BVY 14-009, Request for Exemptions from Portions of 10 CFR 50.47 and 10 CFR 50, Appendix E

Request for Exemptions from Portions of 10 CFR 50.47 and 10 CFR 50, Appendix E

ML14080A141
Person / Time
Site:	Vermont Yankee
Issue date:	03/14/2014
From:	Wamser C Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BVY 14-009
Download: ML14080A141 (78)

Text

Entergy Nuclear Operations, Inc.

Vermont Yankee 320 Governor Hunt Rd.

pEntergy Vernon, VT 802-257-7711 Christopher J. Wamser Site Vice President BVY 14-009 10 CFR 50.12 10 CFR 50.47 10 CFR 50, Appendix E March 14, 2014 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Request for Exemptions from Portions of 10 CFR 50.47 and 10 CFR 50, Appendix E Vermont Yankee Nuclear Power Station Docket No. 50-271 License No. DPR-28

REFERENCES:

1. Letter, Entergy Nuclear Operations, Inc. to USNRC, "Notification of Permanent Cessation of Power Operations," BVY 13-079, dated September 23, 2013

Dear Sir or Madam:

Pursuant to 10 CFR 50.12, Entergy Nuclear Operations, Inc. (ENO) requests exemptions from portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and 10 CFR 50, Appendix E for Vermont Yankee Nuclear Power Station (VY). The requested exemptions would allow VY to reduce emergency planning requirements and subsequently revise the VY Emergency Plan consistent with the anticipated permanently defueled condition of the station.

On September 23, 2013, ENO informed the NRC that VY will permanently cease operations in the fourth quarter of 2014 (Reference 1). Once VY permanently ceases operations and submits the certifications required by 10 CR 50.82(a)(1)(i) and (ii), pursuant to 10 CFR 50.82(a)(2), the 10 CFR Part 50 license for VY will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel.

The requested exemptions are permissible under 10 CFR 50.12 because they will not present an undue risk to the public health and safety, and application of the regulations in this particular circumstance is not necessary to achieve the underlying purpose of the rules. More specifically, application of the portions of the regulations from which exemptions are sought is not necessary to ensure adequate emergency response capability for VY. Furthermore, continued application of these portions of the regulations from which exemptions are sought would impose a burden on ENO and the VY Decommissioning Trust Fund by requiring continued implementation of unnecessary emergency response capability.

The exemption requests are contained in Attachment 1 to this letter. ENO has performed an analysis which shows that, within 15.4 months after shutdown, the spent fuel stored in the spent A4OU

BVY 14-009 / Page 2 of 3 fuel pool will have decayed to the extent that the requested exemptions may be implemented at VY without any additional compensatory actions. Following the VY shutdown, which is expected by the end of 2014 (Reference 1), 15.4 months after shutdown will occur near the middle of April 2016. This analysis is contained in Attachment 2.

VY plans to submit a permanently defueled emergency plan, containing a permanently defueled emergency action level scheme, for NRC review and approval pursuant to 10 CFR 50.54(q)(4) and 10 CFR 50, Appendix E, Section IV.B.2. The proposed emergency plan will be based on the exemptions requested herein.

ENO requests review and approval of this exemption request by December 1, 2015 with an effective date of April 15, 2016. Approval of these exemptions by December 1, 2015 will allow VY adequate time to implement changes to the emergency plan and emergency response organization by the requested effective date of April 15, 2016. of this letter contains new regulatory commitments.

Should you have any questions concerning this letter or require additional information, please contact Mr. Coley Chappell at 802-451-3374.

Sincerely, CJW/pIc Attachments: 1. Request for Exemptions from Portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and 10 CFR Part 50, Appendix E

2. Vermont Yankee Maximum Cladding Temperature Analysis for an Uncovered Spent Fuel Pool with no Air Cooling

3. List of Regulatory Commitments cc list next page

BVY 14-009 / Page 3 of 3 cc: Mr. William M. Dean Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 Mr. James S. Kim, Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 08C2A Washington, DC 20555 USNRC Resident Inspector Entergy Nuclear Vermont Yankee, LLC 320 Governor Hunt Road Vernon, Vermont 05354 Mr. Christopher Recchia, Commissioner Vermont Department of Public Service 112 State Street - Drawer 20 Montpelier, Vermont 05620-2601

BVY 14-009 Docket 50-271 Attachment 1 Vermont Yankee Nuclear Power Station Request for Exemptions from Portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and 10 CFR Part 50, Appendix E

BVY 14-009 / Attachment 1 1 Page 1 of 59 Vermont Yankee Nuclear Power Station Requests for Exemptions from Portions of 10 CFR 50.47(b),

50.47(c)(2) and 10 CFR Part 50, Appendix E I. Summary Description Pursuant to 10 CFR 50.12 "Specific exemptions," Entergy Nuclear Operations, Inc. (ENO) requests exemptions from the following for Vermont Yankee Nuclear Power Station (VY):

" Certain standards in 10 CFR 50.47(b) regarding onsite and offsite emergency response plans for nuclear power reactors;

" Certain requirements of 10 CFR 50.47(c)(2) to establish plume exposure and ingestion pathway emergency planning zones for nuclear power plants; and

• Certain requirements of 10 CFR 50, Appendix E, which establishes the elements that make up the content of emergency plans.

The requested exemptions would allow ENO to reduce emergency planning requirements and subsequently revise the VY Emergency Plan to reflect the permanently defueled condition of the station. The current 10 CFR Part 50 regulatory requirements for emergency planning (developed for operating reactors) ensure safety at VY. However, once the station is permanently shutdown, defueled, and in a state of decommissioning, some of these requirements exceed what is necessary to protect the health and safety of the public.

The requested exemptions and justification for each are based on and consistent with draft Interim Staff Guidance NSIR/DPR-ISG-02, Emergency Planning Exemption Requests for Decommissioning Nuclear Power Plants, which was issued for public comment in December 2013 (Reference 18).

I1. Detailed Description By letter dated September 23, 2013 (Reference 1), pursuant to 10 CFR 50.82(a)(1 )(i), ENO submitted a certification to the NRC indicating its intention to permanently cease power operations at VY at the end of the current operating cycle, which is expected to occur near the end of December 2014. ENO stated its intention to submit a supplement to Reference 1 certifying the date on which operations have ceased, or will cease, in accordance with 10 CFR 50.82(a)(1)(i) and 10 CFR 50.4(b)(8). Once fuel has been permanently removed from the reactor vessel, ENO will submit a written certification to the NRC, in accordance with 10 CFR 50.82(a)(1)(ii) that meets the requirements of 10 CFR 50.4(b)(9). Upon docketing of these certifications, the 10 CFR Part 50 license for VY will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel, as specified in 10 CFR 50.82(a)(2).

In order to allow a reduction in emergency planning requirements commensurate with the hazards associated with VY's permanently defueled condition, exemptions from portions of 10 CFR 50.47(b), 50.47(c)(2), and 10 CFR 50, Appendix E, are needed. ENO has performed an analysis indicating that, within 15.4 months after shutdown, the spent fuel in the spent fuel pool (SFP) will have decayed to the extent that the requested exemptions can be implemented at VY without any compensatory actions. This analysis is included in Attachment 2. Because VY expects shutdown to occur by the end of December 2014, 15.4 months after shutdown will occur near the middle of April 2016. ENO plans to submit a permanently defueled emergency plan by May 31, 2014, including a Permanently Defueled Emergency Action Level scheme, for NRC review and approval pursuant to 10 CFR 50.54(q)(4) and 10 CFR 50, Appendix E, Section IV.B.2. The proposed emergency plan will be based on the exemptions requested herein. ENO

BVY 14-009/ Attachment 1 / Page 2 of 59 requests approval of these exemption requests by December 1, 2015 with an effective date of April 15, 2016. Approval of these exemptions by December 1, 2015 will enable ENO adequate time to implement changes to the emergency preparedness program and emergency response organization by April 15, 2016.

BVY 14-009 / Attachment 1 / Page 3 of 59 EXEMPTIONS TO EMERGENCY PLAN REQUIREMENTS DEFINED BY 10 CFR 50.47 AND APPENDIX E TO PART 50 ENO requests exemptions from portions of 10 CFR 50.47(b) and (c)(2) and Appendix E to 10 CFR Part 50 to the extent that these regulations apply to specific provisions of onsite and offsite emergency planning that will no longer be applicable to VY once the certifications required by 10 CFR 50.82(a)(1 )(i) and (ii) have been submitted and sufficient decay of the spent fuel has occurred. The specific portions of 10 CFR 50.47 and 10 CFR Part 50, Appendix E from which exemptions are being requested are identified using strikethrough text in Table 1 (Exemptions Requested from 10 CFR 50.47(b) and (c)(2)) and Table 2 (Exemptions Requested from 10 CFR Part 50, Appendix E), below. The portions of regulation that are not identified using strikethrough text (i.e., those portions for which exemption is not being requested), will remain applicable to VY. Details related to specific exemption requests are provided in the Basis for Exemption column.

Table 1 Exemptions Requested from 10 CFR 50.47(b) and 50.47(c)(2)

Item # Regulation in 10 CFR 50.47 Basis for Exemption 1 10 CFR 50.47(b): The onsite and, eXcept as provided in paragraph In the Statement of Considerations for the Final Rule for EP requirements (d) of this section, offsitc emergency response plans for nuclear for Independent Spent Fuel Storage Installations (ISFSIs) and for power reactors must meet the following standards: monitored retrievable storage (MRS) facilities (60 FR 32430; June 22, 1995) (Reference 19), the Commission responded to comments concerning offsite emergency planning for ISFSIs or an MRS and concluded that, "the offsite consequences of potential accidents at an ISFSI or a MRS [monitored retrievable storage installation] would not warrant establishing Emergency Planning Zones." In a nuclear power reactor's permanently defueled state, the accident risks are more similar to an ISFSI or MRS than an operating nuclear power plant. The draft proposed rulemaking in SECY-00-0145 (Reference 20) suggested that after at least one year of spent fuel decay time, the decommissioning licensee would be able to reduce its EP program to one similar to that required for an MRS under 10 CFR 72.32(b) and additional EP reductions would occur when: (1) approximately five years of spent fuel decay time has elapsed; or (2) a licensee has demonstrated that the decay heat level of spent fuel in the pool is low enough that the fuel would not be susceptible to a zirconium fire for all spent fuel configurations. The EP program would be similar to that required for an ISFSI under 10 CFR 72.32(a) when fuel stored in the SFP has more than five years of decay

BVY 14-009 / Attachment 1 / Page 4 of 59 Table 1 Exemptions Requested from 10 CFR 50.47(b) and 50.47(c)(2)

Item # Regulation in 10 CFR 50.47 Basis for Exemption time and would not change substantially when all the fuel is transferred from the SFP to an onsite ISFSI. Exemptions from offsite EP requirements have been approved when the specific site analyses show that at least ten hours is available from a partial drain down event where cooling of the spent fuel is not effective until the hottest fuel assembly reaches 9000 C. Because ten hours allows sufficient time to initiate mitigative actions to prevent a zirconium fire in the SFP or to initiate ad hoc offsite protective actions, offsite EP plans are not necessary for these permanently defueled nuclear power plant licensees.

2 10 CFR 50.47(b)(1): Primary responsibilities for emergency See basis for 50.47(b).

response by the nuclear facility licensee and by State and local organizations within the E ,ergency Pl*aning Zones have been assigned, the emergency responsibilities of the various supporting organizations have been specifically established, and each principal response organization has staff to respond and to augment its initial response on a continuous basis.

3 10 CFR 50.47(b)(2): On-shift facility licensee responsibilities for No exemption is requested.

emergency response are unambiguously defined, adequate staffing to provide initial facility accident response in key functional areas is maintained at all times, timely augmentation of response capabilities is available and the interfaces among various onsite response activities and offsite support and response activities are specified.

4 10 CFR 50.47(b)(3): Arrangements for requesting and effectively Discontinuing offsite emergency planning activities and reducing the using assistance resources have been made, eaFrrgernents-tc scope of onsite emergency planning is acceptable given the significantly a.com.*odat State and local staff at the ,,cens.e's Emergency reduced offsite consequences once VY is in the permanently defueled Oporation- FaGciity have been made, and other organizations condition. The VY emergency plan will continue to maintain arrangements capable of augmenting the planned response have been identified. for requesting and using assistance resources from offsite support

BVY 14-009 / Attachment 1 / Page 5 of 59 Table 1 Exemptions Requested from 10 CFR 50.47(b) and 50.47(c)(2)

Item # Regulation in 10 CFR 50.47 1 Basis for Exemption organizations.

Decommissioning power reactors present a low likelihood of any credible accident resulting in radiological releases requiring offsite protective measures because of the permanently shut down and defueled status of the reactor. An emergency operations facility is not required. The control room or other location can provide for the communication and coordination with offsite organizations for the level of support required.

Also see basis for 50.47(b).

5 10 CFR 50.47(b)(4): A standard emergency classification and VY will adopt the Permanently Defueled Emergency Action Levels (EALs) action level scheme, the bases of which include facility system and detailed in Appendix C of Nuclear Energy Institute (NEI) 99-01, effluent parameters, is in use by the nuclear facility licensee-and "Development of EALs for Non-Passive Reactors," Revision 6 (Reference State and local response plans call for reliano. on informnation 2), endorsed by the NRC in a letter dated March 28, 2013. No offsite pro.ied by facility licensces for determinations of minimum initia protective actions are anticipated to be necessary, so classification above re.ponse measures.

ffitc the Alert level will no longer be required.

Also see basis for 50.47(b).

6 10 CFR 50.47(b)(5): Procedures have been established for Per SECY-00-0145 (Reference 20), after approximately 1 year of spent notification, by the licensee, of State and local response fuel decay time [and as supported by the SFP analysis], the NRC staff organizations and for notification of emergency personnel by all believes an exception to the offsite EPA PAG standard is justified for a organizations; the content of initial and followup messages to zirconium fire scenario considering the low likelihood of this event response organizations and the pubtic has been established;-ad together with time available to take mitigative or protective actions m.eans to provide carl.. notification and clear inStruGtion to the between the initiating event and before the onset of a postulated fire. The p.pulaec within the plume exposure pathway Emer-gen.Y Planning spent fuel scoping study (Reference 3) provides that depending on the Zono havc bccn establishd, size of the pool liner leak, releases could start anywhere from eight hours to several days after the leak starts, assuming that mitigation measures are unsuccessful. If 10 CFR 50.54(hh)(2) type of mitigation measures are successful, releases could only occur during the first several days after the fuel came out of the reactor. Therefore, offsite EP plans are not

BVY 14-009 / Attachment 1 / Page 6 of 59 Table 1 Exemptions Requested from 10 CFR 50.47(b) and 50.47(c)(2)

Item # Regulation in 10 CFR 50.47 Basis for Exemption necessary for permanently defueled nuclear power plants.

7 10 CFR 50.47(b)(6): Provisions exist for prompt communications See basis for 50.47(b).

among principal response organizations to emergency personnel and to the public.

8 10 CFR 50.47(b)(7): Information is made available to the public on See basis for 50.47(b).

a pcriodic basis on how they will be notified and what their initial actions should be in an emergency(e.g., listening to a loc-al broadcast station and remaining indoors)., [T]he principal points of contact with the news media for dissemination of information during an emergency (including the physical location*rI-lations)-

are established in advance, and procedures for coordinated dissemination of information to the public are established.

9 10 CFR 50.47(b)(8): Adequate emergency facilities and equipment No exemption is requested.

to support the emergency response are provided and maintained.

10 10 CFR 50.47(b)(9): Adequate methods, systems, and equipment See basis for 50.47(b).

for assessing and monitoring actual or potential eff-site consequences of a radiological emergency condition are in use.

11 10 CFR 50.47(b)(1 0): A range of protective actionRs has been In the unlikely event of a SFP accident, the iodine isotopes which developed for the plume exposure pathway EPZ for erg*cn-G contribute to an off-site dose from an operating reactor accident are not workers and the public. Ideveloping this range of aGti;o&n, present, so potassium iodide (KI) distribution off-site would no longer consideration has been given to evacuation, sheltering, and, as a serve as an effective or necessary supplemental protective action.

supplemen itto those, as*appropriate. the prophylactic Evacuation use of time" esti.+mates potas6iuMmeidide hav been developed

... (KI),

b.**

, TeCmiso epne The Commission responded toocmet comments innisSaeeto its Statement of applicant- an lien.seesG iceses. este .hllupda then devacuatio Considerations for the Final Rule for emergency planning requirements tim estimates ondaeriod.ic basis uidlinesdfor the choice ot for ISFSIs and MRS facilities (60 FR 32435) (Reference 19), and protec"tivactionas upring

, ane.megeny,* c fo*nsiste wth, edera concluded that, "the offsite consequences of potential accidents at an ga,....

