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{{Adams
#REDIRECT [[BVY 14-036, Response to Request for Additional Information on Technical Specifications Proposed Change No. 306, Eliminate Certain ESF Requirements During Movement of Irradiated Fuel - Supplement 1]]
| number = ML14163A008
| issue date = 06/09/2014
| 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 1
| 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 14-036, TAC MF3068
| document type = Letter
| page count = 25
| project = TAC:MF3068
| stage = Response to RAI
}}
 
=Text=
{{#Wiki_filter:v"Entergy Entergy Nuclear Operations, Inc.Vermont Yankee320 Governor Hunt RdVernon, VT 05354Tel 802 257 7711Christopher J. WamserSite Vice President BVY 14-036June 9, 2014ATTN: Document Control DeskU.S. Nuclear Regulatory Commission Washington, DC 20555
 
==SUBJECT:==
 
Technical Specifications Proposed Change No. 306 Eliminate Certain ESFRequirements during Movement of Irradiated Fuel -Supplement 1(TAC No. MF3068)Vermont Yankee Nuclear Power StationDocket No. 50-271License No. DPR-28
 
==REFERENCES:==
: 1. Letter, Entergy Nuclear Operations, Inc. to USNRC, "Technical Specifications Proposed Change No. 306 Eliminate Certain ESFRequirements during Movement of Irradiated Fuel," BVY 13-097,dated November 14, 2013 (TAC No. MF3068) (ADAMS Accession No. ML13323A516)
: 2. Email, USNRC to Entergy Nuclear Operations, Inc. "Vermont YankeeRAI for LAR on Eliminate Certain ESF Requirements DuringMovement of Irradiated Fuel (TAC No. MF3068),"
dated May 19,2014
 
==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 handling irradiated fuel and performing core alterations.
Specifically, the changes would eliminate operability requirements for secondary containment when handling sufficiently decayed irradiated fuel or a fuel cask and while performing corealterations.
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. Attachment 2 ofthis letter provides a revised markup of the VY TS pages affected by the RAI response.
The conclusions of the no significant hazards consideration and the environmental considerations contained in Reference 1 are not affected by, and remain applicable to, this supplement.
This letter contains no new regulatory commitments.
AWo~jz(i BVY 14-036 / page 2 of 2If you have any questions on this transmittal, please contact Mr. Philip Couture at 802-451-3193.
I declare under penalty of perjury that the foregoing is true and correct.Executed on June 9, 2014.Sincerely, CJW/plcAttachments:
: 1. Response to Request for Additional Information
: 2. Markup of Technical Specification Pagescc: Mr. William M. DeanRegion 1 Administrator U.S. Nuclear Regulatory Commission 2100 Renaissance Blvd, Suite 100King of Prussia, PA 19406-2713 Mr. James S. Kim, Project ManagerDivision of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 08D15Washington, DC 20555USNRC Resident Inspector Vermont Yankee Nuclear Power Station320 Governor Hunt RoadVernon, VT 05354Mr. Christopher
: Recchia, Commissioner VT Department of Public Service112 State Street, Drawer 20Montpelier, VT 05620-2601 BVY 14-036Docket 50-271Attachment 1Vermont Yankee Nuclear Power StationResponse to Request for Additional Information BVY 14-036/ Attachment 1 / Page 1 of 10REQUEST FOR ADDITIONAL INFORMATION TECHNICAL SPECIFICATION TASK FORCE TRAVELER 51VERMONT YANKEE NUCLEAR POWER STATIONDOCKET NO. 50-271By application dated November 14, 2013 (Agencywide Documents Access andManagement System (ADAMS) Accession No. ML13323A518),
Entergy NuclearOperations submitted a license amendment for Vermont Yankee (VY). The proposedlicense amendment request (LAR) would eliminate operability requirements for secondary containment when handling sufficiently decayed irradiated fuel and while performing corealterations using Technical Specification Task Force (TSTF) -51, "Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations."
RAI 1Attachment 4, Table 3-2, entitled "VYNPP [VY Nuclear Power Plant] -Re-analysis ofAST/FHA [alternative source term/fuel handling accident]
Radiological Consequences withOpen Containment" (ADAMS Accession No. ML13323A519) of the November 14, 2013application, provides a core inventory based upon a core average maximum burnup of 58giga-watt-days per metric ton of uranium (GWD/MTU).
Attachment 4, Table 3-1 states thatthe FHA uses Table 3 gap fractions from Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear PowerReactors" (Adams Accession Number ML003716792).
Footnote 11 for Table 3 of RG1.183 states that Table 3 is acceptable for use with currently approved reactor light waterfuel with a peak burnup of up to 62,000 mega-watt-days per metric ton of uranium(MWDIMTU)
(equivalent to 62 GWDIMTU) provided that the maximum linear heat generation rate does not exceed 6.3 kilowatts per foot (kW/ft) peak rod average power for burnupsexceeding 54,000 MWD/MTU (equivalent to 54 GWD/MTU).
Since the assumed fuel burnupfor the Attachment 4, Table 3-2 core inventories appear to exceed the RG 1.183, footnote 11limits, please confirm that the VY fuel burnup and linear heat generation rates comply withfootnote
: 11. If not, please justify the use of Table 3 from RG 1.183 with fuel outside theburnup and linear heat generation rates used to derive Table 3.ResponseEntergy Nuclear Operations, Inc. (ENO) confirms that the fuel burnup and linear heat generation rates comply with the RG 1.183 footnote 11 limits on burnup and maximum linear heat generation rate. The RG 1.183 footnote 11 limits were specifically evaluated for the current VY operating cycleduring reload licensing by Global Nuclear Fuel (GNF). Only significantly different control rodpatterns and operation could result in exceeding the 6.3 kW/ft linear heat generation rate for thosefuel rods with burnup greater than 54 GWD/MTU.
Such changes would be evaluated by ENO andGNF before implementation.
The RG 1.183 footnote 11 limits were again evaluated by GNF for the planned extended operating cycle prior to the final shutdown of VY and were likewise found to be met. ENO reviewed theverified summary edits from GNF and found that the limits were met through the end of theplanned cycle with significant margin.
