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{{Adams
#REDIRECT [[BVY 14-059, Response to Request for Additional Information on Technical Specifications Proposed Change No. 306 Eliminate Certain ESF Requirements During Movement of Irradiated Fuel - Supplement 2]]
| number = ML14224A012
| issue date = 08/06/2014
| title = Vermont Yankee, Response to Request for Additional Information on Technical Specifications Proposed Change No. 306 Eliminate Certain ESF Requirements During Movement of Irradiated Fuel - Supplement 2 (TAC No. MF3068)
| author name = Wamser C J
| author affiliation = Entergy Nuclear Operations, Inc
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000271
| license number = DPR-028
| contact person =
| case reference number = BVY 14-059, TAC MF3068
| document type = Letter
| page count = 7
| project = TAC:MF3068
| stage = Response to RAI
}}
 
=Text=
{{#Wiki_filter:m EntergyEntergy Nuclear Operations, Inc.Vermont Yankee320 Governor Hunt RdVernon, VT 05354Tel 802 257 7711Christopher J. WamserSite Vice PresidentBVY 14-059August 6, 2014ATTN: Document Control DeskU.S. Nuclear Regulatory CommissionWashington, DC 20555
 
==SUBJECT:==
Technical Specifications Proposed Change No. 306 Eliminate Certain ESFRequirements during Movement of Irradiated Fuel -Supplement 2(TAC No. MF3068)Vermont Yankee Nuclear Power StationDocket No. 50-271License No. DPR-28
 
==REFERENCES:==
: 1. Letter, Entergy Nuclear Operations, Inc. to USNRC, "TechnicalSpecifications Proposed Change No. 306 Eliminate Certain ESFRequirements during Movement of Irradiated Fuel," BVY 13-097,dated November 14, 2013 (TAC No. MF3068) (ML13323A516)2. Email, USNRC to Entergy Nuclear Operations, Inc. "RAI -EliminateCertain ESF Requirements During Fuel Movement (TAC No.MF3068)," dated July 29, 2014 (ML14210A159)
 
==Dear Sir or Madam:==
By letter dated November 14, 2013 (Reference 1), Entergy Nuclear Operations, Inc. (ENO)proposed an amendment to Renewed Facility Operating License (OL) DPR-28 for Vermont YankeeNuclear Power Station (VY). The proposed amendment would change the Technical Specification(TS) requirements associated with operability requirements for secondary containment whenhandling sufficiently decayed irradiated fuel or a fuel cask.In Reference 2, the NRC provided VY with a Request for Additional Information (RAI) regarding theproposed changes. Attachment 1 of this letter provides the responses to the RAI.The conclusions of the no significant hazards consideration and the environmental considerationscontained in Reference 1 are not affected by, and remain applicable to, this supplement.This letter contains no new regulatory commitments.If you have any questions on this transmittal, please contact Mr. Philip Couture at 802-451-3193.Aw¶c BVY 14-059 / page 2 of 2I declare under penalty of perjury that the foregoing is true and correct.Executed on August 6, 2014.Sincerely,CJW/plc
 
