ML14022A282

Comment (00930) of Michael Keegan on PR-51, Waste Confidence - Continued Storage of Spent Nuclear Fuel

ML14022A282
Person / Time
Site:	Davis Besse, Palisades, Perry, Fermi
Issue date:	12/21/2013
From:	Keegan M J Coalition for a Nuclear-Free Great Lakes
To:	NRC/SECY/RAS
SECY RAS
References
78FR56775 00930, NRC-2012-0246, PR-51
Download: ML14022A282 (63)

Text

1 Rulemaking1CEm Resource From: RulemakingComments Resource Sent: Wednesday, January 22, 2014 4:20 PM To: Rulemaking1CEm Resource Cc: RulemakingComments Resource

Subject:

PR-51 Waste Confidence Attachments:

Comment of Michael Keegan with attachments.pdf DOCKETED BY USNRC-OFFICE OF THE SECRETARY SECY-067 PR#: PR-51 FRN#: 78FR56775 NRC DOCKET#: NRC-2012-0246 SECY DOCKET DATE: 12/20/13 TITLE: Waste Confidence-Continued Storage of Spent Nuclear Fuel COMMENT#: 00930

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PR-51 Waste Confidence Sent Date: 1/22/2014 4:20:13 PM Received Date: 1/22/2014 4:20:37 PM From: RulemakingComments Resource Created By: RulemakingComments.Resource@nrc.gov Recipients: "RulemakingComments Resource" <RulemakingComments.Resource@nrc.gov>

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From: mkeeganj@comcast.net To:RulemakingComments Resource Cc: mkeeganj

Subject:

Comments Nuclear Waste Confidence - Site Specific - Partial Date:Saturday, December 21, 2013 12:03:15 AM Attachments:Site Specific Comments Nuclear Waste Confidence December 20, 2013.docxFermi 2 Spent Fuel Pool Re-rack 11-99.pdfPalisades cask dangle summary report 4406.pdfGenevieve Cook - Connie Kline 1986 Besse document.pdfOne of two with attachments. This one does not contain Davis-Besse 1972 photos.

Dear Rulemaking.Comments@nrc.govWaste Confidence Hearings - Written Comment Docket ID No. NRC-2012-0246Comments of Michael J. Keegan,

Coalition for a Nuclear Free Great Lakes - Site SpecificDavis-Besse, Fermi, Perry, PalisadesSpecifically in this region of the Great Lakes Basin, 20 percent of the world's surface freshwater is in jeopardy from 60 nuclear power plants, 37 of which are directly in the watershed, an accident at any one of which would render 20 percent of the world's precious surface fresh water unusable. And, yet, we go on and do it.Please find attached documents of:Davis-Besse 1972 photos of flooding and 1993 Testimony Fermi 2 Spent Fuel Pool Re-rack and potential boil off at 4.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> into loss of circulation Palisdes Cask Dangle Summary Report - 55 hour6.365741e-4 days <br />0.0153 hours <br />9.093915e-5 weeks <br />2.09275e-5 months <br /> fuel bundle dangle over spent fuel pool

2005 Testimony of Genevieve Cook and Connie Kline on flood concerns at Davis-Besse Comments Nuclear Waste Confidence - Site Specific Michael J. KeeganPlease enter all of these into the record as they are germane to site specific concerns of someGreat Lakes reactors and ISFSIs.Specific to this region andLake Erie, is flooding potential. There is what is known as aseiche where you have straight line winds which blow the water out to Buffalo, and the water sloshes back. One specific concern that applies to Fermi, Davis-Besse, Perry is Lake Erie susceptibility to Seiche / flooding.Lake Michigan and Lake Huron have also had Seiche events.Pasted below are a just a handful of links to Lake Erie Seiche reports.Lake Erie flooding May 31, 2012 - Lake County - Madison, Ohio Lake Erie 1942 - 8 people killed Lake Erie 1882 - Tidal Wavehttp://www.wkyc.com/video/1668105962001/1/3-teenagers-pulled-in-Lake-Erie-by-rare-seiche-wave2003 Seiche in Monroe at Fermi http://www.glerl.noaa.gov/seagrant/glwlphotos/Seiche/1113Storm/November2003.html 1848 Lake Erie Seiche http://oceanservice.noaa.gov/facts/seiche.html http://www.dnr.state.oh.us/OhioGeologicalSurvey/tabid/23637/Default.aspxhttp://downtown.wgrz.com/news/people/71936-lake-erie-seiche-disaster-1844In 1972, the Davis-Besse site was flooded for over a month. If the plant had been operating at that time, it would have been a disaster.Please find attached files of Genevieve Cook testimony in 1986 and then Connie Klinepresentation made in 1993.The aerial photos attached from 1972 flooding beg the question what if HLW was on site /what will be the contingency.To elevate them against flood, is to tee them up for tourists.Connie Kline from Cleveland Ohio notes:"I'm not sure whether you know that the D-B sitewas very badly flooded during construction, but fortunately, prior to fuel loading. I have 1972 professional aerial photos (attached)of the entire plant site submerged, and the flood waters had receded somewhat by the time these photos were taken 3 days after a L. Erie storm caused 2 dikes to break,destroying 300 ft.of dike 3.5 miles up the Toussaint River.

The reactor building was flooded; people had to airlifted from DB & workmen had to reach the site by boat. In the photos, you can see cars submerged in the parking lot, most of which were not salvageable.As you no doubt are aware, DB was built in a marshy wetlands floodplain(in 1968-69, TE traded the small Darby Marsh for the 954 acre Navarre Marsh which is the DB plant site)that has experienced subsequent flooding especially during spring thaws that have made roads leading to and from the plant impassable. I don't know if these photos or some additional documentation I have from 1986-87 would be of any benefit. Let me know, although I should tell you that I am technologically challenged, and I reallydo not want to give up the original photos."Kevin Kamps with Beyond Nuclear has expressed the following concerns, I enter ourworking communications into the record.Kevin Kamps writes:Fukushima scale releases (or order of magnitudes larger) into Lake Erie, from Fermi 2 (or 3) or Davis-Besse, due to a high-level radioactive waste storage pool fire, would be bad, at least for those millions downstream (including in the connecting rivers, Lake Ontario, the St. Lawrence River) who drink its water, boat or swim its waters, eat its fish, etc. Being so shallow, Lake Erie can't really "dilute" the radioactivity, like the Pacific Ocean can This is a site specific risk at Fermi and D-B.Over forevermore, indefinite storage of HLRWs on the Lake Erie shore, the dry casks willerode and release their contents. NRC assumes forevermore institutional control, including once per century complete replacement of the pads, inner canisters, and dry casks -- by using a "Dry Transfer System," DTS, as the pools will be dismantled during decommissioning. No price tag is given for this once per century complete replacement, forevermore, nor where that mysterious amount of money (infinite amount of money, by definition) would supposedly come from. NRC's assumption of forevermore institutional control comes amidst a US federal government shutdown.

As Arjun Makhijani has pointed out, in just the past couple-few hundred years, NorthAmerica has seen multiple wars (the Revolutionary War, the War of 1812 -- including on the Lake Erie shores, where some of the worst battles took place, including in Monroe and Port Clinton -- the Civil War). Institutional control being maintained even just 300 or less years into the future is a huge domain assumption. Please view this link www.stormof1813.comIn addition a 40 foot wave off of Lake Huron, due to the White Hurricane of Nov. 1912 , onehundred years ago.The 1811-1812-1813 New Madrid quakes -- estimated as Magnitude 8 on the Richter scale,the largest in North American history could happen again. These earthquakes created giant waves on the Great Lakes.Even Superstorm Sandy created large waves on Lake Michigan - 30+ footers -- whichfortunately hit Michigan City, IN, and not Covert, MI (Palisades), Bridgman, MI (Cook),

Charlevoix, MI (Big Rock Point, HLRW still there), or Zion, IL (two reactors permanently shutdown, but waste still there). If on Lake MI, why not Lake Erie. With climate destabilization, hurricanes on the Great Lakes could become more common.Then there are tornadoes: water spout near Zion a few weeks back

[1]; Davis-Besse of June 1998; Fermi 2 of June 2010.All risks to unleashing the HLRWs into the environment.Also site-specific: both Fermi 2's Mark I containment, and Davis-Besse's cracked andcracking worse containment, are no containment whatsoever. Look what happened to the Mark Is at Fukushima Daiichi. Davis-Besse's containment could completely fail during a meltdown. These reactor risks are HLRW risks as well -- if the reactors meltdown, that could directly cause the wastes to unleash into the environment, as well. Fukushima Daiichi Unit 4 is such a case in point -- the explosion at the non-operating Unit 4 reactor building could yet lead to a collapse of that reactor building, including the HLRW storage pool, as due to another large quake there. If that were to occur, and the cooling water was lost, then the radiological releases would dwarf what has already occurred to date due to the 3 reactor core meltdowns. Our pools -- as at Fermi 2, as at Davis-Besse -- contain much more waste than Fukushima Daiichi Unit 4.Please review the attached documents which pertain to Spent Fuel Pool concerns at Fermi 2,flooding at Davis-Besse, fuel dangle near disasters at Palisades.Albert Einstein also informs us "To the village square we must carry the facts of atomicenergy. From there must come America's voice."The people have spoken.Stop making it!Cease and desist. Stop making it period. Do not relicense, do not license new ones.

Thank youMichael J. KeeganCoalition for a Nuclear Free Great Lakes P.O. Box 463 Monroe, MI48161 Rulemaking.Comments@nrc.gov Specifically in this region of the Great La kes Basin, 20 percent of the world's surface fresh water is in jeopardy from 60 nuclear power plants, 37 of which are directly in the watershed, an accident at any one of which would render 20 percent of the world's precious surface fresh water unusable. And, yet, we go on and do it.

Please find attached documents of:

Davis-Besse 1972 photos of flooding and 1993 Testimony Fermi 2 Spent Fuel Pool Re-rac k and potential boil off at 4.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> into loss of circulation Palisdes Cas k Dangle Summary Report - 55 hour6.365741e-4 days <br />0.0153 hours <br />9.093915e-5 weeks <br />2.09275e-5 months <br /> fuel bundle dangle over spent fuel pool 2005 Testimony of Genevieve Coo k and Connie Kline on flood concerns at Davis-Besse Comments Nuclear Waste Confidence - Site Specific Michael J. Keegan

Please enter all of these into the record as they are germane to site specific concerns of some

Great La k es reactors and ISFSIs.

