ML21076A439

Attachment 1 - Requests for Exemptions from Portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and 10 CFR 50, Appendix E

ML21076A439
Person / Time
Site:	Dresden
Issue date:	03/17/2021
From:	Exelon Generation Co
To:	Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
Shared Package
ML21076A518	List: ML21076A438 ML21076A439 RS-21-023, Attachment 2 - Calculation DRE20-0008, Rev. 0, Zirconium Fire Analysis for Drained Spent Fuel Pool
References
RS-21-023
Download: ML21076A439 (51)

Text

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 37 of 78 10 CFR 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR 50, APPENDIX E Bold strikethrough text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception .

Item Regulation in 10 CFR 50, APPENDIX E, SECTION IV Basis for Exemption iii. Radiological monitoring teams; iv. Fire control teams (fire brigades);

v. Repair and damage control teams; vi. First aid and rescue teams; vii. Medical support personnel; viii . The number of staff at Dresden during the decommissioning process will be viii. liGensee's headquarters support personnel; small but commensurate with the need to safely store spent fuel at the facility ix. Security personnel. in a manner that is protective of public health and safety. Dresden will In addition , a radiolog ical orientation training program shall maintain a level of emergency response that does not require additional be made available to local services personnel; e.g., local response by headquarters personnel. The on-shift and emergency response emergency serviceslCivil Defense, local law enforcement positions are defined in the Permanently Defueled Emergency Plan and will personnel, loGal news media persons. be regularly tested through drills and exercises , audited , and inspected by Exelon and the NRC.

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

Due to the low probability of design basis accidents or other credible events to exceed the EPA PAGs, offsite emergency measures are limited to support provided by local police, fire departments and medical services , as appropriate .

The term "Civil Defense" is no longer a commonly used term and is no longer applicable as an example in the regulation. Local news media personnel no longer need radiological orientation tra ining since they will not be called upon to support the formal Joint Information Center.

Also refer to the basis for 10 CFR 50.47(b) .

38 F.2. The plan shall describe provisions for the conduct of Dresden , Units 2 and 3, analyses (Reference 5) demonstrate that 348 days after emergency preparedness exercises as follows : Exercises permanent cessation of power operations for Dresden , Unit 2, and 299 days after shall test the adequacy of timing and content of permanent cessation of power operations for Dresden , Unit 3, no remaining implementing procedures and methods, test emergency postulated accidents at Dresden will result in radiological releases requ iring offsite equipment and communications networks, test the publiG protective actions, or in the event of beyond design basis accidents, 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> is alert and notifiGation system, and ensure that available to take mitigative actions, and if needed, implement offsite protective

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 38 of 78 10 CFR 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR 50, APPENDIX E Bold strikethrough text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item Regulation in 10 CFR 50, APPENDIX E, SECTION IV Basis for Exemption emergency organization personnel are familiar with their actions using a CEMP. Therefore, the public alert and notification system will not duties.3 be used , and no testing would be required.

A site specific simulator will not be maintained after the permanent cessation of 3 IJse af site speGifiG simulataFs aF GamputeFs is aGGeptable power operations for Dresden, Unit 2 and Unit 3.

faF any eJ1:eFGise. Also refer to the basis for 10 CFR 50.47(b).

39 F.2.a. A full partisipatiaR 4 eMersise 11i.1hish tests as Dresden will continue to conduct biennial exercises and will invite the State of mush af the liseRsee, State, aRd lasal emergeRsy Illinois and local support organizations (firefighting , law enforcement, and plaRs as is reasaRably ashievable withaut maRdatary ambulance/medical services) to participate in the periodic drills and exercises publis partisipatiaR shall be saRdusted far eash site at conducted to assess their ability to perform responsibilities related to an whish a pawer reastar is lasated. Nuslear pawer emergency at Dresden , to the extent defined by the Dresden emergency response reastar liseRsees shall submit exersise sseRarias plan. Because the need for offsite emergency planning is relaxed due to the low uRder § 50.4 at least GO days befare use iR a full probability of the postulated accident or other credible events that would be partisipatiaR exersise required by this paragraph 2.a. expected to result in an offsite radioactive release that would exceed the EPA F.2.a.(i), (ii), and (iii) are not applicable. PAGs and the available time for event mitigation, no formal offsite radiological emergency response plans will be in place to test.

4 The intent of submitting exercise scenarios for use by power reactor licensees is F1:Jll 13aFtiei13atieA wl=leA 1:JseEl iA eeAjl:JAetieA witl=l emeF§eAey to check that licensees utilize different scenarios in order to prevent the 13Fe13aFeElAess exeFeises feF a 13aFtie1:JlaF site meaAs preconditioning of responders at power reactors. For defueled sites, there are a1313rn13Fiate e#site leeal aAEl State a1:JtheFities aAEl lieeAsee 13eFseAAel 131=lysieally aAEl aetively tal~e 13aFt iA testiA§ tl=leiF limited events that could occur and the previously routine progression to General inte§FateEl ea13asility te aEleEJl:lately assess anEl Fes13eREl te aR Emergency in power reactor site scenarios is not applicable to a decommissioning assiEleAt at a semmeFsial Al:lsleaF 13eweF 13laAt. Fl:lll site.

13aFtiei13atieR iAel1:JEles testiA§ majeF e9seFVa91e 13eFtieRs ef tl=le Exelon considers Dresden to be exempt from 10 CFR 50, Appendix E, Section eAsite aAEl effsite emeF§eAsy 13laAs aAEl mesili;catieA ef State, F.2 .a.(i)-(iii) because Dresden will be exempt from the umbrella provision of lesal aAEl liseAsee 13eFseAAel aAEl etheF Fese1:JFses iA s1:JffisieAt 10 CFR 50, Appendix E, Section IV.F.2.a.

Al:JFAseFs te "eFify the sa13asility te Fes13eAEl te the assiEleAt sseAaFie . Also refer to the basis for 10 CFR 50.47(b).

40 F.2 .b. Each licensee at each site shall conduct a The low probability of design basis accidents or other credible events that would subsequent exercise of its onsite emergency plan every result in an offsite radioactive release that would exceed the EPA PAGs and the 2 years. Nuslear pa1,1i.1er reastar liseRsees shall submit available time for event mitigative actions at Dresden during decommissioning exersise sseRarias URder § 50.4 at least GO days befare render the TSC , OSC and EOF unnecessary. The principal functions required by use iR aR exersise required by this paragraph 2.b. +he

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 39 of 78 10 CFR 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR 50, APPENDIX E Bold strikethrough text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item Regulation in 10 CFR 50, APPENDIX E, SECTION IV Basis for Exemption exerGise may be inGluded in the full partiGipation regulation can be performed at an onsite location that does not meet the biennial exerGise required by paragraph 2.G. of this requirements of the TSC, OSC or EOF.

seGtion. In addition, the licensee shall take actions Dresden will continue to conduct biennial exercises and will invite the State of necessary to ensure that adequate emergency response Illinois and local support organizations (firefighting, law enforcement, and capabilities are maintained during the interval between ambulance/medical services) to participate in the periodic drills and exercises biennial exercises by conducting drills, including at least conducted to assess their ability to perform responsibilities related to an one drill involving a combination of some of the principal emergency at Dresden, to the extent defined by the Dresden emergency response functional areas of the licensee's onsite emergency plan. Because the need for offsite emergency planning is relaxed due to the low response capabilities. The principal functional areas of probability of the postulated accident or other credible events that would be emergency response include activities such as expected to result in an offsite radioactive release that would exceed the EPA management and coordination of emergency response, PAGs and the available time for event mitigation, no formal offsite radiological accident assessment, event classification, notification of emergency response plans will be in place to test.

offsite authorities, and assessment of the onsite and offsite impact of radiological releases, proteGtive aGtion The intent of submitting exercise scenarios for use by power reactor licensees is reGommendation development, proteGtive aGtion to check that licensees utilize different scenarios in order to prevent the deGision making, plant system repair and mitigative preconditioning of responders at power reactors. For defueled sites, there are action implementation . During these drills, activation of all limited events that could occur and the previously routine progression to General of the licensee's emergency response facilities (TeGhniGal Emergency in power reactor site scenarios is not applicable to a decommissioning Support Genter (TSG}, Operations Support Genter site.

(OSG}, and the EmergenGy Operations FaGility (EOFH Also refer to the basis for 10 CFR 50.47(b).

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 have controllers intervene, and the drills may focus on the onsite exercise training objectives.

41 F.2.c. Offsite plans for eaGh site shall be exerGised Dresden will continue to invite the State of Illinois and local support organizations biennially with full partiGipation by eaGh offsite to participate in the periodic drills and exercises conducted to assess their ability authority having a role under the radiologiGal to perform responsibilities related to an emergency at Dresden , to the extent response plan. Where the offsite authority has a role defined by the Dresden emergency response plan . Because the need for offsite under a radiologiGal response plan for more than one emergency planning is relaxed due to the low probability of the postulated

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 40 of 78 10 CFR 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR 50, APPENDIX E Bold strikethrough text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item Regulation in 10 CFR 50, APPENDIX E, SECTION IV Basis for Exemption site, it shall fully partisipate iR one eM:ersise e¥el"}' accident or other credible events that would be expected to result in an offsite two years and shall, at least, partially partisipate in radioactive release that would exceed the EPA PAGs and the available time for other offsite plan eM:ersises iR this period. If two event mitigation, no formal offsite radiological emergency response plans will be in different lisensees eash ha¥e lisensed fasilities place to test.

losated either on the same site or on adjasent, Following submittal of the "Certification of Permanent Removal of Fuel from the sontiguous sites, and share most of the elements Reactor Vessel," in accordance with 10 CFR 50.82(a)(1 )(i) and (ii), Dresden, defining so losated lisensees, then eash lisensee Units 2 and 3, will be permanently shutdown units with irradiated fuel stored in the shalk SFPs and ISFSI. In the EP Final Rule (76 FR 72560, November 23, 2011)

(1) Condust an eM:ersise biennially of its onsite (Reference 12), the NRC defined "hostile action" as, in part, an act directed emergensy plan; toward a nuclear power plant or its personnel. This definition is based on the (2) Partisipate quadrennially in an offsite biennial definition of "hostile action" provided in NRC Bulletin 2005-02, "Emergency full or partial partisipatioR eM:ersise; Preparedness and Response Actions for Security-Based Events," dated July 18, 2005 (Reference 13). NRC Bulletin 2005-02 was not applicable to (3) Condust emergensy preparedness asti*1ities and nuclear power reactors that have permanently ceased operations and have interastions in the years between its partisipatioR in certified that fuel has been removed from the reactor vessels.

the offsite full or partial partisipatioR eM:ersise with offsite authorities, to test and maintain interlase The NRC excluded non-power reactors from the definition of "hostile action" at the among the affested State and losal authorities and time of the rulemaking because, as defined in 10 CFR 50.2, a non-power reactor the lisensee. Co losated lisensees shall also is not considered a nuclear power reactor and a regulatory basis had not been partisipate in emergensy preparedness asti*1ities and developed to support the inclusion of NPR in the definition of "hostile action."

interastioR i.vith offsite authorities for the period Similarly, a decommissioning power reactor or ISFSI is not a "nuclear reactor" as between eM:ersises; defined in the NRC's regulations. A decommissioning power reactor also has a low likelihood of a credible accident resulting in radiological releases requiring (4) Condust a hostile astioR eM:ersise of its GRsite offsite protective measures. For all of these reasons, the NRC has concluded that emergensy plan in eash eM:ersise sysle; aRd a decommissioning power reactor is not a facility that falls within the definition of (5) Partisipate in an offsite biennial full or pamat- "hostile action."

partisipation hostile astioR eM:ersise iR alternating Similarly, for security, risk insights can be used to determine which targets are eM:ersise sysles. important to protect against sabotage. A level of security commensurate with the consequences of a sabotage event is required and is evaluated on a site-specific basis. The severity of the consequences declines as fuel ages and, thereby, removes over time the underlying concern that a sabotage attack, under the current definition, could cause offsite radiological consequences.

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 41 of 78 10 CFR 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR 50, APPENDIX E Bold strikethrough text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item Regulation in 10 CFR 50, APPENDIX E, SECTION IV Basis for Exemption Although , this analysis provides a justification for an exemption to include the definition for a "hostile action" and its related requirements, elements for security-based events would be maintained. The classification of security-based events, notification of offsite authorities and coordination with offsite agencies under a CEMP would still be required. Other security-related requirements in the EP Final Rule would be exempted such as, on-shift staffing analysis, ERO augmentation and alternative facilities, protection of onsite personnel, and challenging drills and exercises due to the reduced radiological risk for a decommissioning power reactor.

The following similarities between Dresden and NPRs show that the Dresden facility should be treated in a similar fashion as an NPR. Similar to NPRs, Dresden will pose lower radiological risks to the public from accidents than do power reactors because: 1) Dresden will be a permanently shutdown facility (with fuel stored in the SFPs and ISFSI) and will no longer generate fission products;

2) fuel stored in the Dresden, Units 2 and 3, SFPs will have lower decay heat resulting in lower risk of fission product release in the event of a beyond design basis boil off or drain down event; and 3) no credible or beyond design basis accident at Dresden will result in radiological releases requiring offsite protective actions.