==Attachment:==
: 1. Response to Request for Additional Informationcc: Mr. William M. DeanRegion 1 AdministratorU.S. Nuclear Regulatory Commission2100 Renaissance Blvd, Suite 100King of Prussia, PA 19406-2713Mr. James S. Kim, Project ManagerDivision of Operating Reactor LicensingOffice of Nuclear Reactor RegulationU.S. Nuclear Regulatory CommissionMail Stop O8C2AWashington, DC 20555USNRC Resident InspectorVermont Yankee Nuclear Power Station320 Governor Hunt RoadVernon, VT 05354Mr. Christopher Recchia, CommissionerVT Department of Public Service112 State Street, Drawer 20Montpelier, VT 05620-2601 BVY 14-059Docket 50-271Attachment 1Vermont Yankee Nuclear Power StationResponse to Request for Additional Information BVY 14-059 / Attachment 1 / Page 1 of 4REQUEST FOR ADDITIONAL INFORMATIONBY CONTAINMENT AND VENTILATION BRANCHREVIEW OF TECHNICAL SPECIFICATIONS PROPOSED CHANGE NO. 306VERMONT YANKEE NUCLEAR POWER STATIONDOCKET NO. 50-271By letter dated November 14, 2013 (Reference 1), Entergy Operations, Inc. (Entergy, or thelicensee) submitted a license amendment request (LAR) for Vermont Yankee Nuclear PowerStation (VY), which proposes to change the requirements associated with handlingirradiated fuel and performing core alterations in its Renewed Facility Operating Licenseand Technical Specifications (TS).The Containment and Ventilation Branch (SCVB) staff has reviewed the LAR and isrequesting responses to the following items to complete its review:SCVB-RAI-1: TSTF-51Section 2.2 of Reference 1, Attachment 1 describes that VY will commit the two guidelinesas described in the "Reviewer's Note" of TSTF-51, "Revise Containment RequirementsDuring Handling Irradiated Fuel and Core Alterations" if the term "recently" is to be addedto the TS. Regarding the commitment, please address the following questions:(a) For the first guideline, has VY assessed the ventilation system and radiation monitoravailability with respect to filtration and monitoring of release from the fuel? Pleaseprovide the assessment.(b) For the second guideline, justify that the proposed contingency plans will meet the"prompt" purpose with respect to enabling ventilation systems to draw the releasefrom a postulated fuel handling accident in the proper direction such that it can betreated and monitored.Response(a) As stated in the regulatory commitment made in Reference 1, an assessment of ventilationsystem and radiation monitor availability will be completed prior to use of the amendment.This means that an assessment would have to be performed prior to any fuel handlingoperations with openings in secondary containment following permanent defueling of theVY reactor. A formal assessment has not been completed because Entergy NuclearOperations, Inc. (ENO) anticipates completion of permanent defueling of VY in January2015. This will allow sufficient time to complete an assessment prior to movement of fuelfollowing implementation of the approved amendment that is based on ventilation systemand radiation monitoring availability and decay of the spent fuel, since as noted in theresponse to RAI 2 in Reference 2, ENO has changed the requested approval date of theproposed changes to be contingent upon the docketing of the certifications for permanentcessation of operations and permanent removal of fuel from the reactor vessel inaccordance with 10 CFR 50.82(a)(1)(i) and (ii) and following a minimum of 13 days after thepermanent cessation of operations. ENO also notes that the assessment may need to bere-performed following extended periods between fuel handling operations. Items that maybe considered during the assessment include:i. Availability of the Standby Gas Treatment System BVY 14-059 / Attachment 1 /Page 2 of 4ii. Availability of the Reactor Building Ventilation Systemiii. Availability of Refuel Floor and Reactor Building Exhaust VentilationMonitorsiv. Availability of the Stack Gas Radiation Monitorv. Potential modifications to change the filtration capability of the ReactorBuilding Ventilation Systemvi. Analysis to determine when sufficient decay of the spent fuel has occurredsuch that the filtration of a main stack release would be inconsequential tothe Total Effective Dose Equivalent at the Exclusion Area Boundary.(b) Contingency methods to ensure prompt closure of openings in secondary containment havenot been finalized. Similar to the response in part (a), the regulatory commitment providedin Reference 1 requires that the contingency methods be implemented prior to use of theamendment. Since ENO anticipates completion of permanent defueling of VY in January2015, there will be sufficient time prior to movement of fuel following implementation of theapproved amendment to develop and provide training on robust measures to ensure thatthe contingency methods will implemented promptly. ENO also notes that there will beinherent variability in each contingency plan that is implemented prior to a period of fuelhandling operations due to the unique set of conditions likely to be present in terms of thenumber, and location, of open penetrations in secondary containment.The contingency methods will include the specific measures that can used to minimize theresponse time to an event requiring implementation of the contingency plans and ensurethat openings are promptly closed. These methods are described in the regulatorycommitment provided in Reference 1 and are repeated here for convenience:Contingency plans for prompt closure of openings will include the following:* Equipment and tools needed to facilitate closure will be staged,* Personnel responsible for closure will be knowledgeable and trained in theprocedures for establishing building integrity,* The closure response team will be accompanied by a Radiation Protection (RP)technician for radiation protection monitoring,* Hoses and cables routed through openings will employ a means to allow rapid, safedisconnect and removal, and* One door in each airlock will be capable of expeditious closureSCVB-RAI-2: Water LevelIn Table 3-1, VYNPP-Design Input for FHA, of Reference 1, Attachment 4, the data forrequired water depth above fuel (B3) is input as 23 ft to support