillances and apply at all times, unless otherwise stated. After the transfer of spent fuel from the SFP to the ISFSI is complete, there will no longer be any applicable Limiting Conditions for Operation or SRs as a result of the proposed changes.

Therefore there are no specific requirements remaining to apply these general requirements to, and deleting the sections in their entirety is acceptable and has no impact on continued safe storage and maintenance of spent fuel in the ISFSI.

BVY 17-004 I Attachment 1 I Page 15 of 46 POTS SECTIONS 3.1/4.1, RADIOACTIVE EFFLUENTS Existing POTS 3.1 and 4.1 Proposed POTS 3.1 and 4.1 3.1 LIMITING CONDITIONS FOR OPERATION Deleted 3.1 RADIOACTIVE EFFLUENTS Applicability:

Applies to the release of all radioactive effluents from the plant.

Objective:

To assure that radioactive effluents are kept "as low as is reasonably achievable" in accordance with 10CFR50, Appendix I and ,

in any event, are within the dose limits for Members of the Public specified in 10CFR20.

Specification :

A. Liqu id Holdup Tanks

1. The quantity of radioactive material contained in any outside tank* shall be limited to less than or equal to 1O curies, excluding tritium and dissolved or entrained noble gases.

2. With the quantity of radioactive material in any outside tank*

exceeding the limit of Specification 3.1 .A.1 , immediately take action to suspend all additions of radioactive material to the tank. Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , reduce the tank contents to within the limit.

BVY 17-004 I Attachment 1 I Page 16 of 46 (Continued) 4.1 SURVEILLANCE REQUIREMENTS 4.1 RADIOACTIVE EFFLUENTS Applicability:

Applies to the required surveillance of all radioactive effluents released from the plant.

Objective:

To ascertain that all radioactive effluents released from the plant are kept "as low as is reasonably achievable" in accordance with 10CFR50, Appendix I and, in any event, are within the dose limits for Members of the Public specified in 10CFR20.

Specification:

A. Liquid Holdup Tanks

1. The quantity of radioactive material contained in each of the liquid holdup tanks* shall be determined to be within the limits of Specification 3.1.A.1 by analyzing a representative sample of the tank's contents within one week following the addition of radioactive materials to the tank. One sample may cover multiple additions.

NOTE: Tanks included in this Specification are only those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank's contents , or that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment svstem.

Basis The requirements of TS LCO 3.1 .A and SR 4.1.A do not satisfy any of the four requirements established in 10 CFR 50.36 (c}(2}(ii) , as described in the General Analysis discussion above and based on the evaluation provided in Section 3.1. Thus , it is appropriate for these sections and requirements to be deleted in their entirety from the POTS and relocated to a licensee controlled procedure due to the controls imposed by regulations including 10 CFR 50.59, Appendix B to 10 CFR Part 50, and the requirements to assure radioactive effluents are maintained "as low as reasonably achievable" in accordance with Appendix I to 10 CFR Part 50, and within the dose limits to the public specified in 10 CFR Part 20. Maintaining the relocated requirements in a procedure subject to the change control process in 10 CFR 50.59 provides adequate control, and has no impact on continued safe storage of spent fuel in the ISFSI.

BVY 17-004 I Attachment 1 I Page 17 of 46 POTS SECTIONS 3.2/4.2, SPENT FUEL STORAGE Existing POTS 3.2 and 4.2 Proposed POTS 3.2 and 4.2 3.2 LIMITING CONDITIONS FOR Deleted OPERATION 3.2 SPENT FUEL STORAGE Applicability:

Applies to storage of spent fuel.

Objective:

To assure sate storage of spent fuel.

Specification:

A. Fuel Storage Pool Water Level Whenever irradiated fuel is stored in the fuel storage pool the pool water level shall be maintained at a level of at least 36 feet.

B. Spent Fuel Pool Water Temperature Whenever irradiated fuel is stored in the spent fuel pool , the pool water temperature shall be maintained below 150°F.

4.2 SURVEILLANCE REQUIREMENTS 4.2 SPENT FUEL STORAGE Applicability:

Applies to the parameters which monitor the storage of spent fuel.

Objective:

To verify that spent fuel is being stored safely.

BVY 17-004 I Attachment 1 I Page 18 of 46 (Continued)

Specification:

A. Fuel Storage Pool Water Level Whenever irradiated fuel is stored in the fuel storage pool, the pool level shall be recorded daily.

B. Spent Fuel Pool Water Temperature Whenever irradiated fuel is in the spent fuel pool, the pool water temperature shall be recorded daily. If the pool water temperature reaches 150°F, all operations tending to raise the pool water temperature shall cease and measures taken immediately to reduce the pool water temperature below 150°F.

Basis The requirements in this section are related to assuring the functional capability of equipment required for safe storage and maintenance of spent fuel stored in the SFP. POTS 3.2.A/4.2.A and 3.2.B/4.2.B do not apply when there are no fuel assemblies stored in the SFP. Therefore ,

these POTS will no longer be needed following the transfer of all spent fuel assemblies from the SFP to the ISFSI. As such , these POTS may be deleted in their entirety with no impact on continued safe storage and maintenance of irradiated fuel in an ISFSI.

POTS SECTION 5.0, DESIGN FEATURES Existing POTS 5.0 and 5.1 Proposed POTS 5.0 and 5.1 5.0 DESIGN FEATURES 5.0 DESIGN FEATURES 5.1 Site 5.1 Site The station is located on the property The station is located on the property on the west bank of the Connecticut on the west bank of the Connecticut River in the Town of Vernon , Vermont , River in the Town of Vernon , Vermont, which Entergy Nuclear Vermont which Entergy Nuclear Vermont Yankee, LLC either owns or to which it Yankee, LLC either owns or to which it has perpetual rights and easements. has perpetual rights and easements.

The site plan showing the exclusion area boundary, boundary for gaseous effluents, boundary for liquid effluents, as well as areas defined per 10CFR20

BVY 17-004 I Attachment 1 I Page 19 of 46 (Continued) as "controlled areas" and "unrestricted areas" are on plant drawing 5920-6245. The minimum distance to the boundary of the exclusion area as defined in 10CFR100.3 is 91 O feet.

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 91 O feet.

This includes the authority to exclude or remove personnel and property within the exclusion area. Only activities related to plant operation are permitted in the exclusion area.

Basis POTS Section 5.0 provides a description of the VY site location. This section is proposed to be revised to reflect the permanently defueled condition of the facility with all fuel assemblies stored within the ISFSI.

A description regarding the minimum distance from the center line of the reactor containment to the site exclusion radius is being deleted. The minimum distance from the center line of the reactor containment to the site exclusion radius is based on requirements contained in § 100.3 regarding reactor accident dose analyses. Because by§ 50.82(a)(2) the VY 10 CFR Part 50 license no longer authorizes operation of the reactor or emplacement or retention of fuel in the reactor vessel, this design feature is no longer needed, and its description may be deleted.

The reference to the site plan drawing that defines the exclusion area boundary, boundary for gaseous effluents, boundary for liquid effluents, as well as areas defined per 10 CFR 20 as "controlled areas" and "unrestricted areas" is proposed to be deleted because the information is more detailed than required to describe the VY site. In addition, the description of licensee authority on the property references the exclusion area defined in§ 100.3, which as stated above, is no longer needed, and does not describe a site location design feature with the reactor permanently defueled, and therefore may be deleted. Removal of these descriptions do not alter any regulatory requirements related to licensee authority over the site location , and thus do not have an impact on continued safe storage and maintenance of irradiated fuel in an ISFSI.

The proposed content of POTS 5.1 describing the site location is consistent with the level of detail provided in the comparable design features site location descriptions in the TS for Zion Units 1 and 2 approved by the NRC on January 14, 2015 (Reference 14) and for Kewaunee approved by the NRC on June 7, 2017 (Reference 15).

BVY 17-004 I Attachment 1 I Page 20 of 46 Existing POTS 5.2 Proposed POTS 5.2 5.2 Spent Fuel Storage 5.2 Spent Fuel Storage A. The K ett of the fuel in the spent Spent fuel shall not be stored in the fuel storage pool shall be less spent fuel pool.

than or equal to 0.95.

B. Spent fuel storage racks may be moved (only) in accordance with written procedures which ensure that no rack modules are moved over fuel assemblies.

C. The number of spent fuel assemblies stored in the spent fuel pool shall not exceed 3353.

0. The maximum core geometry infinite lattice multiplication factor of any segment of the fuel assembly stored in the spent fuel storage pool or the new fuel storage facility shall be less than or equal to 1.31 at 20°C.

Basis POTS 5.2 describes design features associated with fuel storage in the SFP. After all spent fuel is removed from the SFP the existing requirements in POTS 5.2 are no longer applicable and may be deleted. A new design feature for the SFP is proposed to state that spent fuel shall not be stored in the SFP , which reflects the condition upon which this proposed amendment is predicated.

BVY 17-004 I Attachment 1 I Page 21 of 46 POTS SECTION 6.0, ADMINISTRATIVE CONTROLS Existing POTS 6.0 and 6.1 Proposed POTS 6.0 and 6.1 6.0 ADMINISTRATIVE CONTROLS 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1 Deleted A. The plant manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during absences.

B. The plant manager or designee shall approve, prior to implementation, each proposed test, experiment, or modification to systems or equipment that affect nuclear safety.

C. The shift supervisor shall be responsible for the shift command function .

Basis:

POTS Sections 6.1.A and 6.1.B provide a description and requirements regarding certain key operational management responsibilities. The proposed change is to delete these sections from the POTS and relocate them to the VY QAPM administrative controls verbatim with the exception that references to the "plant manager" position will be revised to "manager responsible for overall operational activities" in order to conform to the description of the corresponding management position in Section A, "Management," of the existing VY QAPM. This does not change any requirements, qualifications, or responsibilities of the individual in this position, and is strictly an administrative change. This change is consistent with previously approved changes to the position titles in the TS to allow facility specific titles to be identified in licensee-controlled documents such as the QAPM (Reference 16).

NRC Administrative Letter 95-06 (Reference 6) provides a discussion concerning the relocation of Technical Specification administrative controls to a quality assurance (QA) program. The NRC considers relocating these requirements to the quality assurance program acceptable because of the controls imposed by Appendix B to 10 CFR Part 50 , the existence of an NRC approved quality assurance program, and the quality assurance program change control process in

§ 50.54(a). After these administrative controls are incorporated into the VY QAPM, any future changes are controlled in accordance with§ 50.54(a) . This will provide adequate control for the facility with all spent fuel located within the ISFSI. Therefore , the proposed relocation is acceptable.

Existing POTS 6.1.C requires the shift supervisor to be responsible for the shift command function . This requirement is proposed to be eliminated. As described in the existinQ POTS

BVY 17-004 I Attachment 1 I Page 22 of 46 6.2.8 , the shift command function is focused on operations involving the storage or movement of spent nuclear fuel within the SFP. After all of the spent fuel is permanently removed from the SFP, the need for the shift supervisor and shift command function for spent fuel management no longer exists. The position of shift supervisor described in TS 6.1.C is a holdover from the control room function of supervising multiple functions of an operating nuclear power plant. With the limited requirements for supervision of the passive dry fuel storage system utilized within the ISFSI or with respect to the decommissioning of the former power generation facility, the shift supervisor position and shift command function are no longer required, and the proposed deletion of TS 6.1.C is acceptable.

Administrative controls proposed to be relocated to the VY QAPM are as shown in Attachment 4 to this letter.

Existing POTS 6.2 and 6.2.A Proposed POTS 6.2 and 6.2.A 6.2 ORGANIZATION 6.2 Deleted A. Onsite and Offsite Organizations Organizations shall be established for facility staff and corporate management. These organizations shall include the positions for activities aft ecting sat ety of the nuclear fuel.

1. Lines of authority, responsibility ,

and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating organizational positions. These relationships shall be documented and updated, as appropriate, in the form of organizational charts ,

functional descriptions of departmental responsibilities and relationships , and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Quality Assurance Program Manual. The plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications shall be documented in the Technical Requirements Manual.

BVY 17-004 I Attachment 1 I Page 23 of 46 (Continued)

2. The plant manager shall be responsible for overall facility safe operation and shall have control over those on-site activities necessary for safe storage and maintenance of the nuclear fuel.

3. A specified corporate officer shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure sate management of nuclear fuel.

4. The individuals who train the Certified Fuel Handlers, carry out health physics, or perform quality assurance functions may report to the appropriate on-site manager; however, these individuals shall have sufficient organizational freedom to ensure their ability to perform their assigned functions.

Basis:

The administrative controls describe organizational lines of authority, responsibilities , and requirements for organizational freedom for certain personnel including those performing health physics or quality assurance functions are provided sufficient organizational freedom . The proposed change is to delete POTS Section 6.2.A from the POTS and relocate them to the VY QAPM administrative controls, without altering these basic administrative controls .

Providing onsite and offsite organization descriptions in the QAPM is consistent with NRC Administrative Letter 95-06 (Reference 6), which provides a discussion concerning the relocation of Technical Specification administrative controls to a quality assurance (QA) program . The NRC considers relocating these requirements to the quality assurance program acceptable because of the controls imposed by Appendix B to 10 CFR Part 50, the existence of an NRC approved quality assurance program, and the quality assurance program change control process in

§ 50.54(a). After these administrative controls are incorporated into the VY QAPM , any future changes are controlled in accordance with§ 50.54(a). This will provide adequate control for the facility with all spent fuel located within the ISFSI . Therefore, the proposed relocation is acceptable.

The administrative controls provided in the existing POTS 6.2.A will be incorporated in the VY QAPM verbatim, with the following exceptions:

POTS 6.2.A.1 references to Technical Specifications" and "Technical Requirements Manual" (TRM) will be deleted . This is an administrative chanqe, since once the

BVY 17-004 I Attachment 1 I Page 24 of 46 requirements are incorporated into the QAPM these references would no longer be valid.

The reference to the TRM is removed as unnecessarily restrictive, since this change is consistent with previously approved changes to the TS to allow facility specific titles to be identified in licensee-controlled documents such as the TRM or QAPM (Reference 16).

POTS 6.2.A.2 reference to "plant manager" will be revised to "manager responsible for overall operational activities" in order to conform to the description of the corresponding management position in Section A, "Management," of the existing VY QAPM. This does not change any requirements , qualifications, or responsibilities of the individual in this position, and is strictly an administrative change. This change is consistent with previously approved changes to the position titles in the TS to allow facility specific titles to be identified in licensee-controlled documents such as the QAPM (Reference 16).

POTS 6.2.A.4 reference to individuals who "train the Certified Fuel Handlers" will not be incorporated into the QAPM. After all fuel is stored within the ISFSI and storage of spent fuel in the SFP is prohibited, there will no longer be an organizational need for Certified Fuel Handlers or associated training program , therefore this change will have no impact on continued safe storage and maintenance of spent fuel stored in the ISFSI, and is acceptable.

Administrative controls proposed to be relocated to the VY QAPM are as shown in Attachment 4 to this letter.

Existing POTS 6.2.B Proposed POTS 6.2.B B. Facility Staff Deleted The facility staff organization shall include the following:

1. Each duty shift shall be composed of at least one sh ift supervisor and one Non-certified Operator. The Non-certified Operator position may be filled by a Certified Fuel Handler.

2. At least one person qualified to stand watch in the control room (Non-certified Operator or Certified Fuel Handler) shall be present in the control room when nuclear fuel is stored in the spent fuel pool.

3. All fuel handling operations shall be directly supervised by a Certified Fuel Handler.

BVY 17-004 I Attachment 1 I Page 25 of 46 (Continued)

4. Shift crew composition shall meet the requirements stipulated herein.

Shift crew composition may be less than the minimum requirement of Specification 6.2.8.1 for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on-duty shift crew members, provided immediate action is taken to restore the shift crew composition to within the minimum requirements and all of the following conditions are met:

a. no fuel movements are in progress; and

b. no movement of loads over fuel are in progress; and

c. no unmanned shift positions during shift turnover shall be permitted while the shift crew is less than the minimum.

5. An individual qualified in radiation protection procedures shall be present on-site during the movement of fuel and during the movement of loads over fuel.