Exelon considers Dresden to be exempt from 10 CFR 50, Appendix E, Section F.2 .a.(i)-(iii) because Dresden will be exempt from the umbrella provision of 10 CFR 50 , Appendix E,Section IV.F.2 .a.

Also refer to the basis for 10 CFR 50 .4 7 (b ).

42 F.2 .d. Eash State with responsibility for nuslear po1..1er Dresden has developed an analysis (Reference 5) indicating that 348 days after reastor emergensy preparedness should fully permanent cessation of power operations for Dresden, Unit 2, and 299 days after pal'tisipate iR the ingestion pathway portion of permanent cessation of power operations for Dresden, Unit 3, no credible or exersises at least onse e¥ery exersise sysle. IA States beyond design basis accident at Dresden, Units 2 and 3, will result in radiological with more than one nuslear power reastor plume releases requiring offsite protective actions. The analysis of the potential exposure path1,yay EP2:, the State should rotate this radiological impact of the postulated accident for Dresden , Units 2 and 3, in a pal'tisipation from site to site. Eash State with permanently defueled condition indicates that any releases beyond the site responsibility for nuslear pov1er reastor emergensy boundary are limited to small fractions of the EPA PAG exposure levels .

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 42 of 78 10 CFR 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR 50, APPENDIX E Bold strikethrough text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception .

Item Regulation in 10 CFR 50, APPENDIX E, SECTION IV Basis for Exemption preparedness should fully partiGipate in a hostile In the unlikely event of an SFP accident, the iodine isotopes which contribute to astien exersise at least ense e¥ef¥ sysle and should an offsite dose from an operating reactor accident are not present, so Kl fully partisipate in ene hostile astien exersise by distribution offsite would no longer serve as an effective or necessary Cesember 3~, 20~ 5. States with mere than ene RUGlear supplemental protective action.

power reaster plume exposure path1,1itay EPil sheuld Because it is not possible for PAGs to be exceeded at Dresden, Units 2 and 3, rotate this partisipatien from site ta site. following the Zirc-Fire Window, evacuation planning is not needed since Dresden will meet the criteria for an exemption from offsite emergency preparedness requirements as discussed in the exemption from 10 CFR 50.47(b).

Also refer to the basis for 10 CFR 50 .4 7 (b ).

43 F.2.e. Licensees shall enable any State or local Dresden has developed an analysis (Reference 5) indicating that 348 days after Government lesated within the plume exposure permanent cessation of power operations for Dresden , Unit 2, and 299 days after pathway EPil to participate in the licensee's drills when permanent cessation of power operations for Dresden , Unit 3, no credible or requested by such State or local Government. beyond design basis accident at Dresden , Units 2 and 3, will result in radiological releases requiring offsite protective actions. The analysis of the potential radiological impact of the postulated accident for Dresden , Units 2 and 3, in a permanently defueled condition indicates that any releases beyond the site boundary are limited to small fractions of the EPA PAG exposure levels.

In the unlikely event of an SFP accident, the iodine isotopes which contribute to an offsite dose from an operating reactor accident are not present, so Kl distribution offsite would no longer serve as an effective or necessary supplemental protective action.

Because it is not possible for PAGs to be exceeded at Dresden , Units 2 and 3, following the Zirc-Fire Window, evacuation planning is not needed since Dresden will meet the criteria for an exemption from offsite emergency preparedness requirements as discussed in the exemption from 10 CFR 50.47(b).

Also refer to the basis for 10 CFR 50.47(b).

44 F.2.f. Remedial exercises will be required if the The Federal Emergency Management Agency (FEMA) is responsible for the emergency plan is not satisfactorily tested during the evaluation of an offsite response exercise. No action is expected from State or biennial exercise , such that NRC,--iR sensultatien with F'.EMA, cannot (1) find reasonable assurance that

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 43 of 78 10 CFR 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR 50, APPENDIX E Bold strikethrough text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception .

Item Regulation in 10 CFR 50, APPENDIX E, SECTION IV Basis for Exemption adequate protective measures can and will be taken in the local government organizations in response to an event at a decommissioning site event of a radiological emergency or (2) determine that the other than firefighting , law enforcement, and ambulance/medical services.

Emergency Response Organization (ERO) has maintained Memoranda of understanding will continue to be in place for those services.

key skills specific to emergency response. +he e:ic:teRt af Offsite response organizations will continue to take actions to protect the health State aRd lasal partisipatiaR iR remedial e:ic:ersises and safety of the public as they would at any other industrial site.

must be suffisieRt ta shaw that appropriate sarrestive measures ha11e beeR takeR regardiRg the elemeRts af the plaR Rat properly tested iR the previous e:ic:ersises.

45 F.2.g and F.2.h No exemption requested.

46 F.2.i. Licensees shall use drill and exercise scenarios that At Dresden there will be limited events that could result in radioactive releases provide reasonable assurance that anticipatory responses that exceed the EPA PAGs and the previously routine progression to General will not result from preconditioning of participants. SuGh Emergency in power reactor site scenarios will not be applicable. Therefore, sseRarias far RUGlear pa,11er reaGtar liGeRsees must Dresden is not expected to demonstrate response to a wide spectrum of events.

iRslude a wide speGtrum af radialagiGal releases aRd Also refer to the basis for 10 CFR 50, Appendix E, Section IV.1 regarding "hostile eveRts, iRsludiRg hostile aGtiaR. Exercise and drill action ."

scenarios as appropriate must emphasize coordination among onsite and offsite response organizations.

47 F.2.j. +he exercises GaRdusted uRder paragraph 2 af Dresden , Units 2 and 3, analyses (Reference 5) demonstrate that 348 days after this seGtiaR by Ruslear pawer reaGtar liGeRsees must permanent cessation of power operations for Dresden , Unit 2, and 299 days after provide the appartuRity far the ERO ta demaRstrate permanent cessation of power operations for Dresden , Unit 3, no remaining prefiGieRGy iR the key skills ReGessary ta implemeRt postulated accidents at Dresden will result in radiological releases requ iring offsite the priRGipal fuRGtieRal areas ef emergeRsy respeRse protective actions, or in the event of beyond design basis accidents, 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> is ideRtified iR paragraph 2.b ef this seGtiaR. EaGh available to take mitigative actions, and if needed, implement offsite protective e:ic:ersise must provide the eppartuRity far the ERO ta actions using a CEMP. Therefore , the public alert and notification system will not demeRstrate key skills speGifiG ta emergeRGy respeRse be used , and no testing would be required.

duties iR the saRtrel ream, +SG, OSG, EOF, aRd jeiRt Following submittal of the "Certification of Permanent Removal of Fuel from the iRfermatieR GeRter. AelelitieRally, iR eaGh 8 GaleRelar Reactor Vessel ," in accordance with 10 CFR 50.82(a)(1 )(i) and (ii), Dresden ,

year e:ic:ersise syGle, Ruslear pewer reaster liseRsees Units 2 and 3, will be permanently shutdown units with irradiated fuel stored in the shall vary the GaRteRt ef sGeRaries eluriRg e:ic:erGises SFPs and ISFSI. In the EP Final Rule (76 FR 72560, November 23, 2011)

GaRdusted uRder paragraph 2 ef this seGtieR ta pre11ide (Reference 12), the NRC defined "hostile action" as, in part, an act directed the eppertuRity fer the ERO ta elemeRstrate prefiGieRGY

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 44 of 78 10 CFR 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR 50, APPENDIX E Bold strikethrough text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item Regulation in 10 CFR 50, APPENDIX E, SECTION IV Basis for Exemption in the key skills neGessary to respond to the folloi.*.iing toward a nuclear power plant or its personnel. This definition is based on the sGenario elements: hostile aGtion direGted at the plant definition of "hostile action" provided in NRC Bulletin 2005-02, "Emergency site, no radiologiGal release or an unplanned minimal Preparedness and Response Actions for Security-Based Events," dated radiologiGal release that does not require publiG July 18, 2005 (Reference 13). NRC Bulletin 2005-02 was not applicable to preteGtive aGtions, an initial GlassifiGation of or rapid nuclear power reactors that have permanently ceased operations and have esGalation to a Site .A.rea EmergenGy or General certified that fuel has been removed from the reactor vessels.

EmergenGy, implementation of strategies, proGedures, The NRC excluded non-power reactors from the definition of "hostile action" at the and guidaRGe under§ 50.~ 55{b}{2}, and integration of time of the rulemaking because, as defined in 10 CFR 50.2, a non-power reactor offsite resourGes with onsite response. l=he liGensee is not considered a nuclear power reactor and a regulatory basis had not been shall maintain a reGord of exerGises GonduGted during developed to support the inclusion of NPR in the definition of "hostile action."

eaGh 8 year exerGise GyGle that deGuments the Gentent Similarly, a decommissioning power reactor or ISFSI is not a "nuclear reactor" as of sGenarios used to Gomply ,.vith the requirements of defined in the NRC's regulations. A decommissioning power reactor also has a this paragraph. EaGh liGensee shall GonduGt a hestile low likelihood of a credible accident resulting in radiological releases requiring aGtion exerGise fol' eaGh of its sites no later than offsite protective measures. For all of these reasons, the NRC has concluded that DeGember 3~, 20~ 5. l=he first 8 year exerGise GyGle for a decommissioning power reactor is not a facility that falls within the definition of a site will begin in the Galendar year in whiGh the first "hostile action ."

hostile aGtion exel'Gise is GORduGted. For a site liGensed undel'~ 0 GFR pal't 52, the first 8 yea!' exel'Gise Although, this analysis provides a justification for an exemption to include the GyGle begins in the Galendar year of the initial exerGise definition for a "hostile action" and its related requirements, elements for security-required by SeGtion IV.F.2.a of this appendix. based events would be maintained. The classification of security-based events, notification of offsite authorities and coordination with offsite agencies under a CEMP would still be required. Other security-related requirements in the EP Final Rule would be exempted such as, on-shift staffing analysis, ERO augmentation and alternative facilities, protection of onsite personnel, and challenging drills and exercises due to the reduced radiological risk for a decommissioning power reactor.

Per SECY-00-0145 (Reference 11 ), after approximately 1 year of spent fuel decay time (and as supported by the SFP analysis), the NRC considers an exception to the offsite EPA PAG standard is justified for a zirconium fire scenario considering the low likelihood of this event together with time available to take mitigative or protective actions between the initiating event and before the onset of a postulated fire . If 10 CFR 50.155(b)(2)-type mitigation measures are successful, releases could only occur during the first several days after the fuel was removed

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 45 of 78 10 CFR 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR 50, APPENDIX E Bold strikethrough text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception .

Item Regulation in 10 CFR 50, APPENDIX E, SECTION IV Basis for Exemption from the reactor. As previously indicated, a Dresden analysis (Reference 5) shows that after the spent fuel has decayed for 348 days for Dresden , Unit 2, and 299 days for Dresden , Unit 3, for beyond design basis events where an SFP is drained, and air cooling is not possible , 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> is available to take mitigative or, if needed , offsite protective actions using a comprehensive approach to emergency planning from the time spent fuel cooling is lost until the hottest fuel assembly reaches a temperature of 900°C. Therefore, offsite emergency response plans for the populace within the plume exposure pathway Emergency Planning Zone are not necessary for permanently defueled nuclear power plants.

Also refer to the basis for 10 CFR 50.47(b).

48 G. Maintaining Emergency Preparedness and No exemptions requested.

H. Recovery 49 I. Onsite Protective Actions During Hostile Action Following submittal of the "Certification of Permanent Removal of Fuel from the By J1me 20, 20~ 2, far RUGlear pa111er reaGtar liGeRsees, Reactor Vessel ," in accordance with 10 CFR 50.82(a)(1 )(i) and (ii) , Dresden ,

a raRge af prateGtive aGtiaRs ta prateGt aRsite Units 2 and 3, will be permanently shutdown units with irrad iated fuel stored in the pel"saRRel duriRg hastile aGtiaR must be develaped ta SFPs and ISFSI. In the EP Final Rule (76 FR 72560, November 23 , 2011) eRsure the GaRtiRued ability af the liGeRsee ta safely (Reference 12), the NRC defined "hostile action" as, in part, an act directed shut davm the reaGtar aRd pel"farm the fuRGtiaRs af the toward a nuclear power plant or its personnel. This definition is based on the liGeRsee's emergeRGy plaR. definition of "hostile action" provided in NRC Bulletin 2005-02 , "Emergency Preparedness and Response Actions for Security-Based Events," dated July 18, 2005 (Reference 13). NRC Bulletin 2005-02 was not applicable to nuclear power reactors that have permanently ceased operations and have certified that fuel has been removed from the reactor vessels.

The NRC excluded non-power reactors from the definition of "hostile action" at the time of the rulemaking because, as defined in 10 CFR 50 .2, a non-power reactor is not considered a nuclear power reactor and a regulatory basis had not been developed to support the inclusion of NPR in the definition of "hostile action ."

Similarly, a decommissioning power reactor or ISFSI is not a "nuclear reactor" as defined in the NRC's regulations. A decommissioning power reactor also has a low likelihood of a credible accident resulting in radiological releases requiring offsite protective measures. For all of these reasons , the NRC has concluded that

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 46 of 78 10 CFR 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR 50, APPENDIX E Bold strikethrough text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item Regulation in 10 CFR 50, APPENDIX E, SECTION IV Basis for Exemption a decommissioning power reactor is not a facility that falls within the definition of "hostile action ."