BVY 14-036 / Attachment 1 / Page 2 of 10RAI 2Page 9 of 17 of the application, entitled "Technical Specifications Proposed Change No.306, Eliminate Certain ESF [Engineered Safety Feature]
Requirements during Movement ofIrradiated Fuel," dated November 14, 2013 (ADAMS Accession Number ML13323A518) states:The accidents postulated to occur during core alterations, in additionto the fuel handling
: accident, are [the] inadvertent criticality due tocontrol rod removal error and the inadvertent loading of, andsubsequent operation with, a fuel assembly in an improper location.
These events are not postulated to result in fuel cladding integrity damage. Therefore, the only accident postulated to occur during corealterations that result in significant radioactive release is the FHA [fuelhandling accident].
Thus, the consequence of a FHA envelops theconsequences of potential accidents postulated to occur during corealterations.
Page 14 of 17 of the application also states that the proposed changes follow Technical Specification Task Force traveler 51 (TSTF-51),
Revision 2, "Revise Containment Requirements during Handling Irradiated Fuel and Core Alterations" (ADAMS Accession Number ML040400343).
TSTF-51 states:The addition of the term "recently" associated with handling irradiated fuel in all ofthe containment function Technical Specification requirements is only applicable tothose licensees who have demonstrated by analysis
[emphasis added] that aftersufficient radioactive decay has occurred, off-site doses resulting from a fuelhandling accident remain below the Standard Review Plan limits (well within 10 CFR100) [or 10 CFR 50.67].Standard Review Plan (SRP) 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms," (ADAMS Accession Number ML003734190) states:The models, assumptions, and parameter inputs used by the licensee should bereviewed to ensure that the conservative design basis assumptions outlined in RG-1.183 have been incorporated.
Appendix B of Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms forEvaluating Design Basis Accidents at Nuclear Power Reactors" (ADAMS Accession NumberML003716792),
Regulatory Position 1.1 states:The number of fuel rods damaged during the accident should be based on aconservative analysis that considers the most limiting case.After reviewing the information submitted by the VY submittal to adopt TSTF-51, the NRCstaff needs additional information to verify that the limiting cases have been considered.
: a. Please provide a FHA analysis that evaluates the dropping of loads allowed overirradiated fuel assemblies (i.e. sources or reactivity control components) ontoirradiated fuel assemblies with 24-hours of decay time. The analysis should onlycredit those safety systems required to be operable as required by technical BVY 14-036 / Attachment 1 / Page 3 of 10specification
[TS]. This will provide the staff with reasonable assurance that the FHAdoses remain within regulatory limits when references to Core Alterations areremoved from TSs and ESFs are no longer required during movement of loads suchas sources or reactivity control components.
: b. Page 7 of 17 of the application dated November 14, 2013 states that two mainconfigurations of the Reactor Building during fuel movement were considered.
Thesecond configuration discusses "various
[emphasis added] pre- and post-FHA MainControl Room (CR) ventilation configurations that would support refueling with opencontainment,"
but does not define which configurations are credited in the proposedTS changes.
These ventilation configurations are discussed in the submittal, but theNRC staff needs some clarification regarding these configurations.
Please state theproposed new design basis configuration credited to support the TS changes.Responsea. To address the NRC staff concerns with removal of references to Core Alterations from theTS, ENO is retracting from the license amendment request (Reference
: 1) the proposedremoval of any references to Core Alterations from the VY TS. This retraction is based onthe planned permanent cessation of power operations of VY at the end of the currentoperating cycle, which is expected to occur in the fourth quarter of 2014 (Reference 2).Once the VY reactor is permanently defueled and the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel are docketed inaccordance with 10 CFR 50.82(a)(1)(i) and (ii), per 10 CFR 50.82(a)(2),
the 10 CFR Part 50license no longer will permit operation of the reactor or placement of fuel in the reactorvessel and the term Core Alteration will have no meaning as there will no longer be areactivity concern with the reactor core. Revised markups to the affected TS pages areprovided in Attachment 2 of this letter. As in the LAR, proposed changes to the TS Basesare provided for information in Attachment
: 2. Upon approval of this amendment, changes tothe Bases will be incorporated in accordance with TS 6.7.E, the TS Bases Control Program.To provide additional assurance that the changes proposed in Reference 1 will not beimplemented while there is still fuel in the VY reactor vessel or before 13 days have passedfollowing permanent cessation of operations, ENO is revising the requested approval dateof the proposed changes to be contingent upon the docketing of the certifications forpermanent cessation of operations and permanent removal of fuel from the reactor vesselin accordance with 10 CFR 50.82(a)(1)(i) and (ii) and following a minimum of 13 days afterthe permanent cessation of operations.
ENO no longer requests that the proposed changesbe approved by December 1, 2014.b. No new design basis configuration is proposed as part of TS changes.
The CR ventilation configuration credited to support the proposed TS changes does not differ from the currentVY design associated with the analysis of record. The analysis supporting the proposed TSchange utilizes the following inputs:" No credit for station containment systems (i.e. "open" containment)
* Release point to atmosphere is via a Reactor Building (RB) blowout panel (ground levelrelease)* Release duration to atmosphere is 2 hour release duration per RG 1.183, Appendix B* 3,700 cfm unfiltered Control Room intake* 30 day exposure interval for the Control Room per conservative assumption BVY 14-036 / Attachment 1 / Page 4 of 10* 2 hour exposure interval for the Exclusion Area Boundary (EAB) and Low Populations Zone (LPZ) per RG 1.183, Section 4.1.5Alternate configurations discussed in the FHA analysis are not credited to support theproposed TS changes.
These cases were run for sensitivity studies not associated with theproposed changes.Also note that, as discussed in Section 2.2.4 of the VY Updated Final Safety AnalysisReport (UFSAR),
the EAB is 910 feet from the reactor at the closest point and the LPZ is a5 mile radius and accordingly the dose at the LPZ would be lower. As stated in the currentanalysis of record, calculation of the LPZ dose was considered not to be necessary because the EAB dose is more limiting.