6. Deleted

7. The shift supervisor shall be a Certified Fuel Handler.

8. Deleted Basis POTS 6.2.8 is proposed to be deleted in its entirety, and to not be relocated to the QAPM. These administrative controls pertain to the facility staff organization and requirements when spent fuel is stored or moved within the SFP. Once all spent fuel is located within the ISFSI and spent fuel storage in the SFP is prohibited, it is not necessary for QAPM administrative controls to include these. Therefore, the deletion of TS 6.2.8 after the fuel has been moved from the spent fuel pool to the ISFSI will have no impact on sate storage and maintenance of spent fuel in the ISFSI and is acceptable.

BVY 17-004 I Attachment 1 I Page 26 of 46 Existing POTS 6.2.C Proposed POTS 6.2.C C. Facility Staff Qualifications Deleted

1. Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI/ANS 3.1-1978 for comparable positions with exceptions specified in the Quality Assurance Program Manual (QAPM).

2. An NRC approved training and retraining program for Certified Fuel Handlers shall be maintained .

Basis PDTS 6.2.C.1 will be relocated , verbatim to the QAPM administrative controls. NRC Administrative Letter 95-06 (Reference 6) provides a discussion concerning the relocation of Technical Specification administrative controls to a quality assurance (QA) program. The NRC considers relocating these requirements to the quality assurance program acceptable because of the controls imposed by Appendix 8 to 10 CFR Part 50, the existence of an NRC approved quality assurance program, and the quality assurance program change control process in§ 50.54(a).

After these administrative controls are incorporated into the VY QAPM, any future changes are controlled in accordance with§ 50 .54(a). This will provide adequate control for the facility with all spent fuel located within the ISFSI. Providing the requirements for facility staff qualifications in the QAPM is consistent with Administrative Letter 95-06 and will have no impact on safe storage and maintenance of spent fuel in the ISFSI, therefore the proposed relocation is acceptable.

Existing PDTS 6.2.C.2 specifies that Certified Fuel Handler training programs shall be maintained. As described in the existing PDTS 6.2.B, the shift command function is focused on operations involving the storage or movement of spent nuclear fuel within the SFP. Following the transfer of all spent fuel to the ISFSI , storage of spent fuel in the SFP will be prohibited upon implementation of this amendment, thus there will no longer be a need for Certified Fuel Handlers or the associated training programs. Therefore, this proposed deletion will have no impact on safe storage and maintenance of spent fuel in the ISFSI and is acceptable.

Administrative controls proposed to be relocated to the VY QAPM are as shown in Attachment 4 to this letter.

BVY 17-004 I Attachment 1 I Page 27 of 46 Existing POTS 6.3 and 6.4 Proposed POTS 6.3 and 6.4 6.3 Deleted 6.3 Deleted 6.4 PROCEDURES 6.4 Deleted Written procedures shall be established, implemented, and maintained covering the following activities:

A. Normal startup, operation and shutdown of systems and components needed for the safe storage of nuclear fuel.

B. Fuel handling operations C. Actions to be taken to correct specific and foreseen potential malfunctions of systems or components needed for the sate storage of nuclear fuel.

D. Emergency conditions involving potential or actual release of radioactivity.

E. Preventive and corrective maintenance operations which could have an effect on the safety of the nuclear fuel.

F. Surveillance and testing requirements.

G. Fire protection program implementation.

H. Process Control Program in-plant implementation.

I. Off-Site Dose Calculation Manual implementation .

BVY 17-004 I Attachment 1 I Page 28 of 46 Basis POTS 6.4 requires written procedures to be established , implemented, and maintained for certain activities. The proposed change is to delete POTS 6.4 from the POTS in its entirety, and to relocate the requirements to the VY QAPM administrative controls with the exception of POTS 6.4.B which involves fuel handling operations in the SPF. As discussed above, following the transfer of the spent fuel to the ISFSI , the proposed change to POTS 5.2 will prohibit the storage of spent fuel in the SFP.

After these administrative controls are incorporated into the VY QAPM, any future changes are controlled in accordance with§ 50.54(a). This will provide adequate control for the facility with all spent fuel located within the ISFSI. The relocation of administrative controls for procedures to the QAPM is consistent with NRC Administrative Letter 95-06 (Reference 6) and will have no impact on safe storage and maintenance of spent fuel in the ISFSI , and therefore is acceptable.

Administrative controls proposed to be relocated to the VY QAPM are as shown in Attachment 4 to this letter.

Existing POTS 6.5.A.4.d.1 Proposed POTS 6.5.A.4.d.1

1. Be under the surveillance, as specified in 1. Be under the surveillance, as specified in the RWP or equivalent, while in the area, the RWP or equivalent, while in the area, of an individual qualified in radiation of an individual qualified in radiation protection procedures, equipped with a protection procedures, equipped with a radiation monitoring device that radiation monitoring device that continuously displays radiation does continuously displays radiation dose rates in the area; who is responsible for rates in the area; who is responsible for controlling personnel exposure within the controlling personnel exposure within the area, or area, or Basis POTS 6.5 provides the controls that shall be applied to high radiation areas. POTS 6.5 will remain in the POTS, and there are no proposed changes associated with these controls. The proposed change is to replace with word "does" with the word "dose" in POTS Section 6.5.A.4.d .1 as shown above. This is an administrative change to correct a typographical error.

BVY 17-004 I Attachment 1 I Page 29 of 46 Existing POTS 6.6 Proposed POTS 6.6 6.6 REPORTING REQUIREMENTS Deleted The following reports shall be submitted in accordance with 10 CFR 50.4.

A. Deleted B. Deleted C. Deleted D. Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the facility shall be submitted by May 15 of each year and in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the facility. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM) and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.

E. Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report covering the operation of the facility during the previous calendar year shall be submitted by May 15 of each year.

The report shall include summaries ,

interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM) , and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

BVY 17-004 I Attachment 1 I Page 30 of 46 (Continued)

The Annual Radiological Environmental Operating Report shall include summarized and tabulated results of all radiological environmental samples taken during the report period pursuant to the table and figures in the ODCM. In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.

Basis POTS 6.6 concerns reporting requirements . The proposed change is to delete POTS 6.6 from the POTS in its entirety, and relocate the requirements verbatim to the QAPM. After these administrative controls are incorporated into the QAPM, any future changes are controlled in accordance with§ 50.54(a). This will provide adequate control for the facility with all spent fuel located within the ISFSI. The relocation of administrative controls for reporting requirements to the QAPM is consistent with NRC Administrative Letter 95-06 (Reference 6) and will have no impact on safe storage and maintenance of spent fuel in the ISFSI, and therefore is acceptable.

Administrative controls proposed to be relocated to the VY QAPM are as shown in Attachment 4 to this letter.

Existing POTS 6.7 Proposed POTS 6.7 6.7 PROGRAMS AND MANUALS Deleted The following programs shall be established ,

implemented and maintained:

A. Deleted B. OFF-SITE DOSE CALCULATION MANUAL (ODCM)

An Off-Site Dose Calculation Manual shall contain the current methodology and parameters used in the calculation of off-site doses due to radioactive gaseous and liquid effluents for the purpose of demonstrating compliance with 10 CFR 50, Appendix I, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and in the conduct of the environmental radiological monitoring program.

BVY 17-004 I Attachment 1 I Page 31 of 46 (Continued)

The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities and descriptions of the information that should be included in the Radioactive Effluent Release Report and the Annual Radiological Environmental Operating Report required by Specification 6.6.D and Specification 6.6.E, respectively.

1. Licensee initiated changes to the ODCM:

a. Shall be submitted to the Commission in the Radioactive Effluent Release Report for the period in which the change(s) was made effective. This submittal shall contain:

i. Sufficient information to support the change together with appropriate analyses or evaluations justifying the change(s) and ii. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50, and do not adversely impact the accuracy or reliability of effluent dose or setpoint calculations.

b. Shall become effective upon approval by the plant manager.

c. Shall be submitted to the Commission in the form of a legible copy of the aft ected pages of the ODCM as a part of

BVY 17-004 I Attachment 1 I Page 32 of 46 (Continued) or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g. , month/year) the change was implemented.

C. Deleted D. Radioactive Effluent Controls Program This program conforming to 10 CFR 50.36a provides for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by operating procedures, and shall include remedial actions to be taken whenever the program limits are exceeded . The program shall include the following elements:

a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM ;

b. Limitations on the concentrations of radioactive material released in liquid effluents from the site to unrestricted areas, conforming to 1O times the concentration values in Appendix B, Table 2, Column 2, to 10 CFR 20.1001 -

20.2402;

BVY 17-004 I Attachment 1 I Page 33 of 46 (Continued)

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents pursuant to 10 CFR 20.1302 and with the methodology and parameters in the ODCM;

d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from the facility to unrestricted areas, conforming to 10 CFR 50, Appendix I;

e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days;

f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2 percent of the guidelines for the annual dose or dose commitment ,

conforming to 10 CFR 50, Appendix I;

g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the site boundary shall be limited to the following:

BVY 17-004 I Attachment 1 I Page 34 of 46 (Continued)

1. For noble gases: less than or equal to a dose rate of 500 mrems/yr to the total body and less than or equal to a dose rate of 3000 mrems/yr to the skin , and

2. For iodine-131 , iodine-133, tritium , and for all radionuclides in particulate form with half lives greater than 8 days: less than or equal to a dose rate of 1500 mrems/yr to any organ;

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from the facility to areas at or beyond the site boundary, conforming to 10 CFR 50, Appendix I;

i. Limitations on the annual and quarterly doses to a member of the public from iodine-131 ,

iodine-133, tritium , and all radionuclides in particulate form with half lives greater than 8 days in gaseous effluents released from the facility to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and

j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

E. TECHNICAL SPECIFICATIONS (TS)

BASES CONTROL PROGRAM This prog ram provides a means for processing changes to the Bases of these Technical Specifications.

BVY 17-004 I Attachment 1 I Page 35 of 46 (Continued)

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following :

1. A change in the TS incorporated in the license, or

2. A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

d. Proposed changes that meet the criteria of Specification 6.7. E.b above shall be reviewed and approved by the N RC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50 .71(e).

Basis These specifications require the establishment, implementation and maintenance of specified programs and manuals. POTS 6.7.B specifies how to document, review, and approve changes to the OOCM . The proposed change is to delete POTS 6.7.B from the POTS and relocate the requirements verbatim to the QAPM , with the following exceptions:

References to Specification 6.6.0 and Specification 6.6.E are deleted. This is an administrative change, since these will no longer reside in the POTS once relocated to QAPM , and the references are not necessary to describe the administrative requirements.

A reference to the "plant manager" position will be revised to "manager responsible for overall operational activities" in order to conform to the description of the corresponding management position in Section A, "Management," of the existing QAPM. This does not change any requirements, qualifications, or responsibilities of the individual in this position, and is strictly an administrative change. This change is consistent with previously approved changes to the position titles in the TS to allow facility specific titles to be identified in licensee-controlled documents such as the QAPM (Reference 16).

BVY 17-004 I Attachment 1 I Page 36 of 46 POTS 6.7.0 specifies administrative requirements for the program to control radioactive effluents, and for maintaining doses to the public to within the specified limits. The proposed change is to delete TS 6.7.D from the POTS and relocate the requirements to the QAPM with the following exceptions:

References to iodine-131 and iodine-133 in 6.7.0.g.2 and 6.7.0.i will not be relocated to the QAPM due to the radioactive decay and short half-lives (approximately 8 days and 20.83 hours9.606481e-4 days 0.0231 hours 1.372354e-4 weeks 3.15815e-5 months , respectively) and time since permanent cessation of reactor operation . This is consistent with changes to the OOCM implemented under 10 CFR 50.59.

6.7.0.g.1 and 6.7.D.h will not be relocated to the QAPM since after all spent fuel is transferred to the ISFSI and contained within dry storage casks, there will no longer be a requirement to monitor for noble gases released from the facility.

POTS 6.7.E describes the TS Bases Control Program , and is proposed to be deleted in its entirety from the POTS. As described in this table and shown in Attachment 2 to this letter, all of the Bases in the existing POTS are being eliminated with the proposed changes to the corresponding sections. Since all the Bases will be deleted, there will no longer be a need fo r a TS Bases control program, and it will not be incorporated into the QAPM . Therefore, the proposed deletion of TS 6.7.E is acceptable.

After the administrative controls are incorporated into the QAPM, any future changes are controlled in accordance with§ 50.54(a). This will provide adequate control for the facility with all spent fuel located within the ISFSI. The relocation of these administrative requirements to the QAPM is consistent with NRC Administrative Letter 95-06 (Reference 6) and will have no impact on safe storage and maintenance of spent fuel in the ISFSI , and therefore is acceptable.

Administrative controls proposed to be relocated to the VY QAPM are as shown in Attachment 4.

BVY 17-004 I Attachment 1 I Page 37 of 46

3. REGULATORY EVALUATION 3.1 Applicable Regulatory Requirements 10 CFR 50.2, Definitions. Safety-Related Structures. Systems and Components 10 CFR 50.2 defines safety-related structures, systems, and components (SSCs) as those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary (2) The capability to shut down the reactor and maintain it in a safe shutdown condition ; or (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in § 50.34(a){1) or§ 100.11 of 10 CFR , as applicable.

10 CFR 50.36. Technical Specifications In 10 CFR 50.36, the Commission established its regulatory requirements related to the content of Technical Specifications (TS). In doing so, the Commission placed emphasis on those matters related to the prevention of accidents and mitigation of accident consequences ; the Commission noted that applicants were expected to incorporate into their TS "those items that are directly related to maintaining the integrity of the physical barriers designed to contain radioactivity" (Statement of Consideration, "Technical Specification for Facility Licenses; Safety Analysis Reports ," 33 FR 18610, December 17, 1968). Pursuant to§ 50.36, TS are required to include items in the following five categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs) ; (3) surveillance requirements (SRs) ; (4) design features ; and (5) administrative controls. However, the rule does not specify the particular requirements to be included in a plant's TS.

The final Commission Policy Statement established four criteria to define the scope of equipment and parameters to be included in the improved Standard Technical Specifications. These criteria were developed for licenses authorizing operation (i.e .,

operating reactors) and focused on instrumentation to detect degradation of the reactor coolant system pressure boundary, process variables and equipment, design features ,

or operating restrictions that affect the integrity of fission product barriers during design bases accidents or transients . A fourth criterion refers to the use of operating experience and probabilistic risk assessment to identify and include in the Technical Specifications structures, systems, and components (SSCs) shown to be significant to public health and safety. These criteria, which were subsequently codified in changes to Section 36 of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR 50.36) {60 FR 36953), also pertain to the Technical Specification requirements for safe storage of spent fuel. A general discussion of these considerations is provided below.

BVY 17-004 / Attachment 1 I Page 38 of 46 Criterion 1 of 10 CFR 50.36{c)(2){ii)(A) states that Technical Specification limiting conditions for operation must be established for "installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary." Since the certifications of § 50.82(a)(1) have been docketed for VY, under§ 50.82(a)(2) the VY 10 CFR Part 50 license no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel , therefore this criterion is not applicable.

Criterion 2 of 10 CFR 50.36{c)(2)(ii)(B) states that Technical Specification limiting conditions for operation must be established for a "process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." The purpose of this criterion is to capture those process variables, design features , or operating restrictions that involve an initial condition assumed in the design basis accident and transient analyses, and which are monitored and controlled during power operation. Since VY is no longer licensed to operate, and the existing design basis accident is a fuel handling accident predicated on the storage of spent fuel in the SFP, with all spent fuel stored in the ISFSI, there are no remaining design basis accidents which are credible, and therefore this criterion is not applicable.

Criterion 3 of 10 CFR 50.36{c)(2)(ii)(C) states that Technical Specification limiting conditions for operation must be established for SSCs that are part of the primary success path and which function or actuate to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The intent of this criterion is to capture into Technical Specifications only those SSCs that are part of the primary success path of a safety sequence analysis. Also captured by this criterion are those support and actuation systems that are necessary for items in the primary success path to successfully function. The primary success path of a safety sequence analysis consists of the combination and sequences of equipment needed to operate (including consideration of the single failure criterion), so that the plant response to design basis accidents and transients limits the consequences of these events to within the appropriate acceptance criteria.