Similarly, for security, risk insights can be used to determine which targets are important to protect against sabotage. A level of security commensurate with the consequences of a sabotage event is required and is evaluated on a site-specific basis. The severity of the consequences declines as fuel ages and, thereby, removes over time the underlying concern that a sabotage attack, under the current definition, could cause offsite radiological consequences.

Although, this analysis provides a justification for an exemption to include the definition for a "hostile action" and its related requirements, elements for security-based events would be maintained. The classification of security-based events, notification of offsite authorities and coordination with offsite agencies under a CEMP would still be required. Other security-related requirements in the EP Final Rule would be exempted such as, on-shift staffing analysis, ERO augmentation and alternative facilities, protection of onsite personnel, and challenging drills and exercises due to the reduced radiological risk for a decommissioning power reactor.

The following similarities between Dresden and NPRs show that the Dresden facility should be treated in a similar fashion as an NPR Similar to NPRs, Dresden will pose lower radiological risks to the public from accidents than do power reactors because: 1) Dresden will be a permanently shutdown facility (with fuel stored in the SFPs and ISFSI) and will no longer generate fission products;

2) fuel stored in the Dresden, Units 2 and 3, SFPs will have lower decay heat resulting in lower risk of fission product release in the event of a beyond design basis boil off or drain down event; and 3) no credible or beyond design basis accident at Dresden will result in radiological releases requiring offsite protective actions.

NOTE: Appendix E to 10 CFR 50, Section Vl.2 exempts permanently or indefinitely shutdown plants from the requirement to provide hardware to support the Emergency Response Data System (EROS). Therefore, specific exemptions from Appendix E to 10 CFR 50, Sections Vl.1, 3, 4 and 10 CFR 50.72(a)(4) are not required.

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 47 of 78 10 CFR 50, Appendix E

5.0 TECHNICAL EVALUATION

5.1 Accident Analysis Overview 10 CFR 50.82(a)(2) specifies that the 10 CFR 50 license no longer authorizes operation of the reactor or emplacement or retention of fuel in the reactor vessel after submitting 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 Dresden, Units 2 and 3, and the permanent removal of the fuel from the reactor vessels, the postulated accidents involving failure or malfunction of the reactors 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 Dresden, Units 2 and 3, is presented below and are in accordance with NRC NSIR/DPR-ISG-02 (Reference 1).

Section 5.0 of NSIR/DPR-ISG-02 indicates that site-specific analyses should demonstrate that:

(1) the radiological consequences of the remaining applicable postulated DBAs would not exceed the limits of the EPA PAGs at the EAB; (2) in the event of a beyond design basis event resulting in the drain down of the SFP to the point that cooling is not effective, there is at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> (assuming an adiabatic heat-up) from the time that the fuel is no longer being cooled until the hottest fuel assembly reaches 900°C; (3) adequate physical security is in place to assure implementation of security strategies that protect against spent fuel sabotage; and (4) in the unlikely event of a beyond OBA resulting in a loss of all SFP cooling, there is sufficient time to implement pre-planned mitigation measures to provide makeup or spray to the SFP before the onset of a zirconium cladding ignition.

Table 3 contains a listing of seven analyses that are expected to be evaluated by a decommissioning nuclear power reactor licensee requesting exemption of emergency planning requirements . The table also contains a description of how Dresden addresses each of these analyses.

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 48 of 78 10 CFR 50, Appendix E TABLE 3 INTERIM STAFF GUIDANCE-02 COMPARISON Analysis NSIR/DPR-ISG-02 Description Response Applicable design DBAs (i .e., fuel handling As described in the License Amendment Request for proposed changes to the Dresden accident in the spent fuel storage facility , Technical Specifications and License Conditions reflecting the permanently defueled condition waste gas system release , and cask (Reference 22), the applicable remaining design basis accidents (OBA) were: (1) a Fuel handling accident if the cask handling Handling Accident in the SFP and (2) Postulated Liquid Releases Due to Liquid Tank Failure.

system is not licensed as single-failure- The only postulated OBA that will remain applicable to Dresden that could contribute to dose proof) (Indicates that any radiological release upon implementation of the requested exemptions is the fuel handling accident (FHA) in the would not exceed the limits of EPA PAGs at Reactor Building , where the SFPs are located.

EAB); The result of the analysis indicates that the dose at the Exclusion Area Boundary (EAB) would not exceed:

1) 10 CFR 50 .67 Control Room limits and Regulatory Guide 1.183 EAB and Low Population Zone (LPZ) limits 48 days after permanent cessation of power operations , and

2) Emergency Preparedness Zone (EPZ) Alert limits 72 days after permanent cessation of power operations.

The EPZ Alert limits at the EAB is 0.01 rem TEDE or 0.05 rem COE Thyroid .

Exelon will maintain the version of the EPA EPZ Alert limits as specified in the current proposed Dresden Emergency Response Plan.

The consequences of radioactive liquid tank failures will not result in offsite dose exposures in excess of 10 CFR 100 or EPA limits due to administrative controls placed on the allowable levels of radioactive isotopes contained in the tank.

The analyses of the potential radiological impact of accidents while the plant is in a permanently defueled condition indicate that no design basis accident or reasonably conceivable beyond design basis accident will be expected to result in radioactive releases that exceed U.S.

Environmental Protection Agency (EPA) Protective Action Guidelines (PAGs) beyond the site boundary. Exelon will maintain the version of the EPA PAGs as specified in the current and proposed Dresden Emergency Response Plan. (References 2 and 3)

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 49 of 78 10 CFR 50, Appendix E TABLE 3 INTERIM STAFF GUIDANCE-02 COMPARISON Analysis NSIR/DPR-ISG-02 Description Response 2 Complete loss of SFP water inventory with Exelon performed an analysis (Reference 5) that conservatively evaluated the length of time it no heat loss (adiabatic heat-up) takes for uncovered spent fuel assemblies in the SFPs to reach the temperature at which the demonstrating a minimum of 1O hours is zirconium cladding would fail. The analysis concluded that a decay time of 348 days for available before any fuel cladding Dresden, Unit 2, and 299 days for Dresden, Unit 3, after permanent cessation of power temperature reaches 900 degrees Celsius operations is the period that the hottest fuel assembly would reach 900°C in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after the from the time all cooling is lost assemblies have been uncovered .

(Demonstrates sufficient time to mitigate This analysis is described in Section 5.3 and is included in Attachment 2.

events that could lead to a zirconium cladding fire);

3 Loss of SFP water inventory resulting in Exelon performed an analysis for Dresden (Reference 16) to determine the radiological impact of radiation exposure at the EAB and control a postulated complete loss of SFP water. It was determined that the gamma radiation dose rate room; (Indicates that any release is less than at the EAB and the Control Room would be less than the regulatory defined limits at 348 days EPA PAGs at EAB); after shutdown.

This analysis is described in Section 5.4 .

4 Considering the site-specific seismic hazard, Exelon conducted a seismic evaluation in response to an NRG request for information pursuant either an evaluation demonstrating a high to 10 CFR 50.54(f) regarding Recommendation 2.1 of the Near-Term Task Force (NTTF) Review confidence of a low-probability (less than 1 x of Insights from the Fukushima Dai-ichi Accident (Reference 17). The seismic evaluation, 10-5 per year) of seismic failure of the spent including all structures and the SFPs, was prepared and submitted for NRG review fuel storage pool structure or an analysis (Reference 18). The Exelon submittal documents the seismic evaluation in conformance with demonstrating the fuel has decayed NTTF Recommendation 2.1, including the high-confidence-of-low-probability-of-failure (HCLPF) sufficiently that natural air flow in a values and the 1 x 10-5 per year hazard level.

completely drained pool would maintain peak The NRG review of the NTTF submittal, specifically for the SFP Evaluation associated with the cladding temperature below 565 degrees reevaluated seismic hazard implementing NTTF Recommendation 2.1 is documented in NRG Celsius (the point of incipient cladding letter dated November 8, 2016 (Reference 19). The NRG concluded that the assessment was damage) (Indicates that any release is less performed consistent with the NRG-endorsed (Reference 20) SFP Evaluation Guidance Report than EPA PAGs at EAB). (Reference 21) and provided sufficient information, including the SFP integrity evaluation , to meet the SFP Evaluation Guidance (Item 9 in Enclosure 1 of the NRC's 50.54(f) letter), thus supporting SDA #6 of NUREG-1738 (Reference 14).

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 50 of 78 10 CFR 50, Appendix E TABLE 3 INTERIM STAFF GUIDANCE-02 COMPARISON Analysis NSIR/DPR-ISG-02 Description Response 5 The analyses and conclusions described in IDCs and SDAs are addressed in Section 5.5 and Tables 4 and 5.

NUREG-1738 are predicated on the risk reduction measures identified in the study as Industry Decommissioning Commitments (IDC) and Staff Decommissioning Assumptions (SDA), listed in Tables 4.1-1 and 4.1-2 of that document. The staff should ensure that the licensee has addressed these IDCs and SDAs for the decommissioning site if they are storing fuel in an SFP.

6 Verify that the licensee presents a Dresden maintains actions in site procedures to respond and mitigate a loss of SFP cooling and determination that there is sufficient a loss of inventory. DOA 1900-01 , "Loss of Fuel Pool Cooling ," provides direction to address a resources and adequately trained personnel loss of cooling and to implement mitigating actions, including restart/restore of equipment, available on-shift to initiate mitigative actions initiating additional makeup/blowdown paths, and providing additional makeup sources (e.g.,

within the 10-hour minimum time period that condensate transfer, which will continue to be available for makeup as long as fuel is in the will prevent an offsite radiological release SFPs). TSG-3, "Operational Contingency Action Guidelines," provides direction to stop leakage that exceeds the EPA PAGs at the EAB . and to provide additional makeup sources to the SFPs, including external sources (EDMG).

Validation of TSG-3 activities have shown that these sources can be utilized with the planned decommissioning staffing within the 10-hour minimum time period to prevent an offsite radiological release that exceeds the EPA PAGs at the EAB.

The operation and control of the SFP cooling systems and mitigation of a loss of SFP cooling will be addressed in the Certified Fuel Handler and Non-Certified Operator training programs. The shiftly walkdowns of the SFP systems will remain in place while irradiated fuel remains in the SFPs, and appropriate procedure steps will remain to provide guidance on the capability and availability of onsite and offsite inventory makeup sources.

The following rounds are in place :

SFP Systems:

• 1x/shift general area checks of pumps and heat exchangers area

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 51 of 78 10 CFR 50, Appendix E TABLE 3 INTERIM STAFF GUIDANCE-02 COMPARISON Analysis NSIR/DPR-ISG-02 Description Response

• 1x/shift general area checks of refuel floor and SFP area

• 1x/shift SFP level recorded

• 1x/shift SFP temperature recorded Spent Fuel related equipment:

• 1x/hour SFP filter D/P recorded

• 1x/hour SFP filter flow recorded

• 1x/week EDMG and FLEX equipment general equipment/area checks The following procedures are in place and employed to review and ensure capability and availability of onsite and offsite inventory makeup sources:

• DOA 1900-01, "Loss of Fuel Pool Cooling"

• TSG-03, "Operational Contingency Action Guidelines" 0 TSG-03 contains time-validated actions to ensure time available to initiate these sources for various loss of cooling or inventory events (not tracked as Time Critical or Sensitive per the response time program)

• Surveillances in place to ensure capability and availability 0 DOS 0010-50, "Diesel Driven FLEX Equipment Surveillances" 0 DOS 0010-43, "Operations FLEX Equipment Inventory" 0 DOS 0010-38, "B .5.b Pump Operability Surveillance" 0 DOS 0010-36, "Operations Support Equipment Inspection"

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 52 of 78 10 CFR 50, Appendix E TABLE 3 INTERIM STAFF GUIDANCE-02 COMPARISON Analysis NSIR/DPR-ISG-02 Description Response 7 Verify that mitigation strategies are As discussed in the post permanent shutdown FHA analysis (Reference 2) completed for the consistent with that required by the License Amendment Request for proposed changes to the Dresden Technical Specifications and Permanently Defueled Technical Revised License Conditions reflecting the permanently defueled condition (Reference 22), there Specifications or by retained license are no active systems credited as part of the initial conditions of the analysis or as part of the conditions. primary success path for mitigation of the FHA with the unit permanently defueled. Therefore ,

the use of containment and control room emergency ventilation systems are not credited or required in the FHA for reduction of nuclides or a reduction of onsite or offsite doses after 48 days of decay time. Exelon proposes to prohibit movement of irradiated fuel after the submittal of the certification of permanent removal of fuel from the reactor vessels until 48 days after permanent shutdown through the imposition of new proposed License Conditions 2.C.(22) and 3.EE of the Renewed Facility Operating Licenses for Dresden, Unit 2 and 3, respectively, included in the License Amendment Request for proposed changes to the Dresden Technical Specifications and Revised License Conditions reflecting the permanently defueled condition (Reference 22). The FHA analyses for Dresden concludes that 48 days of decay time after the reactors have shutdown and provided the SFP water level requirement of Technical Specifications LCO 3.7.8 is met, the dose consequences are acceptable without relying on active SSCs to remain functional for accident mitigation during and following the event Dresden maintains procedures and strategies for the movement of any necessary portable equipment that will be relied upon for mitigating the loss of SFP water. TSG-3 , "Operational Contingency Action Guidelines," provides direction to stop leakage and to provide additional makeup sources to the pool, including external sources (EDMGs). Validation of TSG-3 activities have shown that these sources can be utilized with the planned decommissioning staffing within the 10-hour minimum time period to prevent an offsite radiological release that exceeds the EPA PAGs at the EAB. These mitigative strategies were developed in response to 10 CFR 50.155(b)(2) and are maintained in accordance with License Conditions 2.C.(18) and 3.AA of the Renewed Facility Operating Licenses for Dresden , Units 2 and 3, respectively. These diverse strategies provide defense-in-depth and ample time to provide makeup water or spray to the SFP prior to the onset of zirconium cladding ignition when considering very low probability beyond design basis events affecting the SFP.