Specific to this region and La k e Erie, is flooding potential. There is what is k nown as a seiche where you have straight line winds which blow the water out to Buffalo, and the water sloshes

bac k. One specific concern that applies to Fermi, Davis-Besse, Perry is La k e Erie susceptibility to Seiche / flooding. La k e Michigan and La k e Huron have also had Seiche events. Pasted below are a just a handful of lin k s to La ke Erie Seiche reports.

La k e Erie flooding May 31, 2012 - La k e County - Madison, Ohio

La k e Erie 1942 - 8 people k illed La k e Erie 1882 - Tidal Wave

http://www.w kyc.com/video/1668105962001/1/3-teenagers-pulled-in-La k e-Erie-by-rare-seiche-wave 2003 Seiche in Monroe at Fermi http://www.glerl.noaa.gov/seagrant/glwlphotos/Seiche/1113Storm/November2003.html 1848 La k e Erie Seiche http://oceanservice.noaa.gov/facts/seiche.html http://www.dnr.state.oh.us/OhioGeologicalSurvey/tabid/23637/Default.aspx http://downtown.wgrz.com/news/people/71936-la ke-erie-seiche-disaster-1844

In 1972, the Davis-Besse site was flooded for over a month. If the plant had been operating at that time, it would have been a disaster.

Please find attached files of Genevieve Coo k testimony in 1986 and then Connie Kline presentation made in 1993.

The aerial photos attached from 1972 flooding beg the question what if HLW was on site / what will be the contingency. To elevate them against flood, is to tee them up for tourists.

Connie Kline from Cleveland Ohio notes: "I'm not sure whether you know that the D-B site was very badly flooded during construction, but fortunately, prior to fuel loading. I have 1972 professional aerial photos (attached) of the entire plant site submerged, and the flood waters had receded somewhat by the time these photos were ta ken 3 days after a L. Erie storm caused 2 di k es to brea k , destroying 300 ft.of di ke 3.5 miles up the Toussaint River. The reactor building was flooded; people had to airlifted from DB & wor kmen had to reach the site by boat. In the photos, you can see cars submerged in the par king lot, most of which were not salvageable. As you no doubt are aware, DB was built in a marshy wetlands floodplain (in 1968-69, TE traded the small Darby Marsh for the 954 acre Navarre Marsh which is the DB plant site) that has experienced subsequent flooding especially during spring thaws that have made roads leading to and from the plant impassable. I don't know if these photos or some additional documentation I have from 1986-87 would be of any benefit. Let me know, although I should tell you that I am technologically challenged, and I really do not want to give up the original photos."

Kevin Kamps with Beyond Nuclear has expressed the following concerns, I enter our wor k ing communications into the record. Kevin Kamps writes: Fu kushima scale releases (or order of magnitudes larger) into La ke Erie, from Fermi 2 (or 3) or Davis-Besse, due to a high-level radioactive waste storage pool fire, would be bad, at least for those millions downstream (including in the connecting rivers, La ke Ontario, the St. Lawrence River) who drin k its water, boat or swim its waters, eat its fish, etc. Being so shallow, La ke Erie can't really "dilute" the radioactivity, li ke the Pacific Ocean can This is a site specific ris k at Fermi and D-B.

Over forevermore, indefinite storage of HLRWs on the La k e Erie shore, the dry cas k s will erode and release their contents. NRC assumes forevermore institutional control, including once per century complete replacement of the pads, inner canisters, and dry cas k s -- by using a "Dry Transfer System," DTS, as the pools will be dismantled during decommissioning. No price tag is given for this once per century complete replacement, forevermore, nor where that mysterious amount of money (infinite amount of money, by definition) would supposedly come from.

NRC's assumption of forevermore institutional control comes amidst a US federal government

shutdown.

As Arjun Ma khijani has pointed out, in just the past couple-few hundred years, North America has seen multiple wars (the Revolutionary War, the War of 1812 -- including on the La k e Erie shores, where some of the worst battles too k place, including in Monroe and Port Clinton -- the Civil War). Institutional control being maintained even just 300 or less years into the future is a huge domain assumption.

Please view this lin k www.stormof1813.com In addition a 40 foot wave off of La ke Huron, due to the White Hurricane of Nov. 1912 , one hundred years ago.

The 1811-1812-1813 New Madrid qua kes -- estimated as Magnitude 8 on the Richter scale, the largest in North American history could happen again. These earthqua k es created giant waves on the Great La k es. Even Superstorm Sandy created large waves on La k e Michigan - 30+ footers -- which fortunately hit Michigan City, IN, and not Covert, MI (Palisades), Bridgman, MI (Coo k), Charlevoix, MI (Big Roc k Point, HLRW still there), or Zion, IL (two reactors permanently shutdown, but waste still there). If on La k e MI, why not La ke Erie. With climate destabilization, hurricanes on the Great La kes could become more common.

Then there are tornadoes: water spout near Zion a few wee k s bac k ke-michigan-close-to-zion-hlrw-storage-poo.html; Davis-Besse of June 1998; Fermi 2 of June 2010. All ris k s to unleashing the HLRWs into the environment. Also site-specific: both Fermi 2's Mar k I containment, and Davis-Besse's crac k ed and crac k ing worse containment, are no containment whatsoever. Loo k what happened to the Mar k Is at Fu kushima Daiichi. Davis-Besse's containment could completely fail during a meltdown. These reactor ris ks are HLRW ris ks as well -- if the reactors meltdown, that could directly cause the wastes to unleash into the environment, as well. Fu kushima Daiichi Unit 4 is such a case in point

-- the explosion at the non-operating Unit 4 reactor building could yet lead to a collapse of that reactor building, including the HLRW storage pool, as due to another large qua k e there. If that were to occur, and the cooling water was lost, then the radiological releases would dwarf what has already occurred to date due to the 3 reactor core meltdowns. Our pools -- as at Fermi 2, as at Davis-Besse -- contain much more waste than Fu kushima Daiichi Unit 4. Please review the attached documents which pertain to Spent Fuel Pool concerns at Fermi 2, flooding at Davis-Besse, fuel dangle near disasters at Palisades. Albert Einstein also informs us "To the village square we must carry the facts of atomic energy. From there must come America's voice." The people have spo ken. Stop ma k ing it! Cease and desist. Stop ma k ing it period. Do not relicense, do not license new ones.

Than k you Michael J. Keegan

Coalition for a Nuclear Free Great La k es P.O. Box 463

Monroe, MI 48161 Douglas R.Gipson Senior Vice President, Nuclear Generation Fermi 2 6400 North Dixie Hwy., Newport, Michigan 48166 Tel: 313.586.5201 Fax: 313.586.4172 Detroit Edison 1OCFR50.92 November 19, 1999 NRC-99-0084 U.S.Nuclear Regulatory Commission Attention:

DocumentControlDesk Washington D C 20555-0001

Reference:

Fermi 2 NRC Docket No.50-341 NRC License No.NPF-43

Subject:

Proposed Technical Specification Changes (License Amendment)-Design Features/Fuel Storage (Technical Specification 4.3)and Programs and Manuals/High Density Spent Fuel Racks (Technical Specification 5.5.13)Pursuant to 1OCFR50.90, Detroit Edison hereby proposes to amend the Fermi 2 Plant Operating License NPF-43, AppendixA,Technical Specifications (TS), to change: (1)thedesign features description of the fuel storage equipment and configuration to allow an increase in the spent fuel storage capacity and (2)the description of high density spent fuel racks program to clarify that surveillance program is applicable onlytorackscontaining Boraflex as aneutronabsorber. provides a description and evaluation oftheproposed TS changes.Enclosure 2 provides an analysis of the issue of significant hazards consideration using the standards of IOCFR50.92. provides the markeduppages of the existing TS to show the proposed changes and a typed version of the affected TS pageswiththe proposed changes incorporated. provides a licensing A DTE Energy Company USNRC NRC-99-0084 Page 2 report, which discusses the detailed technical evaluations performedtodemonstrate the acceptability of this license change request.Please notethatEnclosure 4 contains some information thatisconsidered proprietary pursuant to 1OCFR2.790.

An affidavit is also included in this enclosuretoattest to the proprietary nature of the material contained within the licensing report.In this regard, Detroit Edison requests that Enclosure 4 be withheld from public viewing.Enclosure 5 provides a non proprietary version, as required by 1OCFR2.790.

Detroit Edison has reviewed the proposed TS changes against the criteria of 1OCFR51.22 for environmental considerations.

The proposed changesdonot involve a significant hazards consideration, nor significantly change thetypesor significantly increase the amounts of effluents that may be released offsite, nor significantly increase individualorcumulative occupational radiation exposures.

Based on the foregoing, Detroit Edison concludes that the proposed TS changes meetthecriteria provided in 10CFR51.22(c)(9) for a categorical exclusion from the requirements for an Environmental Impact Statement or an Environmental Assessment.

Detroit Edison requests that the NRC approve and issue these changes by January 30, 2001 with an implementation period of within 90 days following NRC approval.Should you have any questions or require additional information,pleasecontact Mr.Norman K.Peterson of my staff at (734)586-4258.Sincerely, Enclosures cc: A.J.Kugler A.Vegel NRC Resident Office Regional Administrator, Region III Supervisor, Electric Operators, Michigan Public Service Commission USNRC NRC-99-0084 Page 3 I, DOUGLAS R.GIPSON, do hereby affirm that the foregoing statements are based on facts and circumstances whicharetrue and accurate to the best of my knowledge and belief.DOUGLASR.soN Senior Vice President,NuclearGeneration On this-day of 7 t/Le.-1999 before me personally appeared Douglas R.Gipson, being first duly sworn and says that he executed the foregoing as his free act and deed.NotaryPublic ROSAUE A.ARMEITA_-.->=:

AFFIDAVIT PURSUANT TO 10CFR2.790 I, Michael P.McNamara, being duly sworn, depose and state as follows: (1)I am the Vice President, Nuclear Projects forHoltecInternational and have been delegatedthefunction of reviewing the information described in paragraph (2)which issoughtto bewithheld,and have been authorized to apply for its withholding.