Since fuel will have been removed from the SFP at the VY facility prior to implementation of this amendment, this criterion is not applicable.

Criterion 4 of 10 CFR 50.36(c)(2)(ii)(D) states that Technical Specification limiting conditions for operation must be established for SSCs that operating experience or probabilistic risk assessment has shown to be significant to public health and safety. The intent of this criterion is that risk insights and operating experience be factored into the establishment of Technical Specification limiting conditions for operation. Since fuel will have been removed from the spent fuel pool at the VY facility prior to implementation of this amendment, this criterion is not applicable.

BVY 17-004 I Attachment 1 I Page 39 of 46 Addressing administrative controls, 10 CFR 50.36(c)(5) states that they "... are the provisions relating to organization and management, procedures, recordkeeping , review and audit, and reporting necessary to assure operation of the facility in a safe manner." The particular administrative controls to be included in the TS, therefore, are the provisions that the Commission deems essential for the safe operation of the facility that are not already covered by other regulations. Accordingly, the NRC staff determined that administrative control requirements that are not specifically required under Section 50.36(c)(5) ,

and which are not otherwise necessary to obviate the possibility of an abnormal situation or an event giving rise to an immediate threat to the public health and safety, may be relocated to more appropriate documents (e.g. , Quality Assurance Program Manual, Technical Requirements Manual, Security Plan, or Emergency Plan) , which are subject to regulatory controls. Similarly, while the required content of TS administrative controls is specified in 10 CFR 50.36(c)(5) ,

particular details may be relocated to licensee-controlled documents, where other regulations including § 50.59 and Appendix B to 10 CFR Part 50 provide adequate regulatory control.

10 CFR 50.36(c)(6) , "Decommissioning ," applies only to nuclear power reactor facilities that have submitted the certifications required by§ 50.82(a)(1 ). For such facilities , Technical Specifications involving safety limits, limiting safety system settings, and limiting control system settings; limiting conditions for operation; surveillance requirements; design features; and administrative controls will be developed on a case-by-case basis.

Administrative Letter (AL) 95-06 The Quality Assurance Program Manual is an appropriate candidate for relocations of administrative controls due to the controls imposed by such regulations as Appendix B to 10 CFR Part 50, the existing NRG-approved QA plans and commitments to industry QA standards, and the established QA program change control process of 10 CFR 50.54(a) .

NRC Administrative Letter (AL) 95-06, "Relocation of Technical Specification Administrative Controls Related to Quality Assurance" (Reference 6) , provides guidance to licensees requesting amendments that relocate administrative controls to NRC-approved QA program descriptions, where subsequent changes are controlled pursuant to 10 CFR 50.54(a). AL 95-06 provides specific guidance in the areas of:

(1) independent safety engineering group, (2) reviews and audits, (3) procedure review process, and (4) records and record retention .

Some relocation actions are specifically discussed in AL 95-06, while others are similar in nature. Relocations not specifically discussed in AL 95-06 were assessed with respect to the appropriateness of the relocation . Editorial changes are allowed without basis by 10 CFR 50.54{a)(3) .

BVY 17-004 I Attachment 1 I Page 40 of 46 This proposed amendment deletes the portions of the existing VY POTS that are no longer applicable to a permanently defueled facility with all irradiated fuel in dry storage within an ISFSI, while modifying the remaining portions to correspond to the SAFSTOR decommissioning condition.

10 CFR 50.51. Continuation of License 10 CFR 50.51 (b) states "Each license for a facility that has permanently ceased operations, continues in effect beyond the expiration date to authorize ownership and possession of the production or utilization facility, until the Commission notifies the licensee in writing that the license is terminated. During such period of continued effectiveness the licensee shall--

(1) "Take actions necessary to decommission and decontaminate the facility and continue to maintain the facility, including, where applicable, the storage, control and maintenance of the spent fuel, in a safe condition , and (2) Conduct activities in accordance with all other restrictions applicable to the facility in accordance with the NRC regulations and the provisions of the specific 10 CFR part 50 license for the facility."

10 CFR 50.82. Termination of License 10 CFR 50.82(a)(2) states "Upon docketing of the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel , or when a final legally effective order to permanently cease operations has come into effect, the 10 CFR part 50 license no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel."

3.2 No Significant Hazards Consideration Determination In accordance with 10 CFR 50.90, Entergy Nuclear Operations, Inc. (ENO) is proposing an amendment to Renewed Facility Operating License DPR-28 for Vermont Yankee Nuclear Power Station (VY). The proposed amendment would revise the Operating License and revise the Permanently Defueled Technical Specifications (POTS) to reflect removal of all spent nuclear fuel from the spent fuel pool (SFP) and its transfer to dry cask storage within an Independent Spent Fuel Storage Installation (ISFSI) and the relocation of various requirements to the VY Quality Assurance Program Manual (QAPM).

By letter dated January 12, 2015 (Reference 1), ENO submitted certifications for permanent cessation of reactor operations and permanent removal of fuel from the reactor vessel pursuant to 10 CFR 50.82(a)(1 ). Therefore as specified in 10 CFR 50.82(a)(2) , the 10 CFR Part 50 license for VY no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel. By letter dated April 12, 2017 (Reference 2) , ENO provided notification that off-load of the spent fuel pool (SFP) and transfer to the ISFSI is expected to be completed in late 2018 , provided the

BVY 17-004 I Attachment 1 I Page 41 of 46 criteria for transfer are met, including certain regulatory approvals. In support of this condition , a revision to the VY Operating License and associated POTS is proposed to comport with the requirements for a facility configuration with all spent nuclear fuel in dry storage within an ISFSI.

The discussion below addresses each of these criteria and demonstrates that the proposed amendment does not constitute a significant hazard.

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed amendment would modify the VY Renewed Facility Operating License (Operating License) and Permanently Defueled Technical Specifications (POTS), or Technical Specifications (TS), by deleting the portions of the Operating License and POTS that are no longer applicable to a facility with no spent nuclear fuel stored in the SFP, while modifying the remaining portions to correspond to all nuclear fuel stored within an ISFSI. This amendment will be implemented within 60 days following ENO's notification to the NRC that all spent fuel assemblies have been transferred out of the SFP and placed in dry storage within the ISFSI.

The definition of safety-related structures, systems, and components (SSCs) in 10 CFR 50.2 states that safety-related SSCs are those relied on to remain function al during and following design basis events to assure:

(1) The integrity of the reactor coolant boundary; (2) The capability to shutdown the reactor and maintain it in a safe shutdown condition; or (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in 10 CFR 50.34(a)(1) or 100.11.

The first two criteria (integrity of the reactor coolant pressure boundary and safe shutdown of the reactor) are not applicable to a plant in a permanently defueled condition . The third criterion is related to preventing or mitigating the consequences of accidents that could result in potential offsite exposures exceeding limits. However, after all nuclear spent fuel assemblies have been transferred to dry cask storage within an ISFSI , none of the SSCs at VY are required to be relied on for accident mitigation. Therefore, none of the SSCs at VY meet the definition of a safety-related SSC stated in 10 CFR 50.2. The proposed deletion of requirements in the POTS does not affect systems credited in any accident analysis at VY.

Section 6 of the VY Defueled Safety Analysis Report (DSAR) described the design basis accidents (DBAs) related to the SFP. These postulated accidents

BVY 17-004 I Attachment 1 I Page 42 of 46 are predicated on spent fuel being stored in the SFP. With the removal of the spent fuel from the SFP, there are no remaining spent fuel assemblies to be monitored and there are no credible accidents that require the actions of a Certified Fuel Handler, Shift Manager, or a Non-certified Operator to prevent occurrence or mitigate the consequences of an accident.

The proposed changes do not have an adverse impact on the remaining decommissioning activities or any of their postulated consequences.

The proposed changes related to the relocation of certain administrative requirements do not affect operating procedures or administrative controls that have the function of preventing or mitigating any accidents applicable to the safe management of irradiated fuel or decommissioning of the facility.

Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response : No.

The proposed changes eliminate the operational requirements and certain design requirements associated with the storage of the spent fuel in the SFP, and relocate certain administrative controls to the Quality Assurance Program Manual or other licensee-controlled process.

After the removal of the spent fuel from the SFP and transfer to the ISFSI, there are no spent fuel assemblies that remain in the SFP. Coupled with a prohibition against storage of fuel in the SFP, the potential for fuel related accidents is removed. The proposed changes do not introduce any new failure modes.

Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response : No.

The removal of all spent nuclear fuel from the SFP into storage in casks within an ISFSI , coupled with a prohibition against future storage of fuel within the SFP, removes the potential for fuel related accidents.

The design basis and accident assumptions within the VY DSAR and the POTS relating to safe management and safety of spent fuel in the SFP are no longer

BVY 17-004 I Attachment 1 I Page 43 of 46 applicable. The proposed changes do not affect remaining plant operations, systems, or components supporting decommissioning activities.

The requirements for systems, structures, and components (SSCs) that have been removed from the VY POTS are not credited in the existing accident analysis for any applicable postulated accident; and as such, do not contribute to the margin of safety associated with the accident analysis.

Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

Based on the above, ENO concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c) , and, accordingly, a finding of "no significant hazards consideration" is justified .

3.3 Precedents This proposed amendment is consistent with recently approved Amendments issued for Zion Nuclear Power Station, Units 1 and 2, on January 14, 2015 (Reference 14), an Amendment issued for Kewaunee Power Station on June 7, 2017 (Reference 15) , and an Amendment to Crystal River Nuclear Plant on June 27, 2017 (Reference 17) to revise the license and technical specifications to reflect removal of all spent fuel from the SFP and transfer to dry cask storage within an ISFSI.

3.4 Conclusion Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations , and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

BVY 17-004 I Attachment 1 I Page 44 of 46

4. ENVIRONMENTAL CONSIDERATIONS This proposed amendment is to the VY license issued under 10 CFR Part 50 and includes changes to requirements involving the installation or use of a facility component located within the protected area and changes to recordkeeping, reporting, or administrative procedures or requirements. As such, ENO has evaluated this proposed amendment against the criteria for identification of licensing and regulatory actions requiring an environmental assessment in accordance with 10 CFR 51.21 , and determined that it meets the eligibility criteria for categorical exclusions set forth in

§ 51 .22(c)(9) and§ 51.22(c)(10)(ii). With respect to the criteria set forth in

§ 51.22(c)(9), this determination is made as follows:

(i) The amendment involves no significant hazards consideration.

As described in Section 3.2 above, the proposed change involves no significant hazards consideration.

(ii) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

The proposed amendment does not involve any physical alterations to the facility configuration that could lead to a change in the type or amount of effluent that may be released offsite.

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

The proposed amendment does not involve any physical alterations to the facility configuration and does not involve any changes to regulatory requirements or programs and procedures related to controls for limiting radiation exposure that could lead to a significant increase in individual or cumulative occupational radiation exposure.

Based on the above, ENO concludes that the proposed amendment involves changes that meets the eligibility criteria for categorical exclusion as set forth in § 51 .22(c)(9) or

§ 51.22(c)(1 O)(ii). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this proposed amendment.

BVY 17-004 I Attachment 1 I Page 45 of 46

5. REFERENCES

1. Letter, Entergy Nuclear Operations, Inc. to USNRC , "Certifications of Permanent Cessation of Power Operations and Permanent Removal of Fuel from the Reactor Vessel," BVY 15-001 , dated January 12, 2015 (ML15013A426)

2. Letter, Entergy Nuclear Operations, Inc. to USNRC, "Notification of Schedule Change for Dry Fuel Loading Campaign ," BVY 17-013, dated April 12, 2017 (ML17104A050)

3. Letter, Entergy Nuclear Operations, Inc. to USNRC, "Application for Order Consenting to Direct and Indirect Transfers of Control of Licenses and Approving Conforming License Amendment and Notification of Amendment to Decommissioning Trust Agreement ," BVY 17-005, dated February 9, 2017 (ML17045A140)

4. Letter, Entergy Nuclear Operations, Inc. to USNRC, "License Amendment Request - Independent Spent Fuel Storage Installation (ISFSI) Emergency Plan and Emergency Action Level Scheme," BVY 17-009, dated May 15, 2017 (ML17139D261)

5. Letter, Entergy Nuclear Operations , Inc. to USNRC , "I ndependent Spent Fuel Storage Installation Physical Security Plan , Revision 0, Proposed Change No . 315," BVY 17-003, dated March 29, 2017 (ML17117A421)

6. NRC Administrative Letter 95-06, "Relocation of Technical Specification Administrative Controls Related to Quality Assurance ," dated December 12, 1995 (ADAMS Legacy Library No. 9512060318)

7. Letter, Entergy Nuclear Operations, Inc. to USNRC , "Post Shutdown Decommissioning Activities Report, " BVY 14-078, dated December 19, 2014 (ML14357A110)

8. Letter, Entergy Nuclear Operations , Inc. to USNRC, "Update to Irradiated Fuel Management Program Pursuant to 10 CFR 50.54(bb) ," BVY 14-085, dated December 19, 2014 (ML14358A251)

9. Letter, USN RC to Dominion Energy Kewaunee, Inc. , "Kewaunee Power Station-Issuance of Amendment to Renewed Facility Operating License Related to License Conditions Associated with Extended Operation (TAC No. MF1771 ),"dated June 23, 2014 (ML14008A297)

10. Letter, USN RC to Entergy Nuclear Operations, Inc. , "Vermont Yankee Nuclear Power Station - Issuance of Amendment to Renewed Facility Operating License Re:

License Condition 3.P and 3.0 changes (TAC No. ME8151)," NVY 13-046, dated April 17, 2013(ML13042A272)

BVY 17-004 I Attachment 1 I Page 46 of 46

11. Letter, USNRC to Vermont Yankee Nuclear Power Corporation, "Issuance of Order for Interim Safeguards and Security Compensatory Measures for -

Vermont Yankee Nuclear Power Station ," Order EA-02-026, dated February 25, 2002 (ML020510305)

12. Letter, USNRC to Holders of Licenses for Operating Power Reactors as Listed in the Enclosure, "Rescission or Partial Rescission of Certain Power Reactor Security Orders Applicable to Nuclear Power Plants," dated November 28, 2011 (ML111220447)

13. SECY-16-0142, Draft Final Rule - Mitigation of Beyond-Design-Basis Events, dated December 15, 2016 (ML16291A186), and SECY-16-0142: Enclosure 1 -

Final Rule: Mitigation of Beyond-Design-Basis Events, dated December 15, 2016 (ML16292A026)

14. Letter, USNRC to Zion Solutions LLC, "Issuance of Amendments Relating to the Unloaded Spent Fuel Pool for Zion Nuclear Power Station, Units 1 and 2 {TAC Nos. J52985 and J52986," License Amendment Nos.188 and 175, dated January 14, 2015(ML14295A716)

15. Letter, USNRC to Dominion Energy Kewaunee, Inc. , "Kewaunee Power Station -

Issuance of Amendment for Proposed Changes to License and Technical Specifications to Reflect Permanent Removal of Spent Fuel from Spent Fuel Pool (CAC No. L53079) ," dated June 7, 2017(ML17123A031)

16. Letter, USNRC to Entergy Nuclear Vermont Yankee , LLC, "Vermont Yankee Nuclear Power Station - Issuance of Amendment Re : Generic Position Titles and Other Administrative Changes (TAC No . MB6925) ," NVY 03-19, dated February 27, 2003 (ML030570728)

17. Letter, USNRC to Crystal River Nuclear Plant, "Crystal River Unit 3 Nuclear Generating Plant - Issuance of Amendment 255 for the License and Permanently Defueled Technical Specifications to Reflect Permanent Removal of Spent Fuel from the Spent Fuel Pools {TAC No. L53146) ," dated June 27, 2017 (ML17027A160)

BVY 17-004 Docket No. 50-271 Attachment 2 Vermont Yankee Nuclear Power Station Proposed Change 313 Markup of the Existing Operating License and Technical Specification Pages

1 ---------

NOTE: A License Transfer Application has been submitted separately from this request to transfer ownership of the facility from ENO to Northstar Nuclear Decommissioning, LLC.

Entergy Nuclear Vermont Yankee. LLC and Entergy Nuclear Operations, Inc.