RAI 3Page 10 of 17 of the application, dated November 14, 2013, states:The operability requirements during movement of a fuel cask for ESF mitigation aredeleted as part of this proposed license amendment.
and,Since the FHA resulting from a dropped fuel cask is shown to not be credible, theproposed TS changes omitting operability requirements during movement of a fuelcask ESF mitigation is justified.
SRP 15.7.5, "Spent Fuel Cask Drop Accidents,"
(ADAM Accession No. ML052350315) states:A design basis radiological analysis is performed if a cask drop exceeding 30 feetcan be postulated or if limiting devices are removed during cask handling within theplant so the 30-foot drop height is exceeded.
If the radiological consequences of acask drop accident are to be computed, then information on whether buildingleaktightness can be expected after a cask drop is obtained from ASB [Auxiliary Systems Branch] (e.g., whether the technical specifications require large doors to beclosed during fuel handling or whether ventilation systems should be operating andwhether the building leaktightness would be violated by the cask drop).At VY can a spent fuel cask drop exceed 30 feet or can the limiting devices be removedduring cask handling?
If so, please provide the radiological consequences of a cask dropaccident.
Please justify all answers.ResponseWhile a spent fuel storage cask can be raised to a height exceeding 30 feet, a cask drop is not apostulated event at VY. The basis for this determination is documented in Section 3.6 of the LAR(Reference 1).RAI 4Page 9 of Attachment 4 of the application, dated November 14, 2013, states that the activityreleases from the containment atmosphere over two hours is 98.2%. Appendix B of RG1.183, Regulatory Position 5.3 states if the containment is open during fuel handling, theradioactivity that escapes from the reactor cavity pool to the containment is released to theenvironment over a two-hour time period. RG 1.83, Regulatory Position 5.1.2, "Assignment BVY 14-036 / Attachment 1 / Page 5 of 10of Numeric Input Values,"
states that the numeric values that are chosen as inputs to therequired analysis should be selected with the objective of determining a conservative dose.Please justify why a conservative value of 100% of the activity in containment was notassumed to be released from the containment over the two hour time period.ResponseAs described in Section 2.2 of Attachment 4 of the LAR (Reference 1), an exponential release tothe environment was assumed for the post-FHA radioactive material that escapes the water pool,based on a building air exchange rate of 2.0 air changes per hour. This air exchange rate leads to[1.0 -exp{-2.0 (hr1)
* 2 (hr)}] = 98.17% of the airborne activity within the reactor building gettingreleased within 2 hrs. This analytical approach using the air exchange rate was selected as areasonable conservative assumption employed in the calculation to support the proposed changes.The table below provides updated EAB dose rates for 100% release of the source within a 2 hourinterval.
This is based on simply increasing the doses documented in Table 5-3 of Attachment 4 ofthe LAR by (100 -98.17) = 1.83%. It is seen that the regulatory limit of 6.3 rem is met in all cases.VY FHA WITH OPEN CONTAINMENT
-EAB DOSEDecay Time EAB TEDE Dose (rem)(days) Original Adjusted(98.17% Release)
(100% Release)1 5.895 6.003(a)3 3.643 3.7105 2.953 3.0077 2.451 2.4969 2.042 2.07911 1.705 1.73613 1.424 1.45015 1.190 1.21217 0.9957 1.01419 0.8333 0.8485(a) 5.895 x 1.0183 6.003 remThe CR doses reported in Table 5-3 of Attachment 4 of the LAR do not require a similaradjustment since the dose analysis is for a period of 30 days. This is clarified in Section 2.2 ofAttachment 4 of the LAR, which states:It is noted that, for the MCR 30-day dose computations, the releases from the RB wereassumed to continue for 30 days. Included in the releases beyond 2 hours are the (100 -98.2) = 1.8% still airborne within the RB at 2 hrs, as well as the noble gases generated bythe decay of iodines retained by the pool water.There is no change to the 13 day time period after shutdown required for the fuel to decay suchthat the dose limits are not exceeded, since the CR dose is not impacted by this addedconservatism and this dose was the limiting factor in the determination of the time period.
BVY 14-036 / Attachment 1 / Page 6 of 10RAI 5Page 10 of Attachment 4 of the application, dated November 14, 2013, states that foursensitivity cases make use of several rates to assess the dose impact on the main CR purgeinitiation time. Please describe which case is to be reviewed for the design basis and clarifywhat is meant by the "purge initiation time."ResponseSee response to RAI 2.b. The term "purge initiation time' is associated with the non-design basissensitivity study described in Table 5-6, Page 31 of Attachment 4 of the LAR that is not associated with the proposed TS changes.RAI 6Regulatory Position 5.1.2 of RG 1.183 states: "The single active component failure thatresults in the most limiting radiological consequences should be assumed."
State the mostlimiting single active failure for FHA and justify the answer.ResponseThere are no ESF components employed or credited in the FHA analysis, hence there was norequirement for single-failure assumption.
No single failure is postulated.
It is noted, in particular, that the CR ventilation system has no filtration capability, and that it was assumed to be in thenormal operating mode.RAI 7Attachment 4, Table 3-1, dated November 14, 2013, states that VY assumes an overall pooldecontamination factor (or DF) of 200 based upon Appendix B of RG 1.183. The DF of 200 isbased upon reference B-1 ("Evaluation of Fission Product Release and Transport,"
(ADAMSAccession No. 8402080322))
of RG 1.183. The data upon which the pool DF of 200 is basedwas developed in 1971 and was based on the Westinghouse fuel marketed at the time (theassumed internal fuel pressure of 1200 pounds force per square inch gage (psig) was used).Since higher pressures correlate to lower DFs, the NRC staff would like VY to confirm thatthe fuel VY uses will have an internal fuel pressure of less than 1200 psig. If not, pleaseprovide the experimental data for current fuel types used at VY that justify a DF of 200 forfuel pressures greater than 1200 psig. Also, please provide a detailed justification for usinga DF of 200 for pressures up to 1200 psig.ResponseENO confirms that the fuel in use at VY has an internal pressure of less than 1200 psig.RAI 8Please provide a justification for all changes from the current licensing basis (See Issue I ofNRC Regulatory Issue Summary 2006-04, "Experience with Implementation of Alternative Source Terms," (ADAMS Accession No. ML053460347) for more detail).