(2)The information sought to be withheld is contained in the revised pages to the document entitled"Licensing Report for Enrico Fermi 2 Spent Fuel Pool Rack Installation", Holtec Report HI-992154.

(3)In makingthisapplication for withholding of proprietary information of which it is the owner, Holtec International relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec.552(b)(4)and the Trade Secrets Act, 18 USC Sec.1905,andNRC regulations 10CFR Part 9.17(a)(4), 2.790(a)(4), and 2.790(b)(1) for"trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4).Thematerial for which exemption from disclosure is here sought isall"confidential commercial information", and some portions also qualify under the narrower definition of"trade secret", within the meanings assigned to those terms for purposes ofFOIAExemption 4 in, respectively, Critical Mass Energy Project v.Nuclear Regulatory Commission, 975F2d871 (DC Cir.1992), and Public CitizenHealthResearch Group v.FDA, 704F2d1280 (DC Cir.1983).(4)Some examples of categories of informationwhichfit into the definition of proprietary information are: a.Informationthatdiscloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by Holtec's competitors without license from Holtec International constitutes a competitive economic advantage over other companies; 1

AFFIDAVIT PURSUANT TO 10CFR2.790 b.Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.c.Information which reveals cost or price information, production, capacities, budget levels, or commercial strategies of Holtec International, its customers, or its suppliers; d.Information which reveals aspects of past, present, or future Holtec International customer-funded development plans and programs of potential commercial value to Holtec International; e.Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs 4.a, 4.b, 4.d, and 4.e, above.(5)The information sought to be withheld is being submitted to the NRC in confidence.Theinformation (including that compiled from many sources)is of a sort customarily held in confidence by Holtec International, and is in fact so held.The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by Holtec International.

No public disclosure has been made, and it is not availableinpublic sources.All disclosures tothirdparties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.

Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6)and (7)following.

(6)Initial approval of proprietary treatment of a document is made bythemanager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge.

Access to such documents within Holtec International is limited on a"need to 2 AFFIDAVIT PURSUANT TO 10CFR2.790 know" basis.(7)The procedure for approval ofexternalrelease of such a document typically requiresreviewby the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his designee),andby the Legal Operation, for technical content, competitive effect, and determination oftheaccuracy of the proprietary designation.

Disclosures outside Holtec International are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, andotherswith a legitimate need for the information, and then only in accordance with appropriate regulatory provisionsorproprietary agreements.

(8)The information classified as proprietary was developedandcompiled by Holtec International at a significant cost to Holtec International.

This information is classified as proprietary because it contains detailed historical dataandanalytical results not available elsewhere.

Thisinformationwould provideotherparties, including competitors, with information from Holtec International's technical database and the results of evaluations performed using codes developed byHoltecInternational.

Release of this information would improve a competitor's position without the competitor having to expend similar resources for the development of the database.A substantial effort has been expendedbyHoltec International to develop this information.

(9)Public disclosure of the information sought to be withheld is likely to cause substantial harm to Holtec International's competitive positionandforeclose or reduce the availability of profit-making opportunities.

The information is part of Holtec International's comprehensive spent fuelstoragetechnology base, and its commercial value extends beyond the original development cost.Thevalue of the technology base goes beyondtheextensive physical database and analytical methodology, and includes development of the expertise to determine and apply the appropriate evaluation process.The research, development, engineering,andanalytical costs comprise a substantial investment of time and money by Holtec International.

3 AFFIDAVIT PURSUANT TO 10CFR2.790 The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but itclearlyis substantial.

Holtec International'scompetitiveadvantage will be lost if its competitors are able to use the results of the Holtec International experience to normalize or verifytheirown process or if they are able to claim an equivalent understanding bydemonstratingthat theycanarrive at the same or similar conclusions.Thevalue of thisinformationto HoltecInternationalwould be lost if the information were disclosed to the public.Making such information available tocompetitorswithout their having been required to undertakeasimilar expenditure of resources would unfairly providecompetitorswith a windfall, and deprive Holtec International of the opportunity to exercise its competitive advantage to seek an adequate return on itslargeinvestmentin developing these very valuable analytical tools.STATE OF NEW JERSEY))ss: COUNTY OF BURLINGTON

)Michael P.McNamara,beingduly sworn,deposesand says: Thathehas read the foregoing affidavitandthe matters stated thereinaretrue and correct to the best of his knowledge, information, and belief.Executed at Marlton, New Jersey, this 16th day of Novembe 1999.Michael P.McNamaraHoltecInternational

-999.Subscribed and sworn before me this day of 1999 MId~A C.P0PG.4 NOTARY FUZLIC OF NEW JERSEY My Corydjsson xirm pilI 25.2000 ENCLOSURE 1 TO NRC-99-0084 FERMI 2 NRC DOCKET NO.50-341 OPERATING LICENSE NO.NPF-43 REQUEST TO REVISETECHNICALSPECIFICATIONS:

DESIGN FEATURES-FUEL STORAGE AND PROGRAMS AND MANUALS-HIGH DENSITY SPENT FUEL RACKS DESCRIPTION AND EVALUATION OF THE PROPOSED CHANGES to NRC-99-0084 Page 2 DESCRIPTION AND EVALUATION OF THE PROPOSED CHANGE(S)OVERVIEW: The following is a request to amend Operating License NPF-43 by incorporating the proposed changes identified in Enclosure 3 intotheTechnical Specifications of Fermi 2 to increase the spent fuel storage capacity.Expansion of the fuel storage capacity intheSpent Fuel Pool (SFP)requires a change to the Design Features section of the Technical Specification.

More specifically, Section 4.3, Fuel Storage, discussesthecurrent storage capacity and design features of the existing racks, which ensure adequate design margin with respect to criticality.Arequest to amend TS 5.5.13, High Density Spent Fuel Racks, is also included to clarifythatthe surveillance program is only applicable to racks which utilize Boraflex as a neutron absorber.BACKGROUND:

The SFP at Fermi 2 was reracked in 1982, priortoinitial licensingandoperation of the facility, with high-density freestanding racks using Boraflex as a neutron poison.At that time, the SFP was authorized tostore2305 fuel assemblies in 14 spentfuelracks, out of which two cells are reserved for the high density rack Boraflex surveillance program.Anadditional rack containing 35 cylindrical cells also currently exists to store control rods, up to 31 defective fuel canisters and other miscellaneous fuel related components.TheSFP also contains four low density GE racks with a total of 80 storage locations.

Therefore, the current configuration allows for storage of up to 2383 (2414 includingthedefective fuel locations)fuelassemblies.TheFermi 2 reactor core holds 764 fuel assemblies.

Current projections, based on expected future spent fuel discharges, indicate that loss of Full-Core-Discharge (FCD)capability will occurwhennew fuelisreceived for Cycle 9 in June 2001.Operation of Fermi 2 beyond loss of full-core-discharge capability ispossiblefor Cycles 9 and 10 toprovideapproximately three additional years of operation until 2004.Fermi 2 operating license authorizes plant operationsthroughMarch 20, 2025.The proposed change would increase the SFP storage capacity to permit continued plant operation until approximately 2015.DESCRIPTION OF PROPOSED CHANGES: Currently, Fermi 2 is proposingtoexpandthestorage capacity in the SFP to accommodate a Full-Core-Discharge when new fuel is received for Cycle 9inJune 2001.This proposed modification will be accomplished by installing additional storage racks and replacing existing racks with highdensityracks in a three phased approach.The initial phase will add up to fourracksto the SFP inopenspaces to increase the storage capacity to 3146 assemblies.

The second phasewillremove the four GE racks, the existing defective fuel storage rack, and the 108 cell high density storage rack and install five new high density racks.Thismodification will increase the storage capacity to 3588 assemblies.

The third phase will replace the remaining 13 existing Enclosure1to NRC-99-0084 Page 3 Boraflexrackswith 14 new high densityracksto increase the storage capacityto4608 assemblies.

The completed configuration represents a storage capacity increase of 2194 assemblies.

Thenewstorageracks will be free standing and self supporting.Thenewmodules will be separated byagap of approximately

1.0 inchfromoneanotherandwithgreater

gaps between the newracksandthe existing racks.As a means of providing additional storage space for miscellaneous components, the new storage racks are designed to accommodate two platforms, whichmaybe eachplacedabove, and supported by, an individualrack.The platformsweighapproximately 1,460 and 1,100 pounds, respectively, andmaysupportupto five tonseach(dry weight)of miscellaneous components.Theseplatforms would be installed on an as needed basis.The new racks will contain Boral as the activefixedneutron absorbingpoisonfor primaryreactivitycontrol.TheBoral absorbers are sized tosufficientlyshadow the active fuel height ofallfuel assembly designs stored in the pool.The proposedrackswill allow fuel storage forenrichmentsup to 5.0 wt%U-235 with fuel of thehighestanticipated reactivityandthe pool flooded withunboratedwater at atemperaturecorrespondingto thehighestreactivity.Asthere is no requirement for an in-service surveillance program for Boral, a clarification is being proposed to the Programs and Manuals section oftheTechnical Specifications.

The existing requirement for the High DensitySpentFuel Rack Surveillance Programwillonly be applicable to therackscontaining Boraflex.Toaccommodatetheproposedincreasein capacity,theFermi2Technical Specifications are required tobemodified.

The revised Technical Specification pages are providedinEnclosure 3.Enclosure 4 provides a report, which discusses the features of the new racks alongwiththeevaluationmethodologies used to establishadequatedesign margins with respect to structural, thermal and criticality performance.

SAFETY ASSESSMENT:

Theplannedexpansion ofthestorage capacity involves the installation of additional fuel racksandremoval of existing racks during three separate phases.Evaluations are performed to ensure that all possiblefuelconfigurationsremain safeundernormal andaccidentconditions.

The SFPthermalperformance,criticality, andseismicresponse are re-analyzed considering the increasedstoragecapacity.The results of these analyses have shown that the poolstoragesystems remain adequate to containandcoolthe fuel in asubcriticalcondition.