(Vermont Yankee Nuclear Power Station)

Docket No. 50-271 Renewed Facility Operating License Renewed Operating License No. DPR-28 The U.S. Nuclear Regulatory Commission (NRC or the Commission) , having previously made the findings set forth in Facility Operating License No. DPR-28, dated February 28, 1973, has now found that:

a. This paragraph deleted by Amendment No. 263.

b. The facility is prohibited from operating the reactor in conformity with the application ,

as amended , the provisions of the Act, and the rules and regulations of the Commission ; and

c. There is reasonable assurance (i) that the activities authorized by this license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the rules and regulations of the Commission; and

d. Entergy Nuclear Vermont Yankee, LLC is financially qualified and Entergy Nuclear Operations, Inc. is technically and financially qualified to engage in the activities authorized by this license, in accordance with the rules and regulations of the Commission ; and

e. Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc.

have satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements" of the Commission's regulations; and

f. The issuance of this license will not be inimical to the common defense and security or to the health and safety of the public; and

g. After weighing the environmental , economic, technical and other benefits of the facility against environmental costs and considering available alternatives, the issuance of this license (subject to the conditions for Renewed Facility Operating License No. DPR-28 Amendment No. 263

protection of the environment set forth herein) is in accordance with 10 CFR Part 51 , of the Commission's regulations and all applicable requirements of said Part 51 have been satisfied;--aRG

h. Aotions have been identified and have been or will be taken with respeot to: (1) managing the effects of aging on the funotionality of structures and components that have been identified to require review under 10 GFR 64 .21 (a)(1) during the period of extended operation , and (2) time limited aging analyses that have been identified to require review under 10 GFR 64 .21(0), suoh that there is reasonable assurance that the activities authorized by this license will continue to be conduoted in accordance with the current licensing basis, as defined in 10 GFR 64 . ~ for the faoility , and that any ohanges made to the faoility's ourrent lioensing basis in order to comply 1nith 1O GFR 64 .29(a) are in aocordance with the

.A.st and the Commission's regulations.

Accordingly, Facility Operating License No. DPR-28, as amended , issued to Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. is superseded by Renewed Facility Operating License No. DPR-28 and is hereby amended in its entirety to read :

1. This renewed license applies to the Vermont Yankee Nuclear Power Station (the facility) , a single cycle, boiling water, light water moderated and cooled reactor, and associated electric generating equipment. The facility is located on Entergy Nuclear Vermont Yankee, LLC's site, in the Town of Vernon , Windham County, Vermont; and is described in the application as amended .

2. Subject to the conditions and requirements incorporated herein , the Commission hereby licenses:

A. Pursuant to Sections 104b of the Atom ic Energy Act of 1954, as amended (the Act), and 10 CFR Part 50 , "Licensing of Production and Utilization Facilities," Entergy Nuclear Vermont Yankee , LLC to possess and use, and Entergy Nuclear Operations, Inc., to possess and use the facility as a utilization facility at the designated location on the Entergy Nuclear Vermont Yankee, LLC site.

B. Entergy Nuclear Operations, Inc., pursuant to the Act and 10 CFR Part 70 , to possess at any time special nuclear material that was used as reactor fuel , in accordance with the limitations for storage and amounts requ ired for reactor operation as described in the Final Safety Analysis Report, as supplemented and amended.

C. Entergy Nuclear Operations, Inc., pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources that were used for reactor startup, sealed sources that were used for calibration of reactor instrumentation and are used in radiation monitoring equipment, and as fission detectors in amounts as requ ired .

Renewed Facility Operating License No. DPR-28 Amendment No. ~

D. Entergy Nuclear Operations, Inc., pursuant to the Act and 10 CFR Parts 30, 40 and 70 , to receive , possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form , for sample analysis or instrument calibration or associated with radioactive apparatus or components.

E. Entergy Nuclear Operations, Inc., pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not to separate, such byproduct and special nuclear material as may be produced by operation of the facility.

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40 , Section 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32of10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations , and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A. This paragraph deleted by Amendment No. 263.

B. Technical Specifications The Technical Specif ions contained in Appendix A, as revised through Amendment No. ~. are hereby incorporated in the license. Entergy Nuclear Operations, Inc. shall operate the facility in accordance with the Technical Specifications.

This paragraph deleted by Amendment No. XXX .

C. Reports Entergy Nuolear Operations, Ins. shall mal(e reports in aooordanoe with the requirements of the Teohnioal Speoifioations.

D. This paragraph deleted by Amendment No. 226.

E. Environmental Conditions Pursuant to the Initial Decision of the presiding Atomic Safety and Licensing Board issued February 27, 1973, the following conditions for the protection of the environment are incorporated herein:

1. This paragraph deleted by Amendment No. 206, October 22 , 2001 .

2. This paragraph deleted by Amendment 131 , 10/07/91 .

Renewed Facility Operating License No. DPR-28 Amendment No. 2-W, ~. ~. ~.~

3. This paragraph deleted by Amendment No. 206, October 22 , 2001 .

4. If harmful effects or evidence of irreversible damage in land or water ecosystems as a result of facility operation are detected by Entergy Nuclear Operations, lnc.'s environmental monitoring program , Entergy Nuclear Operations, Inc. shall provide an analysis of the problem to the Commission and to the advisory group for the Technical Specifications, and Entergy Nuclear Operations, Inc. thereafter will provide, subject to the review by the aforesaid advisory group, a course of action to be taken immediately to alleviate the problem .

5. Entergy Nuclear Operations, Inc. will grant authorized representatives of the Massachusetts Department of Public Health (MDPH) and Metropolitan District Commission (MDC) access to records and charts related to discharge of radioactive materials to the Connecticut River.

6. This paragraph deleted by Amendment No. 206, October 22, 2001 .

7. This paragraph deleted by Amendment No. 206, October 22, 2001 .

8. Entergy Nuclear Operations, Inc. will permit authorized representatives of the MDPH and MDC to examine the chemical and radioactivity analyses performed by Entergy Nuclear Operations, Inc.

9. Entergy Nuclear Operations, Inc. shall immediately notify MDPH , or an agency designated by MDPH , in the event concentrations of radioactive materials in liquid effluents, measured at the point of release from the Vermont Yankee facility , exceed the limit set forth in the facility Offsite Dose Calculation Manual. Entergy Nuclear Operations, Inc. will also notify MDPH in writing within 30 days following the release of radioactive materials in liquid effluents in excess of 10 percent of the limit set forth in the faci lity Offsite Dose Calculation Manual.

10. A report shall be submitted to MDPH and MDC by May 15 of each year of plant operation , specifying the total quantities of radioactive materials released to the Connecticut River during the previous calendar year.

The report shall contain the following information:

(a) Total curie activity discharged other than tritium and dissolved gases.

(b) Total curie alpha activity discharged .

(c) Total curies of tritium discharged .

(d) Total curies of dissolved radio-gases discharged .

Renewed Facility Operating License No. DPR-28 jAmendment No. XXX I

(e) Total volume (in gallons) of liquid waste discharged.

(f) Total volume (in gallons) of dilution water.

(g) Average concentration at discharge outfall.

(h) This paragraph deleted by Amendment No. 206, October 22, 2001 .

(i) Total radioactivity (in curies) released by nuclide including dissolved radio-gases .

U) Percent of the facility Offsite Dose Calculation Manual limit for total activity released .

11. This paragraph deleted by Amendment No. 206, October 22 , 2001 .

12. This paragraph deleted by Amendment No. 206, October 22 , 2001 .

of emergency notification to the states of Vermont and Ne't'1 Hampshire, and the Commonwealth of Massachusetts, satisfactory to the appropriate public health and publio safety officials of those states and the Commonwealth , whioh provides for:

This paragraph deleted by a. Notice of site emergencies as well as general emergencies.

Amendment No.

XXX. b. Direst microwave oommunioation with the state police headquarters of the respective states and the Commonwealth when the transmission facilities of the respective states and the Commonwealth so permit, at the expense of Entergy ~Juo l ear Operations, Ins.

&. A verification or coding system for emergency messages between Entergy Nuclear Operations, Ins. and the state police headquarters of the respective states and the Commonwealth .

14. Entergy Nuclear Operations, Inc. shall furnish advance notification to MDPH , or to another Commonwealth agency designated by MDPH , of the time , method and proposed route through the Commonwealth of any shipments of nuclear fuel and wastes to and from the Vermont Yankee facility which will utilize railways or roadways in the Commonwealth.

F. This paragraph deleted by Amendment No. 263.

Renewed Facility Operating License No. DPR-28 Amendment No. ~

NOTE: This paragraph regarding the Security Plan is being address by an independently submitted amendment request NOTE: This Security Plan paragraph regarding the Entergy Nuclear Operations, Inc. shall fy lly implement and maintain in effect Cyber Security all provisions of the Commission-approved physical security , training and Plan is being qualification , and safeguards contingency plans includ ing amendments made pursuant to provisions of the Miscellaneous Amendments and Search address by an Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) , and the independently authority of 10 CFR 50.90 and 10 CFR 50 .54(p) . The combined set of plans1, submitted which contain Safeguards Information protected under 10 CFR 73 .21 , is amendment entitled : "Vermont Yankee Nuclear Power Station Security Plan , Training request. and Qualification Plan , and Safeguards Contingency Plan , Revision O,"

submitted by letter dated October 18, 2004, as supplemented by letter dated

~ May 16, 2006.

Entergy Nuclear Operations, Inc. shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50 .54(p) . Entergy Nuclear Operations, Inc. CSP was approved by License Amendment No. 247 , as supplemented by changes approved by License Amendment Nos. 251 ,

259 , and 265 .

H. This parag raph deleted by Amendment No. 107, 8/25/88 .

I. This paragraph deleted by Amendment No. 131 , 10/7/91 .

J. License Transfer Conditions On the closing date of the transfer of Vermont Yankee Nuclear Power Station (Vermont Yankee) , Entergy Nuclear Vermont Yankee, LLC shall obtain from Vermont Yankee Nuclear Power Corporation all of the accumulated decommissioning trust funds for the facility , and ensure the deposit of such funds into a decommissioning trust for Vermont Yankee established by Entergy Nuclear Vermont Yankee, LLC . If the amount of such funds does not meet or exceed the minimum amount required for the facility pursuant to 10 CFR 50. 75 , Entergy Nuclear Vermont Yankee, LLC shall at such time deposit additional funds into the trust and/or obtain a parent company guarantee (to be updated annually) and/or obtain a surety pursuant to 10 CFR 50 .75(e)(1)(iii) in a form acceptable to the NRC and in an amount or amounts which , when combined with the decommissioning trust funds for the facility that have been obtained and deposited as required above , equals or 1

The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.

Renewed Facility Operating License No. DPR-28 Amendment No.~. ~. 2-W, ~ . 265 Corrected by letter dated November 21, 2012

exceeds the total amount required for the facil ity pursuant to 10 CFR 50.75.

The decommissioning trust, and surety if utilized , shall be subject to or be consistent with the following requirements, as applicable:

a. Decommissioning Trust (i) The decommissioning trust agreement must be in a form acceptable to the NRC.

(ii) With respect to the decommissioning trust funds, investments in the securities or other obligations of Entergy Corporation and its affiliates, successors, or assigns shall be prohibited. In addition, except for investments tied to market indexes or other non-nuclear-sector mutual funds , investments in any entity owning one or more nuclear power plants are prohibited.

(iii) The decommissioning trust agreement must provide that no disbursements or payments from the trust, other than for ordinary administrative expenses, shall be made by the trustee until the trustee has first given the NRC 30 days prior written notice of payment. The decommissioning trust agreement shall further contain a provision that no disbursements or payments from the trust shall be made if the trustee receives prior written notice of objection from the Director of the Office of Nuclear Reactor Regulation .

(iv) The decommissioning trust agreement must provide that the agreement cannot be amended in any material re'$pect without 30 days prior written notification to the Director of the Office of Nuclear Reactor Regulation.

(v) The appropriate section of the decommissioning trust agreement shall state that the trustee, investment advisor, or anyone else directing the investments made in the trust shall adhere to a "prudent investor" standard , as specified in 18 CFR 35.32(a)(3) of the Federal Energy Regulatory Commission 's regulations.

b. Surety (i) The surety agreement must be in a form acceptable to the NRC and be in accordance with all applicable NRC regulations.

(ii) The surety company providing any surety obtained to comply with the Order approving the transfer shall be one of those listed by the U.S.

Department of the Treasury in the most recent edition of Circular 570 and shall have a coverage limit sufficient to cover the amount of the surety.

Renewed Facility Operating License No. DPR-28

(iii) Entergy Nuclear Vermont Yankee , LLC shall establish a standby trust to receive funds from the surety, if a surety is obtained , in the event that Entergy Nuclear Vermont Yankee, LLC defaults on its funding obligations for the decommissioning of Vermont Yankee.

The standby trust agreement must be in a form acceptable to the NRC, and shall conform with all conditions otherwise applicable to the decommissioning trust agreement.

(iv) The surety agreement must provide that the agreement cannot be amended in any material respect, or terminated , without 30 days prior written notification to the Director of the Office of Nuclear Reactor Regulation .

Entergy Nuclear Vermont Yankee, LLC shall take all necessary steps to ensure that the decommissioning trust is maintained in accordance with the application for approval of the transfer of this license to Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc., and the requirements of the Order approving the transfer, and consistent with the safety evaluation supporting the Order.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. shall take no action to cause Entergy Global Investments, Inc., or Entergy International Holdings Ltd . LLC , or their parent companies to void , cancel, or modify the lines of credit to provide funding for Vermont Yankee as represented in the application without prior written consent of the Director of the Office of Nuclear Reactor Regulation.

K. This paragraph deleted by /\mendment No. 263.

L. This paragraph deleted by Amendment No. 263.

M. This paragraph deleted by Amendment No . 263.

N. Mitigation Strategy License Condition Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(a) Fire fighting response strategy with the following elements:

1. Pre defined coordinated fire response strategy and guidance

.2. Assessment of mutual aid fire fighting assets

3. Designated staging areas for equipment and materials

4. Command and control

6. Training of response personnel (b) Operations to mitigate fuel damage considering the following :

1. Protection and use of personnel assets

2. Communications Renewed Facility Operating License No. DPR-28 Amendment ~

3. Minimizing fire spread

4. Prooedures for implementing integrated fire response strategy

5. Identification of readily available pre staged equipment

6. Training on integrated fire response strategy

7. Spent fuel pool mitigation measures (o) Aotions to minimize release to inolude consideration of:

1. V"Jater spray sorubbing

2. Dose to onsite responders 0 . This paragraph deleted by Amendment ~Jo . 263 .

P. The information in the UFSAR Sl:lpplement, Sl:lbmitted pl:lrsuant to 10 CFR 54 .21 (d) , as revised dl:lring the lioense renewal applioation process, and as supplemented by Commitment Nos. 1 5, 6 (as revised by Entergy Nuclear Vermont Yankee , LLC letter dated May 1Q, 2011 ), 7 36 , 3g , 3Q , 42 , 43 , and 46 66 of Appendix A of Supplement 2 of NU REG 1907 shall be inoorporated as part of the UFS/\R whioh will be updated in aooordanoe with 10 CFR 60 .71 (e) .

As Sl:loh , Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, lno. may make ohanges to the programs and aotivities desoribed in the UFSAR supplement and Commitment Nos. 1 6, 6 (as revised by Entergy

~Jl:lolear Vermont Yankee, LLC letter dated May 1Q, 2011 ), 7 36 , 3g , 3Q, 42, 43 ,

and 46 66 of Appendix A of Supplement 2 of NU REG 1Q07 provided Entergy Nl:lclear Vermont Yankee , LLG and Entergy ~Jl:lclear Operations, lno. e;*aluates suoh ohanges pursuant to the oriteria set forth in 10 GrR 60.69 and otherwise complies with the requirements in that section .

Q. This paragraph deleted by Amendment No. 266 , April 17, 2013 .

R. This paragraph deleted by Amendment No. 263 .

S. This paragraph deleted by Amendment No. 263 .

4. This license is effective as of the date of issuance and is effective until the Commission notifies the licensee in writing that the license is terminated.