No justification isneeded for changes that are consistent with Regulatory Guide 1.183 or are provided in thesubmittal dated August 13, 2013 (ADAMS Accession No. ML13247A076) unless requested by these RAIs.
BVY 14-036 / Attachment 1 / Page 7 of 10ResponseJustification for all changes to the current licensing basis was provided in the application for thelicense amendment, as supplemented by the responses to these RAIs.VY has previously received NRC approval for full-scope implementation of AST by letter datedMarch 29, 2005 (Reference 3).The analysis performed in support of the proposed changes and provided in the application wasintended to answer the specific question of how long it would take for used reactor fuel to decay tothe point that the radiological consequences of a FHA would not result in the offsite and controlroom accident dose criteria being exceeded.
The analysis is not intended to supersede the AST-based FHA analysis that was previously reviewed by the NRC staff.The analysis provided in the application addressed all characteristics of the AST related to the FHAand the TEDE criteria as described in the VY design basis. Therefore, the conclusions reached bythe NRC staff in approving full implementation of AST at VY remain valid.RAI 9The changes to TS 3.9.4 allow an "open" containment when moving fuel that is not recentlyirradiated.
Consistent with Regulatory Issue Summary 2006-04, please confirm that allpathways to the environment created by the proposed changes are considered andanalyzed in the FHA analysis.
: a. Please confirm that the most limiting combination of release point and receptor forthe control room were used to determine atmospheric dispersion factors for eachaccident.
: b. State and justify the release points that correlate to the atmospheric dispersion factors used.ResponseThe release point and atmospheric dispersion factors employed in the FHA analysis with opencontainment are identified in Table 3-3 of Attachment 4 of the LAR. The release point is the RBblowout panels. This forms the most limiting combination when paired with the receptor for themain control room based on the proximity of the RB blowout panels to the main control room airintake. The atmospheric dispersion factors were based on ARCON-96, along with the following:
5 year's worth of hourly meteorological data collected on site (1995-1999).
The building area used for the wake correction was the projected area of the reactorbuilding wall facing the control room air intake.The distance from the source (mid-point of the RB blowout panel) to the control room airintake (32 feet), source/receptor elevations, and the wind direction were based on sitedrawings.
The RB siding facing the Control Room intake was treated conservatively as apoint source (USFAR, Table 14.6.10).
BVY 14-036 / Attachment 1 / Page 8 of 10It is noted that the atmospheric dispersion factors used in Attachment 4 of the LAR are the sameas those used in the applicable VY calculation of record for implementation of the ASTmethodology.
RAI 10SRP 16.0, "Technical Specifications,"
(ADAMS Accession No. ML100351425) states: "In TSchange requests for facilities with TS based on previous STS [Standard Technical Specifications],
licensees should comply with comparable provisions in these STS NUREGsto the extent possible or justify deviations from the STS." Please provide a justification fordeviations from the STS created by the proposed changes.ResponseThe following provides a comparison of the proposed changes to the VY TS to the changesproposed to the STS for NUREG-1433 in TSTF-51A.
NUREG-1433, Revision 4 is the STS forGeneral Electric (GE) Boiling Water Reactor/4 (BWR/4) plants (Reference 4). VY is a GE BWR/4plant. The following discussion also identifies any deviations from the proposed STS changes inTSTF-51A.
It is noted here that VY has "custom" TS and has not performed a conversion to theNUREG-1433 STS. Therefore, there are inherent wording differences between the equivalent VYTS and STS.TS Table 3.2.3, Reactor Building Ventilation Isolation and Standby Gas Treatment SystemInitiation Instrumentation:
TS Table 3.2.3 contains the instrumentation equivalent to those listed in STS Table 3.3.6.2-1, Secondary Containment Isolation Instrumentation.
Footnotes (c) and (d), applicable tothe High Reactor Building Ventilation Radiation (Trip Function
: 3) and High Refuel FloorZone Radiation (Trip Function
: 4) trip functions are proposed for revision.
Footnote (c) is proposed to be revised to be consistent with footnote (b) of STS Table3.3.6.2-1, with brackets removed from "recently" and "secondary".
There is no deviation from the STS with the proposed change to footnote (c). The proposed removal of "or fuelcask" from footnote (c) is an additional change not covered by the scope of TSTF-51A.
Footnote (d) [During Alteration of the Reactor Core] was proposed for deletion to beconsistent with the footnotes of STS Table 3.3.6.2-1 as there is no footnote related to CoreAlterations in STS Table 3.3.6.2-1.
Reference to Core Alterations was removed fromfootnote (c) of STS Table 3.3.6.2-1 by TSTF-51A.
: However, based on the response to RAI2 of this letter, ENO is retracting the portions of the proposed changes that would eliminate references to Core Alterations, thereby creating a deviation from the STS.TS 3.7.B.4, Standby Gas Treatment System:Proposed TS 3.7.B.4.b was to be revised to be consistent with the APPLICABILITY sectionof STS 3.6.4.3, Standby Gas Treatment (SGT) System in that reference to Core Alterations is removed and "recently" is added in front of "irradiated fuel" with brackets removed from"recently" and "secondary."
TSTF-51A removed the words "During CORE ALTERATIONS" and added "[recently]"
in front of "irradiated fuel" in the APPLICABILITY section.
Theproposed removal of "or the fuel cask" from TS 3.7.B.4.b is an additional change notcovered by the scope of TSTF-51A.
Based on the response to RAI 2 of this letter, ENO isretracting the portions of the proposed changes that would eliminate references to CoreAlterations, thereby creating a deviation from the STS. There is no deviation from the STS BVY 14-036 / Attachment 1 / Page 9 of 10with the proposed changes to TS 3.7.B.4.b in terms of the addition of "recently" in front of"irradiated fuel."Proposed TS 3.7.B.4.b.i will be revised to be consistent with STS 3.6.4.3 REQUIREDACTION C.2.1 in that "recently" is added in front of "irradiated fuel" with brackets removedfrom "recently" and "secondary."
The proposed removal of "or the fuel cask" from TS3.7.B.4.b.i is an additional change not covered by the scope of TSTF-51A.
There is nodeviation from the STS with the proposed changes to TS 3.7.B.4.b.i.