With theexpandedcapacity, thesystemsavailable toprovidecooling to the SFP will be required to remove an increased heat load while maintaining the SFP bulk temperature below the design limit of 1500 F.Themaximumheatloaddevelopsfrom the residual heat in the poolafterthelastfullcoredischargeatthe end of spent fuel pool storage capacity. to NRC-99-0084 Page 4 As plant operation continues, the SFPheatloadincreasesdue to the addition of spent fuelintroducedinto the SFP followingeachrefueling operation.

Due to theincreaseindecayheat load as the plantcontinuesto operate and perform refuelingoperations,Fermi 2 has determined the maximum normal SFP bulk temperature, presently 1250 F, will increase to a higher value of less than 1500 F.Additionally, due to an increased SFP spent fuelinventory,time-to-boil valueswilldecrease with a corresponding increase in boil-off-rates.

The new time-to-boil and make-up requirements areconsistentwith the industry andwithinthe plant design basis.TheSignificant Hazards Consideration (SHC), contained herein andtheattachedLicensing Report (Enclosure 4)demonstrate the acceptability of theproposedincreasein the SFP storagecapacityand revisions to the Technical Specifications.

The scope of the technical analysis supporting thisevaluationfocused mainly onthefinalconfiguration of the expanded storage space.Analysis of thetransitionto the final configurationinvolvingsome intermediate stages isalsoincluded intheevaluation.

MECHANICAL DESIGN EVALUATION:

The new fuel rack designs areevaluatedwith respect tothemechanicalandmaterial qualifications, neutron poison, fuelhandlingqualifications, fuelinterfacesand accident considerations.Theprincipalconstruction materials for the new racks will beSA240Type304L stainless steel, andSA564-630precipitation hardened stainless steel for the adjustable support spindles.The rackdesigns,material selection andfabricationprocess will comply withtheapplicableASTM Standards A240,A276,A479,A564and others, for nuclear service.The governing quality assurance requirements for fabrication of the racks meet or exceed 1OCFR50, Appendix B requirements.

For primarynuclearcriticalitycontrol in thenewracks, a fixed neutron absorber will be integrated within the rack structure.Theabsorber,tradenameBoral,is a boron carbide and aluminum-composite sandwich.Boralischemicallyinert and has a long history of applicationsintheSFPenvironments where it hasmaintaineditsneutronattenuation capability under thermal loads.Boral ismanufacturedunder the control of a quality assurance program, which meets or exceeds to the requirements of 1OCFR50, Appendix B.Theinstallation of thenewrack modules willpreservespace for thermal expansion and seismicmovement.Thesupportlegs on the racks will allow for remote leveling and alignment of therackmodules to accommodate variations in the floorflatness.A thick bearing pad will beinterposedbetween therackpedestalsandthefloortodistribute the deadloadovera wider support area.The rack structural performancewithrespect to the accidental drop impact andtensileloads, as well asthesubcriticalconfiguration, has been analyzed.The analyses included an accidental to NRC-99-0084 Page 5 drop of a fuelassemblyduring movement to a storage location andtensileloads (vertical and eccentric) ontherackarising from a stuckassemblyin the storage cell.The results of analysis demonstrate thatthestored spent fuelremainsin acoolableand subcritical configuration.

The storage rack structuralintegrity,and thus the fuel configuration, will be maintained.

The fuel will retain its structural integrity andremainsubcritical.

CRITICALITY CONSIDERATIONS:TheNRC guidelines and the ANSI standards specify thatthemargin of safety for criticality be maintained by having the maximum neutron multiplication factor, keff less than or equal to 0.95,includinguncertainties, for all normalandaccident conditions.

The new spent fuel racks aredesignedtomaintain the requiredsubcriticalitymargin when analyzed using conservative design criteria and assumptions.

The racks are considered fully loaded with fuel ofthehighest acceptable reactivity, submerged in unborated water at a temperature corresponding to the highest reactivity.

Neutron absorption in minor structural members is neglected, i.e., spacer grids are replacedbywater.The criticality analyses are based upon theinfinitemultiplication factor (kinf), i.e., lattice of storageracksis assumed infinite in all directions.Nocredit is taken for axial or radial neutron leakage, except in theassessmentof certain abnormal/accidentconditionswhere neutron leakage is inherent.The effects of calculational and manufacturing tolerances are evaluated and added in determiningthemaximum kinfinthe storage rack.Forreactivity control intheracks, Boral panels are used.The panels are sized to sufficiently shadowtheactive fuel height of all assemblydesignsstored inthepool.The panels are held in place and protected against damage by astainlesssteel jacket, whichisstitch welded to the cell walls.Thepanels are mounted on the exterior or on the interior of the cells, in an alternating pattern.Theanalysisshows that the criterion of keff less than or equal to 0.95, including uncertainties, is always maintained under all normal conditions and postulated accidents.

The accidents and malfunctions evaluated included a dropped fuelassemblyon top of thefuelrack;impact on criticality of water temperature and densityeffects;andimpacton criticality of eccentric positioning of a fuel assembly within the rack.ThenewBoral racks are designed to a kinf in the standard cold core geometry (SCCG)of 1.33, as discussed in Chapter 4 of the LicensingReport(Enclosure 4).However, in order to include additional criticality margin for the newracks,theTechnical Specifications will remain consistent for all racks.Thus,the racks maystorefuel with a maximum knf in the standard cold core geometry (SCCG)of 1.31.RefertoTechnical Specification, Section 4.3.1 (a). to NRC-99-0084 Page 6 THERMAL-HYDRAULICS AND POOL COOLING: A comprehensive thermal-hydraulic evaluation of the expanded storage capacity has been performed to analyze the thermal performance of the SFP anditscoolingsystems.

The maximum allowable SFP bulk temperature is 150*F for all scenarios.

The maximum local water temperature mustremainbelow boiling at localsaturationpressure and cladding temperaturesmustbe limited such that nucleateboilingdoes not occur.Thecalculation of theboundinglong-term decayheatfor thermal analysis ofthepool wasperformedin accordance with the provisions of the USNRCBranchTechnical Position ASB 9-2;"ResidualDecayEnergyfor Light Water Reactors forLongTerm Cooling".Thedetermination ofthedecay heat took into account both the pastdischargesandthe predicted future refueling cycles.Theevaluationsconsideredthedecay heat load from three separate discharge scenarios:

• Apartialcore discharge of 260 assemblies is considered, which corresponds to a conditionthatresults in a loss of full corereservein the SFP.Theminimum decay time of the previously discharged fuel assemblies for this scenario is 18 months.Thedecayheat load coincident with the peak temperature for this scenario is 12.20 MBtu/hr, which occurs 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after reactor shutdown.*A normal fullcoredischarge is considered, which produces a stored fuel inventory that conservatively exceeds themaximumpossibleinventory.The764dischargedassembliesare separated intotwodistinct groups: 260 assemblies with a burnup of 50,000 MWD/MTU and504assemblieswith a burnup of 33,333 MWD/MTU.The minimumdecaytime of thepreviouslydischarged fuel assemblies for this scenario is 18 months.The decay heat load coincident with the peak temperature for thisscenariois41.84 MBtu/hr,whichoccurs 160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> after reactor shutdown.*An emergency fullcoredischarge is considered, which is the same as the normal full-core discharge case above, except for thecoolingtimeconsidered.The 764 discharged assemblies are separated intotwodistinct groups: 260 assemblies with a burnup of 50,000 MWD/MTU and504assemblies with a burnup of 33,333 MWD/MTU.Theminimum decay time of the previously discharged fuel assemblies for this scenario is 12 months.The decay heat loadcoincidentwith thepeaktemperature for this scenario is 42.37 MBtu/hr, which occurs 159 hour0.00184 days <br />0.0442 hours <br />2.628968e-4 weeks <br />6.04995e-5 months <br />safterreactorshutdown.

Under normal SFP operations, with a decay heat load of less than 15.83 MBtu/hr, two trains oftheFuelPool Cooling and Cleanup System(FPCCS)provide sufficient cooling to maintain the SFP bulk temperature below 150°F.The 15.83 MBtu/hr decay heat load corresponds to a partial core discharge of 260 assemblies from the reactor shutdown 12 days previouslyandthe remainder of the pool filled with background fuel assemblies.

Enclosure1to NRC-99-0084 Page 7Duringrefueling conditions, supplemental cooling isnormallyprovidedto the SFP by the Residual HeatRemoval(RHR)system.

Infact,during fuel discharge to thepool,supplemental cooling from one division of theRHRsystem will normally be requiredtoprovide a satisfactory margin of safety to maintain the bulk pool temperature below 150'F as thedecayheat load is risinginthepool.Subsequentto confirmationthatonce theSFPdecayheatloadhasfallen below 15.83 MBtu/hr,supplementalcooling is no longerrequiredto be provided to the SFP bytheResidual Heat Removal(RHR)system.

In allscenarios,the cooling water thatremovesheat from the FPCCS and RHRheatexchangers is assumed to be at itsdesigntemperatureandflow rate.Inallcases analyzed, the heat transfermodelconservatively accounted for an additional resistance fromthefouling of the heat transfer surface in theheatexchangersandperformance loss due to plugged tubes.The local water temperature determinations are performed assuming that the pool is at its peak bulk temperature.Theworst location was identified asthecell with thehottestassembly and the mostrestrictiveflow arrangement.Aconservative value for the axialpeakingfactor is used.Thestoragecell hydraulic resistancewasbasedon the most hydraulically limiting fuel assembly type, themostrestrictive water inlet geometry forcellslocated over theracksupport pedestals,andtheeffects ofpartialoutletblockagedue to a dropped fuelassemblylying over the top of the storage cells.The bulk pooltemperatureanalysis determined that the cooling systems havesufficientcapacity to maintain the temperature below 150 0 F during, andsubsequentto, all postulated fuel discharge scenarios.Thecalculated maximum local watertemperatureisdeterminedto be 169°F in the hottest channel and coincides in time with thehighestpool bulktemperature.This water temperature is substantially below the 238 0 F local boiling temperature at the top of the racks.The maximum fuel claddingtemperatureis calculated to be 197.12°F.Therefore, nucleate boiling will not occur.Complete loss of all forced poolcoolingisnot considered acredibleeventin thedesignbasis, asstatedinUFSARsection 9.1.3.3.Nevertheless, a loss-of-cooling event was analyzed for all discharge scenarios.