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 53 of 78 10 CFR 50, Appendix E 5.2 Consequences of Design Basis Events 5.2.1. Dresden . Units 2 and 3 As described in the License Amendment Request for proposed changes to the Dresden Technical Specifications and License Conditions reflecting the permanently defueled condition (Reference 22), the applicable remaining design basis accidents were: (1) a Fuel Handling Accident in the SFP and (2) Postulated Liquid Releases Due to Liquid Tank Failure. A new analysis (Reference 2) was performed to determine the dose to personnel in the Control Room ,

exclusion area boundary (EAB), and Low Population Zone (LPZ) following cessation of power operations, after all fuel has been removed from both reactors. The analysis determined the minimum decay time required , without credit for normal or emergency Reactor Building and Control Room ventilation and filtration systems , to meet 10 CFR 50 .67 Control Room limits and Regulatory Guide 1.183 EAB and LPZ limits. All applicable regulatory limits can be met after 48 days of decay.

The FHA analysis shows that the dose at the EAB 365 days after shutdown (with no credit for safety systems) is 7.9272 x 10-5 rem TEDE (Reference 2) . This is less than 1 rem TEDE, which is less than the EPA PAG threshold of 1 rem for recommended evacuation.

The consequences of radioactive liquid tank failures will not result in offsite dose exposures in excess of 10 CFR 100 or EPA limits due to administrative controls placed on the allowable levels of radioactive isotopes contained in the tank.

5.2.2. Dresden. Unit 1 Dresden, Unit 1, was permanently shutdown October 31, 1978. All fuel assemblies have been removed from the Dresden , Unit 1, reactor and SFP and transferred either to the onsite Independent Spent Fuel Storage Installation (ISFSI) or to the GE Morris facility in Illinois.

Although spent fuel has been completely removed from storage in the Unit 1 SFP, Chapter 6 of the Dresden , Unit 1, Dresden Defueled Safety Analysis Report (DSAR) (Reference 24) describes conservative accident analyses retained for future reference . These analyses demonstrate that postulated accidents will not result in offsite dose exposures in excess of 10 CFR 100 or EPA limits.

5.3 Hottest Fuel Assembly Adiabatic Heat-up (Zirconium Fire)

The analysis (Reference 5) is provided in Attachment 2 to compare the conditions for the hottest fuel assembly stored in the Dresden, Units 2 and 3, SFPs to a criterion proposed in SECY-99-168 , "Improving Decommissioning Regulations for Nuclear Power Plants" (Reference 25), 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°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 planning involving the plant is not necessary.

Based on the limiting fuel assembly for decay heat and adiabatic heat-up analysis presented in , at 348 days after permanent cessation of power operations, of Dresden , Unit 2, and 299 days (approximately 9.8 months) after permanent cessation of power operations of Dresden, Unit 3, the time for the hottest fuel assembly to reach 900°C is 1O hours after the assemblies have been uncovered . As stated in NUREG-1738, "Technical Study of Spent Fuel

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 54 of 78 10 CFR 50, Appendix E Pool Accident Risk at Decommissioning Nuclear Power Plants" (February 2001) (Reference 14),

900°C 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 heat-up to occur, there is ample time to respond to any 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.

5.4 Consequences of Beyond Design Basis Events NUREG-0586, "Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities," (Reference 26) Supplement 1, Section 4.3.9, identifies that an SFP drain down event is a beyond design basis event. The premise of the required adiabatic heat-up analysis (Reference 5) was the rapid drain down that would expose the fuel to air cooling . The requirements of the analysis were to determine the decay time required to limit the heat-up to 900°C at 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, which would define the mitigation window and event duration .

The offsite and Control Room radiological impacts of a postulated complete loss of SFP water were assessed in Technical Evaluation 633491, "Byron I Dresden Spent Fuel Pool Drain Down Shine Dose Rate Evaluation , Revision 0," (Reference 16). A loss of water shielding above the fuel could increase the offsite radiation levels because of the gamma rays streaming up out of the SFP being scattered back to a receptor at the site boundary. With a decay of 300 days from shutdown, the dose rate at the EAB would be 14.43 mrem/hr not crediting the shielding from the Fuel Handling Building (FHB) roof. Crediting the FHB roof structure, the dose rate at the EAB would be 0.48 mrem/hr. A 300-day decay time is chosen as it conservatively bounds the Zirc-Fire Window. The resultant dose rates if taken over the 10-hour accident duration would be less than the EPA PAGs and the Site Area Emergency Fraction provided by NEI 99-01, Rev. 6 (Reference 9).

It should be noted that the EPA PAGs were developed to respond to a mobile airborne plume that could transport and deposit radioactive material over a large area. In contrast, the radiation field formed by gamma scatter from a drained SFP would be stationary rather than moving and would not cause transport or deposition of radioactive materials. The extended period required to exceed the EPA PAG limit of 1 rem TEDE would allow sufficient time to develop and implement onsite mitigative actions and provide confidence that additional offsite measures could be taken without planning if efforts to reestablish shielding over the fuel are delayed . The calculated dose rate to the Dresden Control Room is less than 0.2 mrem/hr, even when considering only 1 day of decay time .

5.5 Comparison to NUREG-1738 Industry Decommissioning Commitments and Staff Decommissioning Assumptions Exelon evaluated the industry decommissioning commitments (IDCs) and NRC decommissioning assumptions (SDAs) contained in NUREG-1738 (Reference 14).

NUREG-1738 contains the results of the NRC's evaluation of the potential accident risk in an SFP at decommissioning plants in the United States. 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 performed analyses and sensitivity

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 55 of 78 10 CFR 50, Appendix E studies on evacuation timing to assess the risk significance of relaxed offsite emergency preparedness requirements during decommissioning . The NRC based its sensitivity assessment on the guidance in Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" (Reference 27).

The NUREG-1738 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. 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 earthquakes that 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 significant impact on risk. For comparison, at operating reactors, additional risk-significant accidents for which EP is expected to provide dose savings are on the order of 1x10-5 per year, while for decommissioning facilities, the largest contributor for which EP would provide dose savings is about two orders of magnitude lower (cask drop sequence at 2x10-1 per year).

The Executive Summary in NUREG-1738 states, in part,

[T]he 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 industry decommissioning commitments (IDCs) and staff decommissioning assumptions (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-1738. The tables below show how the Dresden, Units 2 and 3, SFPs meet or compare with each of these IDCs (Table 4) and SDAs (Table 5).

5.6 Consequences of a Beyond-Design Basis Earthquake NUREG-1738 (Reference 14) identifies beyond design basis seismic events as the dominant contributor to events that could result in a loss of SFP coolant that uncovers fuel for plants in the Central and Eastern United States. Additionally, NUREG-1738 identifies a zirconium fire resulting from a substantial loss-of-water inventory from the SFP, as the only postulated scenario at a decommissioning plant that could result in a significant offsite radiological release.

The scenarios that lead to this condition have very low frequencies of occurrence (i.e. , on the order of one to tens of times in a million years) and are considered beyond design basis events

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 56 of 78 10 CFR 50, Appendix E because the SFP and attached systems are designed to prevent a substantial loss of coolant inventory under accident conditions. However, the consequences of such accidents could potentially lead to an offsite radiological dose in excess of the EPA PAGs (Reference 3) at the EAB.

The risk associated with zirconium cladding fire events decreases as the spent fuel ages, the decay time increases, the decay heat decreases, and the short-lived radionuclides decay away.

As the decay time increases, the overall risk of a zirconium cladding fire continues to decrease due to two factors: (1) the amount of time available for preventative actions increases, which reduces the probability that the actions would not be successful; and (2) the increased likelihood that the fuel is able to be cooled by air, which decreases the reliance on actions to prevent a zirconium fire. The results of the research conducted for NUREG-1738 and NUREG-2161, "Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor," dated September 2014 (Reference 28), suggest that, while other radiological consequences can be extensive, a postulated accident scenario leading to an SFP zirconium fire, where the fuel has had significant decay time, will have little potential to cause offsite early fatalities, regardless of the type of offsite response (i.e., formal offsite radiological emergency response plan or CEMP).

The purpose of NUREG-2161 was to determine if accelerated transfer of older, colder spent fuel from the SFP at a reference plant to dry cask storage significantly reduces the risks to public health and safety. The study states that:

[T]his study's results are consistent with earlier research studies' conclusions that spent fuel pools are robust structures that are likely to withstand severe earthquakes without leaking cooling water and potentially uncovering the spent fuel. The study shows the likelihood of a radiological release from the spent fuel after the analyzed severe earthquake at the reference plant to be about one time in 10 million years or lower.

If a leak and radiological release were to occur, this study shows that the individual cancer fatality risk for a member of the public is several orders of magnitude lower than the Commission's Quantitative Health Objective of two in one million (2 x 10- 6/year). For such a radiological release, this study shows public and environmental effects are generally the same or smaller than earlier studies.

Dresden is similar to the reference plant (i.e., a General Electric Type 4 BWR with a Mark I containment) used in the study. Therefore , the risks and consequences of an event involving the SFPs at Dresden, Units 2 and 3, are similar to those in the NUREG-2161 study.

The final off-load into the SFP will be constrained to ensure that the requirements of Exelon procedure NF-AB-309, "BWR Special Nuclear Material and Core Component Move Sheet Development." Attachment 1, Section 3 - Moves Involving Spent Fuel Pool ONLY, are met.

To protect against excessive gamma heating of the SFP walls, one of the following will be completed:

• Place fuel assemblies that have not cooled at least 1.5 years since operating in the reactors no closer than 27 inches from the SFP walls;

• Verify that recently operated fuel to be loaded closer than 27 inches from the SFP walls has cooled 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown and is shielded with spent nuclear fuel all in a single

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 57 of 78 10 CFR 50, Appendix E row between the recently operated fuel and the SFP wall that meets all the following conditions:

o Spent nuclear fuel has cooled a minimum of 1.5 years o Row of spent nuclear fuel adjacent to the SFP wall (e.g., all in row 1 or all in row 2) spans beyond fuel recently operated in the reactor by two rack locations in both directions) o Spent nuclear fuel is at least 1 inch from the wall or greater

• Verify that other site-specific requirements or vendor provided requirements are met.

In addition , spent nuclear fuel is stored in the spent fuel storage rack locations adjacent to the SFP walls subject to the following criteria, if possible:

• Verify the spent nuclear fuel distance from the wall is greater than or equal to 4 inches;

• Verify the spent nuclear fuel has cooled for at least 1.5 years since operating in the reactor; and

• Verify the spent nuclear fuel has bundle average exposure of less than or equal to 40 GWD/MTU .

Storing spent fuel in such a dispersed pattern in the SFPs promotes air coolability of the spent fuel in the unlikely event of a loss of water.

Dresden conducted a seismic evaluation in response to an NRC request for information in accordance with 10 CFR 50.54(f) regarding Recommendation 2.1 of the Near-Term Task Force (NTTF) Review of Insights from the Fukushima Dai-ichi Accident. The seismic evaluation included all structures, including the SFPs, and was prepared and submitted for NRC review (Reference 18). The Exelon submittal documents the seismic evaluation in conformance with NTTF Recommendation 2.1, including the high-confidence-of-low-probability-of-failure (HCLPF) values and the 1 x 1o-s per year hazard level. The NRC review of this evaluation is documented in Reference 19. The NRC concluded that the assessment was performed consistent with the NRG-endorsed (Reference 20) SFP Evaluation Guidance Report (Reference 21) and provided sufficient information, including the SFP integrity evaluation, to meet the SFP Evaluation Guidance (Item 9 in Enclosure 1 of the NRC's 50.54(f) letter), thus supporting SDA #6 of NUREG-1738 (Reference 14).

6.0 CONCLUSION

Based on the analysis and actions described above, Exelon has concluded that the health and safety of the public are protected once Dresden, Units 2 and 3, are in the permanently defueled condition . 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 Dresden.

Based on the above, Dresden has demonstrated that no credible or beyond design basis 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.