FOR THE NUCLEAR REGULATORY COMMISSION Original Signed By Eric J. Leeds Eric J. Leeds , Director Office of Nuclear Reactor Regulation

Enclosures:

Append ix A - Techn ical Specifications Date of Issuance: March 21 , 20 11 Renewed Facility Operating License No. DPR-28 Amendment 252, ~. ~

APPENDIX A TO OPERATING LICENSE DPR-28 TECHNICAL SPECIFICATIONS AND BA::rn.s FOR VERMONT YANKEE NUCLEAR POWER STATION VERNON , VERMONT ENTE RGY NUCLEAR OPERATIONS , INC .

AND ENTERGY NUCLEAR VERMONT YANKEE , LLC DOCKET NO . 50 - 271 Ame ndme nt No . ~

VYNPS TABLE OF CONTENTS Page No .

DEFINITIONS,,,,,,,, *1 LIMITING COJ>IDITIOJ>IS OF OPERl\TIOJ>I SURVEILLZ'.NCE LIMITING CONDITIONS OF OPER.".TIOJ>I and SURVEILLANCE REQUIREMENT (SR) APPLICABILITY .2 Bl',SES , , , .. , ' 4 R.~DIOl\CTIVE EFFLUENTS

i Liquid Holdup Tanks

i A

7 SPENT FUEL STOR.~GE g Fuel Stora§e Pool Water Level g A Spent Fuel Pool Water Temperature g B

9 5.0 DESIGN FEATURES 6.0 ADMINISTRATIVE CONTROLS Amendment No . ~ -i-

The suooeed ' that a Ge ~ terpret UAiform. iAt:i,Ag frequeAtl atioA of "' +-t h e specifica used term. s aret:i,oAs

. eJ(pl m.a"'

"'o:i,tly A-r J h e d e f :i,Aed

' so rtified Fuel H J "" achieved.

is aA :i,Adividual Hho . c~m.plies *, Jit:Adler p, Gertified e Gertified Fuel HaAdler Im.m.ediate iAitiated the UAit Gperal3le A syste e of the requir d e aet>on .

~ safe operatioA of

,:*.~* .::~1se. sHall o

s pe . ral or3 have lem ., sul3system.

ope . ' traiA ass...,,tion tffat*::;*I* Implisit oa,able of Be no"" OF emeF neoessary att ;s Sefinition or<unq its luBrioatioa o qeaoy eleetFieal :nQaat instr....ent sHall Be tHo syst9"', suBs .' otHer au*iliar ' '":'"' seuroes , sool at>on , oontrols ,

fuActioA ' s ' ystem. , traiA comp"! eq1upm.eAt that are :i,Ag or seal '. Jater f* 1 1 are al 1 oAeAt 0 d reqHir d 1 unotion)s)so ' oa,aBle o fperform.iAg F e v i e th e t . o perform.

e-itsfor tHe o ea relates .

peratiAg G Sllpport iAteAded ~~Ag

- perat

is p errorm.iAg

i,ts UAGt'!,OAS m.eaAs iA " t sthat a System.

required . or com.poAeAt m.aAAer.

".meAdmeAt No. 203 1

LIMITING CG GPER.~Tiow ~ URVEILLANCE ~E ANDITIGNa FGR

.PPLICA:BILITY - - - ==~.,~~P~PL~~IC~P~J3~I~L~I~T~Y~--~Q~U=I=R:E:M:E:~l:T~(~a~R~)*

aR 4 . Q.l aRs shall be m~t during the speoified

.pplioabilit f~s in the n . oondition

r LCGs, unless) tor individual i 0 hOP ' .

n the SR' Fail * .vise stated Surveillano ~re to meet a f ailure * - 0 , **heth er suoh the per f ::m::~:ri~noed during SuP'" eill anoe or bo t the performanoesfe 0

.10en Surveill to anoe , shall the b .

meet the LCG . e failure perform* -a .,,ur" C'

Failure to-the speo ~ f . "ei* 1 lanoe *1ith .

b % ied fro ' in e failure toquenoy shall exoept as meet the LCG Surveillan:rovided es do notin SR 4 ' Q ' 3 '

performed on . have to be equipment inoperable or " ar

b speoified 1 . : ia les outside imits.

SR 4 . Q. 2 Unless oth OP.HSO th .

ese speoif " . St a t ed in SUP'. eillanoe

ioations t .

/ periodio oalibration ests , Gheoks sh a 11 be perf~minations a' and exa . i speoified s ormed *, nthin the

- ur" e

11 -

intervals

These ~ i anoe

e adjust e -d plus 25intervals %. may b

SR 4,Q , 3 If it is disGo" surveillano~~ered that a

its 0spe.. as' f~

"*ith in not performed 61 deol aring

appl ' ied f requenoy Conditions f ioable Limiting '

not met may : : ~peration (LCGs) the time of d ' elayed 1 from 24 h ours o r u~ l8GO" Or J*1 up to the speoifiedpfto the limit of

',1hiGhO ITOr

" - l8. g requeno Jv I delay per

d :reater. -Tfi.-i-8 a 11 o*.1 performrmitted lO l8 po . to sur' eill anoe of tho

' anoo A .

evaluati' o n shall * "risk b for any Surue ' l l o performed g~oator than ~4 anoo delayed risk impaot h hours and tho s all be managed.

Amendment No, 203 2

OPER.7\,TION~~DITIONa LIMITING c FOR

PLICABILITY (aR) aR 4 '0' 3 (Go n t inued)

If t ....

h e surveill .

perfoERed ,N' itfi . ance is not

~-- ln ti=!

~eriod , applicabl e delay immediately be e LGOs must met , and appl

declared not be entered . icable LGOs must Wl=!en t ttO h SUr" Oill performed "N i~fi . ance is h SUF'~

delay

~-- ln ti=!

period and t ttO not met (i ~ ei 1 lance is er i. t aria a ' er

' 'a accept nce a l e not sa t

f

PP icable LGOs : lS ied) ,

immediately b must met , and appl~ declared not be entered . icable LGOs must l'. mendment No' 203

T£ 4 , 0 ens or oth er 4

£R o~*

4 0' 3 B*ses (GoRtiRoeG)

'

sis for this . delay penoG l Qility of * ~ersoAAel, iaol*Ges ooasi -Ge:-*tioa f the t*

of

i.., '°""uom

~*-*

del2v in

' t " conditions comp+/-eting to. pe<fo"'

Qle re****

t*e ,

it the survei+/-+/-an?e , nee ana t e . perfoERea is requi' reg SUF" . ei+/-+/-a

+/-ar survei I ' ,j,,J,anGe being of aRy part*~:t;; t*e reqoiremeats ,

eRoies is e*peoteQ to ooRfoaoaaoe " . .

so.-veill*ooe freq:oG estaQlio*eG hy

+/- , " ith speeifie of the ae+/-ay per used as an f*Hure to * -. ' o;ooueRoe . ""? Rot iRteRded to Qe lo w*ile op to he aR iofreqoeR fle*ihih - > " . " " "'"" i s G oorve><.a . " Roe iRterv*

. proviGeG -t perform

' ,.

n

n*~:

"" 4. 0 . 3 ** a .. ieRoe to e*teR . . G freqoeRoy **

t*e . spes*:~:;G "~~=GdetermiRatioR oarveillaRoe "'

~!Qe "!:!:e:r.~r!e~!:*::~~t'!

0 0 ::0:::!:.r7::!~:~:Y~o"s'd:r~:~r"a:'a:e,1a"t 0

oper*tional ..it of t*at t*e of perfor""'d at Qle opportoRitY .

t*e surveill*Roe a R~e) and impaot OR t*e first rea::":isk (from . delay:":erfo"" t*e oorv?~~::., plaRniR*

config:~:z:na:sumptions~ i~RC impaot OR pla *aR.08 requ1£Qd * . ioR to YR*t SORG> rfo ... t*e .

f ersoRROo ::aa:::t time sho~;aod i"!'lome~~;ore aRy an.. ' requinreaage:o ld Qe "" t::ough tatioo the

pri* :;::: '

av*HaQility oT::. risk i"'!'act its MaioteoaRoe

~

0 sorveillaRoe . Rt 10 GfR 50.i§(a) ( oO Maoa*i** Risk . GaiGe addresses pl*oe to

. latory GoiGe

"!':"""' l. 182, "Assess*o*

Po* oer Plaots '

This Re*ulatory Getermioat>on of G iok iffipaoto ' . Tho ffi>sse Re§o t Noolear " d **re§ate r eRt aotrnR . G in Aotivitieo . a f t-orary aR a ooRsiderat>oR o *otioR thresholdo , aR ome<§eRt ooRdit~aRtit*tive, aRd rial* maRa§em- c oo as disoosse risk maRa§effie:!oolG ho treated, a: .. alaatioR ""'Y sarveillanoe 'Goido, Thens* ",he Ge§Fee of e e of the GO"!'Onen '

"!",:: and ri*or of **:

the Re§olator, hlenGeG mothoGs. " ith the i"!'ortano be aRaly*eG risk

- ~ranees qaalitative , o',G be o-eosoratet ..oomponeRts shoold Gete,..ine t*e evaloation

'Missea quantitativ

survei " Oly increase is signiaction., this: ? , ' Iuficant for importaRf the risk evaloatrn:d to Gete,..ine :::

f the resu+/-ts o i ti' on shou+/-a O" a*oao A+/-+/- misse 9 survei+/-+/-anee

u~ii beB *Hoo be p+/-acea in

~afeensee'

.is st ooorso of . .. Aotion Pro§<'"'"

s Correcti

e . . h . the a+/-+/-o*, 'e? de+/-

G " it - >A * >ts aA*

a~' per>oG , theA the comp+/-etion 1

i~m~IH\~;e~ai~a~t~e~:+/-~:si'~*;1:1~p~hi:e~d~et+/-:!a!)':p:e~r~:i:o:d,2~f; ;:t!h~e:;:l':*:c~ti:o:n; .~£~t;. ,a"e'"*~:f*a:i:+/-:u::r:e:o:f:t:he fai+/-ed *.1ithin t mp+/-etion times 0

+/-imits anCO g the Conditionsegi GO= e

n imme~d i' ate+/-y upon the

~T~ONS, applioahle b . llo.,eG by this sarveillaRoe . .,

thiR the Golay restsree Comp+/-etion survei+/-+/-anse "i tion time of

. t ' of on the or"" ithin the comp+/-e

£pecifica i . ' h £~ 4.0 . 1 .

GOH\p+/-iance *, n t .

J\Hlenarnent ~o. I 203 5

Note: TS 3.1 will be relocated to a VYNPS licensee controlled procedure LIMITING CONQITIONa FOR 4 .1 aURVEILL.",NCE REQUIREMENTa OPERZ\TION RP,QIOP,CTIVE EFFLUK~ITa R.",QIOZ\CTIVE EFFLUENTa

.'\pplioability : .'\pplioability :

Applies to the release of all Applies to the required radioactive effluents from the surveillance of all radioaotive plant. effluents released from the plant ,

Objective :

Objeotive :

To assure that radioactive effluents are kept " as lm: as To ascertain that all is reasonably achievable " in radioactive effluents released accordance *.iith 10CFR50 , from the plant are kept " as 10*.1 Z\ppendix I and , in any event , as is reasonably aohievable " in are 11ithin the dose limits for aooordanoe *,lith 10CFR50 ,

Members of the Public specified Appendix I and , in any event ,

in 10CFR2Q, are 11ithin the dose limits for Members of the Publio specified in 10CFR20 .

apecification: apeoifioation :

A-.- Liquid Woldup Tanks A-.- Liquid Woldup Tanks

.- The quantity of .- The quantity of radioactive material radioactive material contained in any contained in eaoh of outside tank* shall be the liquid holdup limited to less than tanks

shall be or equal to 10 curies , determined to be excluding tritium and *,1ithin the limits of dissolved or entrained Specification 3 . 1.A , l noble gases. by analy~ing a representative sample

~ With the quantity of of the tank ' s oontents radioactive material Hithin one 11eek in any outside tank

follo *1ing the addition exceeding the limit of of radioaotive apecification 3 , 1 , A. 1 , materials to the tank, iffiffiediately take One sample may oover aotion to suspend all multiple additions .

additions of radioaotive material to the tank . Within 4B hours , reduoe the tank contents to Hithin the limit .

NOTE : Tanks included in this apecification are only those outdoor tanks that are not surrounded by liners , dikes , or 11alls capable of holding the tank ' s contents , or that do not have tank overflo11s and surrounding area drains oonneoted to the liquid rad1:aste treatment system .

.'\mendment No, 203 6

.J....-+/-. R.".DIO.".GTIVE EFFLUEWT.S A-.- Liquid Holdup Tanks The tanks listed in this .Specification include all outdoor tanks that contain radioactivity that are not surrounded by liners , dikes , or t1alls capable of holding the tank contents , or that do not have tank overflo*.1s and surrounding- area drains connected to the liquid ra~

aste treatment system .

Restricting- the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks ' contents , the resulting concentrations *,JOuld be less than the limits of lOGFR Part 20 . 1001 20,2402 , ,n.ppendi1c g , Table 2 , GolUJHn 2 , at the nearest potable Hater supply and in the nearest surface *. 1ater supply in an Unrestricted Area.

.".menclrnent No, 20 3 7

LIMITING CONDITION£ FOR £URVEILL~NCE REQUIREMENT£ OPERZ'.TION

£PE~lT FUEL £TORZ',GE SPE~lT FUEL STOR.".GE

.n.pplicability: .n.pplicability:

Applies to storage of spent Applies to tho parameters uhich

.f.ye-1-,... monitor the storage of spent

.f.ye-1-,...

Objective: Objective :

To assure safe storage of spent To verify that spent fuel is

.f.ye-1-,... being stored safely.

Specification : Specification :

A-r Fuel Storage Pool Water A-r Fuel Storage Pool Water

~ ~

Whenever irradiated fuel is Whenever irradiated fuel is stored in the fuel storage stored in the fuel storage pool the pool Hater level pool , the p ool level shail shall be maintained at a be recorded daily .

level of at least 3fi feet.

fh.. Spent Fuel Pool Water .g..... Spent Fuel Pool Water Temperature Temperature Whenever irradiated fuel is Whenever irradiated fuel is stored in the spent fuel in the spent fuel pool, the pool, the pool .1ater 1

pool *, :ater temperature shall temperature shall be be recorded daily . If the maintained belm: 150.l!F, poo 1 1.*ater temperature reaches 150gF, all operations tending to raise the pool 1:ater temperature shall cease and measures taken iHlHlediately to reduce the pool .:ater temperature 1

belm: 150gF, Aiaondrnont No. 2fi 3 8

BP.SES ;

3,2 & 4, 2 .SPE~I T FUEL .STORZ',GE A-r To assure that there is adequate \Jater to shield and sool the irradiated fl10l assemblies stored in the pool , a minimUHt pool Hater level is established . This minimlllfl water level of 36 feet is established besause it \1ould be a significant ohange from the normal level , ',1011 above a level to assure adequate sooling (just above astive fuel) ,

B.- The .Spent Fuel Pool Cooling £ystem is designed to maintain the pool 11ater temperattire belo*.1 125gF dtiring normal operations , If the roaster sore is sompletely dissharged , the temperattire of the peol

\Jater may insrease to greater than 12sar .

Arnendrnent No . 263 9

VYNPS 5.0 DESIGN FEATURES 5.1 Site The station is located on the property on the west bank of t h e Connecticut River in the Town of Vernon , Vermont , wh ich En tergy Nuclear Vermont Yankee , LLC either o wn s or to wh ich it has perpetual rights and easements . The site plan shO\Jing the exolusion area boundary , boundary for gaseous effluents , boundary for liquid effluents , as 1mll as areas defined per 10GFR20 as " oontrolled areas " and " unrestrioted areas " are on plant dra**ing 5920 0245, -T-he minimUHI distanee to the boundary of the eiwlusion area as defined in lOGFRl00 . 3 is 910 feet, The lioensee 11ill at all times retain the oomplete authority to determine and maintain suffioient oontrol of all aotivities through O\mership , easement , oontraet and/or ot h er legal instrUH1ents on property *,1hish is sloser to the roaster senter line than 910 feet ,

This insludes the authority to exolude or remove personnel and property 11ithin the exslusion area . Only astivities related to plant operation are permitted in the exslusion area.