The interruption of the cooling to the poolwasassumedto occur coincident with the SFPpeakdecay heat generation.Theanalysis determined thetimewhen the pool bulkwaterreaches boilingandthe resultant maximum water lossratefromthesurface.

The calculated time to boilis4.20 hour2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />safterthe cooling is lost inthemost severe scenario.However, this is acceptable because the corresponding boil-off rate is less than the makeup capacity of 100 gpm available from thecondensatestorage tanks, andadditionalsources of makeupincludingthe fire protection system and category I systems which can be aligned to supply SFP makeup.Additionally, the4.20hour period allows sufficient time for the operators tointerveneandlineup an alternate source to remove thedecayheat and replenish the pool inventory.

Enclosure I to NRC-99-0084 Page 8 SEISMIC AND STRUCTURAL EVALUATION:

A complete re-evaluation of the mechanical and civil structures, toaddressthestructuralissues resulting from the expansion of the pool storage capacity, has been performed.

The analysis considered the loadsfromseismic, thermal,andmechanical forces to determine themarginof safety in the structural integrity of thefuelracks, the liner,andthe SFP located in the Reactor Building.The loads, load combinations, and acceptance criteria are based on theASMESection III, Subsection NF, and on NUREG-0800, SRP Section3.8.4,Appendix D.a.The Storage Rack Evaluation:Thenew high density racks are analyzed considering the configurations at the end ofeachofthethreephaseswith therackscompletelyfilledwith fuel.Interim configurations and partiallyloadedracks arealsoevaluatedconsideringsingleisolated racks.The seismic analysis is performed using a whole poolmulti-rackanalysis.Theseismic load evaluations consider simulations of the Safe ShutdownEarthquake(SSE) and the Operating Basis Earthquake(OBE)in accordance with SRP 3.7.1.The rack modules are analyzed using a conservative assembly weight of 690 pounds.Thisweightisconservative since the actual weight of the assembly is 680 pounds.Two of the racks arealsoqualified for an additionaloptionalstoragefunction, i.e., the racks are designed to accommodate a storage platform, which has a capacity of storing up to five tons.(dry).The 1,460 poundplatformfor rack B and a smaller 1,100 pound platform for rack G are movable, andcanbe installed ontopoftheracks by inserting its four support legs into emptystoragecells.Theresults indicate that the maximum seismic displacements do not result in any rack-to wall or inter-rack impacts.The resultant member and weld stresses in the racks are all belowtheallowablestresses, with a safety factor of at least 1.05.This minimum calculated safety factor is associated withthepedestal support femalethreadshearstress.Theminimum calculated safety factorassociatedwiththecellmembranesis 3.8.The minimum safety factor for the welds is 1.97.The rackswillremainfunctional during and after all postulated loadingconditions,including the SSE.Fatigue analysiswasperformed onthestorage rackstodetermine thecumulativedamage factorresultingfromtwenty operatingbasisearthquakesfollowed byonedesign basis earthquake.

This analysis showed that the factor of safety is greater than 2.6 for fatigue within therackcomponents.

The rack analysis provides pedestal-to-bearing pad impact loads resulting from lift-off and subsequent resettling during dynamicevents.The pool floor stresses are determined fortheseimpact loads to remainwithinallowablelimits,even when considering the worst case pedestal locationwithrespect to leak chases.

Enclosure I to NRC-99-0084 Page 9 In addition to the seismic evaluations, the storage racks are also analyzed for all postulated structural accident conditions as described in UFSAR section 9.1.2.3.A fuel handling accident involving a fuel assembly dropped from the Refuel Bridge highest possible lift point would not compromise the integrity of the rack.Permanent deformation of the rack would be limited to the top region only.This is acceptable since the rack cross-sectional geometry at the active fuel height is not altered.Thus, the functionality of the rack is not affected.In the event of a stuck fuel assembly in the rack,theresultant load on the members will not affect the rack structural integritytomaintain the fuel storage qualifications.

b.Spent FuelPoolStructural Evaluation:

The SFP is located atthefifth floor of the Fermi 2 Reactor Building, north of the reactor drywell.The SFP consists of cast-in-place monolithic reinforced concrete interior and exterior walls and is designed as a seismic Class I stainless steel lined pool structure that provides space for storage of spent fuel assemblies.Thepool structurehasbeen analyzed using a 3-Dfiniteelement model withrackpedestal and hydrodynamic loads, and seismic loads applied by developing a modal analysis and performing a quasi-static evaluation.

The individual loads and load combinations used are in accordance with NUREG-0800, SRP Section 3.8.4 and based on the"ultimate strength" design method.The primary loads considered are:-the dead weight of the concrete structure and steel liner, fully loaded racks, fully loaded overhead platformsandthe water,-quasi-static seismic loads consistent with the original plant design fortheOBE and SSE cases,-hydrostatic pressure force lateral to the walls,-hydrodynamic coupling forces applied to the lower portion of the wall and water slosh and inertia pressures above the elevation of the top of the racks,-bounding thermal loads from a full-core-discharge and a loss of cooling, producing the largest temperature gradient across the thickness of the wall and the slab,-reactive forces due to live loads, and-seismically induced rack pedestal loads.In addition totheloads described above, the pool structure and liner are also analyzed for mechanical loads under accident conditions.

Analyses are also performed on liner fatigue considering both temperature and seismic excitation.

The results of the analyses performed on to NRC-99-0084 Page 10 the SFP and Reactor Building indicate that under all postulated loadingsthestructural components, floor slabs, poolwalls,supportingwalls, liner and its anchorages will be subjected to stresses orstrainswithinacceptable limits.RADIOLOGICAL CONSIDERATIONS:

Radiological consequences of the proposed change during theinstallationevolutionandduring normalandaccident conditions in the SFP area of the Reactor Building are evaluated.Thetotaldoseto personnelduringthe installation phasesisestimated to be less than 12 person-rem.

This includes removal and cleaning of old racks, diving operations toremoveunderwaterappurtenances,pool cleaning and rack installation.

This dose is comparabletosimilarprojects carried out at other facilities and allows compliance with the radiological limits of 1 OCFR20.Low level solid radwaste will be generated bytheremoval of the existing SFP modules, as well as any interferences or SFP hardwarethatmay have to beremovedfrom theSFPto permit installation of thenewspent fuel rack modules.Theseracks will be cleaned to the maximum extent possible prior to offsite shipment and volume reduction will be performed at a decontamination and processing facility prior todisposaland burial at alowlevelwaste repository.

Therefore,solidradwaste represented bytheserackswill be minimized.

Thus, Fermi2doesnotexpectthatincreasingthestorage capacity of the SFP will result in a significant change inthegeneration of solidradwasteatFermi 2.There are no significant solid, gaseous or liquidradiologicalreleases associated with this storage capacity increase effort.The analysis of the fuel handling accident event shows that the racks remain intact and the resulting fueldamagemaintains gas releases and the corresponding radiological dose belowlevelspreviously determined.

Arackdropinvolvingradiological consequences is precluded, since all rack movement during removal andinstallationphaseswill follow safeloadpathsthatpreventheavyloadsfrombeing transported over the storedspentfuel.

Therehasbeennosteady long-term increase of radiologicalconditionsin the SFP resulting from the radionuclides within the fuel as more spent fuel is added to the pool.Theradiological conditionswithinthe building are typically dominated by the mostrecentbatch ofthespentfuel from a full-core-discharge.

The radioactive inventory of the older fuel that will increase with the expanded storagecapacitywill be insignificant comparedtothat of the recent offload.Sincethenewstoragerackswill be locatedincloser proximity to the SFP walls, an increase intheadjacentradiologicaldosesisexpected.

Radiological analyses have shown thatthedose levels adjacent to all pool areas willremainwithin acceptable levels.

ENCLOSURE 2 TO NRC-99-0084 FERMI 2 NRC DOCKETNO.50-341 NRC LICENSE NO.NPF-43 REQUEST TO REVISE TECHNICAL SPECIFICATIONS:

10CFR50.92 SIGNIFICANT HAZARDS CONSIDERATION to NRC-99-0084 Page 2 10 CFR 50.92 SIGNIFICANT HAZARDS CONSIDERATION BASIS FOR SIGNIFICANT HAZARDS DETERMINATION In accordance with lOCFR50.92,DetroitEdison has reviewed the proposed changes and has concluded that theydonot involve a Significant Hazards Consideration (SHC).Thebasis forthisconclusionis thatthethree criteria of 1OCFR50.92(c) are notcompromised.The proposed changes do not involve a SHC because theywouldnot: 1.Involve a significant increase in the probability or consequences of an accident previously evaluated.Thefollowingpreviously postulated accident scenarios are considered:

a.A spent fuelassemblydrop in the SFP b.Loss of SFPcoolingflow c.A seismic event d.Misplaced fuel assembly The probability that any oftheaccidentsinthe above list can occur is not significantly increased by the modification itself.The probabilities of a seismic event or loss of SFP cooling flow are not influenced bytheproposed changes.The probabilities ofaccidentalfuelassemblydrops or misplacement of a fuel assembly areprimarilyinfluenced bythemethods used to liftandmove these loads.Themethod of handling loads during normal plant operations is not changed, since the same equipment (i.e., Refuel Bridge)andprocedures will beused.Since the methods used tomoveloadsduringnormal operationsremainthesame as thoseusedpreviously, thereisnosignificant increase in the probability of an accident.During rack removal and installation, all work in the pool area will be controlled and performedinstrict accordance with specific written procedures.

Any movement of fuelassembliesrequiredtosupportthe modification (e.g.,removaland installation of racks)will be performed in the same manner as during normalrefuelingoperations.

Spent Fuel shipping cask movements will not be performed during the modification period.Accordingly, the proposed modification does not involve a significant increase in the probability of an accidentpreviouslyevaluated.