____._ are de..o. and inegpRlce, and protecti*v actions ISFSl or a MRS would not warrant establishing Emergency Planning

BVY 14-009 1 Attachment 1 / Page 7 of 59 Table 1 Exemptions Requested from 10 CFR 50.47(b) and 50.47(c)(2)

Item # Regulation in 10 CFR 50.47 Basis for Exemption the ingestion exposur. pathway EPZ appropriate to the locale hay Zones." Additionally, in the Statement of Considerations for the Final Rule bfor EP requirements for ISFSIs and for MRS facilities (60 FR 32430)

(Reference 19), the Commission responded to comments concerning site-specific emergency planning that includes evacuation of surrounding population for an ISFSI not at a reactor site, and concluded that, "The Commission does not agree that as a general matter emergency plans for an ISFSI must include evacuation planning."

Also see basis for 50.47(b).

12 10 CFR 50.47(b)(11): Means for controlling radiological No exemption is requested.

exposures, in an emergency, are established for emergency workers, The means for controlling radiological exposures shall include exposure guidelines consistent with EPA Emergency Worker and Lifesaving Activity Protective Action Guides.

13 10 CFR 50.47(b)(12): Arrangements are made for medical No exemption is requested.

services for contaminated injured individuals.

14 10 CFR 50.47(b)(13): General plans for recovery and reentry are No exemption is requested.

developed.

15 10 CFR 50.47(b)(14): Periodic exercises are (will be) conducted to No exemption is requested.

evaluate major portions of emergency response capabilities, periodic drills are (will be) conducted to develop and maintain key skills, and deficiencies identified as a result of exercises or drills are (will be) corrected.

16 10 CFR 50.47(b)(15): Radiological emergency response training is No exemption is requested.

provided to those who may be called on to assist in an emergency.

17 10 CFR 50.47(b)(16): Responsibilities for plan development and No exemption is requested.

review and for distribution of emergency plans are established, and

BVY 14-009 / Attachment 1 / Page 8 of 59 Table I Exemptions Requested from 10 CFR 50.47(b) and 50.47(c)(2)

Item # Regulation in 10 CFR 50.47 Basis for Exemption planners are properly trained.

18 10 CFR 50.47(c)(2): Generally, the plume exposuWe pathway EPZ Analyses have been developed indicating that, within 15.4 months after fOr nuclear powc. plants shall con.sit of an area about 10 miles (16 shutdown, no credible accident at VY will result in radiological releases kin) in radius and the ingestion pathway EPZ shall consist of an requiring offsite protective actions. The analysis of the potential area about 50 miles (80 kin) in radius. The exact size and radiological impact of an accident for VY in a permanently defueled cnfiguration of the EPZs SUrrounding a partic.ular nuclear power condition indicates that any releases beyond the site boundary are limited reactor shall be determined in relation to*local eFmergency to small fractions of the EPA PAG exposure levels. According to the response needs and capabilities as they a.e affected by such EPA's "Protective Action Guides and Planning Guidance for Radiological conditions as demo.graphy, topography, land characteristiG6, Incidents, Draft for Interim Use and Public Comment," dated March 2013 access rout.s, and jurisdiGtin;a*boundares.

" The size of the EPZs (PAG Manual), "EPZs are not necessary at those facilities where it is not ase may be determined on a case-by-case basis for gas cooled possible for PAGs to be exceeded off-site." (Reference 4).

nuclear reactors and for reactors with an authorized power level less than 250 MW thermal. The plans for the ingestion pathway A s if5 (

shall focus On such actions as are appropriate to protect the food

.ngcstien pathway.

BVY 14-009 / Attachment 1 / Page 9 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption 19 10 CFR 50 App E: Ill. The Final Safety Analysis Report; Site No exemption is requested.

Safety Analysis Report The final safety analysis report or the site safety analysis report for an early site permit that includes complete and integrated emergency plans under § 52.17(b)(2)(ii) of this chapter shall contain the plans for coping with emergencies. The plans shall be an expression of the overall concept of operation; they shall describe the essential elements of advance planning that have been considered and the provisions that have been made to cope with emergency situations. The plans shall incorporate information about the emergency response roles of supporting organizations and offsite agencies. That information shall be sufficient to provide assurance of coordination among the supporting groups and with the licensee. The site safety analysis report for an early site permit which proposes major features must address the relevant provisions of 10 CFR 50.47 and 10 CFR part 50, appendix E, within the scope of emergency preparedness matters addressed in the major features. The plans submitted must include a description of the elements set out in Section IV for the emergency planning zones (EPZs) to an extent sufficient to demonstrate that the plans provide reasonable assurance that adequate protective measures can and will be taken in the event of an emergency.

BVY 14-009 / Attachment 1 / Page 10 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption 20 10 CFR 50 App E Following docketing of its "Certification of Permanent Removal of Fuel from the Reactor Vessel," in accordance with 10 CFR 50.82(a)(1)(ii), VY IV Content of Emergency Plans will become a permanently shutdown facility with spent fuel stored in the SFP. In the EP Final Rule (76 FR 72596, Nov. 23, 2011) (Reference 5),

the NRC defined "hostile action" as, in part, an act directed toward a

1. The applicant's emergency plans shall contain, but not nuclear power plant or its personnel. This definition is based on the necessarily be limited to, information needed to demonstrate definition of "hostile action" provided in NRC Bulletin 2005-02. NRC compliance with the elements set forth below, i.e., organization Bulletin 2005-02 was not applicable to nuclear power reactors that have for coping with radiological emergencies, assessment actions, permanently ceased operations and have certified that fuel has been activation of emergency organization, notification procedures, removed from the reactor vessel. The NRC excluded non-power reactors emergency facilities and equipment, training, maintaining (NPRs) from the definition of "hostile action" at that time because an NPR emergency preparedness, and recovery, and Onsite protective is not a nuclear power plant and a regulatory basis had not been actions during hostile action. In addition, the emergency developed to support the inclusion of NPR in that definition. Likewise, response plans submitted by an applicant for a nuclear power SFPs are not a nuclear power plant.

reactor operating license under this part, or for an early site permit (as applicable) or combined license under 10 CFR part 52, shall contain information needed to demonstrate The following similarities between VY and NPRs show that the VY facility compliance with the standards described in § 50.47(b), and should be treated in a similar fashion as an NPR. Similar to NPRs, VY will they will be evaluated against those standards. pose lower radiological risks to the public from accidents than do power reactors because: (1) VY will be a permanently shutdown facility (with fuel stored in the SFP) and will no longer generate fission products; 2) Fuel stored in the VY SFP will have lower decay heat resulting in lower risk of fission product release in the event of a non-credible boil off or drain down event; and 3) no credible accident at VY will result in radiological releases requiring offsite protective actions.

21 IV. 2 This nuclear power reactor license applicant shall also See basis for 50.47(b)(1 0).

Wetrand d~listneofithinthme plqumFe txo urc pathwYayEPO f providc andanalysi~sofhithe tieleuired toevacupatewa arious transient and peFrmanent populations, using the moest recent U.S.

BVY 14-009 / Attachment 1 / Page 11 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption Cenus Burcau data as of the date the applicant submits its applicatioR to the ISNRC.

22 IV. 3 Nuclear power reactor licensees I . halluse NRC applroe See basis for IV.2.

evacuation time estimates (ETEs) and updates to the ETEG in the formulation Of protective action recommendations and shall providc the ETEs and ETE updates to_Stato and-local governmental auth*,ories f; r use indeveloping offs/te protective aGtion 6statcgies.

23 IV. 4 Within 365 days of the later of the date of the availability of See basis for IV.2.

the most recent decenn..ial ;ensus data from the U.S. Census Bureau Or Decemnber 23, 2011, nuclear power reactor licensees shall develop an ETE analysis using this decennial data and submit it under § 50.1 to the NR.* These 'icensees shall submit this ETE analysis to the NRC at least 180 days before using it to formn protective actio reomedations and providing it to State and loc~al govornmental outhorities for use In developin~g Offsite protective action strategies.

24 IV 5 During the ycars beeen decennial Genu*ses, nuclear power See basis for IV.2.

reacteFr1leensees shall estimate EPZ pefrmaneRt resident population changes once a year, but no later than 365 days fFrom the date of the previous estimate, using the Most Frcent U.S.

Cznsus Bureau annual resident population estimate and State!local government pepulation data, if available. These icensees shall mainrtai these estimates so that they arc available for NRC inspection during the period between decennial censuses

____and shall submit these estimates to the NRC with any updated ________________________________

BVY 14-009 1 Attachment 1 / Page 12 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption EEa-alysis.

25 IV6 if at any time du.rig the dc-nnial period, the EPZ permanent See basis for IV.2.

r*o;d*e*n ppulatior increases such thatit causes the Iongest ETE halue for the 2 iOle zon or 5

-m.. zoIne including all affected Emergenicy Response Planning Areas, Or for the enti,-o10 mile EPZ to inro~ase by 25 percent Or 30 minutes, whichever is less, from the nuclear power reactor licensee's currently NRC appro..d Or updated ETE, the licensee shall update the ETE analysis to reflect the impact of that population increase. The licensee shall submit the updated ETE analysis to the NRc under § 50.1 no later thar365 days after the licensee's determination that the criteria for updating the ETt have been met and at least 180 days before using it toform protective actoon recolmendatioes and providig it to State and lcal governmental authioitcies for us in developirng offsite pnstetiveactnnstrategies.r 26 IV 7 After an applicant for a combined license under part 52 of this No exemption is requested. VY is not an applicant for a combined chapter receives its license, the licensee shall conduct at least one license, and therefore, this regulation is not applicable to VY.

review of any changes in the population of its EPZ at least 365 days prior to its scheduled fuel load. The licensee shall estimate EPZ permanent resident population changes using the most recent U.S. Census Bureau annual resident population estimate and State/local government population data, if available. Ifthe EPZ permanent resident population increases such that it causes the longest ETE value for the 2-mile zone or 5-mile zone, including all affected Emergency Response Planning Areas, or for the entire 10-mile EPZ, to increase by 25 percent or 30 minutes, whichever is

BVY 14-009 / Attachment 1 / Page 13 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption less, from the licensee's currently approved ETE, the licensee shall update the ETE analysis to reflect the impact of that population increase. The licensee shall submit the updated ETE analysis to the NRC for review under § 50.4 of this chapter no later than 365 days before the licensee's scheduled fuel load.

27 A Organization No exemption is requested.

The organization for coping with radiological emergencies shall be described, including definition of authorities, responsibilities, and duties of individuals assigned to the licensee's emergency organization and the means for notification of such individuals in the event of an emergency. Specifically, the following shall be included:

28 A.1. A description of the normal plant epe-ating organization. Appendix A to 10 CFR Part 50, "General Design Criteria for Nuclear Power Plants," states in part: "... there may be water-cooled nuclear power units for which fulfillment of some of the General Design Criteria may not be necessary or appropriate. For plants such as these, departures from the General Design Criteria must be identified and justified." In Appendix A, a nuclear power unit is defined as a nuclear power reactor and associated equipment necessary for electric power generation and includes those structures, systems, and components required to provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public. Following docketing of the certifications required by 10 CFR 50.82(a)(1), VY will not be a facility that can be operated to generate electrical power. Therefore, VY will not have a "plant operating organization." Rather, the station will

BVY 14-009 / Attachment 1 / Page 14 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption be maintained by a defueled on-shift staff.

29 A 2. A description of the onsite emergency response organization No exemption is requested.

(ERO) with a detailed discussion of:

a. Authorities, responsibilities, and duties of the individual(s) who will take charge during an emergency;

b. Plant staff emergency assignments;

c. Authorities, responsibilities, and duties of an onsite emergency coordinator who shall be in charge of the exchange of information with offsite authorities responsible for coordinating and implementing offsite emergency measures.

30 A 3. A dcScripticn, by position and function to be peFformed, of the The number of staff at VY during the decommissioning process will be licensee's h.adquarteF6 p..,....l who will be sent to the plant site small but commensurate with the need to safely store spent fuel at the to augment the E)..*tc emergency organization, facility in a manner that is protective of public health and safety.

Decommissioning sites typically have a level of emergency response that does not require response by headquarters personnel.

31 A 4. Identification, by position and function to be performed, of Analyses have been developed indicting that, within 15.4 months after persons within the licensee organization who will be responsible for shutdown, no credible accident at VY will result in radiological releases making effsite-dose projections and a description of how these requiring offsite protective actions.

projections will be made and the results transmitted to State and local authorities, NRC, and other appropriate governmental VY will still be able to determine ifa radiological release is occurring. If a

BVY 14-009 / Attachment 1 / Page 15 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption entities. release is occurring, then VY will evaluate the need to promptly communicate that information to offsite authorities for their consideration.

The offsite organizations are responsible for deciding what, ifany, protective actions should be taken.

32 A 5. Identification, by position and f*unctin t be perfcrmed, ot The time available to initiate compensatory actions in the event of a loss

,tho .empleyc.s of the licensec with .pecial qualifications fo" of SFP cooling or inventory precludes the need to identify and describe coing with em...geRcY condition that may arise. Other persons the special qualification of these individuals in the emergency plan. The with p.eil qu..alifications, su'h as consultants, who arFe* n number of staff at VY once it is in the permanently defueled state will be empley,.s of the licensee and whomay be called upon for small but will be commensurate with the need to operate the facility in a asi.stance fo emern. ;cies shall also be identified. The special manner that is protective of public health and safety.

qualifications of those persons shall be described.

33 A 6. A description of the local offsite services to be provided in No exemption is requested.

support of the licensee's emergency organization.

34 A 7. By June 23, 2011, identification of, and a desc3rp*tion f the Requiring a licensee for a decommissioning site to provide a description assistance expected from, appropriate State, local, and Federal of the assistance expected from appropriate State, local, and Federal agencies with responsibilities for coping with emergencies, agencies with responsibilities for coping with emergencies is an i*nluding hostile action at the site. For purposes of this appendix, unnecessary burden on the licensee, in light of the low risk of "hostile action" is defined as an act directed toward a nuclear emergencies necessitating offsite assistance.

powe Fplant Or its personnel that inldst eusof Yiolent force to destroy equipment, take hostages, and/or intim.idate the licensee to Requiring an identification and description of the assistance expected achieve an end. This inc.ludes attack by air, land, or water u6sig from appropriate State, local, and Federal agencies with responsibilities guns, explosiv*s, projectiles, vehicles, or other devices Used to for coping with hostile action at the site is unnecessary because, as deli'ver destructive forcoe explained in section IV.1, a decommissioning power reactor is exempt from requirements in Appendix E related to a "hostile action."

BVY 14-009 / Attachment 1 / Page 16 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix .E Item # Regulation in Appendix E to Part 50 Basis for Exemption 35 A 8. Identification of the State and/or local offi.ialS r*sponsiblc for Offsite emergency measures are limited to support provided by local planning for, *d,*i*ng, and controlGing appropriate protective police, fire departments, and ambulance and hospital services as actions, including e.acuation. when necessar,.. appropriate. Because analyses have been developed indicating that within 15.4 months after shutdown, no credible accident at VY will result in radiological releases requiring offsite protective actions, protective actions such as evacuation should not be required.

Also see basis for 50.47(b)(10).

36 A 9. By December 24, 2012, fo*rUnclearc*re. . reactor licensees, a In the EP Final Rule (Reference 5), the NRC acknowledged that the detailedanalysis demonstrating that on shift personnel assigned staffing analysis requirement was not necessary for non-power reactor emergency.plan impn are not assign'ed licensees because staffing at non-power reactors is generally small,

..p.-osibilitis that would prevent the timely peforancc of thei which is commensurate with operating the facility in a manner that is asigned funtions as spe.ified in the emergency plan. protective of the public health and safety. The minimal systems and equipment needed to maintain the spent nuclear fuel in the spent fuel pool or in a dry cask storage system in a safe condition requires minimal personnel and is governed by Technical Specifications. Because of the slow rate of the event scenarios postulated in the design basis accident and postulated beyond design basis accident analyses and because the duties of the on-shift personnel at a decommissioning reactor facility are not as complicated and diverse as those for an operating reactor, significant time is available to complete actions necessary to mitigate an emergency without impeding timely performance of emergency plan functions. For all of these reasons, it can be concluded that a decommissioning NPP is exempt from the requirement of 10 CFR Part 50, Appendix E, Section IV.A.9.