5.2 Spent Fuel Storage The Kerr of the fuel in the spent fuel storage pool shall be less than or equal to 0 , 95 ,

£pent fuel storage rasks may be moved (only ) in assordanse '. tith 11ritten prosedures i,:hish ensure that no rask modules are moved over fuel assemblies .

G-r The number of spent fuel assemblies stored in the spent fuel pool shall not exseed 3353 .

-!;}.... The maicimllHI sore geometry infinite lattise multiplioation faster of any segment of the fuel assembly stored in the spent fuel storage pool or the ne'.1 fuel storage faoility shall be less than or equal to 1 . 31 at 2o~c .

Spent fuel shall not be stored in the spent fuel pool.

Amendment No . ,?..&J 10

BASES :

Excluaion area meana that area aurrounding the reactor , aa meaaured from the reactor center line , in *1hich the reactor licenaee haa the authority to determine all aotivitiea inoluding enoluaion or removal of peraonnel and property from the area. Thia area may be traveraed by a hig~ray ,

railroad , or *.1aten1ay , provided thoae are not ao cloae to the facility aa to interfere *. 1ith normal operationa of the facility and provided appropriate and effective arrangementa are made to control traffic on the high'<lay , railroad , or ""aten.iay , in oaae of an emergenoy , to protest the public health and aafety.

Contract proviaiona for property agreementa in the mrnluaion area Hill enaure that the licenaee retaina aufficient control of all activitiea in the enoluaion area including the authority to enolude or remove peraonnel and property , thereby ( 1) maintaining compliance *. :ith 1 0CFR50 . 07 radiological limita for the excluaion area , and (2) enauring that any and all activitiea , no'<I or in the future , in the excluaion area '<JOuld not negatively affect nuclear aafety , aafe plant operation or violate current plant deaign or lioenaing baais .

P.ny property tranaaction in the excluaion area , aa ia the caae for any activity *, 1hich haa the potential to adveraely affect nuclear aafety or aafe plant operation , requirea a revie11 in accordance 11ith 10CFR50 . 59.

!\dditionally, any property tranaaction *. mulct be required to comply '" 'ith other regulatory requirementa (e . g. , 10CFR5Q , Q3) aa applicable .

rooendment No. 2 03 11

VYNPS

6. 0 ADMINISTRATIVE CONTROLS

6. 1 REaPmrnIBILITY ~

Note: TS A..- The p+/-aHt maHage:i;: sha+/-+/- be :i;:espeHsib+/-e fe:i;: e>.ze:i;:a+/-+/- faci+/-it'!'

6.1.A and epe:i;:atieH aHd sha+/-+/- de+/-egate iH ffitiHg the successieH te this 1

6.1.B will be :i;:espoHsibility du'E'iHg abseHces.

relocated to Ih- The p+/-aHt maHage:i;: e:i;: desigHee shaU app:i;:e,,.e , p:i;:ie:i;: te the QAPM . imp+/-emeHtatieH , each p:i;:epesed test, expe:i;:imeHt , e:i;: medificatieH te S'! Stems e:i;: el'fYipmeHt that affect HUC+/-ea:i;: safety.

1 G-.. The shift supe:i;:vise:i;: shaH be :i;:espeHsib+/-e fe:i;: the shiH cemmaHd fuHctieH.

6.2 ORGP,NI6Z\TION ~

Note: TS OH.site aHd Offsite O:i;:gaHizatieHs 6.2.Awill be O:i;:gaHizatieHs sha+/-+/- be estab+/-ished fe:i;: faci+/-ity staff aHd ce:i;:pe:i;:ate relocated to maHagemeHt. These e:i;:gaHizatieHs sha+/-+/- iHc+/-ude the pesitieHs fe:i;:

the QAPM. activities affectiHg safety ef the Huc+/-ea:i;: fue+/- .

+/--.- LiHes ef autho:i;:ity , :i;:espeHsibi+/-ity, aHd cemmuHicatieH sha+/-+/- be estab+/-ished aHd defiHed fe:i;: the highest maHagemeHt +/-evels th:i;:eugh iHte:i;:mediate +/-eve+/-s te aHd iHc+/-udiHg a+/-+/- epe:i;:atiHg e:i;:gaHizatieHal pesitieHs . These :i;:e+/-atieHships sha+/-l be declH!leHted aHd updated , as app:i;:epriate , iH the fe:i;:m ef ergaHizatieHa+/- cha:i;:ts , fuHctieHa+/- descriptieHs ef departmeHta+/-

respeHsibi+/-ities aHd re+/-atieHships , aHd jeb descriptieHs fer key perseHHe+/- pesitieHs , er iH equiva+/-eHt ferms ef doouraentatioH . These requiremeHts sha+/-+/- be deoumeHted in the Qua+/-ity AssuraHce Pregram MaHua+/-. The p+/-aHt specific tit+/-es ef these perseHHe+/- fu+/-fi+/-+/-iHg the respensibi+/-ities ef the pesitieHS de+/-iHeated iH these TechHica+/- apecifioatieHS sha+/-+/-

be declH!leHted iH the Te chHica+/- RequiremeHts MaHua+/-.

The p+/-aHt maHager sha+/-+/- be respeHsib+/-e fer evera+/-+/- faci+/-ity safe eperatieH aHd sha+/-+/- have ceHtre+/- ever these eH site activities Hecessary fer safe ste:i;:age aHd maiHteHaHce ef the Huc+/-ear fue+/- .

A specified cerperate efficer sha+/-+/- have cerperate respeHsibi+/-ity fer evera+/-+/- p+/-aHt Huc+/-ear safety aHd sha+/-+/- take aHy measures Heeded te eHsure acceptab+/-e perfermaHce of the staff iH eperatiHg , maiHtaiHiHg , aHd previdiHg techHica+/-

suppert to the p+/-aHt to eHsure safe maHagemeHt ef Huc+/-ear

.f.tle-h-The iHdividua+/-s ',Jhe traiH the Certified Fue+/- MaHd+/-ers , carry eut hea+/-th physics , er perferm qua+/-ity assuraHce fuHctieHs may repert te the apprepriate eH site maHager ; he\Jever , these iHdividua+/-s sha+/-+/- haue sufficieHt ergaHizatieHa+/- freedem te eHsure their abi+/-ity te perferm their assigHed fuHctieHs.

Amendment No . ~ 12

ORGMlI6ATION (Cont ' d)

~

The fasility staff 0 rganic.ation

. shall insh1de the follmJing :

Eash duty shift shall e supervisor and one ~lon:e:::~:::d Oof at least one shift 1.

Operator position may ee f "ll db ,perator: . The Non sertified i e Y a Certified Fuel Handler.

2. At least one person qualified to stand ' ratsh in th sontrol room Handler) shall(Nonb serti"f ie~

" d Operator or " Certified eFuel

, 1 . e present in the sontrol room Hhen nus ear fuel is stored in the spent fuel pool .

All fuel handling operations shall ee direstly ey a Certified Fuel Handler. supervised

4. Shift ere'..' composition shall me t th .

herein. Shift ore" sompos . t . e e requirements stipulated

, . ". . i ion may ee less than the *

requirement of Spesifisation i 2 g 1 f . m~nimum to exseed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order t ' ' ' or a period of time not on duty shift srrn: memeers , or:~:::mo?ate ~nexpest~d a~sense of to restore the shift , P v . . d iffiffiediate astion is taken cre.i composition to , ' ithin th . .

requirements and all of the f 0 11 o ,,* "~ ~e

- ing sonditions areminimum met:

no fuel movements are in progress ; and no movement of loads over fuel are in progress ; and permitted IJhile tho shift no unmanned shift position d S~ring . .

Shift turnover Shall be ore,. is less than the minimum.

ai~d~vidual qualifi~d in radiation protestion procedures d' . e present on site during the mo " ement of fuel and uring the movement of loads over fuel.

Deleted The shift super visor

shall ee a Certified Fuel Handler.

Deleted Facility Staff Qualifications 1~

Note: TS 6.2.C.1 minim~ qyalifisations of P.N8I/.".~lS a~il l~;:t for exseed the Eash memeer of the fasility staff sh will be relocated to positions 1ith exseptions s esif . . or somparable 1

the QAPM . Program Manual (QAPM) . P ied in the Quality Assuranse An NRG approved training and retra . .

Fuel Handlers shall ee ~ t : ining program for Certified main- ained .

6.3 Deleted /Note : TS 6.4.A will be 6.4 """""""""" .::---!Deleted I K' relocated to the QAPM.

Written prosedyres shall be established . 1

o<ioq the follo.,ioq aoHvHies ' * >mp """""e"* """ maint>ine8 I Normal startyp , operation and shYtdo' 'n I ~ Fuel handling operations .

Amendment No. 2 i3 13

~ PROCEOURE£ (Cont ' d)

G... Actions to be taken to correct specific and foreseen potential malfunctions of systems or components needed for the safe storage of nuclear fuel .

Emergency conditions involving potential or actual release of radioactivity .

Preventive and corrective maintenance operations * ~ich could have an Note: TS effect on the safety of the nuclear fuel.

6.4.C thru

£urveillance and testing requirements.

6.4.1will be relocated to Fire protection program implementation .

the QAPM.

-.- Process Control Program in plant implementation .

h Off Site Qose Calculation Manual implementation .

6.5 HIGH RADIATION AREA As provided in paragraph 20 . 1601(c) of 10 CFR 20 , the follo wing controls shall be applied to high radiation areas in place of the controls required by paragraphs 20 . 1601(a) and 20 . 1601(b) of 10 CFR 20 :

A. High Radiation Areas with dose rates greater than 0 . 1 rem/hour at 30 centimeters , but not exceeding 1 . 0 rem/hour at 30 centimeters from the radiation source or from any surface penetrated by the radiation:

1. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area . Such barricades may be opened as necessary to permit entry or exit of personnel or equipment .

2. Access to , and activities in , each such area shall be controlled by means of Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures .

3. Individuals qualified in radiation protection procedures and personnel con tinuou sly escorted by such individuals may be exempted from t h e requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to , exit from , and work in such areas .

4. Each individual or group entering such an area shall possess:

a. A radiation monitoring device that continuously displays radiation dose rates in the area , or

b. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device ' s dose alarm setpoint is reached ,

with an appropriate alarm setpoint , or

c. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area , or

.l\mendment No, 2 03 14

VYNPS

d. A self - reading dosimeter (e . g ., pocket ionization chamber or electronic dosimeter) and ,

1. Be under the surveillance , as specified in the RWP or equivalent , while in the area , of an individual qualified in radiation protection procedures , equipped with a radiation monitoring device that continuously displays radiation Gee-&

rates in the area ; who is responsible for ~

controlling personnel exposure within the area , d or ose

2. Be under the surveillance , as specified in the RWP or equivalent , while in the area , by means of closed circuit television , of personnel qualified i n radiation protection procedures , responsible for controlling personnel radiation exposure i n t h e area , and with the means to communicate with individuals in the area who are covered by such surveillance .

5. Except for individuals qualified in radiation protection procedures , or personnel continuously escorted by such individuals , entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them . These continuously escorted personnel will receive a pre - job briefing prior to entry into such areas . This dose rate determination ,

knowledge , and pre - job briefing does not require documentation prior to initial entry .

B. High Radiation Areas with dose rates greater than 1 . 0 rem/hour at 30 centimeters from the radiation source or from any surface penetrated by the radiation , but less than 500 rads/hour at 1 meter from the radiation source or from any surface penetrated by the radiation :

1. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized e n try , and , in addition :

a. All such door and gate keys shall be maintained under the administrative control of the shift supervisor ,

and/or radiation protection manager , or his or her designee .

b. Doors and gates shall remain locked except during periods of personnel or equipment entry or exit .

2. Access to , and activities in , each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation does rates in the immediate work area(s) and other appropriate radiation protection equipment and measures .

3. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following plant radiation protection procedures for entry to , exit from , and work in such areas .

Amendment No. ~ 15

VYNPS

4. Each individual or group entering such an area shall possess one of the following :

a. A radiation monitoring device that continuous l y integrates the radiation rates in the area and alarms when the device ' s dose alarm setpoint is reached , with an appropriate alarm setpoint , or

b. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area , or

c. A self- reading dosimeter (e . g ., pocket ionization chamber or electronic dosimeter) and ,

1. Be under the surveillance , as specified in the RWP or equivalent , while in the area , of an individual qualified in radiation protection procedures , equipped with a radiation monitoring device that continuously displays radiation dose rates in the area ; who is responsible for controlling personnel exposure within the area ,

or

2. Be under the surveillance , as specified in the RWP or equivalent , while in the area , by means of closed circuit television , of personnel qualified in radiation protection procedures , responsible for controlling personnel radiation exposure in the area , and with the means to communicate with and control every individual in the area .

d. In those cases where option (b) and (c) , above , are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable " principle , a radiation monitoring device that continuously displays radiation dose rates in the area .

5. Except for individuals qualified in radiation protection procedures , or personnel continuously escorted by such individuals , entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them . These continuously escorted personnel will receive a pre - job briefing prior to entry into such areas . This dose rate determination ,

knowledge , and pre-job briefing does not require documentation prior to initial entry .

6. Such individual areas that are within a larger area where no enclosure exists for the purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate , nor continuously guarded , but shall be barricaded , conspicuously posted , and a clearly visible flashing light shall be activated at the area as a warning device .

Amendment No . ~ 16

~ REPORTHlG REQUIREMENTS The foll011ing reports shall be submitted in accordance ',1ith 10 CFR 50, 4, A-r Deleted

.Jh- Deleted G.... Deleted Note: TS 6.6.D th- Radi oaet i>.*e Ef f:luent Release Re~ort and 6.6.Ewill The Radi oact i>.*e Effluent Release Report covering the operation of be relocated to the f:acilit,,, shall be submitted by MaJ' 15 of: each ,,,ear and in the QAPM . accordance ',Ji th 10 CFR 50 . 3ea . The report shall include a SUffiffiary of: the ~uantities of: radioactive li~uid and gaseous eff:lt1ents and solid 11aste released f:rom the f:acility . The material pro,.,ided shall be consistent 11 ith the ob:jecti*.zes outlined in the Offsite Dose Calculation Manual (ODCM) and Process Cont+/-:ol Program and in confo+/-:mance 11ith 10 CFR 50.3ea and 10 CFR 50 , l\ppendix I ,Section I V. B. l .

.g..,_ Annual Radiological Environmental Operating Report The l'. nnual Radiological En*.zironmental Operating Report co:i,1 ering the operation of the faeility during the previous ealendar year shall be submitted b,,, May 15 of each year . The report shall include summaries , interpretations , and an analysis of trends of the results of the radiological en*.zironmental sur:i.zeillance activities for the report period. The material pro¥ided shall be eonsistent 11ith the ob:jeeti>.*es outlined i n the Off site Dose Calculation Manual ( ODC~4) , and in 10 CFR 50 , l\ppendi1< I ,

Sections IV , B, 2 , P 7 . B. 3 , and IV . C.

The ,D,nntial Radiological Environmental Operating Report shall inelude summarir;ed and tabulated results of all radiologieal environmental samples tal~en during the report period pursuant to the table and figures in the ODCM, In the e*.zent that some results are not available for inclusion 1,1ith the report , the report shall be submitted noting and explaining the reasons for the missing results . The missing data shall be submitted as soon as possible in a supplementary report .

4-r+ PROGRl\MS AND Ml',~lUl\LS The follo11ing programs shall be established , implemented and maintained :

A-r Deleted Note: TS 6.7.B .Jh- OFF SITE DOSE CALCULATION MANUAL (ODCM) will be l\n Off Site Dose Calculation Manual shall contain the current relocated to methodolog;z and parameters useEi in the calculation of off site the QAPM. Eioses Eiue to raEiioacti*o<e gaseous anEi li~uiEi effluents for the ptirpose of Eiemonstrating compliance >,.*ith 10 CFR 50 , AppenEii1< I , in the calculation of gaseous and li~uid effluent monitoring ala+/-:m,ltrip setpoints , and in the conduct of the environmental radiological monitoring program, The ODCM shall also contain the radioacti¥e effluent controls and raEiiological environmental monitoring acti*.zities anEi descriptions of the info+/-:mation that shoulEi be includeEi in the RaEiioactive Effluent Release Report and the ,:O.nnual RaEiiological En¥ironmental Operating Report re~uireEi by Specification e . e.D and Specification e. e . E, respecti*.zel,,, .