The consequences of thepreviouslypostulated scenarios for an accidental drop of a fuel assemblyinthe SFP havebeenre-evaluated fortheproposedchange.

Theresultsshow that the postulated accident of afuelassemblystriking the top of the storage racks will not distorttherackssufficiently to impair theirfunctionality.Theminimumsubcriticality margin, keff to NRC-99-0084 Page 3 less thanorequal to 0.95, will be maintained.

The structural damage to theReactorBuilding, pool liner, andfuelassembly resultingfroma fuel assembly drop striking thepoolfloor oranotherassemblylocated within theracksis primarilydependenton the mass ofthefalling objectandthe drop height.Sincethese two parameters are not changed by the proposedmodification,the structural damagetotheseitems remains unchanged.

The radiological dose at the exclusion area boundary will not beincreaseddue to the changes.Thus, theresultsof the postulated fuel drop accidents remain acceptable and donotrepresent a significant increase in consequences from any of the same previously evaluated accidentsthathave beenreviewedandfound acceptable by the NRC.Thetime toboilrepresents the onset of loss of poolwaterinventory and is commonly used as a gage for establishing the comparison of consequences before and after a reracking project.The heat up rate in the SFP is a nearlylinearfunction of the fueldecayheat load.The fuel decay heatloadwill increasesubsequentto the proposed changes because of the increase in the number of fuelassembliesstored in the spent fuel pool.Thethermal-hydraulic analysis determined the maximum fuel decayheatloadsandthe corresponding time to boil conditions subsequent to complete loss of forced cooling.These results show that, in the extremelyunlikelyevent of a complete failure of both the FPCCSandRHR System,therewould be at least 4.20hoursavailable for correctiveactions.Themaximumwater boiloff rate is less than 91 gpm.This is less thanthenormal makeup capacity of 100 gpm available from thecondensatestorage tanks,andadditional sources of makeup are available.

It has been determined that this duration provides sufficient time for theoperatorsto provide alternate means of makeup (i.e., fire hoses)before the onset of poolboiling.Therefore, the proposed change represents no increase in the consequences of loss of pool cooling.Theconsequences of a design basis seismic event are not increased.

Theconsequencesof this accident are evaluated on the basis of subsequent fuel damage or compromise of the fuel storage or building configurations leading to radiologicalorcriticality concerns.The racks are analyzedintheirnewconfiguration and foundsafeduringseismic motion.Fuel has been determinedtoremainintact and the storageracksmaintain the fuel and fixed poisonconfigurationssubsequentto a seismic event.The structural capability of the pool and liner will notbeexceededunder the appropriate combinations of deadweight,thermal, and seismic loads.The Reactor Building structure will remainintactduring a seismic event and willcontinuetoadequatelysupportand protect the fuel racks,storagearray,andpoolmoderator/coolant.Thus,theconsequences of aseismicevent are not increased.

A fuel misplacement accident represents afuelassemblyinadvertentlylowered or dropped outside of and adjacent to a storagerack.The consequence of a fuel misplacement accident has been analyzedforthe worst possible storageconfigurationsubsequent to the proposed modification, and it has been shown that theconsequencesremainacceptable with respect totheneutron multiplication factor staying below0.95(i.e.thesameacceptance criteria as used for normal conditions).Therefore,there is no increaseinconsequences. to NRC-99-0084 Page 4 Therefore, itisconcluded that theproposedchanges do not significantly increase the probability or consequences of any accident previously evaluated.

2.Create the possibility of a new or different kind of accident from any previously evaluated.

Load drops were determined to be events that might represent a new or different kind of accident.The new loads that will be required during or subsequenttoinstallation of the new racks include the rack modules, the overhead platforms, and the pool gates.Racks will not be allowed to travel overanyracks containing fuel assemblies, thus a rackdroponto fuel is precluded.

A construction accident of a rack dropping onto the pool floor liner is not a postulated event due to the defense-in-depth approach to be taken, as discussed in detail within Section 10.2 of the attached Licensing Report (Enclosure 4).A new temporary hoist and rack lift rig will be introduced to lift and suspend the racksfromthe bridge of the Reactor Crane.These temporary lift items are designed in accordance with the requirements of NUREG 0612 and ANSI N14.6.Nevertheless, the analysis of a rack dropping to the liner has been performed and shown to be acceptable.

The integrity of the liner will be maintained and no loss of pool coolantwouldoccur subsequent to arackdropping to the liner.Since fuel integrity is maintained and significant loss of coolant does not occur, the drop of a rackisnot considered a new type of accident.A drop of a pool gate is also an extremely unlikely event.Thenew storagerackswill not be located directly beneath the gates.However,thedrop ofagate, weighing approximately 9500 pounds, onto racks containing irradiated fuel assemblies, and the drop of a gate onto the pool liner have been analyzed.The analysis performed for the drop of a pool gate onto fuel demonstrates thatthenumber of fuelrodsdamaged (81)remainsbelow the Fermi 2 fuel handling accident design basis (of 140 rods).The analysis performedforthe drop of a pool gate onto the liner demonstrates that thelinerwould be locally ruptured.However, the underlying concrete slab remains intact and possible leakage would be confined to the leak chase system, which is monitored and controllable.

The kinetic energy associated with the drop of the heaviest (1460 pound)overhead platform is enveloped by the kinetic energy associated with the gate drop.Therefore, the potential structural damage to fuel and the liner wouldbebounded bytheresultsforthe gate.Since the resulting fuel damage does not exceed the previously analyzed design basis condition and significant loss of coolant would not occur, the drops ofagate or an overhead platform are not considered a new type of accident.The additional heat load resulting from additional storage of spentfuelhasbeenevaluated for the possibility of creating a new or different kind of accident.The existing Fermi2SFP cooling system,hasbeen shown by analysis, to be capable of removingthedecay heat generated by the additional spentfuelassemblies.

The pool coolant will not be significantly affected.Thus, the increased heat load does not create the possibility a new or different kind of accident. to NRC-99-0084 Page 5Nounproven technologyhasbeen utilizedinthe design, analysis or in the proposed installation methodology.

The basic technology for the Fermi 2 spent fuel pool capacity increase isconsistentwith other license amendments (over 80)approved by the USNRC.This change hasbeenevaluated in accordance with the USNRCpositionpaper"OT Position for ReviewandAcceptance of Spent FuelStorageand Handling Applications, April 14, 1978 and AdditiondatedJanuary 18, 1979.The proposed change does notaltertheoperating requirements of the plant or of the equipment credited in the mitigation ofthedesign basis accidents.

The proposed change does not affecttheparameters required forsafefuel storage.Therefore, this change does not create the possibility ofanew or different kind of accident from any previously evaluated.

3.Involve a significant reductioninthe margin of safety.The function of the SFP is tostorethe fuel assemblies in a subcriticalandcoolable configuration through all environmentalandabnormal loadings, such as an earthquake or fuel assembly drop.Thenew rack design must meet all applicable requirements for safe storage and be functionallycompatiblewith the SFP.Detroit Edison has addressed the safety issues related to the expanded pool storage capacity in the following areas: 1.Material, mechanical and structural considerations 2.Nuclear criticality 3.Thermal-hydraulic andpoolcooling The mechanical, material, and structural designs of thenewracks are reviewed in accordance with the applicable provisions oftheUSNRC position paper"OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, April 14, 1978 and Addition dated January 18, 1979..The rack materials used are compatible with the spent fuel assembliesandthe SFP environment.

The design of the new racks preserves the proper margin of safety during abnormalloadssuch as a dropped assembly and tensile loads from a stuck assembly.It hasbeenshown thatsuchloads will not invalidatethemechanical designandmaterial selection tosafelystore fuel in a coolable and subcritical configuration.

The methodology used in the criticality analysis of the expanded SFP storage capacity meets the appropriateNRCrequirementsandthe ANSI standards (GDC 62, NUREG 0800, Section 9.1.2, theOTPosition for ReviewandAcceptance of Spent Fuel Storage and Handling Applications, Reg.Guide 1.13, and ANSI ANS 8.17).The margin ofsafetyfor subcriticality is maintained by having the neutron multiplicationfactorequal to, orlessthan, 0.95, including uncertainties, under all accident conditions.

This criterion is the same as that used to NRC-99-0084 Page 6 previouslytoestablishcriticality safety evaluation acceptanceandremainssatisfied for all analyzed accidents.

Therefore, the accepted margin of safety remainsthesame.Thethermal-hydraulicand cooling evaluation of the pooldemonstratedthat the pool can be maintainedbelowthe specified thermallimitsunder the conditions of the maximum heat load and during all credible accidentsequencesandseismicevents.Thebulk pool temperature will not exceed 150°F during any conditionswhenforcedcoolingis available.

The increase fromthecurrent maximum normal SFP bulk temperature of125°Fisnotsignificant, because the existing racks andcoolingsystem werepreviouslyevaluatedfor the 150'F condition, asstatedinUFSARsections 9.1.2.2.2 and 9.1.3.1,respectively.The maximum local water temperature in the hottest rackcellwill remain below the boiling point.The fuel will notundergoanysignificant heat up after an accidental drop ofafuelassemblyontop of the rack blocking the flow path.The time of 4.20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> for the onset of poolboiling,subsequent to total loss of forced cooling allowssufficienttime for the operators to intervene and line up alternate cooling paths and/or the means ofinventorymake-upbefore the onset of pool boiling.Thus, it isconcludedthat thechangesdo not involve a significant reduction inthemargin of safety.TheNRC has provided guidance concerningtheapplication of standards in 1OCFR50.92 by providingcertainexamples (5 1FR775 1, March 6, 1986)ofamendmentsthat areconsiderednot likely to involve a SHC.Theproposed changes for Fermi 2 are similar to Example (x): an expansion ofthestorage capacity of SFP when all of the following are satisfied:

(1)The storage expansionmethodconsists ofeitherreplacing existingrackswith a design that allows closer spacingbetweenstored spent fuel assemblies or placing additional racks of theoriginaldesign on the pool floor if space permits.The Fermi 2 storage expansion modification involves replacement of the existing racks withadesignthatwill allow a greater number of stored fuel assemblies and placement of additional racks in the SFP.(2)The storage expansion method doesnotinvolve rod consolidation or double tiers.The Fermi 2 storageexpansionmodification doesnotinvolve any fuel consolidation processes or storage of consolidated fuel.Theracks will not be double tiered;no fuelassemblieswill bestoredabove other assemblies.