BVY 14-009 / Attachment 1 / Page 17 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption 37 B. Assessment Actions VY will adopt the Permanently Defueled EALs detailed in Appendix C of NEI 99-01, Revision 6. VY proposes to continue to review EALs with the B.1. The means to be used for determining the magnitude of, and State of Vermont on an annual basis. However, based upon the reduced for continually assessing the impact of, the release of radioactive scope of EALs for the permanently defueled facility, the scope of the materials shall be described, including emergency action levels annual review of EALs is expected to be reduced (informal mailings, etc.).

that are to be used as criteria for determining the need for notification and participation of local and State agencies, the Also see basis in section IV.1 for the justification from the requirements in Commission, and other Federal agencies, and the emergency Appendix E related to "hostile action."

action levels that are to be used for determining when and what type of protective measures should be considered within aPA eutside the site boundary to protect health and safety. The emergency action levels shall be based on in-plant conditions and instrumentation in addition to onsite and effsote monitoring. By Junh20, 2012, fFr nU..*r pe*,. r*a*tor licensees, thesc action levels mAust includc hostile action that mnay advcrccly affcct the nu.l.ar powc. plant. The initial emergency action levels shall be discussed and agreed on by the applicant or licensee and state and local governmental authorities, and approved by the NRC.

Thereafter, emergency action levels shall be reviewed with the State and local governmental authorities on an annual basis.

38 B.2. A licensee desiring to change its entire emergency action level No exemption is requested.

scheme shall submit an application for an amendment to its license and receive NRC approval before implementing the change. Licensees shall follow the change process in § 50.54(q) for all other emergency action level changes.

BVY 14-009 / Attachment 1 / Page 18 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption 39 C. Activation of Emergency Organization The Permanently Defueled EALs, detailed in Appendix C of NEI 99-01, Revision 6, will be adopted. This scheme eliminates the Site Area C.1. The entire spectrum of emergency conditions that involve the Emergency and General Emergency event classifications. Additionally, alerting or activating of progressively larger segments of the total the need to base EALs on containment pressure and the response of the emergency organization shall be described. The communication ECCS is no longer appropriate for notification of offsite agencies.

steps to be taken to alert or activate emergency personnel under each class of emergency shall be described. Emergency action Containment parameters do not provide an indication of the conditions at levels (based not only on onsite and-effsite radiation monitoring a defueled facility and emergency core cooling systems are no longer information but also on readings from a number of sensors that required. Other indications such as SFP level or temperature will be used indicate a potential emergency, such as the pressure in while there is spent fuel in the SFP.

containment and the response of the Emergency Core CoolingU System) for notification of offsite agencies shall be described. The In the Statement of Considerations for the Final Rule for EP requirements existence, but not the details, of a message authentication scheme for ISFSIs and for MRS facilities (60 FR 32430) (Reference 19), the shall be noted for such agencies. The emergency classes defined Commission responded to comments concerning a general emergency at shall include: (1) Notification of unusual events, (2) alert, (3)-site an ISFSI and MRS, and concluded that, "...an essential element of a area emFergency, and (4) general emergency. These classes are General Emergency is that a release can be reasonably expected to further discussed in NUREG-0654/FEMA-REP-1. exceed EPA Protective Action Guidelines exposure levels off site for more than the immediate site area." The probability of a condition reaching the level above an emergency classification of alert is very low.

In the event of an accident at a defueled facility that meets the conditions for relaxation of EP requirements, there will be time to take ad hoc measures to protect the public.

As stated in NUREG-1 738, for instances of small SFP leaks or loss of cooling scenarios, these events evolve very slowly and generally leave many days for recovery efforts. Offsite radiation monitoring will be performed as the need arises. Due to the decreased risks associated with

BVY 14-009 / Attachment 1 / Page 19 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption defueled plants, offsite radiation monitoring systems are not required.

40 C.2. By June 20, 201-2, nuclear power reactor licensees shall In the Proposed Rule (74 FR 23254) (Reference 21) to amend certain establish and maintain the capability to assess, classify, and emergency planning requirements for 10 CFR Part 50, the NRC asked for declare an emergency condition within 15 mint after the public comment on whether the NRC should add requirements for non-availability of indications to plant operators that an emergency power reactor licensees to assess, classify, and declare an emergency action level has been exceeded and shall promptly declare the condition within 15 minutes and promptly declare an emergency emergency condition as soon as possible following identification of condition. The NRC received several comments on these issues. The the appropriate emergency classification level. Licensees shall not NRC believed there may be a need for the NRC to be aware of security construe these criteria as a grace period to attempt to restore plant related events early on so that an assessment can be made to consider conditions to avoid declaring an emergency action due to an the likelihood that the event is part of a larger coordinated attack.

emergency action level that has been exceeded. Licensees shall However, the NRC determined that further analysis and stakeholder not construe these criteria as preventing implementation of interactions are needed prior to changing the requirements for non-power response actions deemed by the licensee to be necessary to reactor licensees. Therefore, the NRC did not include requirements in prot*ct public health and safety provided that any delay in the 2011 EP Final Rule (Reference 5) for non-power reactor licensees to deelaration does not deny the State and Iloal authoities the assess, classify, and declare an emergency condition within 15 minutes opportunity to implement measures

.prtectnecessary to the publi. and promptly declare an emergency condition. See basis in section IVA1 health and safety- for discussion on the similarity between a permanently defueled reactor and a non-power reactor for the low likelihood of any credible accident resulting in radiological releases requiring offsite protective measures.

41 D. Notification Procedures See basis for 50.47(b) and 50.47(b)(10).

D.1. Administrative and physical means for notifying local, State, and Federal officials and agencies and agreements reached with these officials and agencieus for the prompt notification of the public-and for public evacuation or other protective measures, should they beo*me necessary, shall be described. This description shall

BVY 14-009 / Attachment 1 / Page 20 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption include identification of the appropriate officials, by title and ageReY, of the State and local government agencies within he 42 D.2. Pro.ision. shall be described fr yearly disse*in*ation to the See basis for section IV.D.1.

public within the plume exposure pathway EPZ of basic mrgny -

plannirn information, such as the mFethods and times required for publ notification and the protectiye actions planned if an acciden occurS, general information as to the nature and effSct otx radiation, and a lmting of local broadcast stations that will be uscd for dissemination of informsation during an emhergency. Signv or other m..asu.es shall also be used to disseminate to any transien population within the plume exposure pathway EPZ appropriate information that would be helpful ifan accident Gocurs-.

43 D.3. A licensee shall have the capability to notify responsible State While the capability needs to exist for the notification of offsite and local governmental agencies within 4-6 minutes after declaring government agencies within a specified time period, previous exemptions an emergency. The licensee shall demonstrate that the appropriate have allowed for extending the State and local government agencies' governmental aulthOorties have the capability to mnake a public notification time up to 60 minutes based on the site-specific justification ale~ting and notification decision promnptly on being informed by the provided.

lienesee of an emnergency condition. Prior to initial operation greater as than ....

eah 5 percent .... th Of rated eaF power te power

÷-- alnse of the

... shall. ....first reactor demnstat at

÷hat.. VY proposes to complete emergency notification within 60 minutes after aste, each.nu.lear.o.er.reactr.licenseeshall.demostrate. an emergency declaration or a change in classification to the State of admini.trative and physical m have been established fo

.eans Vermont. This timeframe is consistent with the 10 CFR 50.72(a)(3) alerting aRd providing prompt instru.tions to the public within thee notification to the NRC and is appropriate because in the permanently plume exposure pathway EPZ. The design objective of the prom.p defueled condition, the rapidly developing scenarios associated with public ale.t and notification system shall be to have the capability to events initiated during reactor power operation are no longer credible and essentially .omplete the initial ale..ing and initiate notification of the there is no need for State or local response organizations to implement

____public within the plume exposure pathway E=PZ within about 15 ________________________________

BVY 14-009 / Attachment 1 / Page 21 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption minutes. The use of this alrti-g and notific*÷ation apability.i.l any protective actions.

rangc from immediate alerting and notiffication of the public (Within 15 minutes of the time that State and locao.. fficials are notifi.d that Also see basis for 50.47(b) and 50.47(b)(10).

a situation exists requiring urgent action) to the more likely events whcre thcrc is substantial tim~e -;Faailabl-Pfor the appropriate governmental authorities to make a judgmetnt;hether or not to activate the public ale. t ad. notification system. The a1Fting and notWifiation capability shall additionally include administrative and physigal mean for method of public alerting and a backup notification capable of being used in the event the primary mnethod of alerting and notification is unavailable during an emnergency to alert Or notify all Or portions of the plume exposure pathway EPZ population. The backup method shall have the capability to alert and notify the public within the plume exposure pathway EPZ=, but does not need to mneet the 15 min ute design objecative for the primar; prompt publicalert and notification systemn. When there is a decision to activate the alert and notification system, thee appropriate governmental authorities will dete~rmnwe whether to activate the entire alert and notification systemn simultaneously or i a graduated orstaged manner. The responsibility foractivating suc~h a public alert andnotification system shallremnafin with the appropriate governmental authorities.

44 0.4. If EEM.A has approved a nuclear power reactor site's alert and See basis for section IV D.3 regarding the alert and notification system notification design report, including the back~up alert and requirements.

notification capability, as of Decemnber 23, 2011, then the backup alert and notificatian capability requirements in Section IV.D.3 must be implemented by Decemnber 24, 2012. if the alert and notification

_____design report does nOt include a backup alert and notification

BVY 14-009 / Attachment 1 / Page 22 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption capabilityOoReecds rcvision to enSUre adequate backup alert and notifi-ationRapability, thenRa rvisiOr of the alert and not'fication design repor-t must be submitted to FEMA for rcview by Junc 24, 2013, and the FEMA approvcd backup alc~rt -and-notification means MUMt be implemented within 365 days after FEMA approval.

Howeycr, the total time pcriod to implement a FEMA approved backup alert and notification mneansmust not excced Junc 22, 45 E. Emergency Facilities and Equipment No exemption is requested.

Adequate provisions shall be made and described for emergency facilities and equipment, including:

E.1. Equipment at the site for personnel monitoring; 46 E.2. Equipment for determining the magnitude of and for No exemption is requested.

continuously assessing the impact of the release of radioactive materials to the environment; 47 E.3. Facilities and supplies at the site for decontamination of onsite No exemption is requested.

individuals; 48 E.4. Facilities and medical supplies at the site for appropriate No exemption is requested.

emergency first aid treatment;

BVY 14-009 / Attachment 1 / Page 23 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption 49 E.5. Arrangements for medical service providers qualified to No exemption is requested.

handle radiological emergencies onsite; 50 E.6. Arrangements for transportation of contaminated injured No exemption is requested.

individuals from the site to specifically identified treatment facilities outside the site boundary; 51 E.7. Arrangements for treatment of individuals injured in support of No exemption is requested.

licensed activities on the site at treatment facilities outside the site boundary; 52 E.8.a. (i) A licensee "nite technical SUPPOt conter and an Due to analyses indicting that, within 15.4 months after shutdown, no omorgency epcratiOns facility from which effective direction can be credible accident at VY will result in radiological releases requiring offsite given and effective control can be exercised during an emergency; protective actions, offsite agency response will not be required at an emergency operations facility (EOF) and onsite actions may be directed from the control room or other location, without the requirements imposed on a Technical Support Center (TSC).

An onsite facility will continue to be maintained, from which effective direction can be given and effective control may be exercised during an emergency. The VY emergency plan will continue to maintain arrangements for requesting assistance and using resources from appropriate offsite support organizations.

53 E.8.a (ii) For n'uccar power rcectOr licensees, a 1icensce onsite NUREG-0696, "Functional Criteria for Emergency Response Facilities,"

operational SUPPOrt GRcter; (Reference 22) provides that the operational support center (OSC) is an onsite area separate from the control room and the TSC where licensee operations support personnel will assemble in an emergency. For a

BVY 14-009 / Attachment 1 / Page 24 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption defueled power plant, an OSC is no longer required to meet its original purpose of an assembly area for plant logistical support during an emergency. A single onsite facility will continue to be maintained at VY, from which control room support, emergency mitigation, radiation monitoring, and effective control may be exercised during an emergency.

54 E.8.b. For a nuclear power reactor licensee's emergency No exemption is requested.

operations facility required by paragraph 8.a of this section, either a facility located between 10 miles and 25 miles of the nuclear power reactor site(s), or a primary facility located less than 10 miles from the nuclear power reactor site(s) and a backup facility located between 10 miles and 25 miles of the nuclear power reactor site(s). An emergency operations facility may serve more than one nuclear power reactor site. A licensee desiring to locate an emergency operations facility more than 25 miles from a nuclear power reactor site shall request prior Commission approval by submitting an application for an amendment to its license. For an emergency operations facility located more than 25 miles from a nuclear power reactor site, provisions must be made for locating NRC and offsite responders closer to the nuclear power reactor site so that NRC and offsite responders can interact face-to-face with emergency response personnel entering and leaving the nuclear power reactor site. Provisions for locating NRC and offsite responders closer to a nuclear power reactor site that is more than 25 miles from the emergency operations facility must include the following:

55 E.8.b. (1) Space for members of an NRC site team and Federal,

BVY 14-009 / Attachment 1 / Page 25 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption State, and local responders 56 E.8.b. (2) Additional space for conducting briefings with emergency response personnel; 57 E.8.b.(3) Communication with other licensee and offsite emergency response facilities; 58 E.8.b.(4) Access to plant data and radiological information; and 59 E.8.b.(5) Access to copying equipment and office supplies; 60 E.8.c. By June 20, 2012, for a nuclear power reactor licensce's See basis for 50.47(b)(3).

emAergency pcratiens facilit' required by paragraph 8.a of this section, a facility having the following capabilities:

(1) The capability for obtaining and displaying plant data and radiological information for each reactor at a nuclear pewer reactor site and foreach nuclearpower eactor cte that the facility sewes; 61 E.8.c (2)The capability to analyzc plant technical informnation and prWie technical briefingS on event conditions and pFroSgnoi to lien-sPe Mand OffSite response organizations for each reactor at a nuclear power reactor site and for each nuelear power reactor site that the facility scres; and

BVY 14-009 / Attachment 1 / Page 26 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption 62 E.8.c (3) The capability to SUPPOrt response to eventS occurring iultancously at mAore than one nuclIear power reartor cite if the cernegoncY opcrationS facilit' ser.'es moro than ono site; and 63 E.84 For nuclear power reactor licensees, an alternative fa.Gc*i' See basis for section IV.1 regarding hostile action.

(Or faGilities) that would be ac-ssible even ifthe site is under threat Of Or expdrienGi*g hostile actior, t function as a staging area for augmentation of emnergency responsc staff and 64 lE.8tively haVing the follonot g characteristics: the capability for oEmmunieation with the emaergeny operations facility, control roaom, and plant security; the capability to pe.rm offsitc a notifications; and the eapability fOr engineeing assesmnt activities, including damage control teamn planning and preparation, for use when onsite em ergeney facilities cannot be safel accessed during hostile action. The requirements in this paragraph 8Ad must be implemented no later than Decemnber 23, 2014, with the exception of the capability for staging emergency response organization personnel at the alternative facility (Or facilities) and the capability fOr commnAUications with the emnergency operations facility, control room, and plant security, which must be implemented no later than June 20, 2012.

64 E.8.e. A Wiczsee shall not be subject to the requiremnents ot See basis for 50.47(b)(3).

paragraph 8.b of this sectiGn for an existing emnergency operations facility approved as, of December 23, 2011;,

65 E.9. At least one onsite and one offsite communications system; See basis for 50.47(b) and (b)(1 0).

each system shall have a backup power source. All communication

BVY 14-009 / Attachment 1 / Page 27 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption plans shall have arrangements for emergencies, including titles VY will maintain communications with the State of Vermont and the NRC.

and alternates for those in charge at both ends of the The onsite response facilities will be combined into a single facility.

communication links and the primary and backup means of communication. Where consistent with the function of the governmental agency, these arrangements will include:

E.9.a. Provision for communications with contiguous State/local governments within the plume exposure pathway EPZ.. Such communications shall be tested monthly.

66 E.9.b. Provision for communications with Federal emergency No exemption is requested.

response organizations. Such communications systems shall be tested annually.

67 E.9.c. Provision for communications among the RurleaF-peweF Due to analyses indicting that, within 15.4 months after shutdown, no reactor control room, the onsite technical support Ge*ter, and the credible accident at VY will result in radiological releases requiring offsite em.ergency operations fa-ili'-; and among teh nuclear facility, the protective actions, there is no need for the TSC, EOF or field assessment principal State and local emergency operations centers, and-the teams. An onsite facility will continue to be maintained, from which field assessment teams. Such communications systems shall be effective direction can be given and effective control can be exercised tested annually. during an emergency. VY will also continue to test communication systems used to contact the State EOCs on an annual basis.

Also see justification for 50.47(b)(3).

BVY 14-009 / Attachment 1 / Page 28 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption 68 E.9.d. Provisions for communications by the licensee with NRC The functions of the control room, EOF, TSC and OSC may be combined Headquarters and the appropriate NRC Regional Office into one or more locations due to the smaller facility staff and the greatly Operations Center from the nuclear power reactor control room, reduced required interaction with State and local emergency response theO

.. itc th*, ,ical SUPPOrt center, and te .m..gen.y operations facilities. An onsite facility will continue to be maintained, from which facility. Such communications shall be tested monthly. effective direction can be given and effective control may be exercised during an emergency. VY will maintain communications with the NRC.