Amendment No. 203 17

.frrl PROGRl\MS AND MANU.7'.LS (Cont ' d)

Note: TS 6.7.B Licensee initiated changes to the ODCM :

will be relocated to the Shall be submitted to the CoJRIRission in the Radioactive Effluent Release Report for the period in '.Jhich the QAPM . change (s) *. 1as made effective . This submittal shall contain :

-i-.- Sufficient information to support the change together *.1ith appropriate analyses or evaluations justifying the change(s) and ii. A determination that the change \Jill maintain the level of radioactive effluent control required by 10 CFR 20.1302 , 40 CFR 190 , 10 CFR 50.3ea , and Appendix I to 10 CFR Part 50 , and do not adversely impact the accuracy or reliability of effluent dose or setpoint calculations.

Shall become effective upon approval by the plant manager .

Shall be submitted to the CoJRIRission in the form of a legible copy of the affected pages of the ODCM as a part of or concurrent 11ith the Radioactive Effluent Release Report for the period of the report in t1hich any change to the ODCM 1.ias made, Each change shall be identified by markings in the margin of the affected pages , clearly indicating the area of the page that 11as changed , and shall indicate the date (e.g. , month/year) the change

. 1as implemented .

~ Deleted Note: TS 6. 7.D -Ih- Radioactive Effluent Controls Program will be This program conforming to 10 CFR 50.3ea provides for the control relocated to of radioactive effluents and for maintaining the doses to members the QAPM . of the public from radioactive effluents as 1011 as reasonably achievable, The program shall be contained in the ODCM , shall be implemented by operating prooedures , and shall include remedial actions to be taken Hhenever the program limits are exceeded. -T-he program shall inolude the foll0t1ing elements:

a-.- Limitations on the functional oapability of radioaotive liquid and gaseous monitoring instrUH1entation inoluding surveillanoe tests and setpoint determination in accordance

,iith the methodology in the ODCM ;

&r Limitations on the conoentrations of radioactive material released in liquid effluents from the site to unrestrioted areas , conforming to 10 times the concentration " alues in Appendix g, Table 2, Colliffln 2, to 10 CFR 20.1001 20.2402 ;

B-r Monitoring, sampling , and analysis of radioactive liquid and gaseous effluents pursuant to 10 CFR 20 . 1302 and ""ith the methodology and parameters in the ODCM ;

Q.... Limitations on the annual and quarterly doses or dose coJRIRitment to a member of the public from radioactive materials in liquid effluents released from the facility to unrestricted areas , conforming to 10 CFR 50, Appendix I; Amendment No. 203 18

Note: TS 6.7.D Determination of ctlHlulative and projected dose contributions will be froffi radioactive effluents for the current calendar quarter and current calendar year in accordance 11ith the ffiethodology relocated to the and paraffieters in the ODCM at least every 31 days; QAPM.

.f....- Liffiitations on the functional capability and use of the liquid and gaseous effluent treatffient systeffis to ensure that appropriate portions of these systeffis are used to reduce releases of radioactivity *.1hen the proj acted doses in a period of 31 days .10uld mrneed 2 percent of the guidelines 1

for the annual dose or dose coffiffiitffient , conforffiing to 10 CFR 5 0 1 Appendix I ;

~ Liffiitations on the dose rate resulting from radioactive ffiaterial released in gaseous effluents froffi the site to areas at or beyond the site boundary shall be liffiited to the follm1ing :

1. For noble gases: less than or equal to a dose rate of 500 ffireffis/yr to the total body and less than or equal to a dose rate of 3 000 ffirems/yr to the skin , and

2. For iodine 131, iodine 133 , tritium , and for all radionuclides in particulate forffi 11ith half lives greater than B days: less than or equal to a dose rate of 1500 ffireffis/yr to any organ ;

fh.. Liffiitations on the annual and quarterly air doses resulting froffi noble gases released in gaseous effluents froffi the facility to areas at or beyond the site boundary , conforming to 10 CFR 50 , Z\ppendi1( I ;

h Liffiitations on the annual and quarterly doses to a ffieffiber of the public froffi iodine 131 , iodine 133 , tritiuffi , and all radionuclides in particulate forffi ', 1ith half lives greater than B days in gaseous effluents released froffi the facility to areas beyond the site boundary , conforffiing to 10 CFR 50 ,

Z\ppendix I; and j--.- Liffiitations on the annual dose or dose coffiffiitffient to any ffieffiber of the public , beyond the site boundary , due to releases of radioactivity and t o radiation froffi uranillffi fuel cycle sources, conforffiing to 40 CFR 190, TECWNICAL £PECIFICZ\TION£ (T£) .BZ',£E£ CONTROL PROCR1\M This prograffi provides a ffieans for processing changes to the .Bases of these Technical £pacifications.

a-,... Changes to the .Bases of the T£ shall be ffiade under appropriate adffiinistrative controls and revie*.1s, b.... Licensees ffiay make changes to .Bases 11ithout prior NRG approual provided the changes do not require either of the follo11ing;

1. A change in the T£ incorporated in the license, or

2. A change to the updated F£AR or .Bases that requires NRG approval pursuant to 10 CFR 50.59 Affiendffient No. 2e3 19

VYNPS

.fr,.:+ PROCR."i4£ mm MmlUAL£ (Cont ' d)

&... The Bases Control Program shall oontain provisions to ensure that the Bases are maintained oonsistent 11ith the F£AR.

~ Proposed ehanges that meet the eriteria of

£peoifioation 6 . 7.E . b above shall be reviewed and approved by the NRG prior to implementation. Changes to the Bases implemented 11ithout prior NRG approval shall be provided to the "1-lRG on a frequenoy oonsistent 1.'ith 10 GFR 50, 71 (e),

Amendment No, 263 20

BVY 17-004 Docket No. 50-271 Attachment 3 Vermont Yankee Nuclear Power Station Proposed Change 313 Retyped Operating License and Technical Specification Pages

Entergy Nuclear Vermont Yankee. LLC and Entergy Nuclear Operations . Inc.

(Vermont Yankee Nuclear Power Station)

Docket No. 50-271 Renewed Facility Operating License Renewed Operating License No. DPR-28 The U.S. Nuclear Regulatory Commission (NRC or the Commission), having previously made the findings set forth in Facility Operating License No. DPR-28 , dated February 28, 1973, has now found that:

a. This paragraph deleted by Amendment No. 263.

b. The facility is prohibited from operating the reactor in conformity with the application ,

as amended , the provisions of the Act, and the rules and regulations of the Commission; and

c. There is reasonable assurance (i) that the activities authorized by this license can be conducted without endangering the health and safety of the public , and (ii) that such activities will be conducted in compliance with the rules and regulations of the Commission; and

d. Entergy Nuclear Vermont Yankee , LLC is financially qualified and Entergy Nuclear Operations , Inc. is technically and financially qualified to engage in the activities authorized by this license, in accordance with the rules and regulations of the Commission ; and

e. Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations , Inc.

have satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements" of the Commission's regulations ; and

f. The issuance of this license will not be inimical to the common defense and security or to the health and safety of the public; and

g. After weighing the environmental, economic, technical and other benefits of the facility against environmental costs and considering available alternatives, the issuance of this license (subject to the conditions for protection of the environment set forth herein) is in accordance with 10 CFR Part 51, of the Commission's regulations and all applicable requirements of said Part 51 have been satisfied .

Renewed Facility Operating License No. DPR-28 Amendment~

Accordingly, Facility Operating License No. DPR-28, as amended , issued to Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. is superseded by Renewed Facility Operating License No. DPR-28 and is hereby amended in its entirety to read:

1. This renewed license applies to the Vermont Yankee Nuclear Power Station (the facility), a single cycle, boiling water, light water moderated and cooled reactor, and associated electric generating equipment. The facility is located on Entergy Nuclear Vermont Yankee, LLC's site, in the Town of Vernon, Windham County, Vermont, and is described in the application as amended .

2. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses:

A. Pursuant to Sections 104b of the Atomic Energy Act of 1954, as amended (the Act) , and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," Entergy Nuclear Vermont Yankee, LLC to possess and use, and Entergy Nuclear Operations , Inc. , to possess and use the facility as a utilization facility at the designated location on the Entergy Nuclear Vermont Yankee , LLC site.

B. Entergy Nuclear Operations, Inc. , pursuant to the Act and 10 CFR Part 70 , to possess at any time special nuclear material that was used as reactor fuel , in accordance with the limitations for storage and amounts required for reactor operation as described in the Final Safety Analysis Report, as supplemented and amended .

C. Entergy Nuclear Operations, Inc., pursuant to the Act and 10 CFR Parts 30, 40 and 70 , to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources that were used for reactor startup , sealed sources that were used for calibration of reactor instrumentation and are used in radiation monitoring equipment, and as fission detectors in amounts as required.

D. Entergy Nuclear Operations, Inc., pursuant to the Act and 10 CFR Parts 30 ,

40 and 70 , to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form , for sample analysis or instrument calibration or associated with radioactive apparatus or components.

E. Entergy Nuclear Operations , Inc. , pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not to separate, such byproduct and special nuclear material as may be produced by operation of the facility .

Renewed Facility Operating License No. DPR-28 Amendment~

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41of10 CFR Part 40, Section 50.54 and 50 .59 of 10 CFR Part 50 , and Section 70 .32 of 10 CFR Part 70 ; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A. This paragraph deleted by Amendment No. 263.

8. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. X.XX, are hereby incorporated in the license. Entergy Nuclear Operations , Inc. shall operate the facility in accordance with the Technical Specifications.

C. This paragraph deleted by Amendment No . .XXX.

D. This paragraph deleted by Amendment No. 226.

E. Environmental Conditions Pursuant to the Initial Decision of the presiding Atomic Safety and Licensing Board issued February 27, 1973, the following conditions for the protection of the environment are incorporated herein:

1. This paragraph deleted by Amendment No. 206, October 22 , 2001 .

2. This paragraph deleted by Amendment 131 , 10/07/91.

3. This paragraph deleted by Amendment No. 206, October 22, 2001 .

4. If harmful effects or evidence of irreversible damage in land or water ecosystems as a result of facility operation are detected by Entergy Nuclear Operations, lnc.'s environmental monitoring program, Entergy Nuclear Operations , Inc. shall provide an analysis of the problem to the Commission and to the advisory group for the Technical Specifications ,

and Entergy Nuclear Operations, Inc. thereafter will provide, subject to the review by the aforesaid advisory group, a course of action to be taken immediately to alleviate the problem.

5. Entergy Nuclear Operations, Inc. will grant authorized representatives of the Massachusetts Department of Public Health (MDPH) and Metropolitan District Commission (MDC) access to records and charts related to discharge of radioactive materials to the Connecticut River.

6. This paragraph deleted by Amendment No. 206 , October 22 , 2001.

Renewed Facility Operating License No. DPR-28 Amendment~ . ~ . ~ . ~ . ~

7. This paragraph deleted by Amendment No. 206, October 22, 2001 .

8. Entergy Nuclear Operations , Inc. will permit authorized representatives of the MDPH and MDC to examine the chemical and radioactivity analyses performed by Entergy Nuclear Operations, Inc.

9. Entergy Nuclear Operations, Inc. shall immediately notify MDPH , or an agency designated by MDPH , in the event concentrations of radioactive materials in liquid effluents, measured at the point of release from the Vermont Yankee facility, exceed the limit set forth in the facility Offsite Dose Calculation Manual. Entergy Nuclear Operations, Inc. will also notify MDPH in writing within 30 days following the release of radioactive materials in liquid effluents in excess of 10 percent of the limit set forth in the facility Offsite Dose Calculation Manual.

10. A report shall be submitted to MDPH and MDC by May 15 of each year, specifying the total quantities of radioactive materials released to the Connecticut River during the previous calendar year.

The report shall contain the following information:

(a) Total curie activity discharged other than tritium and dissolved gases.

(b) Total curie alpha activity discharged.

(c) Total curies of tritium discharged.

(d) Total curies of dissolved radio-gases discharged.

(e) Total volume (in gallons) of liquid waste discharged.

(f) Total volume (in gallons) of dilution water.

(g) Average concentration at discharge outfall.

(h) This paragraph deleted by Amendment No. 206, October 22, 2001.

(i) Total radioactivity (in curies) released by nuclide including dissolved radio-gases.

U) Percent of the facility Offsite Dose Calculation Manual limit for total activity released.

11. This paragraph deleted by Amendment No. 206, October 22, 2001 .

12. This paragraph deleted by Amendment No. 206, October 22, 2001 .

13. This paragraph deleted by Amendment No. XXX.

Renewed Facility Operating License No. DPR-28 Amendment XXX

14. Entergy Nuclear Operations, Inc. shall furnish advance notification to MDPH, or to another Commonwealth agency designated by MDPH , of the time, method and proposed route through the Commonwealth of any shipments of nuclear fuel and wastes to and from the Vermont Yankee facility which will utilize railways or roadways in the Commonwealth.

F. This paragraph deleted by Amendment No. 263.

G. Security Plan Entergy Nuclear Operations , Inc. shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans 1, which contain Safeguards Information protected under 10 CFR 73.21 , is entitled : "Vermont Yankee Nuclear Power Station Security Plan, Training and Qualification Plan , and Safeguards Contingency Plan , Revision O,"

submitted by letter dated October 18, 2004, as supplemented by letter dated May 16, 2006.

Entergy Nuclear Operations , Inc. shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). Entergy Nuclear Operations, Inc. CSP was approved by License Amendment No. 247, as supplemented by changes approved by License Amendment Nos. 251 ,

259 , and 265.

H. This paragraph deleted by Amendment No. 107, 8/25/88.

I. This paragraph deleted by Amendment No. 131 , 10/7/91 .

1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan .

Renewed Facility Operating License No. DPR-28 Amendment~. ~. ~ . ~. 265 Corrected by lotter dated November 21 , 2012

J. License Transfer Conditions On the closing date of the transfer of Vermont Yankee Nuclear Power Station (Vermont Yankee), Entergy Nuclear Vermont Yankee, LLC shall obtain from Vermont Yankee Nuclear Power Corporation all of the accumulated decommissioning trust funds for the facility , and ensure the deposit of such funds into a decommissioning trust for Vermont Yankee established by Entergy Nuclear Vermont Yankee, LLC. If the amount of such funds does not meet or exceed the minimum amount required for the facility pursuant to 10 CFR 50. 75, Entergy Nuclear Vermont Yankee, LLC shall at such time deposit additional funds into the trust and/or obtain a parent company guarantee (to be updated annually) and/or obtain a surety pursuant to 10 CFR 50.75(e)(1 )(iii) in a form acceptable to the NRC and in an amount or amounts which , when combined with the decommissioning trust funds for the facility that have been obtained and deposited as required above , equals or exceeds the total amount required for the facility pursuant to 10 CFR 50.75.

The decommissioning trust, and surety if utilized, shall be subject to or be consistent with the following requirements, as applicable:

a. Decommissioning Trust (i) The decommissioning trust agreement must be in a form acceptable to the NRC.

(ii) With respect to the decommissioning trust funds , investments in the securities or other obligations of Entergy Corporation and its affiliates, successors, or assigns shall be prohibited. In addition ,

except for investments tied to market indexes or other non-nuclear-sector mutual funds, investments in any entity owning one or more nuclear power plants are prohibited.

(iii) The decommissioning trust agreement must provide that no disbursements or payments from the trust, other than for ordinary administrative expenses , shall be made by the trustee until the trustee has first given the NRC 30 days prior written notice of payment. The decommissioning trust agreement shall further contain a provision that no disbursements or payments from the trust shall be made if the trustee receives prior written notice of objection from the Director of the Office of Nuclear Reactor Regulation.

(iv) The decommissioning trust agreement must provide that the agreement cannot be amended in any material respect without 30 days prior written notification to the Director of the Office of Nuclear Reactor Regulation.

Renewed Facility Operating License No. DPR-28

(v) The appropriate section of the decommissioning trust agreement shall state that the trustee, investment advisor, or anyone else directing the investments made in the trust shall adhere to a "prudent investor" standard, as specified in 18 CFR 35 .32(a)(3) of the Federal Energy Regulatory Commission's regulations.

b. Surety (i) The surety agreement must be in a form acceptable to the NRC and be in accordance with all applicable NRC regulations .