Proposed TS 3.7.B.4.b.ii was proposed for deletion to be consistent with the requiredactions of STS 3.6.4.3 as reference to Core Alterations was removed from STS 3.6.4.3CONDITION C by TSTF-51A.
: However, based on the response to RAI 2 of this letter, ENOis retracting the portions of the proposed changes that would eliminate references to CoreAlterations, thereby creating a deviation from the STS.TS 3.7.C, Secondary Containment System:Proposed TS 3.7.C.1.b and TS 3.7.C.1.c were proposed to be revised to be consistent withthe APPLICABILITY section of STS 3.6.4.1, Secondary Containment, in that reference toCore Alterations is removed from TS 3.7.C.1.c and, for TS 3.7.C.1.b, "recently" is added infront of "irradiated fuel" with brackets removed from "recently" and "secondary."
TSTF-51Aremoved the words "During CORE ALTERATIONS" and added "[recently]"
in front of"irradiated fuel" in the APPLICABILITY section.
The proposed removal of "or the fuel cask"from TS 3.7.C.1.b is an additional change not covered by the scope of TSTF-51A.
Basedon the response to RAI 2 of this letter, ENO is retracting the portions of the proposedchanges that would eliminate references to Core Alterations, thereby creating a deviation from the STS. There is no deviation from the STS with the proposed changes to TS3.7.0C.1.b.
Proposed TS 3.7.C.4 will be revised to be consistent with STS 3.6.4.1 CONDITION C inthat reference to Core Alterations is removed from TS 3.7.C.4 and "recently" is added infront of "irradiated fuel" with brackets removed from "recently" and "secondary."
Theproposed removal of "or the fuel cask" from TS 3.7.C.4 is an additional change not coveredby the scope of TSTF-51A.
Based on the response to RAI 2 of this letter, ENO is retracting the portions of the proposed changes that would eliminate references to Core Alterations, thereby creating a deviation from the STS. There is no deviation from the STS with theproposed change to TS 3.7.C.4 in terms of the addition of "recently" in front of "irradiated fuel."Proposed TS 3.7.C.4.a will be revised to be consistent with the STS 3.6.4.1 REQUIREDACTION C. 1, in that "recently" is added in front of "irradiated fuel" with brackets removedfrom "recently" and "secondary."
TSTF-51A added "[recently]"
in front of "irradiated fuel" inREQUIRED ACTION C.1. The proposed removal of "and the fuel cask" from TS 3.7.C.4.a isan additional change not covered by the scope of TSTF-51A.
There is no deviation from theSTS with the proposed changes to TS 3.7.C.4.a.
Proposed TS 3.7.C.4.b was proposed for deletion to be consistent with STS 3.6.4.1REQUIRED ACTION C.2. TSTF-51A removed "Suspend CORE ALTERATIONS" fromREQUIRED ACTION C.2. However, based on the response to RAI 2 of this letter, ENO isretracting the portions of the proposed changes that would eliminate references to CoreAlterations, thereby creating a deviation from the STS.
BVY 14-036 / Attachment 1 / Page 10 of 10TSTF-51A also included changes to the following STS sections applicable to NUREG-1433:
* AC Sources -Shutdown (STS 3.8.2)* DC Sources -Shutdown (STS 3.8.5)* Inverters
-Shutdown (STS 3.8.8)" Distribution Systems -Shutdown (STS 3.8.10)" Secondary Containment Isolation Valves (STS 3.6.4.2)* Main Control Room Environmental Control System Instrumentation (STS 3.3.7.1)* Primary Containment Isolation Valves (STS 3.6.1.3)* Main Control Room Environmental Control System (STS 3.7.4)* Control Room Air Conditioning System (STS 3.7.5)VY reviewed those STS sections for applicability to the VY TS and determined that no additional changes to the VY TS were required in order to implement the proposed changes.
In particular TS3.7.E (Reactor Building Automatic Ventilation System Isolation Valves (RBAVSIVs))
was notincluded in the scope of the proposed changes on the basis that this specification only applieswhen secondary containment integrity is required.
TS 3.7.E is consistent with the applicable STSsection, 3.6.4.2 (Secondary Containment Isolation Valves).
Omission of TS 3.7.E from theproposed changes is acceptable because once VY dockets the certifications of permanent cessation of power operations and permanent defueling of the reactor required by 10 CFR50.82(a)(1)(i) and (ii), respectively, pursuant to 10 CFR 50.82(a)(2),
the Part 50 license will nolonger authorize operation of the VY reactor or emplacement or retention of fuel within the VYreactor vessel. Since the VY reactor will not be allowed to be refueled and operated again, new"recently" irradiated fuel will not be able to be generated.
Upon approval of the proposed changesto TS 3.7.C, secondary containment integrity will not be required once the required fuel decay timepasses and, accordingly, it will no longer be possible for TS 3.7.E to be applicable.
The fact that TSTF-51A included changes to the STS that were determined to not be applicable tothe VY TS does not by itself represent a deviation from the STS.No justification for changes to the VY TS Bases based on corresponding changes made to theSTS Bases by TSTF-51A is provided since the changes to the TS Bases were provided forinformation only and will be incorporated in accordance with TS 6.7.E, the VY TS Bases ControlProgram.