Also see basis for 50.47(b).

69 F. Training No exemption is requested.

F.1. The program to provide for: (a) The training of employees and exercising, by periodic drills, of emergency plans to ensure that employees of the licensee are familiar with their specific emergency response duties, and (b) The participation in the training and drills by other persons whose assistance may be needed in the event of a radiological emergency shall be described. This shall include a description of specialized initial training and periodic retraining programs to be provided to each of the following categories of emergency personnel 70 F.1. i. Directors and/or coordinators of the plant emergency organization; 71 F.1. ii.Personnel responsible for accident assessment, including control room shift personnel;

BVY 14-009 / Attachment 1 / Page 29 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption 72 F.1. iii Radiological monitoring teams; 73 F.1. iv. Fire control teams (fire brigades);

74 F.1. v. Repair and damage control teams; 75 F.1. vi. First aid and rescue teams; 76 F.1. vii. Medical support personnel; 77 F.1. viii. LiGeRcs headquarterSpp

, . t*,perSOnnel; The number of staff at VY during the decommissioning process will be small but commensurate with the need to safely store spent fuel at the facility in a manner that is protective of public health and safety.

Decommissioning sites typically have a level of emergency response that does not require additional response by headquarters personnel.

Therefore, exempting licensee's headquarters personnel from training requirements is considered to be reasonable.

78 F.1. ix. Security personnel. No exemption is requested.

79 F.A In addition, a radiological orientation training program shall be Because there will no longer be any expected actions that must be taken made available to local services personnel; e.g., local emergency by the public during an emergency, it is no longer necessary to pre-plan services/GqvM Defense, local law enforcement personnel,-ecGaI the dissemination of this information to the public or to provide news m..dia pe"S.. S. radiological orientation training to local news media persons.

BVY 14-009 / Attachment 1 / Page 30 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption The phrase "Civil Defense" is no longer a commonly used term and is no longer applicable as an example in the regulation.

80 F.2. The plan shall describe provisions for the conduct of Due to analyses indicting that, within 15.4 months after shutdown, no emergency preparedness exercises as follows: credible accident at VY will result in radiological releases requiring offsite protective actions, the public alert and notification system will not be used Exercises shall test the adequacy of timing and content of and therefore requires no testing.

implementing procedures and methods, test emergency equipment and communications networks, test the publi. .-lert

• d*,- n,-oificatien Also see basis for 50.47(b).

system, and ensure that emergency organization personnel are familiar with their duties.

81 F.2.a. A full participation exorci.e.. hich tests as m.uch of the VY will continue to invite the State of Vermont and local support licensee, State, and local emerg.ncy plan a .i. reasonabl" organizations to participate in the periodic drills and exercises conducted achwablc -'without mandator; public, partiipati* n ,hallbe to assess its ability to perform responsibilities related to an emergency at conduct, d for each site at which a power reactor is located. VY to the extent defined by the VY emergency plan and State emergency Nuelr*, power

.. actor licens..ees shall submit exerci.. . c c... - plans. Because the need for off-site emergency planning is relaxed due to under § 50.4 at least 60 days before use in a full participation the low probability of design-basis accidents or other credible events that e..rci. re.uireby this paragraph 2.a. would be expected to result in an offsite radioactive release that would exceed the limits of EPA PAGs and the available time for event 82 F.2.a(i) For an operating license issued under this part, this mitigation, no off-site emergency plans will be in place to test.

exercise must be conducted within two years before the issuance of the first operating license for full power (one authorizing The intent of submitting exercise scenarios at power reactors is to check operation above 5 percent of rated power) of the first reactor and that licensees utilize different scenarios in order to prevent the shall include participation by each State and local government preconditioning of responders at power reactors. For defueled sites, there within the plume exposure pathway EPZ and each state within the are limited events that could occur and the previously routine progression ingestion exposure pathway EPZ. Ifthe full participation exercise is to General Emergency in power reactor site scenarios is not applicable to conducted more than 1 year prior to issuance of an operating

BVY 14-009 / Attachment 1 / Page 31 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption licensee for full power, an exercise which tests the licensee's a decommissioning site.

onsite emergency plans must be conducted within one year before issuance of an operating license for full power. This exercise need ENO considers VY to be exempt from F.2.a.(i)-(iii) because VY will be not have State or local government participation. exempt from the umbrella provision of F.2.a.

83 F 2.a.(ii) For a combined license issued under part 52 of this chapter, this exercise must be conducted within two years of the scheduled date for initial loading of fuel. If the first full participation exercise is conducted more than one year before the scheduled date for initial loading of fuel, an exercise which tests the licensee's onsite emergency plans must be conducted within one year before the scheduled date for initial loading of fuel. This exercise need not have State or local government participation. If FEMA identifies one or more deficiencies in the state of offsite emergency preparedness as the result of the first full participation exercise, or if the Commission finds that the state of emergency preparedness does not provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency, the provisions of § 50.54(gg) apply.

84 F 2.a (iii) For a combined license issued under part 52 of this chapter, if the applicant currently has an operating reactor at the site, an exercise, either full or partial participation, shall be conducted for each subsequent reactor constructed on the site.

This exercise may be incorporated in the exercise requirements of Sections IV.F.2.b. and c. in this appendix. If FEMA identifies one or more deficiencies in the state of offsite emergency preparedness as the result of this exercise for the new reactor, or if the Commission finds that the state of emergency preparedness does

BVY 14-009 / Attachment 1 / Page 32 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption not provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency, the provisions of § 50.54(gg) apply.

85 F 2.b. Each licensee at each site shall conduct a subsequent See basis for section IV.F.2.a.

exercise of its onsite emergency plan every 2 years.-Nuelear-power reactor lceneRes snail submit exercise Scenarios unaer ý 50.4 at The low probability of a design-basis accident or other credible events least 60 days bcfore use in an exrcs re-quired by this paragraph that would result in an offsite radioactive release that would exceed the 2.b. The exercise mnay be included in the full participation biennial EPA PAGs and the available time for event mitigation at VY during exercis requirFed by paragraph 2.e. of this section. In addition, the decommissioning render TSCs, OSCs and EOFs unnecessary. The licensee shall take actions necessary to ensure that adequate principal functions required by regulation can be performed at an onsite emergency response capabilities are maintained during the interval location that does not meet the requirements of the TSC, OSC or EOF.

between biennial exercises by conducting drills, including-at least The onsite response facilities at VY will be combined into a single facility.

one drill involving a combination of some of the principal functional areas of the licensee's onsite emergency response capabilities.

VY will continue to conduct biennial exercises and will invite the State of The principal functional areas of emergency response include Vermont and local support organizations (firefighting, law enforcement, activities such as management and coordination of emergency and ambulance/medical services) to participate in periodic drills and response, accident assessment, event classification, notification of exercises to assess its ability to perform responsibilities related to an offsite authorities, assessment of the onsite an,d-effsite-impact of emergency at VY to the extent defined by the VY and State emergency radiological releases, pretcctive action rFecmmendation plans.

devclopment, pr"tective actio.n decisieon making, plant system repair and mitigative action implementation. During these drills, activation of all of the licensee's emergency response facilities (TehR~hic.-, Siinnnt Crentre (T-SC) I Onerntiens' Siinnnrt Ccente~

(C)SC), and the Em..ergency Operations Facility (-tOF)) would not be necessary, licensees would have the opportunity to consider accident management strategies, supervised instruction would be permitted, operating staff in all participating facilities would have the opportunity to resolve problems (success paths) rather than

BVY 14-009 / Attachment 1 / Page 33 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption have controllers intervene, and the drills may focus on the onsite exercise training objectives.

86 F 2.c. Offsite plans for each site shall be exercFsec Diennally WitR See basis for section IV.F.2.a.

radio'logcal respo*ns plan. Where the offsite authority has a role under a radiological response plan for more than one site, it shall fuly participate in one exercise ever; two years and shall, at least, partially participate in other offsite plan exercises in this period. It two difFe*rt licensees each have licensed facilities located either on the same site Or On adjacent, contiguous sites, and share most of the elements defining Go located licensees, then each licensee 87 F 2.c.(1) Conduct, an exercise biennially of its onsite emergenc 88 F 2.c.(2) Participate quad.ennially in an offsite biennial full Or par"tial participation eerIse 89 F 9 r (3'1 nnd.,z. A +; ;+; A interactions in the years between its participation in the offsite full Or partial *a*tic*iation exercise with .offitc authorities, to test and maintafin interface amonEg the affected State and local authorities and the licensee. Go loeated licensees shall also participate in emnergency preparedness activitfies and interaction with offsitc authorities for the period between exercises;

BVY 14-009 / Attachment 1 / Page 34 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption 90 F 2.c.(4) Conduct a hostile action exercise of its onsite emergency plan in each cxcrcisce cycle; and 91 F 2.c.(5) Participate in an offsite bicnnial full or partial pa*t*i*pat*Gn hostile action exercise on altornating exercise cycles.

92 F 2.d. Each State with responsibility for nuc!ear power reactor See basis for section IV.2.

emnergency preparedness should fully participate in the ingestion pathway portion of exercises at least once e Iey exrc.;se cycle. in States with more than one nuclear power reactor plume expoSUr pathway EPZ, the State should rotate this participation from site to site. Each State with responsibility for nuclear power reactor em~ergency prcparcdncss should fully participate in a hostile action ex(ercise at least ance everY cycle and should fully participate in one hostile action exercise by December 31, 2015. States with mnore than one RueleaF power reactorF plume exposure pathway EPZ should rotate this participation from site to site.

93 F 2.e. Licensees shall enable any State or local government See basis for section IV.2.

located within the plume exposure pathway EPZ to participate in the licensee's drills when requested by such State or local government.

94 F 2.f. Remedial exercises will be required if the emergency plan is The U.S. Federal Emergency Management Agency (FEMA) is not satisfactorily tested during the biennial exercise, such that responsible for the evaluation of an offsite response exercise. No action NRC, in consultat"on with FEM- , cannot (1) find reasonable is expected from State or local government organizations in response to assurance that adequate protective measures can and will be an event at a decommissioning site other than firefighting, law taken in the event of a radiological emergency or (2) determine that enforcement and ambulance/medical services. Memoranda of

BVY 14-009 / Attachment 1 / Page 35 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption the Emergency Response Organization (ERO) has maintained key understanding will continue to be in place for those services. Offsite skills specific to emergency response. The extent of State and response organizations will continue to take ad hoc actions to protect the local participation in remedial exercises must be sufficient to she. health and safety of the public as they would at any other industrial site.

that appropriate corrective mneaSUres have been ta;ken regarding the elements of the plan not properly tested in the previous 95 F 2.g. All exercises, drills, and training that provide performance No exemption is requested.

opportunities to develop, maintain, or demonstrate key skills must provide for formal critiques in order to identify weak or deficient areas that need correction. Any weaknesses or deficiencies that are identified in a critique of exercises, drills, or training must be corrected.

96 F 2.h. The participation of State and local governments in an No exemption is requested.

emergency exercise is not required to the extent that the applicant has identified those governments as refusing to participate further in emergency planning activities, pursuant to § 50.47(c)(1). In such cases, an exercise shall be held with the applicant or licensee and such governmental entities as elect to participate in the emergency planning process.

97 F 2.i. Licensees shall use drill and exercise scenarios that provide For defueled sites, there are limited events that could occur and the reasonable assurance that anticipatory responses will not result previously routine progression to General Emergency in power reactor from preconditioning of participants. Su1ch scenarios for nuclear site scenarios is not applicable to a decommissioning site. Therefore, power reac*to lien.sees must include a wide spectrum .. defueled sites are not expected to demonstrate response to a wide radiological relcases and eventS, including hostile act*i. Exercise spectrum of events.

and drill scenarios as appropriate must emphasize coordination

BVY 14-009 / Attachment 1 / Page 36 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption among onsite and offsite response organizations. Also see basis for section IV.1 regarding hostile action.

m.

98 F 2.j. The exercises condlucted under paragraph 2 of this section See basis for section IV.F.2.

by nuclear power rcactor iensccs must provide the eppeotunity Periodic drills and exercises will be completed to demonstrate ERO fOr the ERO to demonStrate proficiency in the key skills necsessar proficiency in key skills necessary to implement the principal functional to ipleentthe principal functional areas of emnergency response areas of emergency response as applicable for the permanently defueled plant status. Critiques will follow each drill or exercise activity. The VY provid the opprunity for the E-RO to demo.enstrate key skills emergency plan will discuss exercise and drill types and frequencies of spe-ific-to emergency response duties in the control room. TS-",

k÷ occurrence. Scenarios will be developed to test all major elements of the calendar yea eecie cycle, nuclear power reactor licensees sh-all emergency plan within an eight (8) year period. These elements include management and coordination of emergency response, accident Yar 3' the content Of scenarios during exerciseS conducted under assessment, and system repair and corrective action. VY will continue to paragraph 2 of this section to provide the opportunoty for the ERO include the State of Vermont and local support organizations in the to demoenstrate profiiency in the key skills necessar; to respondn- to0 periodic drills and exercises to assess its ability to perform responsibilities the foellowfing scenario elements: hostile action directed at the plant related to an emergency at VY to the extent defined by the VY and State s;+ite, -0 radilo'giGal release or an unplanned mi;rnial rladiolomg*al emergency plans.

release that does not require public protective actions, an initial

,lass.,- ;f o forrapid OFat- escalation to a Site Area Emergency or General Emergency, implementation of strategies, prcedures, and guidance developed under § 50.54(hh)(2), and integration ot offs; e resources with onsite response. The licensee shall maintain a-record of e-xercises conducted dur~ing each eight yeareecs cycle that documents the content Of scenarios used to comRply With the ire-quirm-ements of this paragraph. Each licensee shall conduct a hostile action exercise for each of its sites no later than Decemnber 341 .2015. The first eight yea ercs cycle for a site will begin i the calendar year in which the first hostile action exercis is conduct_'ed. For a site licensed under Part 52, the first eight year exercise cycle begins in the calendar year of the initial exercisee

BVY 14-009 / Attachment 1 / Page 37 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption required by Secstion IV.F.2.a-.

99 G. Maintaining Emergency Preparedness No exemption is requested.

Provisions to be employed to ensure that the emergency plan, its implementing procedures, and emergency equipment and supplies are maintained up to date shall be described.

100 H. Recovery No exemption is requested.

Criteria to be used to determine when, following an accident, reentry of the facility would be appropriate or when operation could be resumed shall be described.

101 I. Onsite Protective Actions During Hostile Action See basis for section IV.1.

By June 20, 2012, for nuclear power reactor licensees, a Fange of pro~tective actions to protect onsite personnel during hostile action mrust be developed to ensure the continued ability of the licensce to safely shut down the reactor and pe~ferm the functions of the Iensee's cmeF-egnY plan-.

102 10CFR 50 App E No exemption is requested.

V. Implementing Procedures No less than 180 days before the scheduled issuance of an operating license for a nuclear power reactor or a license to possess nuclear material, or the scheduled date for initial loading of fuel for a combined license under part 52 of this chapter, the applicant's or licensee's detailed implementing procedures for its emergency plan shall be submitted to the

BVY 14-009 1 Attachment 1 / Page 38 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Appendix E to Part 50 Basis for Exemption Commission as specified in § 50.4. Licensees who are authorized to operate a nuclear power facility shall submit any changes to the emergency plan or procedures to the Commission, as specified in

§ 50.4, within 30 days of such changes.

103 10CFR 50 App E The regulation that identifies the requirement to maintain the Emergency Response Data System (ERDS) is not applicable to nuclear power VI. Emergency Response Data System facilities that are permanently shutdown.

1. The Emergency Response Data System (ERDS) is a direct near real-time electronic data link between the licensee's onsite Once VY is permanently defueled, this system will no longer be computer system and the NRC Operations Center that provides for necessary to transmit safety system parameter data. No exemption is the automated transmission of a limited data set of selected requested since this change in the ERDS data requirement is identified in parameters. The ERDS supplements the existing voice 10 CFR 50 Appendix E, VI. 2.

transmission over the Emergency Notification System (ENS) by providing the NRC Operations Center with timely and accurate updates of a limited set of parameters from the licensee's installed onsite computer system in the event of an emergency. When selected plant data are not available on the licensee's onsite computer system, retrofitting of data points is not required. The licensee shall test the ERDS periodically to verify system availability and operability. The frequency of ERDS testing will be quarterly unless otherwise set by NRC based on demonstrated system performance.

2. Except for Big Rock Point and all nuclear power facilities that are shut down permanently or indefinitely, onsite hardware shall be provided at each unit by the licensee to interface with the NRC receiving system. Software, which will be made available by the NRC, will assemble the data to be transmitted and transmit data

BVY 14-009 / Attachment 1 / Page 39 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Basis for Exemption

BVY 14-009 / Attachment 1 / Page 40 of 59 Based on the analysis detailed below, ENO has concluded that the portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and 10 CFR 50, Appendix E, identified in the above tables will not be necessary to protect the health and safety of the public once VY is in the permanently defueled condition, and would be unduly burdensome. Approval of the exemptions requested above would not present an undue risk to the public or prevent appropriate response in the event of an emergency at VY.