(3)The keff of thepoolis maintained less than, or equal to, 0.95.The design of thenewracksintegrates aneutronabsorber,Boral,withintheracks to allow close storage of spentfuelassemblies while ensuring that keff remainslessthan or equal to 0.95 under all conditions. to NRC-99-0084 Page 7 (4)Nonewtechnology orunproventechnology isutilizedin either the construction process or the analyticaltechniquesnecessary to justifytheexpansion.

Therackvendor hassuccessfullyparticipated in the licensing of numerous other racks of a similar design.The constructionprocessandthe analytical techniques oftheFermi 2 modifications arethesame as in the other completedpoolstorage capacity expansionprojects.Thus,nonew or unproven technology is used in the Fermi 2 storage expansion modification.

ENVIRONMENTAL CONSIDERATIONS:DetroitEdison has reviewed the proposed license amendmentagainstthe criteria of 1OCFR51.22 forenvironmentalconsiderations.

The proposed changes do not significantly increase the typesandamounts of effluents that maybereleased offsite nor significantly increase individual or cumulative occupational radiation exposures.

Based on the foregoing, Detroit Edison concludes that the proposed changesmeetthe criteria delineated in 1OCFR51.22(c)(9) for a categoricalexclusionfrom the requirements for an environmentalimpactstatement.

CONCLUSION:

As discussed herein, theproposedchangesto the Technical Specificationsdonot involve a SHC pursuant to 1OCFR50.92.

Storage expansionatthe Fermi 2 SFP has been determined tobesafe.Additionally,DetroitEdison hasdeterminedthatthis license amendment meets the criteriadelineatedin 1OCFR51.22(c)(9) for a categorical exclusionfromtherequirements for an environmentalimpactstatement.

ENCLOSURE 3 TO NRC-99-0084 FERMI 2 NRC DOCKET NO.50-341 OPERATING LICENSE NPF-43 REQUEST TOREVISETECHNICAL SPECIFICATIONSDESIGNFEATURES AND PROGRAMS AND MANUALS Attached is a mark-up of the existingTechnicalSpecifications(TSs), indicatingtheproposed changes (Part 1)andatypedversion of the TSsincorporatingthe proposedchangeswith a list of included pages (Part 2).

ENCLOSURE 3-PART 1 TO NRC-99-0084 PROPOSED TECHNICAL SPECIFICATION MARKED UP PAGES INCLUDED PAGE(S): 4.0-2 5.0-19 Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued) b.kff z 0.95 if fully flooded with unborated water.which.includes an allowance for uncertainties as described in Section 9.1 of the UFSAR: and c.,-.ninal G.22 inch ccnte.to center dirstanee.

t,.-...fuol-asemblies pleeed intehg density stAorege raeks nd-O.......1......x ,., inbcente ,.eto fuel.esse,,-lies piaced in the-IC-CISEOS lazi"tvrQ*a.4.3.2 Drainage-#The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 660 ft 11.5 inches.4.3.3 Capacity The spent fuel with a storage assemblies.

storage pool is designed and shall be maintained capacity limited to no more than-E414-fuel (46ob FERMI-UNIT 2 Amendment No.134 4.0-2 INSERT#1 The following nominal center tocenterdistances between fuel assemblies placed in the various storage rack types, as applicable Spacinaq Rack Type (Inches)6.22 High density storage racks that contain Boraflex as the neutron absorbing material 6.23 High density storage racks that contain Boral as the neutron absorbing material 11.9x6.6 Low density storage racks 10.5 Defective fuel assembly storage rack Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Primary Containment Leakage Rate Testing Program (continued) e.The provisions of SR 3.0.2 do not apply to the test frequencies in the Primary Containment Leakage Rate Testing Program.f.The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.5.5.13 High Density Spent Fuel Racks unnticipate-a....of the h4gh density spent fmtel racks-bll be*,.dctcOcd and 4l3 ntwi sem o the in t..ityof the Amendment No.-134-FERMI-UNIT 2 5.0-19 INSERT#2 A program shall be provided, for the high density storageTacks containing-Boraflex as the neutron absorber, which will ensure that any unanticipated degradation of the Boraflex will be detected and will not compromise the integrity of the racks.

ENCLOSURE 3-PART 2 TO NRC-99-0084 PROPOSED TECHNICAL SPECIFICATION REVISED PAGES INCLUDED PAGE(S): 4.0-2 5.0-19 Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued) b.k,,: 0.95 if fullyfloodedwith unborated water, which includes an allowance for uncertainties as described in Section 9.1 of the UFSAR;and c.The following nominal center to center distances betweenfuelassembliesplacedin the various storage rack types, as applicable Spacing (inches)6.22 6.23 11.9x6.6 10.5 Rack Type High density storage racks that contain Boraflex as the neutron absorbing material High density storage racks that contain Boral as theneutronabsorbing material Low density storage racksDefectivefuel assembly storage rack 4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to preventinadvertentdraining of the pool below elevation 660 ft 11.5 inches.4.3.3 Capacity The spent fuel with a storage assemblies.

storage pool is designed and shall be maintained capacity limited to no more than 4608 fuel Amendment No.J1, FERMI-UNIT 2 4.0-2 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12-Primary Containment Leakaqe-Rate-estinQ-Program (continued) e.The provisions of SR 3.0.2 do not apply to the test frequencies in the Primary Containment Leakage Rate Testing Program.f.The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.5.5.13 High Density Spent Fuel Racks A program shall be provided, for the high density storage racks containing Boraflex as the neutron absorber, which will ensure that any unanticipated degradation of the Boraflex will be detected and will not compromise the integrity of the racks.Amendment No.J1 , FERMI-UNIT 2 5.0-19 Coalition for a Nuclear-Free Great Lakes

• Don't Waste Michigan

• Nuclear Information and Resource Service (NIRS)

• Summary Report High-Level Atomic Waste Mishap at Palisades Nuclear Reactor Risks Radioactive Inferno with Casualty Potential of Thousands of Deaths DownwindBased Upon U.S. Nuclear Regulatory Commission Freedom of Information Act (FOIA)

Response Documents April 4, 2006 Prepared byKevin KampsNuclear Waste Specialist NIRS 6930 Carroll Avenue, Suite 340 Takoma Park, MD 20912 Office 301.270.6477x14 Cell 240.462.3216 kevin@nirs.org www.nirs.orgOn March 18, 2006 the Detroit Free Press ran a front page article entitled "Nuclear safety left hanging as crane dangled fuel rods: Michigan incident got warning but no fine," by Hugh McDiarmid, Jr., Free Press Staff Writer. The article revealed a previously unreported October 2005 incident at the Palisades nuclear power plant on the Lake Michigan shoreline in southwest Michigan. According to a U.S. Nuclear Regulatory Commission (NRC) inspection report, a container weighing 110 tons, fully loaded with high-level radioactive waste, dangled for 55 hours6.365741e-4 days <br />0.0153 hours <br />9.093915e-5 weeks <br />2.09275e-5 months <br /> from a stuck crane above the reactor's

irradiated fuel storage pool. Plant personnel, lacking proper knowledge about the crane, and without permission from plant management, mishandled the crane's emergency brake, increasing the risk of the heavy load crashing, out of control, back down into the pool. The falling container could have severely damaged the pool, draining the cooling

water. A radioactive waste fire could have followed, resulting in tens of thousands of

cancer deaths from radiation exposure to a distance of 500 miles downwind, according to a separate NRC report.

1 Internal Palisades and NRC documents, received by NIRS via FOIA, reveal the mistakes that led up to this incident, and the potentially catastrophic consequences that could have resulted. The Cask Dangle First Comes to Light The Palisades nuclear power plant is located in Covert, Michigan, on the Lake Michigan shoreline. Living up to the name of its hometown - Covert -- Consumers Energy's Palisades nuclear power plant, with help from the NRC, managed to keep the public in the dark for months

about an incident that could have led to a Chernobyl-scale radiation release on the Lake Michigan shoreline. Tens of thousands of people, out to a distance of 500 miles downwind, could

have died immediately or due to later cancer, according to NRC reports. Coalition for a Nuclear-Free Great Lakes, Don't Waste Michigan, and NIRS first learned of the cask dangle on December 21, 2005 while attending an NRC/Palisadestechnical meeting at NRC's Region III office in Lisle, Illinois. An NRC official revealed that, while lifting a fully loaded waste container out of its storage pool, Palisades experienced a brake engagement which left the cask suspended over the pool from October 11 to 13, 2005. It was also admitted that no event report had been published, thus having kept the public in the dark for over two months at that point.

NIRS filed a FOIA request on January 9, 2006. Although NRC stated that it would respond to the FOIA request in two to four weeks, the FOIA response was not received by NIRS until March 20 th, over two months later. However, a few days earlier, researchersfrom the Coalition for a Nuclear-Free Great Lakes uncovered an NRC quarterly inspection report issued January 25, 2005 (with an erratum dated February 2, 2006). This NRC inspection report revealed, at page 9:

"The [NRC] inspectors concluded that working outside the bounds of a work package on a crane with a suspended load that if dropped would damage the spent fuel pool warranted a safety significance determination-Had the load dropped, the spent fuel pool could have sustained severe damage. The inspectors were also aware that the individuals involved in the work activity were not fully knowledgeable of the crane's design, operation, and failure modes at the time the work occurred. In order to compensate for the gap in knowledge, the licensee [the owner, Consumers Energy, and operator, Nuclear Management Company] obtained telephonic support from the crane vendor. Therefore, the inspectors concluded working outside the bounds of the approved work package and manipulating the brake release represented an increase in the risk of a load drop. This increase in risk is directly associated with the reactor safety cornerstone objective of the spent fuel cooling system as a radiological barrier."(1)

In other words, the crashing cask, fully loaded with high-level radioactive waste and weighing 107 tons, could have cracked the bottom of the pool and drained out the cooling water.