 
==References:==
: 1. Letter, Entergy Nuclear Operations, Inc. to USNRC, "Technical Specifications ProposedChange No. 306 Eliminate Certain ESF Requirements during Movement of Irradiated Fuel(TAC No. MF3068),"
BVY 13-097, dated November 14, 2013 (ML13323A518)
: 2. Letter, Entergy Nuclear Operations, Inc. to USNRC, "Notification of Permanent Cessation ofPower Operations,"
BVY 13-079, dated September 23, 2013 (ML13273A204)
: 3. Letter, USNRC to Entergy Nuclear Operations, Inc., "Vermont Yankee Nuclear PowerStation Issuance of Amendment RE: Alternative Source Term (TAC No. MC0253),"
NVY05-045, dated March 29, 20054. NUREG-1433, Standard Technical Specifications, General Electric BWR/4 Plants, Revision4.0, published April 2012 BVY 14-036Docket 50-271Attachment 2Vermont Yankee Nuclear Power StationMarkup of Technical Specification Pages VYNPSTable 3.2.3 (page 1 of 1)Reactor Building Ventilation Isolation and Standby Gas Treatment SystemInitiation Instrumentation ACTIONSWHENREQUIRED REQUIREDAPPLICABLE MODES OR CHANNELS CHANNELSOTHER SPECIFIED PER TRIP ARETRIP FUNCTION CONDITIONS SYSTEM INOPERABLE TRIP SETTING1. Low Reactor RUN, STARTUP/HOT 2 Note 1 127.0 inchesVessel Water STANDBY, HOT SHUTDOWN, Level Refuel(a),
(b)2. High Drywell RUN, STARTUP/HOT 2 Note 1  2.5 psigPressure
: STANDBY, HOT SHUTDOWN, Refuel (a)3. High Reactor RUN, STARTUP/HOT 1 Note 1 ! 14 mR/hrBuilding
: STANDBY, HOT SHUTDOWN, Ventilation Refuel(a),
(b), Co), (d)Radiation
: 4. High Refueling RUN, STARTUP/HOT 1 Note 1 1 100 mR/hrFloor Zone STANDBY, HOT SHUTDOWN, Radiation Refuel (a), (b), cc, (d)(a) With reactor coolant temperature
> 212 'F.(b) During operations with potential for draining the reactor vessel.(c) During movement of irradiated fuel assemblies 6r fuel cask in secondary containment.
r---recently]
(d) During Alteration of the Reactor Core.Amendment No. 2-3451 VYNPS3.7 LIMITING CONDITIONS FOROPERATION 4.7 SURVEILLANCE REQUIREMENTS shutdown condition, the actions andcompletion times ofSpecification 3.7.B.4.b shallapply. After sevendays with aninoperable train ofthe Standby GasTreatment Systemduring refueling orcold shutdownconditions requiring secondary containment integrity, theoperable train ofthe Standby GasTreatment Systemshall be placed inoperation and itsassociated dieselgenerator shall beoperable, or theactions andcompletion times ofSpecification 3.7.B.4.b shallapply.4. With two trains of theStandby Gas Treatment System inoperable, or asmade applicable bySpecification 3.7.B.3:a. With the reactor inthe run mode,startup mode, or hotshutdown condition, the reactor shall beplaced in hotshutdown within 12hours and coldshutdown within 36hours.b. During movement of4--irradiated fuelassemblies A~- #1;Afel oaek in thesecondary containment, duringcore alterations, orduring operations with the potential for draining thereactor vessel,immediately:
recentlyAmendment No. 4-95155 VYNPS3.7 LIMITING CONDITIONS FOROPERATION
: i. Suspendmovement ofIrecently
>- irradiated fuelassemblies an4in secondary containment; andii. Suspend corealterations; andiii. Initiateaction tosuspendoperations with thepotential fordraining thereactorvessel.C. Secondary Containment System1. Secondary Containment Integrity shall bemaintained during thefollowing modes orconditions:
: a. Whenever the reactoris in the Run Mode,Startup Mode, or HotShutdown condition*;
or4.7 SURVEILLANCE REQUIREMENTS C. Secondary Containment System1. Secondary containment capability to maintain a0.15 inch of water vacuumunder calm wind(2<0<5 mph) conditions with a filter train flowrate of not more than1,550 cfm, shall bedemonstrated at leastquarterly.
* NOTE: The reactor mode switch may be changed to either the Run or Startup/Hot Standby position, and operation not considered to be in the Run Mode or StartupMode, to allow testing of instrumentation associated with the reactor modeswitch interlock functions, provided:
: 1. Reactor coolant temperature is < 212&deg;F;2. All control rods remain fully inserted in core cells containing one ormore fuel assemblies; and3. No core alterations are in progress.
Amendment No. 14-4, 4-4-, 4-9-, 2-2-3, 2-2-4155a VYNPS3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION
: b. During movement ofIrecently fuelassemblies er-thefuel-ea ^ insecondary containment; orc. During alteration ofthe Reactor Core; ord. During operations with the potential for draining thereactor vessel.Amendment No. 4-44, 4-9, 2-2, -26 1156 VYNPS3.7 LIMITING CONDITIONS FOROPERATION
: 2. With Secondary Containment Integrity notmaintained with thereactor in the Run Mode,Startup Mode, or HotShutdown condition, restore Secondary Containment Integrity within four (4) hours.4.7 SURVEILLANCE REQUIREMENTS
: 2. Intentionally blank.3. Intentionally blank.4. Intentionally blank.3.4.If Specification 3.7.C.2cannot be met, place thereactor in the HotShutdown condition within12 hours and in the ColdShutdown condition withinthe following 24 hours. recentlyWith Secondary Containment Int rity notmaintained d ingmovement of irradiated fuel assemblies er--tefuel eask in secondary containment, duringalteration of the ReactorCore, or duringoperations with thepotential for drainingthe reactor vessel,immediately perform thefollowing actions:a. Suspend movement ofIrecently
>-irradiated fuelassemblies aid-th4efuel -eas insecondary containment; andb. Suspend alteration of the Reactor Core;andc. Initiate action tosuspend operations with the potential for draining thereactor vessel.Amendment No. 44-4, 2-2-6157 VYNPSBASES: 3.2.C/4.2.C REACTOR BUILDING VENTILATION ISOLATION AND STANDBY GASTREATMENT SYSTEM INITIATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) instrumentation are implicitly assumed in the safety analyses of References 2,3, and 4, to initiate closure of the RBAVSIVs and start the SGT System to limitoffsite doses.Reactor building ventilation isolation and Standby Gas Treatment Systeminitiation instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)
(2) (ii).The operability of the reactor building ventilation isolation and Standby GasTreatment System initiation instrumentation is dependent on the operability ofthe individual instrumentation channel Trip Functions specified in Table 3.2.3.Each Trip Function must have the required number of operable channels in eachtrip system, with their trip setpoints within the calculational as-foundtolerances specified in plant procedures.
Operation with actual trip setpoints within calculational as-found tolerances provides reasonable assurance that,under worst case design basis conditions, the associated trip will occur withinthe Trip Settings specified in Table 3.2.3. As a result, a channel isconsidered inoperable if its actual trip setpoint is not within thecalculational as-found tolerances specified in plant procedures.