Ill. Background VY is located in the town of Vernon, Vermont in Windham County on the west shore of the Connecticut River immediately upstream of the Vernon Hydrostation. VY is a boiling water reactor with a rated thermal power of 1912 MWt. The station is located on approximately 125 acres in Windham County, and is owned by Entergy Nuclear Vermont Yankee, LLC, with the exception of a narrow strip of land between the Connecticut River and the VY property for which Entergy Nuclear Vermont Yankee, LLC has perpetual rights and easements from the owner.

The site is bounded by the Connecticut River (Vernon Pond) on the east, by farm and pasture land mixed with wooded areas on the north and south, and by the town of Vernon on the west.

Warwick and Northfield State Forests (approximately 8 miles southwest of the site), Green Mountain National Forest (approximately 18 miles southwest of the site) and the Pisgah Mountain Range (northeast of the site) limit the population density and land use within a 50-mile radius of the site. Most of the land around the site is undeveloped. The developed land is used for agricultural, dairying, and for residential areas within small villages. The primary agricultural crop is silage corn, which is stored for year-round feed for milk cows.

The area within 10 miles of the site has only one urban area, the city of Brattleboro, Vermont (2000 population 12,005), which is located about 5 miles upriver. The remainder of this area is rural and contains several small villages with populations between 1,000 and 3,000. The average population density within a 10-mile radius of VY for 2000 was estimated to be 126 people per square mile.

Section 14 of the VY Updated Final Safety Analysis Report (UFSAR) describes the design basis accident (DBA) scenarios that are applicable to VY during power operations and describe the accidents with the greatest potential for radiation exposure of any accident considered under the same assumptions. The most severe postulated accidents for nuclear power plants involve damage to the nuclear reactor core and the release of large quantities of fission products. The UFSAR accident scenarios include a Control Rod Drop Accident (CRDA), a Loss-of-Coolant Accident (LOCA), a Refueling/Fuel Handling Accident (RA/FHA) and a Main Steam Line Break Accident.

Following docketing of its "Certification of Permanent Removal of Fuel from the Reactor Vessel," in accordance with 10 CFR 50.82(a)(1)(ii), and within two years following cessation of operations, VY will submit a Post-Shutdown Decommissioning Activities Report (PSDAR),

which will identify VYs selected method of decommissioning. At the end of the current operating cycle, the VY reactor will be permanently shut down. After the reactor is shut down, all fuel assemblies will be removed from the reactor vessel and placed in the SFP. The irradiated fuel will be stored in the SFP and the ISFSI until it is shipped off-site in accordance with the schedules described in the PSDAR and updated Irradiated Fuel Management Plan. The PSDAR will identify most fluid systems drained and the plant in a stable condition until final decontamination and dismantlement.

BVY 14-009 / Attachment 1 / Page 41 of 59 When the reactor is permanently defueled, the SFP and its supporting systems will be modified and dedicated only to spent fuel storage. With the reactor defueled, the reactor vessel assembly and supporting structures and systems are no longer in operation and have no function related to the safe storage and management of irradiated fuel in the SFPs. A fuel pool cooling and clean-up system is provided to remove decay heat from spent fuel stored in the fuel pool and to maintain a specified water temperature, purity, clarity, and level.

IV. Technical Evaluation A. Accident Analysis Overview 10 CFR 50.82(a)(2) specifies that the 10 CFR Part 50 license no longer authorizes operation of the reactor or emplacement or retention of fuel in the reactor vessel after docketing the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel in accordance with 10 CFR 50.82(a)(1). Following the termination of reactor operations at VY and the permanent removal of the fuel from the reactor vessel, the postulated accidents involving failure or malfunction of the reactor and supporting structures, systems and components are no longer applicable.

A summary of the postulated radiological accidents analyzed for the permanently shutdown and defueled condition of VY is presented below. According to the EPA, "Protective Action Guides and Planning Guidance for Radiological Incidents, Draft for Interim Use and Public Comment,"

dated March 2013 (Reference 4), Section 2.3.5, "PAGs and Nuclear Facilities Emergency Planning Zones (EPZ)," EPZs are not necessary at those facilities where it is not possible for PAGs to be exceeded off-site.

A. Consequences of Design Basis Events The postulated design basis accident that will remain applicable to VY in its permanently shutdown and defueled condition is the FHA in the reactor building where the SFP is located. A new analysis based on the FHA was performed to determine the dose to operators in the control room and the public at the Exclusion Area Boundary (EAB or "Site Boundary") and Low Population Zone (LPZ) as a function of time after shutdown. The analysis shows that the dose at the EAB and LPZ 17 days after shutdown (with open containment) is less than 1 rem TEDE, which is below the EPA PAG threshold of 1 rem for recommended evacuation.

Due to the amount of decay assumed (17 days), the results of this analysis may be applied after January 17, 2015, assuming a VY shutdown by the end of December 2014. The analysis was submitted for NRC review in Reference 6.

B. Consequences of Beyond Design Basis Events

a. Hottest Fuel Assembly Adiabatic Heatup - Beyond Design Basis Event The analysis in Attachment 2 is provided to compare the conditions for the hottest fuel assembly stored in the VY fuel pools to a criterion proposed in SECY-99-168 (Reference 7) applicable to offsite emergency response for the unit in the decommissioning process. This criterion considers the time for the hottest assembly to heat up from 30 degrees Celsius (°C) to 900°C adiabatically. If the heat up time is greater than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, then offsite emergency preplanning involving the plant is not necessary.

BVY 14-009 / Attachment 1 / Page 42 of 59 Based on the limiting fuel assembly for decay heat and adiabatic heatup analysis, at 15.4 months after shutdown (15.4 months decay time), the time for the hottest fuel assembly to reach 900°C is 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after the assemblies have been uncovered. As stated in NUREG-1 738, "Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants" (February 2001) (Reference 8), 9000C is an acceptable temperature to use for assessing onset of fission product release under transient conditions (to establish the critical decay time for determining availability of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> to evacuate) if fuel and cladding oxidation occurs in air.

Because of the length of time it would take for the adiabatic heatup to occur, there is ample time to respond to any partial drain down event that might cause such an occurrence by restoring cooling or makeup, or providing spray. As a result, the likelihood that such a scenario would progress to a zirconium fire is not deemed credible.

C. Consequences of Other Analyzed Events

a. Loss of Spent Fuel Pool Normal Cooling This analysis assesses the time available to initiate compensatory measures in the event of a loss of spent fuel pool inventory as well as the radiological impact. From Engineering Change 47710, the initiating event is postulated to be an external event that results in a prolonged loss of all Alternating Current (AC) power. In this scenario, there is no active cooling of the spent fuel pool, nor is there the ability to maintain pool water inventory with normal plant systems. This evaluation determined that 15.4 months following shutdown, the time to reach 212 degrees Fahrenheit will be 74 hours8.564815e-4 days <br />0.0206 hours <br />1.223545e-4 weeks <br />2.8157e-5 months <br />, and the total time from the loss of cooling to boil off inventory to 3 feet above the top of the fuel assemblies will be 16 days. Although no fuel damage is expected while the water level remains above the top of the fuel, a level of 3 feet above the top of the fuel was chosen for ease of comparison to the corresponding information contained in NUREG-1 738.

Three feet of water continues to provide sufficient shielding from radiation to any personnel involved in responding to the event. Due to the slow rate of spent fuel pool water boil-off, adequate time will be available to restore cooling or makeup, either through restoration of normal systems or through readily available mitigation measures, without significant radiological consequences for plant workers in the Reactor Building.

b. Radioactive Waste Handling Accident This analysis evaluated the drop of a high integrity container (HIC). The accident evaluated the drop of the largest liner containing the highest concentration of radioactive materials (dewatered resin containing 19,415 curies of 25 various radionuclides representing the highest activity waste at the facility). The calculation postulates that the container is dropped 250 meters (820 feet) from the closest site boundary with subsequent container failure with 1% of the liner contents released and 0.5% of the release becoming aerosolized and carried in the direction of the closest Site Boundary.

The resulting two hour integrated dose at the Site Boundary is projected to be 16.1 millirem TEDE, which is below the EAB limit of 1 rem TEDE.

B. Comparison to NUREG-1738 Industry Decommissioning Commitments and Staff Decommissioning Assumptions

BVY 14-009/ Attachment 1 / Page 43 of 59 Although the limited scope of design and beyond design basis accidents that remain applicable to VY justify a reduction in the necessary scope of emergency response capabilities, ENO also evaluated the industry decommissioning commitments (IDCs) and staff decommissioning assumptions (SDAs) contained in NUREG-1738.

NUREG-1738 contains the results of the NRC staff's evaluation of the potential accident risk in spent fuel pools at decommissioning plants in the United States. As stated therein, the study was undertaken to support development of a risk-informed technical basis for reviewing exemption requests and a regulatory framework for integrated rulemaking. The NRC staff performed analyses and sensitivity studies on evacuation timing to assess the risk significance of relaxed offsite emergency preparedness requirements during decommissioning. The staff based its sensitivity assessment on the guidance in Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" (Reference 9). The staff's analyses and conclusions apply to decommissioning facilities with SFPs that meet the design and operational characteristics assumed in the risk analysis.

The study found that the risk at decommissioning plants is low and well within the Commission's Safety Goals. The risk is low because of the very low likelihood of a zirconium fire (resulting from a postulated irrecoverable loss of SFP cooling water inventory) even though the consequences from a zirconium fire could be serious.

The study provided the following assessment:

"The staff found that the event sequences important to risk at decommissioning plants are limited to large earthquakes and cask drop events. For emergency planning (EP)assessments, this is an important difference relative to operating plants where typically a large number of different sequences make significant contributions to risk. Relaxation of offsite EP a few months after shutdown resulted in only a "small change"in risk, consistent with the guidance of RG 1.174. Figures ES- I and ES-2 [in NUREG- 1738] illustrate this finding. The change in risk due to relaxation of offsite EP is small because the overall risk is low, and because even under current EP requirements,EP was judged to have marginal impact on evacuation effectiveness in the severe earthquakesthat dominate SFP risk. All other sequences including cask drops (for which emergency planning is expected to be more effective) are too low in likelihood to have a significantimpact on risk.

For comparison, at operatingreactors,additionalrisk-significantaccidents for which EP is expected to provide dose savings are on the orderof lx10-5 per year, while for decommissioning facilities, the largestcontributorfor which EP would provide dose savings is about two orders of magnitude lower (cask drop sequence at 2x10-7 per year)."

The Executive Summary in NUREG-1 738 states, in part, "the staff's analyses and conclusions apply to decommissioning facilities with SFPs that meet the design and operational characteristics assumed in the risk analysis. These characteristics are identified in the study as IDCs and SDAs. Provisions for confirmation of these characteristics would need to be an integral part of rulemaking." The IDCs and SDAs are listed in Tables 4.1-1 and 4.1-2, respectively, of NUREG-1 738. The following tables show how the VY SFP meets or compares with each of these IDCs (Table 3) and SDAs (Table 4). Attachment 3 includes a new regulatory commitment to update the VY UFSAR with this information.

BVY 14-009/ Attachment 1 / Page 44 of 59 C. Consequences of a Beyond-Design Basis Earthquake In June 2013, a draft study, entitled "Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark 1 Boiling Water Reactor," was published for public comment (Reference 17). The purpose of the consequence study was to determine if accelerated transfer of older, colder spent fuel from the SFP at a reference plant to dry cask storage significantly reduces risks to public health and safety. The specific reference plant used for the study was a General Electric (GE) Type 4 BWR with a Mark I containment.

VY is a GE BWR/4 with a Mark I containment.

The study states: "Past risk studies have shown that storage of spent fuel in a high-density configuration is safe and risk of a large release due to an accident is very low. This study's results are consistent with earlier research conclusions that spent fuel pools are robust structures that are likely to withstand severe earthquakes without leaking. The NRC continues to believe, based on this study and previous studies that spent fuel pools protect public health and safety."

The study also estimated that the likelihood of a radiological release from the SFP resulting from the selected severe seismic event analyzed in the study was on the order of one time in 10 million years or lower. The study analyzed two cases for each scenario: one where mitigation measures of 10 CFR 50.54(hh)(2) were credited, and one where they were not used or were unsuccessful. It showed that successful mitigation reduces the likelihood of a release and that the likelihood of a release was equally low for both high- and low-density loading in the SFP.

The study did not consider the post-Fukushima mitigation measures required by Orders EA 049 (Mitigating Strategies Order) and EA-12-051 (Reliable Hardened Containment Vents Order).

D. Conclusion Based on the above, VY has demonstrated that no credible accident will result in radiological releases requiring offsite protective actions. Additionally, there is sufficient time, resources and personnel available to initiate mitigative actions that will prevent an offsite release that exceeds EPA PAGs.

BVY 14-009 / Attachment 1 / Page 45 of 59 TABLE 3 Industry Decommissioning Commitments (IDCs) Comparison IDC Industry Commitments Response 1 Cask drop analyses will be performed The VY design is in alignment with this description. Program Procedure PP-7023 "Control of or single failure-proof cranes will be in Heavy Loads Program Document" controls the handling of heavy loads to meet the guidance use for handling of heavy loads (i.e., provided in NUREG-0612. The cask handling crane (i.e. reactor building bridge crane) trolley was phase II of NUREG-0612 will be upgraded to address Phase I requirements of NUREG-0612 to provide redundancy in the load implemented). carrying path from the cask to the crane trolley itself, so that no single failure would allow the cask to drop. In addition to the trolley replacement, a comprehensive maintenance program and strict administrative control of all cask handling was implemented. The NRC Safety Evaluation Report for the acceptance of NUREG-0612 Phase I actions is documented in Reference 10.

As documented in Reference 11 the NRC considered Phase II to be an enhancement and completed without requiring completion of implementation actions identified during the Phase II review.

2 Procedures and training of personnel VY procedures are in place to ensure onsite and offsite resources can be brought to bear during will be in place to ensure that onsite and an event, including: ON-3177, Operations Response to an Aircraft Threat, OPOP-SECU-3132, offsite resources can be brought to bear OperationsDepartmentResponse to Security Events, OPOP-PHEN-3127, NaturalPhenomena, during an event, and ON-3157, Loss of Fuel Pool Level/Cooling. Additionally, EPOP-CR-3540, Control Room Actions During an Emergency, directs the Shift Manager to notify the Security Shift Supervisor to implement OP 3547, Security Actions during an Emergency, which activates the VY Pager System. This action notifies Emergency Personnel to report to the proper location. EPOP-CR-3540 also directs the ERO Notification System Activation.

These procedures are not specifically referenced in the existing VY Emergency Plan and will not be included in the planned Permanently Defueled Emergency Plan (to be submitted for NRC approval). These procedures are required by Technical Specification 6.4. Therefore, it is not necessary for them to be specifically referenced in the Emergency Plan. Equipment requirements are specified in the pertinent procedures.

Once VY is shutdown and defueled, the on-shift plant operators, including Certified Fuel Handlers (CFH), and fire brigade members will be appropriately trained on the various actions needed to provide makeup to the SFP based on a systematic approach to training. Once VY is no longer operating, maintaining SFP cooling and inventory would be the highest priority activity; therefore, the personnel needed to perform these actions will be available at all times. The VY CFH training program was submitted for NRC review and approval by letter dated October 31, 2013

BVY 14-009 / Attachment 1 / Page 46 of 59 (Reference 12).

Finally, quarterly Emergency Plan drills are conducted with frequent participation of the Offsite Response Organizations to maintain proficiency in response to a plant event.

3 Procedures will be in place to establish Procedures OP 3504, Emergency Communications, EPOP-CR-3540, Control Room Actions communication between onsite and During an Emergency, OPOP-PHEN-3127, NaturalPhenomena, AP 0153, Operations offsite organizations during severe Department Communicationsand Log Maintenance,and AP 3150, Extensive Damage Mitigation weather and seismic events. Strategy for Establishing Command and Control, provide guidance for initiating and maintaining communications between offsite agencies and the onsite Emergency Response Organization during severe weather and seismic events.

4 An offsite resource plan will be Appendix G, Loss of Large Areas of the Plant Due to Fire or Explosion, of Procedure PP 7019, developed which will include access to Severe Accident Management Program,contains an offsite resource list which shows providers, portable pumps and emergency power their capabilities, and a contact telephone number as well as actions for spent fuel pool damage to supplement onsite resources. The and supplying water to off-site fire support for external make-up.

plan would principally identify organizations or suppliers where offsite resources could be obtained in a timely manner.