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IDC Industry Commitments Response 1 Cask drop analyses will be performed or The main hoist of the Reactor Building Overhead Crane (RBOC) has a 125-ton capacity and is single failure-proof cranes will be in use for designated as a single failure proof crane for 110-ton loads as described in UFSAR Section handling of heavy loads (i.e. , phase II of 9.6.2. SFP cask lifts are made with the main hoist of the RBOC and weigh less than 110-tons.

NUREG-0612 will be implemented). All analyses for handling spent fuel casks relative to the RBOC have been based on the National Lead (NL) 10/24 spent fuel shipping cask which weighs 100 tons and the HI-TRAC 100 transfer cask which weighs less than 100 tons. Fuel cask handling above the 545-foot level of the Reactor Building is considered a "restricted load" and must be performed, to the extent practical, in the RBOC RESTRICTED mode, in which the crane bridge and trolley movement is restricted to ensure the crane remains within a predefined pathway. Limit switches are provided for both bridge and trolley forward and reverse directions of travel. When the limits are reached, the limit switches deenergize the control circuits.

2 Procedures and training of personnel will be The following procedures and training are in place and require actions that will remain in place to in place to ensure that onsite and offsite ensure onsite and offsite resources can be brought to bear during an event until the applicable resources can be brought to bear during an system(s) are no longer required:

event. . Dresden Operating Abnormal (DOA) Procedures:

0 DOA 1900-01 , "Loss of Fuel Pool Cooling" 0 DOA 0800-01, "Spent Fuel Cask Abnormal Conditions" 0 DOA 0010-13, "Security Threat" DOA 0010-18, "Escalated Security Event- Hostile Force Intrusion" 0

Dresden General Abnormal (DGA) Procedures:

0 DGA-12, "Loss of Offsite Power" DGA-22, "Station Blackout" 0

Extensive Damage Mitigation Guidelines (EDMG) Procedures:

0 FSG-06, "FLEX Strategy for Aligning Power to U2(3) 480 Volt Safety Related Busses 28(38) and 29(39)"

0 FSG-10, "FLEX Spent Fuel Pool Make-up" 0 FSG-12, "FLEX Spent Fuel Pool Spray" 0 FSG-30, "FLEX Strategy for Obtaining Alternate Readings"

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IDC Industry Commitments Response

• Technical Support Guidelines (TSG) Procedures:

o TSG-03, "Operational Contingency Action Guidelines"

• Dresden Fuel Procedure (DFP):

o DFP 0850-01, "Slow or Rapid Water Level Loss in Fuel Pool I Reactor Cavity" o DFP 0850-02, "New/Irradiated Fuel Damage"

• Additionally, other DOAs (Tornado/High Winds, Earthquakes, Floods) provide direction to enter any number of the above procedures:

o EP-DR-1000, "Exelon Nuclear Dresden Station Radiological Emergency Plan" These procedures are required by NRC regulations and will be implemented as necessary depending on the type of event Once Dresden is shutdown, defueled, and is beyond the zirconium cladding fire period, the on-shift plant operators, Certified Fuel Handlers and Non-Certified Operators, will continue to be appropriately trained on the various actions needed to provide makeup to the SFPs based on a systematic approach to training. Once Dresden 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 when irradiated fuel is stored in the SFPs. Training will be required in accordance with the proposed changes to Dresden Technical Specifications Section 5.0, Administrative Controls, submitted to the NRC on September 24, 2020 (Reference 29). The Dresden Certified Fuel Handler training and retraining program was submitted to the NRC by letter dated September 24, 2020 (Reference 30).

Emergency Response Plan drills will be conducted to maintain proficiency in response to a plant event as described in the PDEP. The EP implementing procedures will be revised in accordance with 10 CFR 50.54(q).

3 Procedures will be in place to establish The following procedures are in place and required actions to establish communication between communication between onsite and offsite onsite and offsite organizations during severe weather and seismic events will remain in place:

organizations during severe weather and

• Dresden Operating Abnormal (DOA) Procedures:

seismic events.

o DOA 0010-02, "Tornado Warning-Severe Winds" o DOA 0010-03, "Earthquakes"

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IDC Industry Commitments Response o DOA 0010-04, "Floods" o DOA 0010-14, "Loss of Off-Site Telephone Communication Systems" o DOA 0010-S1 , "Key Phone Numbers for DOA 0010 Block Procedures"

• FLEX Strategy Guidelines (FSG):

o FSG-39 , "FLEX Communication Options"

• Corporate Guidance (OP-AA):

o OP-AA-108-111-1001 , "Severe Weather and Natural Disaster Guidelines" These procedures provide direction for additional actions and communications with onsite and offsite stakeholders in the event of loss of normal communication methods. Additionally, the procedures provide direction on how/when to reach out to external support organizations during an event.

If the severity of the event requires entry into the PDEP, communications with onsite and offsite organizations will be directed by the Dresden PDEP and associated procedures.

These procedures are required by NRC regulations and will be utilized as necessary depending on the type of event. Communications are described in the procedures for onsite and offsite communications. These procedures are not listed in the existing Dresden Emergency Response Plan and will not be included in the planned Permanently Defueled Emergency Plan (to be submitted for NRC approval). Therefore, it is not necessary for them to be specifically referenced in the Emergency Plan.

4 An offsite resource plan will be developed Dresden has an offsite resource plan built into the Emergency Response Plan. This plan which will include access to portable pumps provides access to and delivery of additional portable pumps, generators, and personnel should it and emergency power to supplement onsite be required to support the multiple portable pumps and emergency generators that meet resources. The plan would principally Extensive Damage Mitigation Guidelines (EDMG) and FLEX requirements already onsite.

identify organizations or suppliers where Various procedures identify the organizations and suppliers of offsite resources and direct when offsite resources could be obtained in a to contact.

timely manner.

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IDC Industry Commitments Response 5 SFP instrumentation will include readouts Temperature of the SFPs are monitored in the Control Room. With the SFP cooling system and alarms in the control room (or where running, temperatures of the process water are available on Control Room indications and personnel are stationed) for SFP annunciation. If the system were to be lost, operators would be dispatched to monitor local temperature, water level, and area radiation temperature indication.

levels. Water level of the SFPs are monitored in the Control Room through annunciation of abnormal level. The skimmer surge tank is replenished by the SFP. If a leak or loss of water inventory was to occur in an SFP, the skimmer would no longer receive this input and would quickly lower due to the suction of the SFP cooling system, bringing in skimmer surge tank low level alarms.

Operators would then be dispatched to monitor local level from the Refueling Floor locally.

Diverse indications and annunciation are available in the Control Room for area and process radiation monitoring.

However, in the Statement of Considerations (SOCs) for the final rule for Mitigation of Beyond-Design-Basis Events (80 FR 70610; November 13, 2015) (Reference 7) and in the associated Final Rule (84 FR 39684; August 9, 2019) (Reference 8), once the certifications of permanent removal of fuel from the reactor vessels have been submitted in accordance with 10 CFR 50.82(a)(1 ), the licensee does not need to comply with the requirement in 10 CFR 50.155(e) to provide reliable means to remotely monitor wide-range SFP levels.

6 SFP seals that could cause leakage leading Each Dresden SFP is contained in the Reactor Building and is connected to the reactor vessel to fuel uncovery in the event of seal failure cavity and steam separator/dryer pit via a refueling transfer canal. To avoid unintentional shall be self-limiting to leakage or otherwise draining of the SFP, there are no penetrations that would permit the SFP to be drained below the engineered so that drainage cannot occur. top of the SFP racks (elevation 589'). The top of fuel is an additional 13 inches lower at elevation 587'-11 ". There are no seals in the SFP that would be subject to leakage other than the gaskets used on the SFP gates.

The lowest penetration in each SFP is the refueling transfer canal that is protected by double-sealed gates. Installed between the gates is a drain line with a flow indicating switch to monitor any leakage past the SFP gates. This arrangement permits detection of leaks and repair of the gates in the event of such leakage. Any leakage past a gate seal is anticipated to be well within the capacity of the makeup sources.

The SFP gates are static, and there is no credible catastrophic failure mechanism for these double-sealed gates. If SFP inventory were to leak due to gate seal rupture or degradation, SFP water level would not go below the top of the spent fuel racks. The elevation of the bottom of the

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IDC Industry Commitments Response refueling transfer canal between the SFP and the reactor vessel cavity is at elevation 589'-2W',

which is above the elevation of the top of the SFP racks and stored fuel. Therefore, leakage past a gate seal could not lead to fuel uncovery.

Failure of the SFP cooling pump seals will not cause any drain-down of an SFP.

7 Procedures or administrative controls to Administrative controls are in place to protect and maintain SFP inventory. This includes controls reduce the likelihood of rapid drain down to ensure movement of heavy loads both over the SFPs and in the Reactor Building follow events will include (1) prohibitions on the prescribed pathways to prevent damage to the SFPs and supporting systems. Procedural use of pumps that lack adequate siphon controls are in place to prevent inadvertent damage to the SFPs from fuel , cask, or equipment protection or (2) controls for pump suction mishandling. Additionally, the ISFSI equipment design is such that there are no ISFSI related and discharge points. The functionality of SFP operations that have the potential to cause a rapid drain down.

anti-siphon devices will be periodically Procedure HU-AA-104-101 , "Procedure Use and Adherence ," establishes the expectations and verified. requirements for procedure adherence and usage for all personnel performing 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 main hoist of the Reactor Building Overhead Crane has a 125-ton capacity and is designated as a single failure proof crane for 110-ton loads as described in UFSAR Section 9.6.2 . As discussed in IDC-04, in addition to the diverse strategies to address lowering SFP level with onsite equipment, Dresden has an offsite resource plan built into the Emergency Response Plan.

The SFP cooling system is incapable of draining or pumping out SFP water based on its design.

Water from the SFP overflows to the skimmer surge tanks. This then supplies the SFP cooling system , and as discussed in Table 4 IDC-05, is alarmed to alert the operators of abnormal level.

The two return lines have openings in the pipe about 6 inches below the SFP surface to break any potential siphon .

8 An onsite restoration plan will be in place to An onsite restoration plan is in place in Dresden procedures to provide repair of the SFP cooling provide repair of the SFP cooling systems systems and/or to provide access for makeup water to the SFPs, including makeup without or to provide access for makeup water to requiring entry to the refuel floor.

the SFP. The plan will provide for remote alignment of the makeup source to the SFP without requiring entry to the refuel floor.

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IDC Industry Commitments Response TSG-03 , "Operational Contingency Action Guidelines," provides a number of methods by which to provide alternate SFP cooling and makeup water to the SFPs. The FP system provides 500 gpm, and a EDMG portable (B.5.b) pump provides 500 gpm.

DOP 1900-01 , "Fuel Pool Cooling and Cleanup System Startup," and DOA 1900-01 , "Loss of Fuel Pool Cooling," provide direction for SFP feed and bleed from multiple locations, including shutdown cooling, Radwaste, and other various Reactor Building locations.

9 Procedures will be in place to control SFP WC-DC-100 , "Decommissioning Work Control Process," dictates the review and approval of work operations that have the potential to rapidly conducted while in decommissioning. This procedure directs performance of integrated risk decrease SFP inventory. These assessment per OP-DC-104, "Decommissioning Integrated Risk Management," that provides for administrative controls may require evaluation of potential operational risk.

additional operations or management Heavy loads are controlled through MA-AA-716-022 , "Control of Heavy Loads Program." Fuel review, management physical presence for moves and heavy load moves that could affect the safe handling and storage of irradiated fuel designated operations or administrative require approval by the Shift Manager.

limitations such as restrictions on heavy load movements. Additionally, the ISFSI transfer equipment is designed such that there is no ISFSI related SFP operations that have the potential to cause a rapid drain down of the SFP.

10 Routine testing of the alternative fuel pool Maintenance and surveillance strategies are in place for alternative SFP makeup, and system makeup system components will be components, and administrative controls are in place for response equipment.

performed and administrative controls for Dresden has multiple systems and sources to provide alternate makeup to the SFPs. The FP equipment out of service will be system provides 500 gpm, and a EDMG portable (B.5.b) pump provides 500 gpm. Additionally, implemented to provide added assurance the FLEX/SAWA submersible pump may be used to pump water from the ultimate heat sink to that the components would be available, if the FLEX I SAWA manifold and directed to the SFPs.

needed .

The fire equipment, EDMG equipment, and supporting equipment provides defense-in-depth and have testing and out of service requirements controlled by their program procedures and administrative program (clearance and tagging , risk assessment, work control) processes.

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SDA Staff Assumptions Response 1 Licensee's SFP cooling design will be at The Dresden SFP cooling system design has two independent trains of SFP cooling. Each train least as capable as that assumed in the risk of SFP cooling rejects heat to the Reactor Building Closed Cooling Water System, which in turn assessment, including instrumentation. rejects its heat to the Service Water system, which then goes to the Illinois River, evaporation, or Licensees will have at least one motor- conduction into the ground depending on the canal configuration. SFP Cooling and support driven and one diesel-driven fire pump systems will remain in place as designed for each unit's pool has fuel in it. Shutdown Cooling capable of delivering inventory to the SFP. System trains (for Fuel Pool Assist augmented cooling) will remain in standby or in operation until a single Fuel Pool Cooling (FPC) train is able to cool its respective pool and has an operational spare train of FPC to meet redundancy requirements.