The actualtrip setpoint is calibrated consistent with applicable setpoint methodology assumptions and recentlyIn genera , the individual Trip Functions are required to be OPERABLE in RUN,STARTUP/H T STANDBY, HOT SHUT OWN, Refuel (with reactor coolant temperature 212 F),\uring operations th the potential for draining the reactor vessel(OPDRVs).-_during movement of irradiated fuel assemblies rr fuel rask insecondary containment, and during Alteration of the Reactor Core; consistent with the Applicability for the SGT System and secondary containment requirements in Specifications 3.7.B and 3.7.C. Trip Functions that havedifferent Applicabilities are discussed below in the individual Trip Functions discussion.
Thk specific Applicable Safety Analyses, LCO, and Applicability discussions arelis ed below on a Trip Function by Trip Function basis.Insert 11. Low Reactor Vessel Water LevelLow reactor pressure vessel (RPV) water level indicates that the capability tocool the fuel may be threatened.
Should RPV water level decrease too far, fueldamage could result. An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of anoffsite release.
The Low Reactor Vessel Water Level Trip Function is one ofthe Trip Functions assumed to be operable and capable of providing isolation and initiation signals.
The isolation and initiation of systems on Low ReactorVessel Water Level support actions to ensure that any offsite releases arewithin the limits calculated in the safety analysis.
Low Reactor Vessel Water Level signals are initiated from level transmitters that sense the difference between the pressure due to a constant column ofwater (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Low Reactor Vessel Water Level TripFunction are available and are required to be operable to ensure that no singleinstrument failure can preclude the isolation and initiation function.
Amendment No. 4-7676p VYNPSBASES: 3.2.C/4.2.C REACTOR BUILDING VENTILATION ISOLATION AND STANDBY GASTREATMENT SYSTEM INITIATION APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) have originated from the primary containment due to a break in the RCPB or therefueling floor due to a fuel handling accident.
When High Reactor BuildingVentilation Radiation or High Refueling Floor Zone Radiation is detected, secondary containment isolation and actuation of the SGT System are initiated to support actions to limit the release of fission products as assumed in theUFSAR safety analyses (Ref. 4).The High Reactor Building Ventilation Radiation and High Refueling Floor ZoneRadiation signals are initiated from radiation detectors that are located onthe ventilation exhaust duct coming from the reactor building and the refueling floor zones, respectively.
Two channels of High Reactor Building Ventilation Radiation Trip Function and two channels of High Refueling Floor Radiation TripFunction are available and are required to be operable to ensure that no singleinstrument failure can preclude the isolation and initiation function.
The Trip Settings are chosen to promptly detect gross failure of the fuelcladding.
The High Reactor Building Ventilation Radiation and High Refueling Floor ZoneRadiation Trip Functions are required to be operable in RUN, STARTUP/HOT
: STANDBY, HOT SHUTDOWN, Refuel (with reactor coolant temperature
> 2120F) whereconsiderable energy exists in the RCS; thus, there is a possibility of pipebreaks resulting in significant releases of radioactive steam and gas. In COLDSHUTDOWN and Refuel (with reactor coolant temperature
< 2120F), the probability and consequences of these events are low due to the RCS pressure andtemperature limitations of these Modes; thus, these Trip Functions are notrequired.
In addition, the Trip Functions are also required to be operableduring OPDRVs.7 during movement of irradiated fuel assemblies cr fucl caR: inthe secondary ntainment, and du ing Alteration of the Reactor Core, becausethe capability f detecting radiaton releases due to fuel failures (due tofuel uncovery o dropped fuel asse lies) must be provided to ensure thatoffsite dose li its are not exceedand recentlyACTIONInsert 1Table 3.2. Note 1Because of the diversity of sensors available to provide isolation signals andthe redundancy of the isolation design, an allowable out of service time of12 hours or 24 hours depending on the Trip Function (12 hours for those TripFunctions that have channel components common to RPS instrumentation, i.e.,Trip Functions 1 and 2, and 24 hours for those Trip Functions that do not havechannel components common to RPS instrumentation, i.e., all other TripFunctions),
has been shown to be acceptable (Refs. 5 and 6) to permitrestoration of any inoperable channel to operable status. This out of servicetime is only acceptable provided the associated Trip Function is stillmaintaining isolation capability (refer to next paragraph).
If the inoperable channel cannot be restored to operable status within the allowable out ofservice time, the channel must be placed in the tripped condition per Table3.2.3 Note l.a.l) or l.a.2), as applicable.
Placing the inoperable channel intrip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue.
Alternately, Amendment No. 2-44 76r VYNPSBASES: 3.7 (Cont'd)surveillances such as monthly torus to drywell vacuum breaker tests.Procedurally, when AC-6A is open, AC-6 and AC-7 are closed to preventoverpressurization of the SBGT system or the reactor building
: ductwork, should a LOCA occur. For this and similar analyses performed, aspurious opening of AC-6 or AC-7 (one of the closed containment isolation valves) is not assumed as a failure simultaneous with apostulated LOCA. Analyses demonstrate that for normal plant operation system alignments, including surveillances such as those described above, that SBGT integrity would be maintained if a LOCA waspostulated.
Therefore, during normal plant operations, the 90 hourclock does not apply. Accordingly, opening of the 18 inch atmospheric control isolation valves AC-7A, AC-7B, AC-8 and AC-10 will be limitedto 90 hours per calendar year (except for performance of the subjectvalve stroke time surveillances
-in which case the appropriate corresponding valves are closed to protect equipment should a LOCAoccur). This restriction will apply whenever primary containment integrity is required.
The 90 hour clock will apply anytime purge andvent evolutions can not assure the integrity of the SBGT trains orrelated equipment.
B. and C. Standby Gas Treatment System and Secondary Containment SystemThe secondary containment is designed to minimize any ground levelrelease of radioactive materials which might result from a seriousaccident.
The Reactor Building provides secondary containment duringreactor operation, when the drywell is sealed and in service; theReactor Building provides primary containment when the reactor isshutdown and the drywell is open, as during refueling.