5 SFP instrumentation will include VY design meets the intent of this IDC. There are two narrow range channels of continuous readouts and alarms in the control room remote indication of spent fuel pool water level in the control room. Each of these channels (or where personnel are stationed) for provides high and low annunciation on the Fuel Pool Filter Demineralizer Panel with a system SFP temperature, water level, and area trouble alarm and high and low level indicator lights in the control room. In addition, each of these radiation levels, channels provides input to the plant computer and there is local water level indication on the side of the SFP.

There are two channels of continuous remote indication of the spent fuel pool water temperature in the control room. There are two channels of SFP water temperature to a common recorder that provides high temperature annunciation in the control room. In addition, there are two channels of SFP water temperature that provide input to the plant computer.

There are two channels of continuous remote indication of Refueling Floor area radiation in the control room. Each of these channels provide high area radiation annunciation in the control room. A local alarm to notify personnel of high area radiation levels is also in place. In addition, each of these channels provides input to the plant computer.

BVY 14-009 / Attachment 1 / Page 47 of 59 6 SFP seals that could cause leakage The VY SFP gate has static seals between the inner and outer gate. There is no credible leading to fuel uncovery in the event of catastrophic failure mechanism for these seals. If SFP inventory were to leak due to seal rupture seal failure shall be self limiting to or degradation, level would not go below the top of the spent fuel racks. The fixed top elevation of leakage or otherwise engineered so the refueling slot between the SFP and reactor vessel where a removable refueling slot plug is that drainage cannot occur. placed over is at elevation 321.5 feet. The top elevation of a spent fuel rack in the SFP is 321.29 feet.

7 Procedures or administrative controls to VY procedure, OPOP-NFPC-2184, Normal Fuel Pool Cooling System, allows specified volumes reduce the likelihood of rapid draindown to be pumped or letdown from the SFP. The procedure meets the requirements of this IDC by events will include (1) prohibitions on controlling the suction and discharge points. Additionally, the ISFSI equipment design is such that the use of pumps that lack adequate there are no ISFSI related SFP operations that have the potential to cause a rapid draindown.

siphon protection or (2) controls for pump suction and discharge points. The Procedure EN-HU-1 06, Procedureand Work Instruction Use and Adherence, establishes the functionality of anti-siphon devices will expectations and requirements for procedure adherence and usage for all personnel performing be periodically verified- activities. Additionally, all work activities are subject to the work process controls and integrated risk management where the activities are analyzed and managed for risk. (e.g. address SFP activities.)

The VY SFP does not have anti-siphon devices.

8 An onsite restoration plan will be in The onsite restoration plan is incorporated into procedure ON-3157, Loss of Fuel Pool place to provide repair of the SFP Level/Cooling, and PP-7019, Appendix G, Loss of Large Areas of the Plant Due to Fire or cooling systems or to provide access Explosion. ON-3157 establishes multiple makeup sources from onsite and offsite that includes:

for makeup water to the SFP. The plan

• Fire Water system will provide for remote alignment of the 0 Service Water (SW) system via the Residual Heat Removal system makeup source to the SFP without 0 Cooling Tower #2 deep basin via an engine driven emergency makeup pump requiring entry to the refuel floor.

There are multiple ways to add makeup water to the SFP with or without entry to the refuel floor.

9 Procedures will be in place to control VY procedure OPOP-NFPC-2184, Normal Fuel Pool Cooling System, allows specified volumes to SFP operations that have the potential be pumped or letdown from the spent fuel pool. The procedure meets the requirements of this to rapidly decrease SFP inventory. IDC by controlling the suction and discharge points. Additionally, the Independent Spent Fuel These administrative controls may Storage Installation (ISFSI) equipment design is such that there are no ISFSI related SFP require additional operations or operations that have the potential to cause a rapid draindown.

I management review, management

BVY 14-009 1 Attachment 1 / Page 48 of 59 physical presence for designated Procedure EN-HU-106, Procedure and Work Instruction Use and Adherence, establishes the operations or administrative limitations expectations and requirements for procedure adherence and usage for all personnel performing such as restrictions on heavy load activities. Additionally, all work activities are subject to the work process controls and integrated movements, risk management where the activities are analyzed and managed for risk. (e.g. address SFP activities.)

Heavy loads requirements are controlled under the procedures EN-MA-1 19, MaterialHandling Program,and the VY heavy loads program PP-7023, Control of Heavy Loads Program Document. Heavy Loads in the vicinity of the SFP are addressed in PP-7023. Fuel moves and heavy load moves that could affect the safe handling and storage of nuclear fuel require approval by the Shift Manager.

10 Routine testing of the alternative fuel VY practices align with this description. The SW system has redundant pumping capability and pool makeup system components will power supplies to ensure alternative fuel pool makeup function. The SW system runs be performed and administrative continuously thus allowing for constant monitoring. Additionally, there is an electric-driven fire controls for equipment out of service pump and a diesel-driven fire pump that can supply makeup water to the SFP via the SW system will be implemented to provide added or the Fire Water system. The VY Technical Requirements Manual (TRM) contains administrative assurance that the components would controls for the fire pumps that address operation with equipment out of service and periodic be available, if needed. functionality testing. All sources discussed above take suction from the Connecticut River.

VY also has an engine driven emergency makeup pump capable of taking suction from the Cooling Tower #2 deep basin to provide an alternate source of makeup water to the SFP. The systems provide defense-in-depth. The systems and components necessary to implement the emergency makeup strategies are routinely tested to ensure capability is maintained.

BVY 14-009 / Attachment 1 / Page 49 of 59 TABLE 4 Staff Decommissioning Assumptions (SDAs) Comparison SDA Staff Assumptions Response 1 Licensee's SFP cooling design will be at The VY design aligns with the intent of this description. The VY SFP cooling system design is least as capable as that assumed in the based, in part, on Regulatory Guide 1.13 which included, in part, a Seismic Category I makeup risk assessment, including system to add coolant to the SFP. The design basis requirement for SFP cooling is provided by instrumentation. Licensees will have at the SW (SW) system, which is a Nuclear Safety Design Class I system (i.e. it is designed to least one motor-driven and one diesel- withstand design basis earthquake seismically induced load) protected by a Nuclear Safety driven fire pump capable of delivering Design Class I structure inventory to the SFP.

The SFP cooling system heat exchangers are cooled by SW which has redundant pumping capacity and is provided by redundant power sources adequate to provide makeup at the required capacity. The SW pumps are normally powered from offsite power, but can be supplied from an alternate reliable power source.

The stations design also includes an electric-driven fire pump and a diesel-driven fire pump, both of which will be maintained until all fuel is removed from the SFP. Each fire pump has the capability to deliver 500 gallons per minute (gpm) of makeup water to the SFP.

All sources discussed above take suction from the Connecticut River. The SW system runs continuously thus allowing for constant monitoring.

VY also has an engine driven emergency makeup pump capable of taking suction from the Cooling Tower #2 deep basin to provide an alternate source of makeup water to the SFP at a minimum of 300 gpm. When the discharge of this pump is connected to a pumper type fire truck, the fire truck is able to provide a minimum of 200 gpm of makeup water to the SFP.

2 Walk-downs of SFP systems will be Currently VY performs a walk-down of SFP systems once per day due to dose considerations performed at least once per shift by the associated with an operating reactor. The frequency of these walk-downs may be increased operators. Procedures will be following final plant shutdown and permanent defueling of the reactor. There are other methods developed for and employed by the available in the control room to alert operators to potential SFP events, such as annunciators and operators to provide guidance on the level indication.

capability and availability of onsite and offsite inventory makeup sources and VY procedures meet the requirements of this SDA by providing the guidance on the capability time available to initiate these sources and availability of onsite and offsite makeup sources. OPOP-PHEN-3127, Natural Phenomena, for various loss of cooling or inventory directs the inspection of the SFP and cooling systems following a seismic event. ON-3157, Loss events. of Fuel Pool Level/Cooling, establishes multiple makeup sources from onsite and offsite that

BVY 14-009 / Attachment 1 / Page 50 of 59 includes:

• Fire Water system

• SW system via the Residual Heat Removal system

• Cooling Tower #2 deep basin via an engine driven emergency makeup pump Prior to final shutdown, VY will establish the timelines required to initiate the various onsite and offsite SFP makeup sources based on expected system configurations and availability.

3 Control room instrumentation that VY design meets the intent of this SDA. There are two narrow range channels of continuous monitors SFP temperature and water remote indication of SFP water level in the control room. Level is determined by measuring the level will directly measure the hydrostatic pressure from a sensor located in the SFP. Each of these channels provides high and parameters involved. Level low annunciation on the Fuel Pool Filter Demineralizer Panel with a System Trouble Alarm and instrumentation will provide alarms at high and low level indicator lights in the control room. In addition, each of these channels levels associated with calling in offsite provides input to the plant computer and there is local water level indication on the side of the resources and with declaring a general SFP.

emergency.

There are two channels of continuous remote indication of the SFP water temperature in the control room from temperature elements located in the SFP. There are two channels of SFP water temperature to a common recorder that provides high temperature annunciation in the control room. In addition, there are two channels of SFP water temperature that provide input to the plant computer.

VY has procedures in place to respond to an abnormally low level in the SFP to direct the plant staff to take appropriate actions to provide the necessary SFP makeup; first through normal means, then by utilizing all available onsite resources, including both design basis and defense-in-depth capabilities. Refer to the VY responses for IDC 2 and IDC 4 for details associated with calling in offsite resources.

Regarding the declaration of a general emergency, VY will be employing Shutdown EALs using an approved NRC EAL Scheme. Based on Appendix C of NEI 99-01, "Development of Emergency Action Levels for Non-Passive Reactors," Revision 6, it is expected that station conditions will not have the capacity to reach any threshold requiring the declaration of a general emergency.

4 Licensee determines that there are no The VY SFP design is consistent with this SDA.

drain paths in the SFP that could lower

• The VY normal SFP cooling system suction lines are located at elevation 336'-6", which the pool level (by draining, suction, or is approximately 7'-6" below normal SFP water level.

pumping) more than 15 feet below the 9 The VY maximum depth of siphon path is at elevation 334'-0", which is approximately 10'

BVY 14-009 / Attachment 1 / Page 51 of 59 normal pool operating level and that below normal SFP water level. However, this line is isolated with a normally locked-licensee must initiate recovery using closed valve to prevent possible siphoning. This is seismically rated piping.

offsite sources. The VY maximum drain path is via the 3" drain line located between in the inboard and outboard SFP gates, located at elevation 319.85' which is located approximately 24' below normal SFP water level. Drain-down to this elevation would result in uncovery of the top 1.5' of the spent fuel racks. However, there would need to be a gross failure of the inboard gate sealing gasket as well as failure of the 3" drain line. Therefore, this drain path is not considered to be a credible failure mode for inventory loss given the assumption that inventory loss is not the result of catastrophic failures.

5 Load Drop consequence analyses will The VY design is in alignment with this description. The VY heavy loads program PP-7023, be performed for facilities with Control of Heavy Loads ProgramDocument, controls the handling of heavy loads to meet the nonsingle failure-proof systems. The guidance provided in NUREG-0612. The cask handling crane (i.e. reactor building bridge crane) analyses and any mitigative actions trolley was upgraded to address Phase I requirements of the NUREG to provide redundancy in necessary to preclude catastrophic the load carrying path from the cask to the crane trolley itself, so that no single failure would allow damage to the SFP that would lead to a the cask to drop. In addition to the trolley replacement, a comprehensive maintenance program rapid pool draining would be sufficient and strict administrative control of all cask handling was implemented via implementation of PP-to demonstrate that there is high 7023. The NRC Safety Evaluation Report for the acceptance of NUREG-0612 Phase I actions is confidence in the facilities ability to documented in Reference 10.

withstand a heavy load drop.

The VY TRM contains functionality requirements for the reactor building crane, specifically, TRM 3.12, "Refueling and Spent Fuel Handling," which is subject to the provisions of 10 CFR 50.59.

6 Each decommissioning plant will A VY-specific seismic risk estimate of the SFP was performed in support of development of successfully complete the seismic NUREG-1 738. The risk estimate and results are documented in Attachment 2 to Appendix 2B of checklist provided in Appendix2B to this the NUREG. Table 2 of the attachment shows that the risk of a seismic-induced SFP failure was study [NUREG-1738]. If the checklist estimated to be 8.9x10-7 per year. This value is less than the NRC Pool Performance Guideline cannot be successfully completed, the (PPG) of 1x10,5 per year specified in Section 2 of Appendix 2B.

decommissioning plant will perform a plant specific seismic risk assessment Also, a separate study performed under NUREG/CR-5176 that is referenced in Appendix 2B of of the SFP and demonstrate that SFP NUREG-1 738 includes a VY-specific seismic risk assessment of the SFP. The results of this seismically induced structural failure assessment indicate SFP failure was 6.7x10-6 per year, which is below the PPG guideline as well.

and rapid loss of inventory is less than the generic bounding estimates Finally, item 10 of the seismic checklist provides an alternative in which the licensee delays provided in this study (<1 x10-5 per year requesting of a waiver (i.e. EP exemptions) until the plant specific danger of a zirconium fire is no including non-seismic events), longer a concern. As detailed in Section II of this submittal, VY is requesting that the exemptions become effective on April 15, 2016, which occurs approximately 15.4 months following the final VY reactor shutdown scheduled for December 29, 2014. VY's zirconium fire analysis (Attachment

BVY 14-009 / Attachment 1 / Page 52 of 59

2) determined that 15.4 months after shutdown, it would take 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> for the spent fuel to reach 900'C following uncovering of the fuel.

7 Licensees will maintain a program to The VY spent fuel racks utilize Boral, rather than Boraflex, as the neutron absorbing material. As provide surveillance and monitoring of described in Section 15.2.40 of the VY UFSAR, an aging management program is in place to Boraflex in high-density spent fuel racks manage loss of material and reduction of neutron absorption capacity of Boral neutron absorption until such time as spent fuel is no panels in the spent fuel racks. The loss of material and the reduction of the neutron-absorbing longer stored in these high-density capacity will be determined through coupon testing, direct in situ testing or both. Such testing will racks. include periodic verification of boron loss through areal density measurement of coupons or through direct in situ techniques, such as measurement of boron areal density, measurement of geometric changes in the material (blistering, pitting and bulging), and detection of gaps through blackness testing.

As part of License Renewal Commitment 52, VY plans to perform neutron attenuation testing using an in-situ method prior to the end of 2014.

BVY 14-009 / Attachment 1 / Page 53 of 59 V. JUSTIFICATION FOR EXEMPTIONS AND SPECIAL CIRCUMSTANCES 10 CFR 50.12 states that the Commission may, upon application by any interested person or upon its own initiative, grant exemptions, from the requirements of the regulations of Part 50 which are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the defense and security. 10 CFR 50.12 also states that the Commission will not consider granting an exemption unless special circumstances are present. As discussed below, this exemption request satisfies the provisions of Section 50.12.

A. The exemptions are authorized by law 10 CFR 50.12 allows the NRC to grant exemptions from the requirements of 10 CFR Part 50. The proposed exemption would not result in a violation of the Atomic Energy Act of 1954, as amended, or the Commission's regulations. Therefore, the exemption is authorized by law.

B. The exemptions will not present an undue risk to public health and safety The underlying purpose of 10 CFR 50.47(b), 10 CFR 50.47(c)(2), 10 CFR 50, Appendix E, Section IV is to ensure that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency, to establish plume exposure and ingestion pathway emergency planning zones for nuclear power plants, and to ensure that licensees maintain effective offsite and onsite emergency plans.

As discussed in this request, revised radiological analyses have been developed that show that, 17 days after shutdown, the radiological consequences of design basis accidents will not exceed the limits of the Environmental Protection Agency (EPA)

Protective Action Guides at the EAB. In addition, analyses have been developed for beyond design basis events related to the SFP which show that, within 15.4 months after shutdown, the analyzed event is either not credible, is capable of being mitigated, or the radiological consequences of the event will not exceed the limits of the EPA Protective Action Guides at the exclusion area boundary (EAB).

Therefore, offsite emergency response plans will no longer be needed for protection of the public beyond the EAB. Based on the reduced consequences of radiological events possible at the site when it is in the permanently defueled condition, the scope of the onsite emergency preparedness organization and corresponding requirements in the emergency plan may be accordingly reduced without an undue risk to the public health and safety.

Therefore, the underlying purpose of the regulations will continue to be met. Since the underlying purpose of the rules will continue to be met, the exemptions will not present an undue risk to the public health and safety.

C. The exemptions are consistent with the common defense and security The reduced consequences of radiological events that will remain possible at the site once it is in the permanently defueled condition allows for a corresponding reduction in the scope of the onsite emergency preparedness organization and associated reduction

BVY 14-009 / Attachment 1 / Page 54 of 59 of requirements in the emergency plan. These reductions will not adversely affect VY's ability to physically secure the site or protect special nuclear material. Physical security measures at VY are not affected by the requested exemption. Therefore, the proposed exemptions are consistent with the common defense and security.