SFP Cooling instrumentation and instrumentation on support systems will remain in place as designed while those systems are in service to support fuel in the SFPs. The installed instrumentation is diverse and provides high level of redundancy for operators to continue to monitor those systems. The discussion provided for SDA-3 provides more elaboration on instrumentation.

Normal makeup to the SFPs will continue to provide for evaporation losses by reclaimed water.

To provide makeup to address abnormal loss in the SFPs, there are multiple means available .

The primary method would be to use Fire Protection (FP) system water to provide makeup via hoses to the SFPs. The FP System includes two diesel fire pumps (U1 and U2/3) that both take suction from the Kankakee River. The FP system provides 500 gpm, and a EDMG portable (B .5.b) pump provides 500 gpm. The fact that Dresden has two diesels (versus one electric and one diesel) is of no consequence. Dresden does not have any common-mode-failure scenarios that would make two diesels less reliable than one electric and one diesel.

The Dresden design associated with level instrumentation, cooling and makeup meets and exceeds the capabilities assumed in the risk assessment (NUREG 1738, Section 3.2, "Characteristics of SFP Design and Operations for a Decommissioning Plant").

2 Walk-downs of SFP systems will be The operation and control of the SFP cooling systems and mitigation of a loss of SFP cooling will performed at least once per shift by the be addressed in the Certified Fuel Handler and Non-Certified Operator training programs. The operators . Procedures will be developed shiftly walkdowns of the SFP systems will remain in place while irradiated fuel remains in the for and employed by the operators to SFPs, and appropriate procedure steps will remain to provide guidance on the capability and provide guidance on the capability and availability of onsite and offsite inventory makeup sources.

availability of onsite and offsite inventory The following rounds are in place:

makeup sources and time available to

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SDA Staff Assumptions Response initiate these sources for various loss of SFP Systems:

cooling or inventory events.

• 1x/shift general area checks of pumps and heat exchangers area

• 1x/shift general area checks of refuel floor and SFP area

• 1x/shift SFP level recorded

• 1x/shift SFP temperature recorded Spent Fuel related equipment:

• 1x/hour SFP filter D/P recorded

• 1x/hour SFP filter flow recorded

• 1x/week EDMG and FLEX equipment general equipment/area checks The following procedures are in place and employed to review and ensure capability and availability of onsite and offsite inventory makeup sources :

• DOA 1900-01 , "Loss of Fuel Pool Cooling"

• TSG-03, "Operational Contingency Action Guidelines" 0 TSG-03 contains time-validated actions to ensure time available to initiate these sources for various loss of cooling or inventory events (not tracked as Time Critical or Sensitive per the response time program)

• Surveillances in place to ensure capability and availability 0 DOS 0010-50 , "Diesel Driven FLEX Equipment Surveillances" 0 DOS 0010-43, "Operations FLEX Equipment Inventory" 0 DOS 0010-38, "B .5.b Pump Operability Surveillance" 0 DOS 0010-36 , "Operations Support Equipment Inspection" 3 Control room instrumentation that monitors Dresden design meets the intent of this SDA. SFP levels are monitored in the Main Control Room SFP temperature and water level will by redundant low-level alarm switches LS 2(3)-1901-104 & -105 in the skimmer surge tank.

directly measure the parameters involved. Additionally, the SFPs are marked for visual confirmation of level.

Level instrumentation will provide alarms at A separate alarm provides alarm function for high SFP levels.

levels associated with calling in offsite

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SDA Staff Assumptions Response resources and with declaring a general Temperature of the SFPs are monitored on a recorder and alarm by redundant temperature emergency. elements TE 2(3)-1901-110 &-117 in the Control Room . Additionally, the SFPs have a local temperature indicator for visual confirmation of SFP temperature.

Alarm setpoints are set with adequate margin to allow response and establishment of mitigating actions prior to risk of fuel damage. Processes and procedures in place to ensure offsite resources are notified at appropriate thresholds, including emergency classifications.

Regarding the declaration of a General Emergency, the results of the dose calculations for both the FHA and the beyond design basis event of a total loss of water inventory in the SFP do not approach the Protective Action Guidelines to support a classification of greater than an Alert.

Dresden will be employing Permanently Defueled 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. Consistent with the NEI 99-01 Permanently Defueled EALs scheme, it is expected that station conditions will not have the capacity to reach any threshold requiring the declaration of a Site Area Emergency nor General Emergency.

However, in the Statement of Considerations (SOCs) for the final rule for Mitigation of Beyond-Design-Basis Events (80 FR 70610; November 13, 2015) (Reference 7) and in the associated Final Rule (84 FR 39684; August 9, 2019) (Reference 8), once the certifications of permanent removal of fuel from the reactor vessels have been submitted in accordance with 10 CFR 50.82(a)(1 ), the licensee does not need to comply with the requirement in 10 CFR 50 .155(e) to provide reliable means to remotely monitor wide-range SFP levels.

4 Licensee determines that there are no drain The Dresden SFPs have a number of design considerations to provide adequate protection paths in the SFP that could lower the pool against the loss of water from the SFPs. Each SFP is a reinforced concrete structure, completely level (by draining, suction, or pumping) lined with seam-welded 3/16 -inch stainless steel plates. In the unlikely event the concrete and more than 15 feet below the normal pool liner develop cracks, a drainage trough directs leakage to a drain system with monitored flow operating level and that licensee must glasses. This system can handle 100 gpm, which is greater than the anticipated amount of initiate recovery using offsite sources. leakage from cracks . To avoid unintentional draining of an SFP, there are no penetrations that would permit the SFP to be drained below the top of the SFP racks (elevation 589'). The top of fuel is an additional 13 inches lower at elevation 587'-11 ". The lowest penetration is the refueling transfer canal (589'-2% elevation) that is protected by double-sealed gates and has a monitored drain between the gates for leakage. Any leakage past a gate seal is anticipated to be well within the capacity of the makeup sources.

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SDA Staff Assumptions Response The SFP cooling system is incapable of draining or pumping out SFP water based on its design.

Water from the SFP overflows to the skimmer surge tanks. This then supplies the SFP cooling system, and as discussed in Table 4 IDC-05, is alarmed to alert the operators of abnormal level.

Makeup to the SFP , both normal and alternate, provide over double(> 200 gpm) the anticipated amount of leakage from cracks. The normal depth of water in the SFP is 37'-9" (elevation 612'),

and the depth of water in the transfer canal during refueling is 22'-9". Top of fuel installed in the SFP storage rack is at elevation 587'-11 ". The water in the SFP returns to the SFP cooling system via a skimmer weir that can be set to maintain SFP level as low as elevation 611 '-4" .

There are no lower elevation piping penetrations in the SFP than the transfer canal. The bottom of the fuel transfer gate connecting the SFP to the reactor cavity is at elevation 589'-2W'. The SFP cooling return is normally supplied through two 6" pipes that discharge into the SFP via diffusion holes 6 inches below the SFP surface of the water to break any potential siphon .

The SFP lowest drain path is via the 3-inch drain line located between the inboard and outboard SFP gates. As previously discussed in IDC-06, the passage between the SFP and the refueling cavity above the reactor vessel is provided with two double sealed gates with a monitored drain between the gates. This arrangement permits detection of leaks from the passage and repair of the gates in the event of such leakage. If SFP inventory were to leak due to seal rupture or degradation , the level would not go below the top of the SFP fuel racks.

Administrative controls are in place to protect and maintain the SFP inventory. This includes controls to ensure movement of heavy loads both over the SFPs and in the Reactor Building follow prescribed pathways to prevent damage to the SFP and supporting systems. Procedural controls are in place to prevent inadvertent damage to the SFPs from fuel, cask, or equipment mishandling. The main hoist of the Reactor Building Overhead Crane has a 125-ton capacity and is designated as a single failure proof crane for 110-ton loads as described in UFSAR Section 9.6.2 . As discussed in IDC-04, in addition to the diverse strategies to address lowering SFP level with onsite equipment, Dresden has an offsite resource plan built into the Emergency Response Plan.

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SDA Staff Assumptions Response 5 Load Drop consequence analyses will be The Dresden design is in alignment with this description.

performed for facilities with non-single The main hoist of the Reactor Building Overhead Crane (RBOC) has a 125-ton capacity and is failure-proof systems. The analyses and designated as a single failure proof crane for 110-ton loads as described in UFSAR Section 9.6.2.

any mitigative actions necessary to SFP cask lifts are made with the main hoist of the RBOC and weigh less than 110-tons. All preclude catastrophic damage to the SFP analyses for handling spent fuel casks relative to the RBOC have been based on the National that would lead to a rapid pool draining Lead (NL) 10/24 spent fuel shipping cask which weighs 100 tons and the HI-TRAC 100 transfer would be sufficient to demonstrate that cask which weighs less than 100 tons . Fuel cask handling above the 545-foot level of the Reactor there is high confidence in the facilities Building is considered a "restricted load" and must be performed, to the extent practical, in the ability to withstand a heavy load drop. RBOC RESTRICTED mode, in which the crane bridge and trolley movement is restricted to ensure the crane remains within a predefined pathway. Limit switches are provided for both bridge and trolley forward and reverse directions of travel. When the limits are reached, the limit switches deenergize the control circuits.

6 Each decommissioning plant will Dresden has performed a plant specific seismic risk assessment of the SFP and demonstrated successfully complete the seismic checklist that SFP seismically induced structural failure and rapid loss of inventory is less than the generic provided in Appendix 2B to this study bounding estimates provided in NUREG-1738 (<1 x10- 5 per year including non-seismic events).

[NUREG-1738]. If the checklist cannot be Exelon conducted a seismic evaluation in response to an NRC request for information pursuant to successfully completed, the 10 CFR 50.54(f) regarding Recommendation 2.1 of the Near-Term Task Force (NTTF) Review of decommissioning plant will perform a plant Insights from the Fukushima Dai-ichi Accident (Reference 17). The seismic evaluation, including specific seismic risk assessment of the SFP all structures and the SFPs, was prepared and submitted for NRC review (Reference 18).

and demonstrate that SFP seismically induced structural failure and rapid loss of The Exelon submittal documents the seismic evaluation in conformance with NTTF inventory is less than the generic bounding Recommendation 2.1, including the high-confidence-of-low-probability-of-failure (HCLPF) values estimates provided in this study (<1 x10- 5 and the 1 x 10-5 per year hazard level.

per year including non-seismic events). The NRC review of the NTTF submittal, specifically for the SFP Evaluation associated with the reevaluated seismic hazard implementing NTTF Recommendation 2.1 is documented in NRC letter dated November 8, 2016 (Reference 19). The NRC concluded that the assessment was performed consistent with the NRG-endorsed (Reference 20) SFP Evaluation Guidance Report (Reference 21) and provided sufficient information, including the SFP integrity evaluation, to meet the SFP Evaluation Guidance (Item 9 in Enclosure 1 of the NRC's 50 .54(f) letter), thus supporting SDA #6 of NUREG-1738 (Reference 14).

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SDA Staff Assumptions Response 7 Licensees will maintain a program to Baral is the neutron absorber material used in the high density spent fuel storage racks at provide surveillance and monitoring of Dresden. Baral , which is manufactured by Brooks and Perkins Incorporated, is approximately Boraflex in high-density spent fuel racks 48% by weight of boron carbide particles embedded in a matrix of 1100 aluminum .

until such time as spent fuel is no longer This boron carbide-aluminum material is formed into a plate, clad with 1100 aluminum on both stored in these high-density racks. sides. Periodic testing and destructive examination of the boral specimens is done in accordance with plant procedures and determine whether the predicted behavior is occurring in the specimens. The behavior of these specimens can then be used to judge the performance of storage racks. This surveillance is currently performed every 10 years with the next surveillance due October 4, 2021 , and the last performed October 5, 2011. This is an aging management commitment required by License Conditions 2.C.(21) and 3.DD of the Renewed Facility Operating Licenses for Dresden, Units 2 and 3, respectively, and controlled by plant procedures.

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 70 of 78 10 CFR 50, Appendix E 7.0 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 10 CFR 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.

7.1 Exemptions A. The exemptions are authorized by law 10 CFR 50 .12 allows the NRC to grant exemptions from the requirements of 10 CFR 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 response plans.

The requested exemptions and justification for each are based on and consistent with Interim Staff Guidance NSIR/DPR-ISG-02, "Emergency Planning Exemption Requests for Decommissioning Nuclear Power Plants," (Reference 1).

As discussed in this request, revised radiological analyses have been developed that show that, 365 days after shutdown, the radiological consequences of design basis accidents will not exceed the limits of the Environmental Protection Agency (EPA)

Protective Action Guides (PAGs) at the exclusion area boundary (EAB). In addition, analyses have been developed for beyond design basis events related to the spent fuel pool (SFP) which show that, 348 days after permanent cessation of power operation of Dresden, Unit 2 (i.e., the most limiting unit), and 299 days after permanent cessation of power operations of Dresden, Unit 3, 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 PAGs at the EAB.

Additionally, the offsite and Control Room radiological impacts of a postulated complete loss of SFP water were assessed. It was determined that the gamma radiation dose rate at the EAB would be limited to small fractions of the EPA PAG exposure levels and the dose rate in the Control Room will be less than 0.2 mRem/hr.