the overall pooldecontamination factor data as input (B2). Is there any plant procedure in place to assuresuch a minimum water level of spent fuel pool or reactor cavity pool to be maintained?ResponseSite procedure OP 1101, Management of Refueling Activities and Fuel Assembly Movement,contains a prerequisite to ensure that the reactor cavity is flooded, fuel pool gates removed andwater level maintained greater than or equal to 36 feet 10 inches in the spent fuel pool (SFP)(measured from the bottom of the SFP) prior to fuel movement. It is noted that maintaining the SFPlevel of 36 feet 10 inches during fuel movement does not ensure that 23 feet of water is maintainedabove a postulated dropped and damaged fuel assembly within the SFP (36 feet 10 inches BVY 14-059/ Attachment 1 / Page 3 of 4corresponds to the SFP low level alarm). However, as discussed below, the radiologicalconsequences of a FHA in the SFP are considered to be bounded by the design basis FHA overthe reactor core. During normal plant operations, water level is maintained 23 feet above the top ofthe active fuel in the SFP storage racks.Having the reactor cavity flooded during refueling operations ensures that 23 feet of water ismaintained over the fuel in the reactor core, given that the existing VY design basis FHA(Calculation VYC-2299, Radiological AST Fuel Handling Accident) is based on the drop of a fuelassembly onto the core. The design basis FHA also considers a FHA 24 hours after shutdown. Theanalysis utilized damaged rods from drop heights of 34 feet utilizing the General Electric StandardApplication for Reactor Fuel, GESTAR II method and 30 feet based on the maximum heightallowed by VY refueling equipment. A fuel assembly drop height of 34 feet was used in conjunctionwith a decontamination factor of 200 (associated with 23 feet of water above damaged fuel). TheFHA analysis of record is based on a fuel assembly drop of 34 feet onto the core, 24 hours aftershutdown. This is considered to be bounding compared to a drop of significantly less height of afuel assembly over the spent fuel pool. The license amendment request (Reference 1) andaccompanying analysis to allow fuel moves with an open containment with a period of sufficientradioactive decay (fuel moves at 13 days) is considered to be bounded by the existing analysis ofrecord (fuel moves at 24 hours).SCVB-RAI-3: Release PathIn Table 3-3, Atmospheric Dispersion Factors for the Postulated FHA, of Reference 1,Attachment 4, it indicates that the only release path from reactor building is via the reactorbuilding blowout panel. Is there any other leakage path to plant personnel, main controlroom, existing?ResponseThe release point selected for the analysis was a RB blowout panel located near the main controlroom air intake as this creates the most limiting combination of release point and the receptorpoint. The use of this limiting combination in the analysis ensures that any other potential leakagepaths to the main control room are bounded by the results.SCVB-RAI-4: Fuel Assembly Drop or Fuel Cask Drop?In Section 3.6, Fuel Cask Movement, of Reference 1, the statement starting from the thirdsentence in the first paragraph looks like described for the drop of fuel assembly not forfuel cask drop. Clarify the information presented in this paragraph is for fuel assembly dropor fuel cask drop.ResponseThe information in the paragraph was intended to show that a drop of a fuel cask is not a credibleevent. The first paragraph of Section 3.6 of Reference 1 should read as follows (changes areshown in underline/strikethrough format):The operability requirements during movement of a fuel cask for ESF mitigation are deletedas part of this proposed license amendment. There is no The-ony accident postulatedduring handling of a fuel cask is-the-F-HA, as the crane is considered to be single failureproof as described below. The design basis accident FHA only assumes an irradiated fuelassembly is dropped onto the reactor core from the maximum height allowed by the fuel BVY 14-059 / Attachment 1 / Page 4 of 4handling equipment. The analysis assumes that the entire amount of potential energy isavailable for application to the fuel assemblies involved in the accident. Also, none of theenergy associated with the dropped fuel assembly is absorbed by the fuel material.SCVB-RAI-5: AuditOther than RG 1.183, there are four major references providing the input data andinformation for the re-analysis of AST/FHA (Attachment 4 of Reference 1). To have acomplete review and facilitate the review, provide the following references for audit viaelectronic reading room:3.* ENTERGY Calculation VYC-2299, "Radiological AST Fuel Handling Accident Analysis[PSAT 3019CF.QA.05, Rev. 0]" (Jun. 2003)4. AREVA NP Document 32-9053350-001, "ELISA-2 -A Software Package for theRadiologicalEvaluation of Licensing and Severe Accidents at Light-Water NuclearPower Plants Based on the Classical and Alternative-Source-Term Methodologies"(Aug. 2008) [See also AREVA NP Document 2A4.26-2A4-ELISA2-2.4_UsersManual-000, "ELISA-2 Version 2.4 User's Manual -Revision 2".]9.* ENTERGY Calculation VYC-2260, "Bounding Core Inventories of Actinides andFission Products for Design-Basis Applications at 1950 MWt" (Rev. 0, Feb. 2003)10.* ENTERGY Calculation VYC-2275, "Control Room Air Intake X/Q Due to Release fromReactor Building Blowout Panel Using Arcon96 Methodology" (Rev. 0, April 2003)ResponseThe listed documents will be made available to the NRC staff via electronic reading room or otherappropriate means.REFERENCES1. Letter, Entergy Nuclear Operations, Inc. to NRC, "Technical Specifications ProposedChange No. 306, Eliminate Certain ESF Requirements during Movement of Irradiated Fuel,Vermont Yankee Nuclear Power Station, Docket No. 50-271 License No. DPR-28", BVY 13-097, dated November 14, 2013 (ML13323A516) (TAC No. MF3068)2. Letter, Entergy Nuclear Operations, Inc. to NRC, "Technical Specifications ProposedChange No. 306 Eliminate Certain ESF Requirements during Movement of Irradiated Fuel -Supplement 1 (TAC No. MF 3068)," BVY 14-036, dated June 9, 2014}}

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