(ii) The surety company providing any surety obtained to comply with the Order approving the transfer shall be one of those listed by the U.S.

Department of the Treasury in the most recent edition of Circular 570 and shall have a coverage limit sufficient to cover the amount of the surety.

(iii) Entergy Nuclear Vermont Yankee, LLC shall establish a standby trust to receive funds from the surety, if a surety is obtained, in the event that Entergy Nuclear Vermont Yankee, LLC defaults on its funding obligations for the decommissioning of Vermont Yankee.

The standby trust agreement must be in a form acceptable to the NRC, and shall conform with all conditions otherwise applicable to the decommissioning trust agreement.

(iv) The surety agreement must provide that the agreement cannot be amended in any material respect, or terminated , without 30 days prior written notification to the Director of the Office of Nuclear Reactor Regulation.

Entergy Nuclear Vermont Yankee, LLC shall take all necessary steps to ensure that the decommissioning trust is maintained in accordance with the application for approval of the transfer of this license to Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc., and the requirements of the Order approving the transfer, and consistent with the safety evaluation supporting the Order.

Entergy Nuclear Vermont Yankee , LLC and Entergy Nuclear Operations, Inc. shall take no action to cause Entergy Global Investments, Inc., or Entergy International Holdings Ltd. LLC , or their parent companies to void, cancel , or modify the lines of credit to provide funding for Vermont Yankee as represented in the application without prior written consent of the Director of the Office of Nuclear Reactor Regulation.

Renewed Facility Operating License No. DPR-28 Amendment 2eJ

4. This license is effective as of the date of issuance and is effective until the Commission notifies the licensee in writing that the license is terminated .

FOR THE NUCLEAR REGULATORY COMMISSION Original Signed By Eric J. Leeds Eric J. Leeds , Director Office of Nuclear Reactor Regulation

Enclosures:

Appendix A - Technical Specifications Date of Issuance: March 21, 2011 Renewed Facility Operating License No. DPR-28 Amendment 263

APPENDIX A TO OPERATING LICENSE DPR-28 TECHNICAL SPECIFICATIONS FOR VERMONT YANKEE NUCLEAR POWER STATION VERNON , VERMONT ENTERGY NUCLEAR OPERATIONS, INC.

AND ENTERGY NUCLEAR VERMONT YANKEE, LLC DOCKET NO. 50-271 Amendmen t No . ~

VYNPS TABLE OF CONT ENTS Page No .

5.0 DESIGN FEATURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 6.0 ADMINISTRATIVE CONTROLS . . . . . . . . . . . . . . . . . . . . . . . . . . 2 Amendment No. ~ -i-

VYNPS 5.0 DESIGN FEATURES 5.1 Site The station is located on the property on the west bank of the Connecticut River in the Town of Ve rnon , Vermont , which Entergy Nuclear Vermont Yankee , LLC either owns or to which it has perpetual rights and easements.

5.2 Spent Fuel Storage Spent Fuel shall not be stored in the Spent Fuel Pool.

Amendment No. ~ 1

VYNPS 6.0 ADMINISTRATIVE CONTROLS 6.1 Deleted 6.2 Deleted 6.3 Deleted 6.4 Deleted 6.5 HIGH RADIATION AREA As provided in paragraph 20 . 1601(c) of 10 CFR 20 , the following controls shall be applied to high radiation areas in place of the controls required by paragraphs 20 . 1601(a) and 20 . 1601(b) of 10 CFR 20:

A. High Radiation Areas with dose rates greater than 0.1 rem/hour at 30 centimeters , but not exceeding 1 . 0 rem/hour at 30 centimeters from the radiati o n source or from any surface penetrated by the radiation :

1. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area . Such barricades may be opened as necessary to permit entry or exit of pers onnel or equipment .

2. Access to , and activities in , each such area shall be controlled by means of Radiation Work Permit (RWP) or equivalent that includes specification of radiati on dose rates in the immed iate work area(s) a nd other appropriate radiation protection equipment and measures .

3. Individuals qualified in radiatio n protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to , exit from , and work in such areas .

4. Each individual or group e ntering such an area shall possess :

a. A radiation monitoring device that continuously displays radiation dose rates in the area , or

b. A radiation monitoring device th at continuously integrates the radiation dose rates in the area and alarms when the device ' s dose alarm setpoint is reached ,

with an appropriate alarm setpoint , or

c. A radiation monitoring device that continuously transmits dose rate and cumulative dose informat i on to a remote receiver monitored by radiation protection personnel responsible for cont rolling personnel radiation exposure within the area , or Amendment No . ~ 2

VYNPS

d. A self - reading dosimeter (e . g. , pocket ionization chamber or electronic dosimete r ) and ,

1. Be under the surveillance , as specified in the RWP or equivalent , while in the area , of an individual qualified in radiation protection procedures , equipped with a radiation monitoring device that continuously displays radiat i on dose rates in the area; who is responsible for controlling personnel exposure within the area ,

or

2. Be under the surveillance , as specified in the RWP or equivalent , while in the area , by means of closed circuit television , of personnel qualified in radiation protection procedures , responsible for controlling personnel radiation exposure in the area , and with the means to communicate with individuals in the area who are covered by such surveillance .

5. Except for individuals qualified in radiation protection procedures , or personnel continuously esco r ted by such individuals , entry into such areas shall be made only after dose rates in the area have been de t ermined and entry personnel are knowledgeable of them . These continuous l y escorted personnel will receive a pre - job briefing prior to entry into such areas . This dose rate determination ,

knowledge , and pre - job briefing does not require documentation prior to initial entry .

B. High Radiation Areas with dose rates greater than 1 . 0 rem/hour at 30 centimeters from the radiation source or from any surface penetrated by the radiation , but less than 500 rads/hour at 1 meter from the radiation source or from any surface penetrated by the radiation:

1. Each entryway to such an area sha l l be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthor i zed entry , and , in addition :

a. All such door and gate keys shall be maintained under the administrative control of the shift supervisor ,

and/or radiation protection manager , or his or her designee .

b. Doors and gates shall remain locked except during periods of personnel or equipment entry or exit .

2. Access to , and activities in , each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation does rates in the immediate work area(s) and other appropriate radiation protection equipment and measures .

3. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following plant radiation protection procedures for entry to , exit from , and work in such areas .

Amendment No. 263 3

VYNPS

4. Each individual or group entering such an area sha l l possess one of the following :

a. A radiation monitoring dev i ce that co n tinuously integrates the radiation rates in the area and alarms when the device ' s dose alarm setpo i nt is reached , with an appropriate alarm setpoint , or

b. A radi ation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radi a t ion pro t ec t ion personnel responsible for controlling personne l radiation exposure within the area with the means to commu nicate wi th and cont r ol every individual in the area , or

c. A self- reading dosimeter (e . g. , pocket ionization chamber or electronic dosime t er) and ,

1. Be under the survei l lance , as specified in the RWP or equivalent , while in the area , of a n individual qualified in radiation protection procedures , equipped with a rad i ation monitoring device that continuous l y displays radiation dose rates in the area ; who is responsible for control l ing personne l exposure within the area ,

or

2. Be under the surveillance , as specified in the RWP or equivalent , while in the area , by means of closed ci r cuit television , o f personnel qualified in radiation protection procedures , responsible for controlling personnel radia t ion exposure in the area , and with the means to communicate with and contro l every individual in the area .

d. In those cases where option (b) and (c) , above , are imp r actical or determined to be inconsistent with the

" As Low As is Reasonably Achievable " principle , a rad i ation monitoring device that continuously displays radiation dose rates in the area.

5. Except for i ndividua l s qualified in radiation protection procedures , or personnel continuously escorted by such individuals , entry into such areas sha l l be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them . These continuously escorted personnel will receive a pre - job briefing prior to entry into such areas . This dose rate determination ,

knowledge , and pre - job briefing does not require documentation prior to initial entry .

6. Such individual areas that are within a larger area where no enclosure exists for the purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate , nor continuously guarded , but shal l be barr i caded , conspicuously posted , and a clearly visible flashing light shall be activated at the area as a warning device .

Amendment No. 263 4

BVY 17-004 Docket No. 50-271 Attachment 4 Vermont Yankee Nuclear Power Station Proposed Change 313 Administrative Controls Relocated to the QAPM

BVY 17-004 I Attachment 4 I Page 1of5 Administrative Controls Relocated to the QAPM This attachment provides the administrative controls proposed to be relocated from the Vermont Yankee Nuclear Power Station (VY) Permanently Defueled Technical Specifications (POTS) , Section 6.0 , "Administrative Controls," to the VY Quality Assurance Program Manual (QAPM), as described and evaluated in Attachment 1 to this letter. Refer to Attachment 2 to this letter for a mark-up of the POTS Administrative Controls with additional clarifications.

On implementation of the approved amendment, the administrative controls and other requirements shown below will be incorporated into the QAPM. Text formatted in red strikethrough is shown for information only to clarify deletions from the existing POTS and will not be included in the QAPM. Text formatted in underline has been added to the text relocated from the POTS and will be incorporated into the QAPM. Changes to section and paragraph numbers/letters and other format changes from the existing POTS are not indicated, and may be altered or re-formatted as needed for incorporation into the QAPM.

3.0 ADMINISTRATIVE CONTROLS RELOCATED FROM TECHNICAL SPECIFICATIONS 3.1 RESPONSIBILITY A. The ~ manager shall be responsible for overall facility operation and operational activities shall delegate in writing the succession to this responsibility during absences.

B. The ~ manager responsible for overall operational activities or designee shall approve, prior to implementation , each proposed test , experiment , or modification to systems or equipment that affect nuclear safety.

3.2 ORGANIZATION A. Onsite and Offsite Organizations Organizations shall be established for facility staff and corporate management.

These organizations shall include the positions for activities affecting safety of the nuclear fuel.

1. Lines of authority, responsibility, and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating organizational positions . These relationships shall be documented and updated, as appropriate, in the form of organizational charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions , or in equivalent forms of documentation. These requirements shall be documented in the Quality Assurance Program Manual. The plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications requirements shall be documented in the Technical l=lequirements Manual.

BVY 17-004 I Attachment 4 I Page 2 of 5

2. The ~ manager shall be responsible for overall facility safe operation aoo operational activities shall have control over those on-site activities necessary for sate storage and maintenance .:>f the nuclear fuel.

3. A specified corporate officer shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure sate management of nuclear fuel.

4. The individuals who train the Certified Fuel Handlers, carry out health physics, or perform quality assurance functions may report to the appropriate on-site manager; however, these individuals shall have sufficient organizational freedom to ensure their ability to perform their assigned functions.

B. Facility Staff Qualifications

1. Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI/ANS 3. 1-1978 for comparable positions with exceptions specified in the Quality Assurance Program Manual (QAPM).

3.3 PROCEDURES Written procedures shall be established , implemented, and maintained covering the following activities:

A. Normal startup, operation and shutdown of systems and components needed for the safe storage of nuclear fuel.

B. Fuel handling operations.

C. Actions to be taken to correct specific and foreseen potential malfunctions of systems or components needed for the safe storage of nuclear fuel.

D. Emergency conditions involving potential or actual release of radioactivity.

E. Preventive and corrective maintenance operations which could have an effect on the sat ety of the nuclear fuel.

F. Surveillance and testing requirements.

G. Fire protection program implementation.

H. Process Control Program in-plant implementation.

I. Off-Site Dose Calculation Manual implementation.

3.4 REPORTING REQUIREMENTS

BVY 17-004 I Attachment 4 I Page 3 of 5 The following reports shall be submitted in accordance with 10 CFR 50.4.

A. Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the facility shall be submitted by May 15 of each year and in accordance with 10 CFR 50 .36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the facility. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM) and Process Control Program and in conformance with 10CFR50.36a and 10 CFR 50, Appendix I, Section IV.B.1.

B. Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report covering the operation of the facility during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries , interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period . The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM) , and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

The Annual Radiological Environmental Operating Report shall include summarized and tabulated results of all radiological environmental samples taken during the report period pursuant to the table and figures in the ODCM.

In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results . The missing data shall be submitted as soon as possible in a supplementary report.

3.5 PROGRAMS AND MANUALS The following programs shall be established , implemented and maintained:

A. OFF-SITE DOSE CALCULATION MANUAL (ODCM)

An Off-Site Dose Calculation Manual shall contain the current methodology and parameters used in the calculation of off-site doses due to radioactive gaseous and liquid effluents for the purpose of demonstrating compliance with 10 CFR 50 , Appendix I, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints , and in the conduct of the environmental radiological monitoring program.

The ODCM shall also contain the radioactive effluent controls and rad iological environmental monitoring activities and descriptions of the information that should be included in the Radioactive Effluent Release Report and the Annual Radiolog ical Environmental Operating Report required by Speoifioation 6.6 .D and Speoifioation 6.6.E, respeotively .

BVY 17-004 I Attachment 4 I Page 4 of 5

1. Licensee initiated changes to the ODCM:

a. Shall be submitted to the Commission in the Radioactive Effluent Release Report for the period in which the change(s) was made effective. This submittal shall contain:

i. Sufficient information to support the change together with appropriate analyses or evaluations justifying the change(s) and ii. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50, and do not adversely impact the accuracy or reliability of effluent dose or setpoint calculations.

b. Shall become effective upon approval by the f:}taRt manager responsible for overall operational activities.

c. Shall be submitted to the Commission in the form of a legible copy of the affected pages of the ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed , and shall indicate the date (e.g., month/year) the change was implemented.

B. Radioactive Effluent Controls Program This program conforming to 10 CFR 50.36a provides for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by operating procedures, and shall include remedial actions to be taken whenever the program limits are exceeded . The program shall include the following elements:

a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM;

b. Limitations on the concentrations of radioactive material released in liquid effluents from the site to unrestricted areas, conforming to 1O times the concentration values in Appendix B, Table 2, Column 2, to 10 CFR 20.1001 - 20.2402;

c. Monitoring , sampling , and analysis of radioactive liquid and gaseous effluents pursuant to 10 CFR 20.1302 and with the methodology and parameters in the ODCM;

BVY 17-004 I Attachment 4 I Page 5 of 5

d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from the facility to unrestricted areas, conforming to 10 CFR 50,

Appendix I;

e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days;

f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2 percent of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I;

g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the site boundary shall be limited to the following:

1. For noble gases: less than or equal to a dose rate of 500 mrems/yr to tho total body and loss than or equal to a dose rate of 3000 mrems/yr to the skin , and

2. For iodine 131 , iodine 133, tritium, and for all radionuclides in particulate form with half lives greater than 8 days: less than or equal to a dose rate of 1500 mrems/yr to any organ ;

h. Limitations on tho annual and quarterly air doses resulting from noble gases released in gaseous effluents from tho faoility to areas at or beyond the site boundary, oonforming to 10 CFR 50, Appendix I;

i. Limitations on the annual and quarterly doses to a member of the public from iodine 131 , iodine 133, tritium, and all radionuclides in particulate form with half lives greater than 8 days in gaseous effluents released from the facility to areas beyond the site boundary, conforming to 10 CFR 50 , Appendix I; and

j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

BVY 17-004 Docket No. 50-271 Attachment 5 Vermont Yankee Nuclear Power Station Proposed Change 313 Regulatory Commitment

BVY 17-004 I Attachment 5 I Page 1of1 Regulatory Commitment This table identifies actions discussed in this letter for which ENO commits to perform . Any other actions discussed in this submittal are described for the NRC's information and are not commitments.

TYPE (Check one) Scheduled COMMITMENT Completion Date ONE-TIME CONTINUING (If Required)

ACTION COMPLIANCE The administrative controls relocated x On implementation from the POTS will be in incorporated of the approved into the VY QAPM as shown and amendment described in Attachment 4 to this letter.

BVY 17-004, Technical Specifications Proposed Change No. 313: Revision to License and Permanently Defueled Technical Specifications to Reflect Permanent Removal of Spent Fuel from the Spent Fuel Pool

Contents

Text

SUBJECT:

REFERENCES:

Dear Sir or Madam:

SUMMARY

Enclosures:

Enclosures:

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