Because thesecondary containment is an integral part of the complete containment system, secondary containment is required at all times that primarycontainment is required except, however, for initial fuel loading andlow power physics testing.In the Cold Shutdown condition or the Refuel Mode,the probability andconsequences of the LOCA are reduced due to the pressure andtemperature limitations in these conditions.
Therefore, maintaining Secondary Containment Integrity is not required in the Cold Shutdowncondition or the Refuel Mode, except for other situations for whichsignificant releases of radioactive material can be postulated, such asduring operations with a potential for draining the reactor vessel,during alteration of the Reactor Core, or during movement of irradiated ln-s:er l fuel assemblies er the fuel cack in the secondary containmen&
recent1y1 n order for secondary containment integrity to be met, the secondary containment must function properly in conjunction with the operation ofthe Standby Gas Treatment System to ensure that the required vacuum canbe established and maintained.
This means that the reactor building isintact with at least one door in each access opening closed, and allreactor building automatic ventilation system isolation valves areoperable or the affected penetration flow path is isolated.
With the reactor in the Run Mode, the Startup Mode, or the Hot Shutdowncondition, if Secondary Containment Integrity is not maintained, Secondary Containment Integrity must be restored within 4 hours. The4 hours provides a period of time to correct the problem that iscommensurate with the importance of maintaining secondary containment during the Run Mode, the Startup Mode, and the Hot Shutdown condition.
This time period also ensures that the probability of an accident(requiring Secondary Containment Integrity) occurring during periodswhere Secondary Containment Integrity is not maintained, is minimal.Amendment No. 4-9, 44--, 4-7, 149-1,1-9-,
165a VYNPSBASES: 3.7 (Cont'd) recentlyIf Secondary ontainment Integrity cannot be restored within therequired tim period, the plant must be brought to a mode or condition in which the rLCO does not apply.Movement of irradiated fuel assemblies zr the fucl eask in thesecondary containment, alteration of the Reactor Core, and operations recently h the potential for draining the reactor vessel can be postulated tocaus fission product release to the secondary containment.
In suchcases, secondary containment is the only barrier to release offission pro cts to the environment.
Alteration of the Reactor Coreand movement o irradiated fuel assemblies
#;nd th fuel eask must beimmediately suspended if Secondary Containment Integrity is notmaintained.
Suspension of these activities shall not precludecompleting an action that involves moving a component to a safeposition.
Also, action must be immediately initiated to suspendoperations with the potential for draining the reactor vessel tominimize the probability of a vessel draindown and subsequent potential for fission product release.
Actions must continue until operations with the potential for draining the reactor vessel are suspended.
Amendment No. 4-4-3, -47, -11165b VYNPSBASES: 3.7 (Cont'd)The Standby Gas Treatment System (SGTS) is designed to filter andexhaust the Reactor Building atmosphere to the stack during secondary containment isolation conditions, with a minimum release of radioactive materials from the Reactor Building to the environs.
To insure thatthe standby gas treatment system will be effective in removingradioactive contaminates from the Reactor Building air, the system istested periodically to meet the intent of ANSI N510-1975.
Laboratory charcoal testing will be performed in accordance with ASTM D3803-1989, except, as allowed by GL 99-02, testing can be performed at 70%relative humidity for systems with humidity control.
Both standby gastreatment fans are designed to automatically start upon containment isolation and to maintain the Reactor Building pressure toapproximately a negative 0.15 inch water gauge pressure; all leakageshould be in-leakage.
Should the fan fail to start, the redundant alternate fan and filter system is designed to start automatically.
Each of the two fans has 100% capacity.
This substantiates theavailability of the operable train and results in no added risk; thus,reactor operation or refueling operation can continue.
If neithertrain is operable, the plant is brought to a condition where the systemis not required.
/-ecently]
When the reactor is ' cold shutdown or refueling the drywell may beopen and the Reactor Building becomes the only containment system.During cold shutdown the probability and consequences of a DBA LOCA aresubstantially reduc due to the pressure and temperature limitations in this mode. Howe er, for other situations under which significant radioactive release can be postulated, such as during operations with apotential for drai.'ng the reactor vessel, during core alterations, orduring movement of irradiated fuel in the secondary containment, operability of standby gas treatment is required.
Both trains of the Standby Gas Treatment System are normally operablewhen secondary containment integrity is required.
However,Specification 3.7.B.3 provides Limiting Conditions for Operation whenone train of the Standby Gas Treatment System is inoperable.
Provisional, continued operation is permitted since the remaining operable train is adequate to perform the required radioactivity release control function.
If the applicable conditions ofSpecification 3.7.B.3 cannot be met, the plant must be placed in a moderecently or condition where the Limiting Conditions for Operation do not apply.E try into a refueling condition with one train of SBGTS inoperable isceptable and there is no prohibition on mode or condition entry inhis situation.
In this case, the requirements of TS 3.7.B.3.b areufficient to ensure that adequate controls are in place. Duringefueling conditions, accident risk is significantly
: reduced, and therrimary activities of concern involve core alterations, movement ofirradiated fuel assemblies7-and OPDRVs.During refueling and cold shutdown conditions Specification 3.7.B.3.b provides for the indefinite continuance of refueling operations withone train of the Standby Gas Treatment System inoperable.
When theseven-day completion time associated with Specification 3.7.B.3.b isnot met and secondary containment integrity is .required, the operabletrain of the Standby Gas Treatment System should immediately be placedinto operation.
This action ensures that the remaining train isoperable, that no failures that could prevent automatic actuation haveoccurred, and that any other failure would be readily detected.
Analternative to placing the operable train of Standby Gas Treatment inoperation is to immediately suspend activities that represent apotential for releasing radioactive material to the secondary containment, thus placing the plant in a condition that minimizes risk.Amendment No. -4, 4-4, 14-3, 189, 1-9-1166 Insert 1"Recently irradiated" fuel is defined as fuel that has occupied part of a critical reactor core withinthe previous 13 days, i.e. reactor fuel that has decayed less than 13 days following reactorshutdown.
This minimum decay period is enforced to maintain the validity of the Fuel HandlingAccident dose consequence analysis.}}

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