D. Special Circumstances Pursuant to 10 CFR 50.12(a)(2), the NRC will not consider granting an exemption to its regulations unless special circumstances are present. ENO has determined that special circumstances are present as discussed below.

Special circumstances will exist at VY because the plant will be permanently shutdown and defueled and the radiological source term at the site will be reduced from that associated with reactor power operation. With the reactor power plant permanently shutdown and defueled, the design basis accidents and transients postulated to occur during reactor operation will no longer be possible. In particular, the potential for a release of a large radiological source term to the environment from the high pressures and temperatures associated with reactor operation will no longer exist.

1. Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule. (10 CFR 50.12(a)(2)(ii))

The underlying purpose of 10 CFR 50.47(b), 10 CFR 50.47(c)(2), and 10 CFR 50, Appendix E, Section IV is to ensure that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency, to establish plume exposure and ingestion pathway emergency planning zones for nuclear power plants, and to ensure that licensees maintain effective offsite and onsite emergency plans.

The standards and requirements in 10 CFR 50.47(b), 10 CFR 50.47(c)(2), and 10 CFR 50, Appendix E, Section IV were developed taking into consideration the risks associated with operation of a nuclear power reactor at its licensed full power level.

These risks include the potential for a reactor accident with offsite radiological dose consequences.

The radiological consequences of accidents that will remain possible at VY are substantially lower than those at an operating plant. The upper bound of offsite dose consequences limits the highest attainable emergency class to the alert level. In addition, because of the reduced consequences of radiological events that will still be possible at the site, the scope of the onsite emergency preparedness organization may be reduced accordingly. Thus, the underlying purpose of the regulations will not be adversely affected by eliminating offsite emergency planning activities or reducing the scope of onsite emergency planning.

Revised radiological analyses have been developed that show that, 17 days after shutdown, the radiological consequences of design basis accidents will not exceed the limits of the EPA Protective Action Guides at the EAB. In addition, analyses have been developed for beyond design basis events related to the SFP which show that, within 15.4 months after shutdown, the analyzed event is either not credible, is capable of

BVY 14-009 / Attachment 1 / Page 55 of 59 being mitigated, or the radiological consequences of the event will not exceed the limits of the EPA Protective Action Guides at the EAB. Therefore, application of all of the standards and requirements in 10 CFR 50.47(b), 10 CFR 50.47(c)(2), and 10 CFR 50, Appendix E, Section IV are not necessary to achieve the underlying purpose of those rules.

Since the underlying purposes of the rules would continue to be achieved even with VY being permitted to reduce the scope of emergency preparedness requirements consistent with placing the facility in the permanently defueled condition, the special circumstances are present as defined in 10 CFR 50.12(a)(2)(ii).

2. Compliance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated. (10 CFR 50.12(a)(2)(iii))

Application of all of the standards and requirements in 10 CFR 50.47(b), 10 CFR 50.47(c)(2), and 10 CFR 50, Appendix E, Section IV is not needed for adequate emergency response capability and is excessive for a permanently defueled facility.

Application of all of these standards and requirements would result in undue costs being incurred for the maintenance of an emergency response organization in excess of that actually needed to respond to the diminished scope of credible events. Other licensees similarly situated, such as Zion, have been granted similar exemptions.

Therefore, compliance with the rule would result in an undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated and the special circumstances required by 10 CFR 50.12(a)(2)(iii) exist.

3. The exemptions would result in benefit to the public health and safety that compensates for any decrease in safety that may result from the grant of the exemptions. (10 CFR 50.12(a)(2)(iv))

The plant will be permanently shutdown and defueled and the radiological source term at the site will be reduced from that associated with reactor power operation. With the reactor power plant permanently shutdown and defueled, the design basis accidents and transients postulated to occur during reactor operation will no longer be possible. In particular, the potential for a release of a large radiological source term to the environment from the high pressures and temperatures associated with reactor operation will no longer exist.

The proposed exemptions would allow VY to revise the station emergency plan to correspond to the reduced scope of remaining accidents and events. As such, the plan would no longer need to address response actions for events that would no longer be possible. The revised plan would thereby enhance the ability of the emergency response organization to respond to those scenarios that remain credible since emergency preparedness training and drills would focus only on applicable activities. Elimination of requirements for classification of emergency action levels for events that were no longer possible would enhance the ability of the ERO to correctly classify those events that remain credible. As the proposed exemption will enhance the ability of the organization

BVY 14-009 / Attachment 1 / Page 56 of 59 to respond to credible events, a resultant benefit to the public health and safety is realized.

Therefore, since the granting the exemptions would result in benefit to the public health and safety and would not result in a decrease in safety, the special circumstances required by 10 CFR 50.12(a)(2)(iv) exist.

E. Precedents The exemption requests for 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and 10 CFR Part 50, Appendix E, requirements are consistent with changes to emergency plans approved by the NRC for transition to a permanently defueled condition, implemented in 1998, as identified in References 13 and 14 (with respect to exemptions requested for regulations that were in place in 1998). Specific exemption requests for regulations that involve hostile action and offsite planning are consistent with exemptions approved by the NRC for a shutdown facility with an Independent Spent Fuel Storage Installation by letter dated May 2, 2013 (Reference 15). ENO proposes that VY should not be required to plan for an offsite impact resulting from hostile action because (1) the facility poses a lower radiological risk to the public than does a power reactor, and (2) the facility has a low likelihood of a postulated accident resulting in radiological releases requiring offsite protective measures.

Additionally, the specific exemption request for the regulation that involves a shift staffing analysis is consistent with the exemption approved by the NRC for a shutdown facility with an Independent Spent Fuel Storage Installation by letter dated March 18, 2013 (Reference 16).

VI. ENVIRONMENTAL ASSESSMENT The proposed exemption meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(25), because the proposed exemption involves: (i) no significant hazards consideration; (ii) no significant change in the types or significant increase in the amounts of any effluents that may be released offsite; (iii) no significant increase in individual or cumulative public or occupational radiation exposure; (iv) no significant construction impact; (v) no significant increase in the potential for or consequences from radiological accidents; and (vi) the requirements from which the exemption is sought involve requirements of an administrative, managerial, or organizational nature. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed exemption.

(i) No Significant Hazards Consideration Determination ENO has evaluated the proposed exemption to determine whether or not a significant hazards consideration is involved by focusing on the three standards set forth in 10 CFR 50.92 as discussed below:

1. Does the proposed exemption involve a siqnificant increase in the probability or consequences of an accident Previously evaluated?

The proposed exemptions have no effect on structures, systems, and components (SSCs) and no effect on the capability of any plant SSC to perform

BVY 14-009 / Attachment 1 / Page 57 of 59 its design function. The proposed exemptions would not increase the likelihood of the malfunction of any plant SSC.

When the exemptions become effective, there will be no credible events that would result in doses to the public beyond the exclusion area boundary that would exceed the EPA PAGs. The probability of occurrence of previously evaluated accidents is not increased, since most previously analyzed accidents will no longer be able to occur and the probability and consequences of the remaining Fuel Handling Accident are unaffected by the proposed amendment.

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

2. Do the Proposed exemptions create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed exemption does not involve a physical alteration of the plant. No new or different type of equipment will be installed and there are no physical modifications to existing equipment associated with the proposed exemption.

Similarly, the proposed exemption will not physically change any SSCs involved in the mitigation of any accidents. Thus, no new initiators or precursors of a new or different kind of accident are created. Furthermore, the proposed exemption does not create the possibility of a new accident as a result of new failure modes associated with any equipment or personnel failures. No changes are being made to parameters within which the plant is normally operated, or in the setpoints which initiate protective or mitigative actions, and no new failure modes are being introduced.

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

3. Do the proposed exemptions involve a significant reduction in a mar-gin of safety?

The proposed exemption does not alter the design basis or any safety limits for the plant. The proposed exemption does not impact station operation or any plant SSC that is relied upon for accident mitigation.

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

Based on the above, ENO concludes that the proposed exemption presents no significant hazards consideration, and, accordingly, a finding of "no significant hazards consideration" is justified.

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

There are no expected changes in the types, characteristics, or quantities of effluents discharged to the environment associated with the proposed exemption. There are no

BVY 14-009/ Attachment 1 / Page 58 of 59 materials or chemicals introduced into the plant that could affect the characteristics or types of effluents released offsite. In addition, the method of operation of waste processing systems will not be affected by the exemption. The proposed exemption will not result in changes to the design basis requirements of SSCs that function to limit or monitor the release of effluents. All the SSCs associated with limiting the release of effluents will continue to be able to perform their functions. Therefore, the proposed exemption will result in no significant change to the types or significant increase in the amounts of any effluents that may be released offsite.

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

The exemption will result in no expected increases in individual or cumulative occupational radiation exposure on either the workforce or the public. There are no expected changes in normal occupational doses. Likewise, design basis accident dose is not impacted by the proposed exemption.

(iv)There is no significant construction impact.

No construction activities are associated with the proposed exemption.

(v) There is no significant increase in the potential for or consequences from radiological accidents.

See the no significant hazards considerations discussion in Item (i)(1) above.

(vi) Requirements of an administrative, managerial, or organizational nature.

The proposed exemptions will form the basis for a reduction in size of the VY emergency response organization commensurate with the reduction in consequences of radiological events that will be possible at VY once the facility is in the permanently defueled condition.

References

1. Letter, Entergy Nuclear Operations, Inc. to USNRC, "Notification of Permanent Cessation of Power Operations," BVY 13-079, dated September 23, 2013

2. NEI 99-01, "Development of Emergency Action Levels for Non-Passive Reactors,"

Revision 6

3. Commission Paper SECY-1 3-0112, "Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor,"

October, 2013

4. Environmental Protection Agency Protective Action Guides and Planning Guidance for Radiological Incidents, Draft for Interim Use and Public Comment, dated March 2013

5. Federal Register Notice, Vol. 76, No. 226 (76 FR 72596), Enhancements to Emergency Preparedness Regulations, dated November 23, 2011

6. Letter, Entergy Nuclear Operations, Inc. to USNRC, "Technical Specifications Proposed Change No. 306, Eliminate Certain ESF Requirements during Movement of Irradiated Fuel," BVY 13-097, dated November 14, 2013

BVY 14-009 / Attachment 1 / Page 59 of 59

7. Commission Paper SECY-99-168, Improving Decommissioning Regulations for Nuclear Power Plants, dated June 30, 1999

8. NUREG-1 738, "Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants," dated February 2001

9. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis"

10. Letter, USNRC to Vermont Yankee Nuclear Power Corporation (VYNPC), "Control of Heavy Loads (Phase I)," NVY 84-139, dated June 27, 1984

11. Letter, USNRC to VYNPC, "Completion of Phase II of 'Control of Heavy Loads at Nuclear Power Plants' NUREG 0612. (Generic Letter 85-11)," NVY 85-138, dated June 28, 1985

12. Letter, Entergy Nuclear Operations, Inc. to USNRC, "Request for Approval of Certified Fuel Handler Training Program," BVY 13-095, dated October 31, 2013

13. Letter from USNRC to Zion Nuclear Power Station, Unit Nos. 1 and 2, "Request For Approval of Defueled Station Emergency Plan and Exemption from Certain Requirements of 10 CFR 50.47, "Emergency Plans"- Zion Nuclear Power Station, Unit Nos. 1 and 2 (TAC NOS MA5253 and MA5254)," dated August 31, 1999. (ADAMS Accession No. 9909070079)

14. Letter from USNRC to Maine Yankee Atomic Power Company, "Exemption from Certain Requirements of 10 CFR 50.54(q), 10 CFR 50.47(b) and (c), And Appendix E to 10 CFR Part 50 at Maine Yankee Atomic Power Station (TAC No. MA0069)," dated September 3, 1998. (ADAMS Accession No. 9809140214)

15. Letter from USNRC to Maine Yankee Atomic Electric Company, "Response to Exemption Request For Portions of Title 10 of the Code of Federal Regulations Part 50, Appendix E, and Section 50.47 of Title 10 of the Code of Federal Regulations for the Maine Yankee Atomic Power Station (TAC NO. L24661 )," dated May 2, 2013. (ADAMS Accession No. ML13112A842)

16. Letter from USNRC to Connecticut Yankee Atomic Power Company, "Response to Exemption Request for Portions of Title 10 of the Code of Federal Regulations Part 50 Appendix E, and Title 10 of the Code of Federal Regulations Part 50.47 for the Haddam Neck Plant (TAC No. L24663)," dated March 18, 2013. (ADAMS Accession No. ML13064A374)

17. USNRC, "Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor" (Draft Report for Comment)

June 2013, ADAMS Accession No. ML13133A132.

18. Draft NSIR/DPR-ISG-02, Interim Staff Guidance, Emergency Planning Exemption Requests for Decommissioning Nuclear Power Plants

19. Federal Register Notice, Vol. 60, No. 120, Emergency Planning Licensing Requirements for Independent Spent Fuel Storage Facilities (ISFSI) and Monitored Retrievable Storage Facilities (MRS), dated June 22, 1995

20. USNRC, "Integrated Rulemaking Plan for Nuclear Power Plant Decommissioning,"

Commission Paper SECY-00-01 45, June 28, 2000 (ADAMS Accession No. ML003721626)

21. Federal Register Notice, Vol. 74, No. 94, Enhancements to Emergency Preparedness Regulations, dated May 18, 2009

22. NUREG-0696, "Functional Criteria for Emergency Response Facilities," February 1981

BVY 14-009 Docket 50-271 Attachment 2 Vermont Yankee Nuclear Power Station Vermont Yankee Maximum Cladding Temperature Analysis for an Uncovered Spent Fuel Pool with no Air Cooling

!SE11rgerut: S, LUiridyII DESIGN CONTROL

SUMMARY

CLIENT: Entergy UNIT_ 1, 2 PAGE NO.: 1 PROJECT NAME: Vermont Yankee Nuclear Power Plant S&L NUCLEAR QA PROGRAM PROJECT NO.: 12680-007 APPLICABLE 0 YES El NO CALC. NO.: 2013-13016 SAFETY RELATED _ YES El NO TITLE: Vermont Yankee Maximum Cladding Temperature Analysis for an Uncovered Spent Fuel Pool with no Air Cooling EQUIPMENT NO.:

IDENTIFICATION OF PAGES ADDED/REVISEDISUPERSEDEDNOIDED &REVIEW METHOD Initial Issue. The main body is pages 1 through 10. The final page is INPUTS] ASSUMPTIONS Attachment B, page B2. 0 VERIFIED

___UNVERIFIED REVIEW METHOD: Detailed Review REV.: 0 STATUS: 0Z APPROVED El SUPERSEDED BY CALCULATION NQ. E VOID DATE FOR REV.: -T -

PREPARER: Matthew M. Ross DATE: 11A/?c4$

REVIEWER: Paul J. Szymiczek DATE:

APPROVER: Robert J. Peterson DATE:

IDENTIFICATION OF PAGES ADDEDIREVISED/SUPERSEDED/VOIDED &REVIEW METHOD INPUTS/ ASSUMPTIONS

[I VERIFIED El UNVERIFIED REVIEW METHOD: REV.: 1 STATUS: El APPROVED El SUPERSEDED BY CALCULATION NO. El VOID DATE FOR REV.:

PREPARER: DATE:

REVIEWER: DATE:

APPROVER: DATE:

IDENTIFICATION OF PAGES ADDEDIREVISED/SUPERSEDED/VOIDED &REVIEW METHOD INPUTS/ ASSUMPTIONS El VERIFIED

_E UNVERIFIED REVIEW METHOD: REV.: 2 STATUS: El APPROVED El SUPERSEDED BY CALCULATION NO. El VOID DATE FOR REV.:

PREPARER: DATE:

REVIEWER: DATE:

APPROVER: DATE:

SM.Y*iJIrI l-rdy-L L

Document No. 2013-13016 Vermont Yankee Nuclear Power Plant Revision 0 Paae 2 of 12

1. Purpose and Scope ............................................................................................................................. 3

2. References ........................................................................................................................................... 4

3. Definitions ........................................................................................................................................... 4

4. Input D .ata........................................................................................................................................... 5

5. Assum ptions ........................................................................................................................................ 7

6. M ethodology ....................................................................................................................................... 8

7. Results ............................................................................................................................................... 11

8. Conclusions and Recom m endations .......................................................................................... 12 Attachments: No. of Pages:

Attachment A : Design Input Record (Ref. 2.3) ....................................................................................... 22 Attachm ent B: Analysis ................................................................................................................................ 2 SaraorttLundr'

Document No. 2013-13016 Vermont Yankee Nuclear Power Plant Revision 0 Page 3 of 12

1. Purpose and Scope 1.1. Purpose The purpose of this calculation is to conservatively evaluate the length of time (number of hours) it takes for uncovered spent fuel assemblies to reach the temperature at which the zirconium cladding would fail. The time to failure is calculated for various decay times after shutdown. This analysis conservatively assumes that there is no radiative or air cooling of the assemblies: the flow paths that would provide natural circulation cooling are assumed to be blocked.