For these reasons, 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 response plan may be accordingly reduced without an

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 71 of 78 10 CFR 50, Appendix E 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 of requirements in the emergency response plan. These reductions will not adversely affect Dresden's ability to physically secure the site or protect special nuclear material.

Physical security measures at Dresden are not affected by the requested exemption.

Therefore, the proposed exemptions are consistent with the common defense and security.

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

Special circumstances will exist at Dresden 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.

A. 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 response 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 Dresden 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

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 72 of 78 10 CFR 50, Appendix E 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 as described in this request.

Dresden has an analysis (Reference 2) that demonstrates that 365 days after shutdown, the radiological consequences of design basis accidents will not exceed the limits of the EPA PAGs at the EAB. In addition, analyses have been developed for beyond design basis events related to the SFP which show that, 348 days after permanent cessation of power operations of Dresden , Unit 2 (i.e., the most limiting unit), and 299 days after permanent cessation of power operations of Dresden, Unit 3, 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 PAGs at the EAB (Reference 5). Therefore, application of all 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 Dresden 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).

B. 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 resu lt in undue costs being incurred for the maintenance of an emergency response organization (ERO) in excess of that actually needed to respond to the diminished scope of credible events. Other licensees similarly situated , such as Exelon's Three Mile Island, Unit 1, 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.

C. 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 rad iological 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.

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 73 of 78 10 CFR 50, Appendix E The proposed exemptions would allow Dresden to revise the station emergency response 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 ERO 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 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.

8.0 PRECEDENT The exemption requests for 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and 10 CFR 50, Appendix E, requirements are consistent with exemptions for the same emergency planning requirements that recently have been issued by the NRC for Three Mile Island, Unit 1 (Reference 31) and that were submitted for NRC approval by Duane Arnold Energy Center (Reference 32). Similar to the current request for Dresden, these precedents resulted or will result in exemptions from certain emergency planning requirements in 10 CFR 50.47(b); 10 CFR 50.47(c)(2); and 10 CFR 50, Appendix E, related to the elimination of offsite radiological emergency response plans and reduction in the scope of the onsite emergency planning activities.

9.0 ENVIRONMENTAL ASSESSMENT The proposed exemptions meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(25), because the proposed exemptions involve: (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 exemptions are 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 exemptions.

(i) No Significant Hazards Consideration Determination Exelon has evaluated the proposed exemptions 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. Do the proposed exemptions involve a significant 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 its design function. The proposed exemptions would not increase the likelihood of the

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 74 of 78 10 CFR 50, Appendix E 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 Environmental Protection Agency (EPA) Protective Action Guides (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 (FHA) and Postulated Liquid Releases Due to Liquid Tank Failure are unaffected by the proposed exemptions.

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 margin of safety?

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

Therefore, the proposed exemptions do not involve a significant reduction in a margin of safety.

Based on the above, Exelon concludes that the proposed exemptions present 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 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

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 75 of 78 10 CFR 50, Appendix E 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 Dresden ERO commensurate with the reduction in consequences of radiological events that will be possible at Dresden once the facility is in the permanently defueled condition. They also will modify the requirements for emergency planning. Therefore, the exemptions address requirements of an administrative, managerial, or organizational nature.

10.0 REFERENCES

1. NSIR/DPR-ISG-02, Interim Staff Guidance, "Emergency Planning Exemption Requests for Decommissioning Nuclear Power Plants," dated May 11 , 2015 (ADAMS Accession No. ML14106A057)

2. DRE20-0005, "Fuel Handling Accident Dose Consequence (Post Permanent Shutdown)," Revision 0, dated October 9, 2020

3. U.S. Environmental Protection Agency (EPA), EPA-400-R-92-001 , "Manual of Protective Action Guides and Protective Actions Guidelines for Nuclear Incidents," dated May 1992

4. Letter from J. Bradley Fewell (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission , "Certification of Permanent Cessation of Power Operations for Dresden Nuclear Power Station, Units 2 and 3," dated September 2, 2020 (NRC Accession No. ML20246G627)

5. Dresden Calculation DRE20-0008, Revision 0, "Zirconium Fire Analysis for Drained Spent Fuel Pool ," dated January 2, 2021

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 76 of 78 10 CFR 50, Appendix E

6. Federal Register Notice, Vol. 60, No. 120, (60 FR 32430), "Emergency Planning Licensing Requirements for Independent Spent Fuel Storage Facilities (ISFSI) and Monitored Retrievable Storage Facilities (MRS)," dated June 22, 1995

7. Federal Register Notice, Vol. 80, No. 219, (80 FR 70610), "Mitigation of Beyond-Design-Basis Events- Proposed Rule ," dated November 13, 2015

8. Federal Register Notice, Vol. 84, No. 154, (84 FR 39684 ), "Mitigation of Beyond-Design-Basis Events - Final rule," dated August 9, 2019

9. Nuclear Energy Institute (NEI) 99-01, Revision 6, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors," dated November 2012 (ADAMS Accession No. ML12326A805).

10. Letter from M. Thaggard (U .S. Nuclear Regulatory Commission) to S. Perkins-Grew (Nuclear Energy Institute), "U .S. Nuclear Regulatory Commission Review and Endorsement of NEI 99-01, Revision 6, Dated November 2012 (TAC No. D92368),"

dated March 28, 2013 (ADAMS Accession No. ML12346A463) 11 . Commission Paper (SECY) from W. D. Travers to The Commissioners (U.S. Nuclear Regulatory Commission), SECY-00-0145, "Integrated Rulemaking Plan for Nuclear Power Plant Decommissioning," dated June 28, 2000 (ADAMS Accession No. ML003721626)

12. Federal Register Notice, Vol. 76, No. 226 (76 FR 72560), "Enhancements to Emergency Preparedness Regulations," dated November 23, 2011

13. NRC Bulletin 2005-02, "Emergency Preparedness and Response Actions for Security-Based Events," dated July 18, 2005 (ADAMS Accession No. ML051740058)

14. NUREG-1738, "Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants," dated February 2001 (ADAMS Accession No. ML010430066)

15. NUREG-0696, "Functional Criteria for Emergency Response Facilities," dated February 1981 (ADAMS Accession No. ML051390358)

16. Technical Evaluation 633491 , "Byron I Dresden Spent Fuel Pool Drain Down Shine Dose Rate Evaluation , Revision O," dated February 24, 2021

17. Letter from E. J. Leeds (U .S. Nuclear Regulatory Commission) to All Power Reactor Licensees, "Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Review of Insights from The Fukushima Dai-lchi Accident," dated March 12, 2012 (ADAMS Accession No. ML12053A340)

18. Letter from G. T. Kaegi (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Spent Fuel Pool Evaluation Supplemental Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident,"

dated August 31, 2016 (ADAMS Accession No. ML16244A801)

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 77 of 78 10 CFR 50, Appendix E

19. Letter from F. Vega (U.S . Nuclear Regulatory Commission) to B. C. Hanson (Exelon Generation Company, LLC), "Dresden Nuclear Power Station, Units 2 and 3 - Staff Review of Spent Fuel Pool Evaluation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1 (CAC Nos. MF3877 and MF3878)," dated November 8, 2016 (ADAMS Accession No. ML16291A021)

20. Letter from J. R. Davis (U.S. Nuclear Regulatory Commission) to J. E. Pollock (Nuclear Energy Institute), "Endorsement of Electric Power Research Institute Report 3002007148, 'Seismic Evaluation Guidance: Spent Fuel Pool Integrity Evaluation,"'

dated March 17, 2016 (ADAMS Accession No. ML15350A158) 21 . Electric Power Research Institute Report 3002007148, "Seismic Evaluation Guidance:

Spent Fuel Pool Integrity Evaluation," dated February 2016 (ADAMS Accession No. ML16055A021)

22. Letter from P. R. Simpson (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission - "License Amendment Request - Proposed Defueled Technical Specifications and Revised License Conditions for Permanently Defueled Condition," dated October 29, 2020 (ADAMS Accession No. ML20303A313)

23. U.S. Nuclear Regulatory Commission, Regu latory Guide 1.183, Revision 0, "Alternative Radiolog ical Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," dated July 2000 (ADAMS Accession No. ML003716792)

24. Dresden Nuclear Power Station, Unit 1, Defueled Safety Analysis Report, Revision 10, dated June 2020 (ADAMS Accession No. ML20174A543)

25. U.S. Nuclear Regulatory Commission , Commission Paper SECY-99-168, "Improving Decommissioning Regulations for Nuclear Power Plants," dated June 30, 1999 (ADAMS Accession No. ML992800087)

26. U.S. Nuclear Regulatory Commission , NUREG-0586, "Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities," dated October 2002

27. U.S. Nuclear Regulatory Commission, Regu latory Guide 1.174, Revision 2, "An Approach for Using Probabilistic Risk Assessment in Risk Informed Decisions on Plant-Specific Changes to the Licensing Basis," dated May 2011 (ADAMS Accession No. ML100910006)

28. U.S. Nuclear Regulatory Commission, NUREG-2161 , "Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor," dated September 2014 (ADAMS Accession No. ML14255A365)

29. Letter from Patrick R. Simpson (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission , "License Amendment Request - Proposed Changes to Unit 1 Technical Specifications Section 6.1, 'Responsibility,' and Units 2 and 3 Technical Specifications 1.1 , 'Definitions,' and 5.0, 'Administrative Controls,"' for Permanently Defueled Condition," dated September 24, 2020 (ADAMS Accession No. ML20269A404)

30. Letter from Patrick R. Simpson (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission , "Request for Approval of Certified Fuel Handler Training and Retraining Program" for Byron Station , Units 1and 2 and Dresden Nuclear Power Station Units 1, 2, and 3, dated September 24 , 2020 (ADAMS Accession No. ML20269A233)

Request for Exemptions from Portions of Attachment 1 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and Page 78 of 78 10 CFR 50, Appendix E

31. Letter from T. B. Smith (U.S. Nuclear Regulatory Commission) to B. C. Hanson (Exelon Generation Company, LLC), "Three Mile Island Nuclear Station, Units 1 and 2 -

Exemptions from Certain Emergency Planning Requirements and Related Safety Evaluation (EPID L-2019-LLE-0016)," dated December 1, 2020 (ML20244A292 and ML20244A293)

32. Letter from D. Curtland (Duane Arnold Energy Center) to U.S. Nuclear Regulatory Commission, "Request for Exemption from Portions of 10 CFR 50.47 and 10 CFR 50, Appendix E," dated April 2, 2020 (ADAMS Accession No. ML20101M779)

Attachment 2 Dresden Calculation DRE20-0008, Revision 0, "Zirconium Fire Analysis for Drained Spent Fuel Pool" Dresden Nuclear Power Station, Units 1, 2, and 3 Amended Facility Operating License No. DPR-2 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos.50-010, 50-237, 50-249, and 72-037 57 pages follow

CALCULATION NO. DRE20-0008 REV. No. 0 PAGE NO. 1 CC-AA-309-1001-F-01 Revision 0 ATTACHMENT 1 Design Analysis Cover Sheet Design Analysis I Last Page No. 6 57 Analysis No.: 1 DRE20-0008 Revision : 2 0 Major [gl MinorD

Title:

3 Zirconium Fire Analysis for Drained Spent Fuel Pool EC/ECR No.: 4 632551 Revision: 5 0

Station(s): 7 Dresden Component(s): 14 Unit No. : 8 02 and 03 NIA Discipline: 9 NUDC Descrip. Code/Keyword: 10 N02 Safety/QA Class: 11 SR System Code: 12 NIA Structure: 13 NIA CONTROLLED DOCUMENT REFERENCES 15 Document No.: !From/To Document No. : !From/To NIA Is this Design Analysis Safeguards Information? 16 YesD No [gl If yes, see SY-AA-101-106 If yes, Does this Design Analysis contain Unverified Assumptions? 17 YesO No [gl NIA ATl/AR#:

This Design Analysis SUPERCEDES: 18 NIA in its entirety.

Description of Revision (list changed pages when all pages of original analysis were not changed): 19 Original issue.

Preparer: 20 Annie Wong I JT Markland Print Name Sign Name Date Method of Review: 21 Detailed Review [8] Alternate Calculations (attached) D Testing D Reviewer: 22 Chris Staum Print Name Sign Name Date Review Notes: 23 Independent review [8] Peer review D (For External Analyses Only) NIA External Approver: 24 NIA NIA Print Name Sign Name Date Exelon Reviewer: 25 NIA NIA NIA Print Name Sign Name Date Independent 3rd Party Review Reqd? 26 YesD No [gl Exelon Approver: 27 Tamara Stathes Print Name Sign Name Date

CALCULATION NO. DRE20-0008 REV.No. 0 PAGE NO. 2 Revision History Revision Description 0 Initial issue to evaluate the time required before an uncovered fuel bundle in the spent fuel pool would fail its zirconium cladding.