1.2. Scope The number of hours that it takes for the fuel to reach the failure temperature (the heat-up time) is determined as a function of the decay date after shutdown (the decay time).

The zirconium cladding must remain below the failure temperature. Per NUREG/CR-6451 (Ref. 2.1, see Design Input 4.1), 565 'C (1049 'F) is the lowest temperature where incipient cladding failure might occur with an expected failure at 671'C (1240'F). Per SECY-99-168 (Ref. 2.4), 800'C (1472°F) is the lowest temperature where self-sustained oxidation would occur and 565°C is the minimum temperature where clad swelling might occur. NUREG- 1738 (Ref. 2.7, pgs. 3-7, A IB-5) states that runaway oxidation of zirconium occurs at 900 'C and 565°C is associated with the 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> creep rupture time. For this analysis, the NUREG/CR-6451 temperature (565 °C, 1049 'F) and the NUREG-1738 temperature (900 'C, 1652 'F) are the temperatures of interest for the zirconium cladding.

There are no specific acceptance criteria for this analysis, however, SECY-99-168 (Ref.

2.4) suggests that "10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> (is) sufficient time to take mitigative action." SECY 168 also performed a generic analysis that found that for BWRs, 2 years is expected to be the decay time needed to reach a 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> heat-up time from 30 'C to 900 'C.

NUREG-1738 shows that a 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> heat up time to 900 'C for a PWR would occur at less than 2 years (Ref. 2.7, Fig. 2-2). A 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> heat up time for a BWR would occur sooner (Ref. 2.7, Fig. 2-1).

SmgenZ.& LundyV

Document No. 2013-13016 Vermont Yankee Nuclear Power Plant Revision 0 Page 4 of 12

2. References 2.1. NUREG/CR-6451, "A Safety and Regulatory Assessment of Generic BWR and PWR Permanently Shutdown Nuclear Power Plants," August 1997.

2.2. Incropera, Frank P., and David P. DeWitt, Introduction to Heat Transfer, Fourth Edition, John Wiley & Sons.

2.3. Design Input Record EC47013, "Heat Generation Rate After Shutdown and Adiabatic Heatup Analysis," Rev. 0, dated 10/17/2013. (Included as Attachment A) 2.4. SECY-99-168, "Improving Decommissioning Regulations for Nuclear Power Plants,"

June 30, 1999.

2.5. Sargent & Lundy Calculation 2013-13824, "Decay Heat Rate Analysis for a Bounding Discharged Fuel Assembly," Rev. 0.

2.6. ARS 21002, CRP 9-4, "Alarm Response Sheets," Rev. 14.

2.7. NUREG-1738, "Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants," February 2001.

2.8. NUREG/CR-7024, "Material Property Correlations: Comparisons between FRAPCOM-3.4, FRAPTRAN 1.4, and MATPRO," March 2011.

3. Definitions 3.1. Decay Time The decay time is the time since the reactor was shut down.

3.2. Heat-up Time The heat-up time is the amount of time between when the fuel becomes uncovered and when the zirconium cladding reaches the failure temperatures of interest, 565 'C (1049

'F) and 900 'C (1652 'F).

Swargont Ltindv..

Document No. 2013-13016 Vermont Yankee Nuclear Power Plant Revision 0 Paaqe 5 of 12

4. Input Data 4.1. Maximum Zirconium Temperature Several studies are presented in NUREG/CR-6451 (Ref. 2.1) discussing the maximum allowable temperature of zirconium cladding that will ensure that failure of the zirconium cladding will not occur. Per NUREG/CR-6451 (Ref. 2.1), 565 'C (1049 OF) is the lowest temperature where incipient cladding failure might occur with an expected failure at 671 0 C (1240'F). Per SECY-99-168 (Ref. 2.4), 800 0 C (1472 0 F) is the lowest temperature where self-sustained oxidation would occur and 5650C is the minimum temperature where clad swelling might occur. NUREG-1738 uses 900 °C (1652 OF) as the temperature where "runaway oxidation" is expected to occur and 565°C is associated with the 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> creep rupture time (Ref. 2.7, pgs. 3-7, AIB-5).

Temperatures of 565 0 C and 9000 C are the failure temperatures of interest for this calculation 4.2. Zirconium Properties The specific heat of zirconium at 640 K (692 OF) is 331 J/kg-K (Ref. 2.3, pg. 11). A temperature of 692 OF is in the temperature range (roughly the midpoint for both ranges) of this analysis. From References 2.2 and 2.3, the specific heat slightly increases with an increase in temperature for most of the range of temperatures in this analysis. At higher temperatures, the zirconium would heat up more slowly. This temperature is representative of the full temperature range for this analysis.

4.3. Spent Fuel Pool Temperature The spent fuel pool cooling system alarm response is set at 120 OF (Ref. 2.6). This is the maximum long term temperature in the SFP. As the decay time increases (i.e. as the plant has been shutdown longer) the fuel heat generation rate will be lower and the maximum spent fuel pool temperature would likely be lower.

4.4. Geometry for Limiting Assemblies The table below shows the geometry inputs for the fuel assemblies used in this analysis.

Reference 2.3 shows data for GE14 and GNF2 fuel. Both fuel types are evaluated in this analysis and the results for the worse of the two are presented.

Table 4-1: Fuel Assembly Inputs (from Ref. 2.3)

Fuel Type GNF2 GE14 Fuel Pellet Diameter (mm) 8.88 8.76 Outer Diameter of Cladding (mm) 10.26 10.26 Inner Diameter of Cladding (mm) 9.1 8.94 Number of Full Length Rods 78 78 Number of Partial Length Rods 8 long, 6 short 14 eAor,0tC LUr,*V*'.

Document No. 2013-13016 Vermont Yankee Nuclear Power Plant Revision 0 Y Paae 6 of 12 FueiType GNF2 GE14 Length of Full Length Rods (mm) 3810 3810 Length of Partial Length Rods (mm) 2591, 1372 2134 Number of Water Rods 2 2 Outer Diameter of Water Rods (cm) 2.489 2.489 Inner Diameter of Water Rods (cm) 2.337 2.337 Fuel Density (g/cm3 ) 10.6 10.6 Cladding Density (lb/in 3) 0.237 0.237 4.5. Heat Load Reference 2.5 determines the maximum heat load from a single assembly. Per Reference 2.5, a representative assembly was used for both Vermont Yankee fuel types.

The assembly with the highest heat load will have the shortest heat-up time. The table showing the maximum fuel assembly heat generation rate for several years is below.

TnhkIP 4-2: Heat (,enerated hy a~chpdt Heat Iorac Aemhiv (Ref 2 5 Days Since Decay Heat Days Since Decay Heat Shutdown (watts) Shutdown (watts) 30 8,278 630 1,335 60 6,038 660 1,282 90 4,972 690 1,230 120 4,276 720 1,183 150 3,761 750 1,139 180 3,359 780 1,097 210 3,038 810 1,058 240 2,778 840 1,021 270 2,561 870 987 300 2,379 900 954 330 2,222 930 923 360 2,087 960 894 390 1,967 990 867 420 1,860 1020 841 450 1,764 1050 817 480 1,677 1080 794 510 1,597 1110 772 540 1,524 1140 752 570 1,456 1170 732 600 1,393 1200 714 Per Reference 2.3, the reactor contains 368 fuel assemblies and has a 100% reactor power level of 1912 MWt. The average heat generation for an assembly in the reactor is 5.2 megawatts (= 1912 / 368). The heat loads in the table above are on the order of 0.1% to 0.01% of the peak reactor power. A heat load of 1000 watts is very low compared with heat loads commonly analyzed at nuclear power plants.

SýVgont; S.~ LJuridV..

Document No. 2013-13016 Vermont Yankee Nuclear Power Plant Revision 0 Paae 7 of 12

5. Assumptions 5.1. The properties of pure zirconium are used for the specific heat and density of the zirconium alloy cladding. Based on an examination of alloys of some metals (e.g.

aluminum, nickel, or steel) in Table A. 1 of Reference 2.2, the density and specific heat are not significantly impacted by alloying.

5.2. The heat-up time is assumed to start when the spent fuel pool has been completely drained. This is conservative. It is likely that site personnel will start to respond to an incident when draindown starts.

S~ot~Lundy'-

Document No. 2013-13016 Vermont Yankee Nuclear Power Plant Revision 0 Paae 8 of 12

6. Methodology and Numerical Analysis This analysis determines the heat-up time of the fuel assembly using the thermal capacity of materials (Based on Section 2.3 of Ref. 2.2).

PVXCpXAT q=pxVXCpX-T Equation 6-1 t

Where:

4 is the heat generation rate in BTU/hr 3

p is the density of the material in lb/ft 3

Vis the volume of the material in ft cp is the specific heat in BTU/Ib-0 F AT is the temperature increase in 'F t is the heat-up time in hr For this analysis, there are two materials being heated: the uranium dioxide fuel pellets and the zirconium alloy cladding. The zirconium is in the cladding and the guide tubes, which are also being heated. The zirconium and the uranium dioxide are modeled as heating up at the same rate, so the AT/t will be the same for both materials.

q AT X(p XVUXC' + pZ xV_ X P Equation 6-2 t

Where:

X,, signifies the property is for uranium dioxide X_- signifies the property is for zirconium This calculation seeks the heat-up time, so Equation 6-2 is solved for t.

Equation 6-3 The volume of uranium dioxide is given below.

v. =71x P Nhrx L Equation 6-4 Where:

Dp is the diameter of the uranium dioxide pellet in ft Nir is the number of heated rods per assembly L is the heated length of the rods in ft n..-... & an V

Document No. 2013-13016 Vermont Yankee Nuclear Power Plant Revision 0 Page 9 of 12 The volumes of zirconium in the heated rods and in the guide tubes are given below. The length of the cladding and guide tubes that are heated is conservatively modeled as being the same as the heated length of uranium dioxide. In reality, the guide tubes and cladding are longer than the length of the uranium dioxide pellets (i.e. the heated length of the fuel is less than the total length of the assembly).

I=T tx D,' DN xL Equation 6-5 V g = 7x - Dgi 2 Ng, xL Equation 6-6

• ' 4 V. =Vg +V Equation 6-7 Where:

K, is the volume of zirconium in the cladding of heated tubes in ft 3 3

V-.g is the volume of zirconium in the guide tubes in ft D,.o is the outer diameter of the cladding in ft Dcj is the inner diameter of the cladding in ft Dgo is the outer diameter of the guide tubes in ft Dg., is the inner diameter of the guide tubes in ft Ngt is the number of guide tubes per assembly The temperature increase (AT) for this analysis is from the initial temperature of the pool, 120 'F (Input 4.3), to the zirconium cladding failure temperatures of interest, 1049 'F and 1652 °F (Input 4.1). Since the heat generation rate is low (see Input 4.5), the diameter of the rod is small, and the spacing between rods is small, temperatures in an assembly can be modeled as uniform during the event. The heat-up time is calculated as a function of the decay time.

The use of the maximum initial pool temperature is appropriate to use as the starting cladding temperature for this analysis. In the generic analyses in both SECY-99-168 (Ref. 2.4) and NUREG-1738 (Ref. 2.7), the starting water temperature was set at 30 'C (86 'F). Both documents state that the analysis starts at the time of fuel uncovery. As stated above, the heat generation rate is low (see Input 4.5), the diameter of the rod is small, and the spacing between rods is small. Therefore, temperatures in an assembly can be modeled as uniform before the event. Therefore, using 120 'F as the starting temperature of the analysis is conservative compared to the methods used in the guidance documents.

The specific heat of uranium dioxide (U0 2) is calculated using Equation 2.2-1 in Reference 2.8.

S~ot&Ltendy"

Document No. 2013-13016 Vermont Yankee Nuclear Power Plant Revision 0 Paae 10 of 12 KI, 2o2 exp(1T) E

,YK CP= +2 Kexp(/T)_3 2D exp(-ED/RT) Equation 6-8 T 2[exp(O/T) -112 2RT Where:

Cp is the specific heat in J/kg-K KI, K 2, K 3, are constants for U0 2 of 296.7 J/kg-K, 2.43x 102 J/kg-K 2, and 8.745x10 7 J/kg 0 is the Einstein temperature and is 535.285 K for U0 2 T is the temperature of the U0 2, in K Yis the oxygen to metal ratio (2 for U0 2)

ED is the activation energy for Frenkel Defects and is 1.577x10 5 I/mol R is the universal gas constant 8.3143 J/mol-K The third term in this summation is negligible at temperatures below 900 K and is ignored.

Ignoring the term is conservative because it very slightly reduces the specific heat, which reduces the heat-up time. Per Figure 2.2-1 in Reference 2.8, the heat capacity of uranium is concave over the range of temperatures of this analysis (120 'F is 322 K and 1049 'F is 838 K). Therefore, a conservatively low value for specific heat for the range of temperatures considered is selected as the average of the specific heats at those two temperatures.

Cp = 296.7 x 535.285 2 exp(535.285/322) + 2.43 x 10-2 x 322 = 245---

3222 [exp(535.285/322)- 142 kg-K Cp= 296.7 x 535.2852 exp(535.285/838) + 2.43 x 10-2 x 838 = 307 1g-X 8382 [exp(535.285/838)- 1]2 kg-K Cp=(245 j -/+307 j +- =276~-j Sewgonv ýr LurtdyV..

Document No. 2013-13016 Vermont Yankee Nuclear Power Plant Revision 0 Paae 11 of 12

7. Results The detailed analysis is documented in Attachment B. Per Attachment B, GNF2 fuel heats up slightly faster than GE14 fuel. The results shown in Table 7-1 below are for GNF2 fuel which bound the results for GE14 fuel.

Table 7-1: Results Decay Time Heat-Up Time to Heat-Up Time to (Days since Shutdown) 565 'C (hours) 900 'C hours) 240 3.7 6.2 300 4.4 7.2 360 5.0 8.2 420 5.6 9.2 480 6.2 10.2 540 6.8 11.3 600 7.5 12.3 660 8.1 13.4 720 8.8 14.5 780 9.5 15.6 840 10.2 16.8 900 10.9 18.0 960 11.6 19.2 By interpolating, the heat-up time to 565 'F is 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> at a decay time of 823 days (27 months) after shutdown. By interpolating the heat-up time to 900 'C is 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> at a decay time of 468 days (15.4 months) after shutdown.

The 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> heat-up time to a temperature of 565 'C (1049 'F) occurs at a decay time of under 2.5 years, while the 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> heat-up time to a temperature of 900 'C (1652 'F) occurred at a decay time of under 1.5 years.

As stated above, SECY-99-168 performed a generic analysis that found that for BWRs, 2 years is expected to be the decay time needed to reach a 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> heat-up time from 30 'C to 900 'C. NUREG-1738 shows that a 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> heat up time to 900 'C for a PWR would occur at less than 2 years (Ref. 2.7, Fig. 2-2) and that the 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> heat up time for a BWR would occur sooner (Ref. 2.7, Fig. 2-1). The results calculated here are more favorable than these generic analyses since the calculated decay time for a 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> heat-up time is less than what the generic analyses predict.

A plot showing the heat-up time to the temperatures of interest as a function of decay time is Figure 7-1.

Mwaant k LAu, V..

Document No. 2013-13016 Vermont Yankee Nuclear Power Plant Revision 0 Paae 12 of 12 Figure 7-1: Heat-Up Time vs. Decay Time 20

-- Temperature = 900 C (1652 F)

Temperature = 565 C (1049 F)

0) 15 0

rE 0 10 E0 0) 0._

" 5 0) 0 5 10 15 20 25 30 35 40 Months Since Shutdown

8. Conclusions and Recommendations The Vermont Yankee results are more favorable than the generic analyses performed for SECY-99-168 (Ref. 2.4) and NUREG-1738 (Ref. 2.7). There are no acceptance criteria for this analysis. There are no specific recommendations for this analysis.

S~~gont ~ LJ~.ndy~

BVY 14-009 Docket 50-271 Attachment 3 Vermont Yankee Nuclear Power Station List of Regulatory Commitments

BVY 14-009 / Attachment 3 / page 1 of 1 List of Regulatory Commitments This table identifies actions discussed in this letter for which Entergy 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 ONE-TIME CONTINUING COMPLETION DATE COMMITMENT ACTION COMPLIANCE (If Required)

Revise UFSAR to include a description of x Complete in how the VY spent fuel pool design and accordance with operational characteristics meets or next scheduled compares with the NUREG-1 738 Industry UFSAR update Decommissioning Commitments (IDC) and following exemption Staff Decommissioning Assumptions approval (SDA).

VY will establish the timelines required to x Prior to permanent initiate the various onsite and offsite SFP cessation of makeup sources based on expected operations system configurations and availability.

(SDA 2)