CALCULATION NO. DRE20-0008 REV. No. 0 PAGE NO. 3 Contents Revision History ... .. ... ......... ... ..... .. ....... ... ..... .. ..... .. .......... .. ... .... ... ..... .. ...... ...... ..... ........ .. ....... .. ... ... ......... .... ... .. 2

1.0 Purpose and Scope

.... .......... .. ... .. ..... .. ............... .. ..... .. ... .. .................................. .. .......... ..... .. .............. 4 2.0 Input Data ... .. ............ .. ............. .. ..... .. ............... .. ... ....... .. ..... .. .......... .. ............... .. ... .... ........ .. ..... .. ....... 5 2.1 Zircaloy-2 Properties .... ... .. ..... .......................... ... .. ........................ ... ............ .. ..... ..... ................... .. 5 2.2 Uranium Dioxide Properties ... .. ... .. .. ... .... ... ..... .. ..... ... .. ............ .. ....... ......... ... ... ..... .. ... .... ..... ... ....... .. 5 2.3 Geometry for Lim iting Assembl ies ..... .... .... .... .. ... ....... .. ... .... ..... .... ...... ..... .. ... .. ..... .. .... ...... ..... .. ....... 6 2.4 Heat Load .. ....... ..... .. ..... ....... ... ....... ....... ........ .. ....... ... .. ....... ....... ........ ... .... ..... .. ..... .... ...... .. ....... ... .... 7 3.0 Assumptions ................... .............. .............. ..................................... ....... .......... ................................. 8 4.0 References ......... .... ... .. ... ... .... .. ... ..... ....... ....... ... ......... ... .. .. ... ....... ..... ..... ... .... .. ... ..... ....... .. ... .. ..... ....... 10 5.0 Identification of Computer Programs ................ ................ ... ............... ....... ............ .......... .. ............ 11 6.0 Methodology and Nume ric Analysis .. .... ............ .. ... ....... .. ... ....... ... .. ..... .. ..... .. ... .. ...... ......... .. ... .. ....... 12 7.0 Results and Conclusions ... ....... ... ..... .. ..... ..... ... .... ..... .. .......... .. ....... ... ........ ......... .. ..... .. ... .. ..... ....... ..... 17 Attachment A: ORIG EN-ARP Decay Heat Results Through 3,650 Days ..... ... ... ....... .. ... ........... ... ... .... ..... ... .. 18 Attachment B: ORIGEN-ARP Input for Dresden 2, ATRIUM-10XM Fuel .. ... ... .. .. .. .......... .. ..... ... ... .. .. .. .......... 19 Attachment C: ORIGEN-ARP Input for Dresden 2, Opt ima-2 Fuel .... ...................... .......... .... .......... ..... ....... 22 Attachment D: ORIG EN-ARP Input for Dresden 3, ATRIUM -10XM Fuel ............. ...... ........ ............... ...... ..... 25 Attachment E: Microsoft Excel Calculations and Reference Data ....... .. .......... .. ..... .......... ............ .. ..... .. ..... 28

CALCULATION NO. DRE20-0008 REV. No. 0 PAGE NO. 4

1.0 Purpose and Scope

The purpose of this calculation is to determine the decay time required post-permanent shutdown such that the hottest assembly in the spent fuel pool will not reach the temperature at which zirconium cladding fails within ten hours. This analysis conservatively assumes that there is no air cooling of the assemblies nor heat transfer by conduction, convection, or radiation from the assembly to the other structures in the spent fuel pool and/or environments.

NUREG/CR-6451 (Reference 1} presents several studies discussing the maximum allowable temperature of zirconium cladding that will ensure that failure of the zirconium cladding will not occur. NUREG/CR-6451 states that 565°C (1049°F} is the lowest temperature where incipient cladding fai lure might occur, while NUREG-1738 (Reference 2) states that runaway oxidation (zirconium fire} occurs at 900°(

(1652°F}. For decommissioning purposes, 900°C is the temperature of interest in this calculation.

The results of this analysis are used to support decommissioning of Dresden Station, specifically, when station emergency planning staff may be reduced . This reduction is based on when it can be demonstrated that sufficient decay time has elapsed such that the hottest fuel assembly will take 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> (or more} to reach 900°C. The 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> is based on a criterion identified in Reference 3, Section 5.0, which states: "Complete loss of SFP water inventory with no heat loss (adiabatic heatup}

demonstrating a minimum of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> is available before any fuel cladding temperature reaches 900 degrees Celsius from the time all cooling is lost (Demonstrates sufficient time to mitigate events that could lead to a zirconium cladding fire}."

There are no specific acceptance criteria for this analysis . NUREG-1738 concludes that a 10-hour heat-up time for BWRs requires less than two years of cooling time. This analysis will provide data showing the decay time required to produce a 10-hour heat-up time to 900°C ("runaway oxidation" of zirconium}.

CALCULATION NO. DRE20-0008 REV. No. 0 PAGE NO. 5 2.0 Input Data 2.1 Zircaloy-2 Properties The specific heat and density of Zircaloy-2 were determined using NUREG/CR-6150 {Reference 4) . The density of Zircaloy-2 is 6,551 kg/m 3 {409 lb/ ft 3 ) at 300K. The specific heat values for Zircaloy-2 used in this analysis are adapted from Table 4-2 of Reference 4 and presented in Table 1.

Table 1: Zircaloy-2 Specific Heat Values Temperature Specific Heat Temperature Specific Heat (K) (J/kg-K) (oF) (Btu/I b- °F}

300 281 80.33 0.0671 400 302 260.33 0.0721 640 331 692.33 0.0791 1090 375 1502.33 0.0896 1093 502 1507.73 0.1199 1113 590 1543.73 0.1409 1133 615 1579.73 0.1469 1153 719 1615.73 0.1717 1173 816 1651.73 0.1949 2.2 Uranium Dioxide Properties The specific heat and density of Uranium Dioxide were determined using Reference 5. The density of Uranium Dioxide is 10,673 kg/m 3 {666 lb/ ft 3 ) and is based on linear inter polation of the data in Table 4.3 of Reference 5 at a temperature of 1173K (1652°F). The specific heat va lues for Uranium Dioxide used in this analysis are adapted from Table 4.3 of Reference 5 and presented in Table 2.

Table 2: Uranium Dioxide Specific Heat Values Temperature Specific Heat Temperature Specific Heat (K} (J/kg-K} (oF} (Btu/I b- °F}

300 237 80.33 0.0565 400 264 260.33 0.0631 500 282 440.33 0.0672 600 293 620.33 0.0699 700 300 800.33 0.0718 800 306 980.33 0.0730 900 309 1160.33 0.0738 1000 311 1340.33 0.0744 1100 313 1520.33 0.0748

CALCULATION NO. DRE20-0008 REV. No. 0 PAGE NO. 6 2.3 Geometry for Limiting Assemblies Table 3 and Table 4 reflects the fuel geometries for AREVA ATRIUM lOXM fuel assemblies and Westinghouse Optima-2 fuel assemblies, respectively. The ATRIUM lOXM product is currently loaded in both Dresden Units 2 and 3 (Reference 6, 7}. The Optima-2 product is currently only loaded in Dresden Unit 2 (Reference 6) . Fuel previously discharged to the Spent Fuel Pool has already undergone some decay; therefore, it is reasonably assured those assemblies are not thermally limiting and are excluded from this evaluation.

Table 3: Fuel Assembly Inputs for AREVA ATRIUM lOXM Fuel Number of Full-Length Rods 79 rods Reference 8 Heated Length of Full-Length Rod 145.24 inches Reference 8 Number of Part-Length Rods 12 rods Reference 8 Heated Length of Part-Length Rod 75.00 inches Reference 8 Outer Diameter of Cladding 0.4047 inches Reference 8 Inner Diameter of Cladding 0.3559 inches Reference 8 Uranium Pellet Diameter 0.3492 inches Reference 8 Spacer Weight (sum of 9} 3.59 lb Reference 9 Water Box Outside Width 1.378 inches Reference 8 Water Box Wall Thickness 0.0315 inches Reference 8 Water Box Inside Width 1.315 inches Derived from Outside Dimension and Wa ll Thickness Water Box Length 145.098 inches Reference 10 Channel Thickness (non-corners} 0.075 inches Reference 8 Channel Inside Dimension 5.278 inches Reference 8 Table 4: Fuel Assembly Inputs for Westinghouse Optima-2 Fuel Number of Full-Length Rods 84 rods Reference 11 Heated Length of Full-Length Rod 145.27 inches Reference 11 Number of 2/3-Length Rods 8 rods Reference 11 Heated Length of 2/3-Length Rod 99.61 inches Reference 11 Number of 1/3-Length Rods 4 rods Reference 11 Heated Length of 1/3-Length Rod 50.39 inches Reference 11 Outer Diameter of Cladding 0.3874 inches Reference 11 Inner Diameter of Cladding 0.3398 inches Reference 11 Uranium Pellet Diameter 0.3338 inches Reference 11 Spacer Weight (sum of 32} 1.904 lb Reference 11 Water Cross Inside Width 1.078 inches Reference 11 Water Cross Wall Thickness 0.031 inches Reference 11 Water Cross Outside Width 1.140 inches Derived from Outside Dimension and Wa ll Thickness Channel Thickness (non-corners} 0.055 inches Reference 11 Channel Outside Dimension 5.456 inches Reference 11

CALCULATION NO. DRE20-0008 REV. No. 0 PAGE NO. 7 2.4 Heat Load The limiting decay heat load for a bounding assembly is determined using ORIG EN-ARP. The bounding assembly- for each unit and, in the case of Dresden 2, each fuel type - is established by using the minimum enrichment of any bundle currently operating, the maximum weight of any bundle currently operating, and the maximum exposure of any bundle currently operating (Reference 12). The power on a per assembly basis is the same when considering either unit and is defined as:

Equation 2-1:

2957MWt q = 724 Assem bl ies. = 4.08MW /assembly The operating time, in Effective Full Power Days (EFPD), assumed for the limiting assembly is defined as :

Equation 2-2:

Max .Assembly Exposure (~~i) x Max . Assembly Weight (MTU)

EFPD =

Assembly Power (MW)

Table 5 represents the ORIG EN-ARP inputs used to calculate decay heat for the limiting assembly. Note that for all cases a moderator density of 0.429 g/cc is used and Light Element contributions to decay heat are based on the conservative values presented in NUREG/CR-6999, Table 4.3 (Reference 13).

Table 5: ORIGEN-ARP Inputs for Decay Heat Max. Assembly Minimum Max. Assembly Unit Fuel Type Exposure EFPD Enrichment Weight (MTU}

(GWd[MTU}

ATRIUM 10XM 3.945% 0.176644 52.2259 2258.77 2

Optima-2 3.945% 0.176644 52 .2259 2258 .77 3 ATRIUM 10XM 3.940% 0.177051 48.1176 2085.88

CALCULATION NO. DRE20-0008 REV. No. 0 PAGE NO. 8 3.0 Assumptions 3.1 The heat-up time is assumed to start when the spent fuel pool has been completely drained. This is conservative, as it is likely that site personnel will start to respond to an incident when draindown initiates.

3.2 This analysis conservatively assumes that there is no air cooling of the assemblies (i.e.

adiabatic conditions): the flow paths that would provide natural circulation cooling are assumed to be blocked.

3.3 Gadolinium, which is a burnable poison and can be used in some fuel rods, is conservatively not included in this analysis.

3.4 The specific heat capacity for Uranium Dioxide and Zircaloy-2 are determined at the initial temperature of each time step. Linear interpolation is used to establish the specific heat capacities if the initial temperature at a given time step is between the values presented in Section 2.1 and Section 2.2. An exception to this is used for specific heats above the highest temperature reported in Section 2.1 and Section 2.2. In those instances, the specific heat capacity associated with the highest temperature is utilized. Note that this only occurs for final temperatures above 900°C/1652°F, which exceeds the temperature range of interest for this analysis.

3.5 Some of the assembly materials considered as part of the heat-up are fabricated from Zircaloy-4. In these cases the specific heat and density values used for Zircaloy-2 are also applied to the Zircaloy-4, as any potential differences in the values are negligible.

3.6 The spacer grids are fabricated from Alloy X-750 (Westinghouse Optima-2) and Alloy 718 (AREVA ATRIUM lOXM). For this analysis, they are grouped with the other assembly components that are fabricated from Zircaloy-2. The material properties of these two alloys do vary slightly from that of Zircaloy-2; however, their contribution to the mass of non-fuel components present is relatively low (less than 2.5% by weight) . As such, they are assumed to have the same density and specific heat properties as Zircaloy-2 as any variations would be negligible to the overall calculation .

3.7 The structural components in the fuel assembly are not credited to absorb any of the heat with the exception of the water box/water cross, the spacers, and the portion of the channel parallel to the heated length of the full-length rods .

3.8 The fuel assembly materials in the active fuel region are assumed to start at a uniform temperature and heat up at a uniform rate. While there are temperature gradients throughout the assemblies, these are small relative to the total heat up to 900°C for a zirconium fire to initiate. Furthermore, this simplified approach is still conservative overall since convective heat transfer is not considered, nor is conduction to the upper and lower tie plates or other structural materials, such as the spent fuel pool racks.

3.9 The starting temperature for the heat-up analysis is assumed to be uniform and 150°F (6S.6°C). 150°F is the maximum temperature permitted in the Spent Fuel Pool following a full core offload (Reference 14). NUREG-1738 (Reference 2) sets the starting water temperature at 86°F (30°C); however, the maximum allowable temperature will be used, as this reduces the time to 900°C and ensures the analysis remains bounding.