In a matter of hours or less, the years and decades worth of accumulated high-level radioactive

wastes stored in the pool could have gotten so hot that it would have ignited into a radioactive

conflagration.

2 The Potentially Catastrophic Consequences Had the Cask Dropped Another NRC report, NUREG-1738, "Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants," published February, 2001, examined just such heavy load drops causing the collapse of the waste storage pool floor. At page 3-16, NRC

reports:"The analysis exclusively considered drops severe enough to catastrophically damage the SFP [spent fuel pool] so that pool inventory [of cooling water] would be lost rapidly and it would be impossible to refill the pool using onsite or offsite resources. There is no possibility of mitigating the damage, only preventing it-The staff assumes a catastrophic heavy load drop (creating a large [cooling water] leakage path in the pool) would lead directly to a zirconium fire.The time from the load drop until a fire varies depending on fuel age, burn up, and configuration. The dose rates in the pool area before any zirconium fire are tens of thousands of rem per hour, making any recovery actions (such as temporary large inventory [of replacement cooling water] addition) very difficult. Based on discussions with [NRC] staff structural engineers, it is assumed that only spent fuel casks are heavy enough to catastrophically damage the pool if dropped."(2) Given that Palisades is an operating reactor, the wastes in its storage pool are even hotter - radioactively and thermally - than wastes at a decommissioning, or permanently shut down, nuclear power plant. In fact, NRC has reported that "-the possibility of a zirconium fire leading to a large fission product release cannot be ruled out even many years after final shutdown..."(3)

Thus, these NRC reports reveal that a Chernobyl-scale nuclear catastrophe could have occurred on the Lake Michigan shoreline last October. NRC admits that once the cask had cracked the pool and drained the cooling water away, radiation doses near the pool would have killed any emergency responders who approached too near after just a few minutes exposure time. Firemen would have had to sacrifice their lives in any attempt to stop the quickly unfolding disaster. But NRC chillingly stated such an accident must be prevented in the first place, because once it starts, it is impossible to put the deadly radioactive genie back in the bottle. Palisades' irradiated nuclear fuel rods are clad in zirconium metal. Zirconium, an ingredient in cluster bombs and old-fashioned camera flash bulbs, spontaneously combusts at a high-enough temperature. The thermal heat generated by radioactive decay occurring in Palisades' hotter wastes, morerecently discharged from the reactor core, could initiate the fire, which would then likely spread to the entire waste inventory in the

pool.According to the U.S. Department of Energy's February 2002 EnvironmentalImpact Statement for the proposed national high-level radioactive waste dump at Yucca Mountain, Nevada, Palisades currently has over 188 tons of highly radioactive nuclear fuel stored in its pool.(4). Palisades pool thus contains significantly more dangerous and deadly long-lasting radioactive poisons, such as Cesium-137, than were released by the 3

Chernobyl nuclear catastrophe in 1986.(5) Whereas the shorter lasting radioactive poisons that were present in Chernobyl's reactor core would have decayed away over the

years and decades in the Palisades pool, the longer-lived radioactive poisons, such as Cesium-137 (hazardous for 10 to 20 "half lives," that is, 300 to 600 years), would still be present in very large quantities in the Palisades pool.Alvarez et al. stated "Spent fuel recently discharged from a reactor could heat up relatively rapidly to temperatures at which the zircaloy fuel cladding could catch fire and the fuel's volatile fission products, including 30-year half-life [Cesium-137] would be released. The fire could well spread to older spent fuel. The long-term land-contamination consequences of such an event could be significantly worse than those from Chernobyl."(6)Citing a United Nations report from 2000, the Alvarez report went on to state that: "The damage that can be done by a large release of fission products was demonstrated by the April 1986 Chernobyl accident. More than 100,000 residents from 187 settlements were permanently evacuated because of contamination by [Cesium-137]. Strict radiation-dose control measures were imposed in areas contaminated to levels greater than [15 curies per square kilometer, or 555 kilo-Bequerels per square meter] of [Cesium-137].

The total area of this radiation-control zone is huge: [10,000 square kilometers], equal to half the area of the State of New Jersey. During the following decade, the population of this area declined by almost half because of migration to areas of lower contamination."(7)10,000 square kilometers equals 3,800 square miles, nearly 7% of the total land area of the State of Michigan.NRC goes on to report in NUREG-1738, in "Appendix 4: ConsequenceAssessment from Zirconium Fire," that tens of thousands of people could have then died, either promptly from radiation poisoning, or from latent cancers, up to 500 miles

downwind. Table A4-7, "Mean [Average] Consequences for the Base Case," shows over 26,800 deaths possible. Table A4-15, assuming a larger population density per square mile, estimates a long-term consequence of 44,900 cancer fatalities downwind.(8)

Palisades and NRC report that the population and population density surrounding Palisades is relatively small. In 2000, 118,667 people were living within 20 miles of Palisades, for a density of 238 persons/square mile; 1,287,558 persons were living within 50 miles of the plant, for a density of 283 persons/square mile.(9) However, it must be pointed out that the bulk of Michigan's second largest city - Grand Rapids - lies outside that 50 mile zone. And the largest cities in Michigan and Illinois - Detroit and Chicago -

fall within 500 miles of Palisades. These large populations would worsen casualty rates downwind of a major radiation release from Palisades. Of course, not only could tens of thousands of people have died from radiation poisoning and cancer, but Michigan's entire tourism and agricultural industries could have been ruined, as 4

well. And Palisades is located on the shore of Lake Michigan, whose waters -- and the waters of the Great Lakes downstream into which Lake Michigan flows -- provide drinking water for millions of people in the U.S. and Canada. Thus, the consequences of a large radiation releasefrom Palisades would be dire indeed. Despite this, the NRC quarterly inspectionreport stated "because the actions by the worker did not result in any load motion and both crane brakes remained set, NRC management determined the finding to be of very low safety significance."(10) Incredibly, NRC has let Palisades off with a slap on the wrist. It's not unlike the Davis-Besse nuclear power plant near-meltdown in 2002 near Toledo, in which the NRC's own inspector general reported that both NRC and the nuclear utility put company profits over public safety. In that case, we almost lost Toledo. In this case, we almost lost west Michigan, and Lake

Michigan as well.(11)Coalition for a Nuclear-Free Great Lakes also uncovered an NRC event report fromOctober 12, 2005 - the exact timeframe for the cask dangle - revealing that "Portions of the Palisades Plant Process Computer (PPC) including the Emergency Response Data System (ERDS) became inoperable due to failure of a plant inverter-"(12) This begs the question, could this computer failure have been yet another straw to break the camel's back that day, resulting in a radiological catastrophe downwind and downstream of Palisades?Citing many of the NRC documents previously referenced, The Detroit Free Press reported the Palisades cask dangle, and its potentially catastrophic consequences, on March 18, 2006.Revelations from NRC's FOIA Response to NIRS NRC's "Partial" FOIA response, although dated March 8, 2006 - two full monthsafter the FOIA request was made - did not reach NIRS until nearly two weeks later. NRC and Palisades internal documents reveal that many mistakes led up to the cask dangle, and also that other short cuts on safety could have made the incident event more

dangerous.For example, on October 6, 2005 - just five days before the cask dangle - NRC granted Palisades an exemption from "Criticality Accident Requirements" for loading of independent spent fuel storage installation casks. This despite NRC's admission that "NMC's [Nuclear Management Company's, Palisades operator] request for exemption-proposes to permit NMC to perform spent fuel loading, unloading, and handling operations related to dry cask storage without being subcritical under the most adverse moderation conditions feasible by unborated water." NRC also assumed that the spent fuel would be kept in "a geometrically safe configuration," and that "appropriate, conservative criticality margins during handling and storage of spent fuel" would be

applied.(13) 5 But the cask dangle involved a containerwhose lid had not yet been bolted shut. If the cask had dropped into the pool, the waste within could have fallen out, forming a critical mass. The still-fissile components in the waste - uranium-235 and plutonium-239

- could have caused a nuclearchain reaction in the pool. This would be all the more

likely if unborated water were added to the pool - such as to replenish cooling water in the event of a pool leak from the cask drop. Boron in the pool water serves as an anti-

criticality measure.

Palisades upgraded its irradiated nuclear fuel storage pool crane to "single-failure- proof" in June, 2004, just 16 months before the cask dangle incident. NRC, in granting the crane upgrade, stated "[s]ince the new main hoist for the upgraded crane is of the single-failure-proof design, the cask drop analysis is no longer required for load drops from the main hoist. As a result of the impact-limiting pads previously installed in the spent fuel pool to protect the pool structure from the postulated transfer-cask-drop accident during dry fuel storage operations is being eliminated." Thus, as the crane was upgraded, the pool was allowed to lose a layer of protection against a cask drop. Was this to free up more space in the pool, so that more waste could be stored there? However, in allowing the impact-limiting pads to be removed, NRC was assuming that Palisades would provide proper "training and qualification of crane operators," as well as "inspection, testing and maintenance of cranes." But it was just such failures that led to the cask dangle.An internal Palisades documents reveal that "[crane vendor] Ederer procedure 260 was apparently used a reference, but not followed completely to determine torque value for the EATL [energy absorbing torque limiter,an emergency brake]." It also revealed that "OE [operational experience] from Big Rock [a Consumer's Energy nuclear power plant in Charlevoix, MI that permanentlyshut down in August 1997] showed similar occurrences from an improperly set EATL that needed additional adjustments. This had occurred twice on a crane of the same design as Palisades."(14) So, despite previous company experience with just such crane malfunctions, the "lessons learned" were not enough to prevent the incident at Palisades. And despite a company pledge to NRC in its

Final Safety Analysis Report, that "[t]he design and construction of the [irradiated fuel handling] system includes interlocks, travel and load limitingdevices and other protective measures to minimize the possibility of mishandling or equipment malfunction that could cause damage to the fuel and potential fission product release,"(15) internal company documents reveal an alarming lack of understanding of the crane. In an internal Palisades document tellingly titled "Intent of WO [Work Order] Task Exceeded During Troubleshooting," the teamsent to inspect the stuck crane and its dangling cask admitted:"The team members had all been trained to perform mechanicalinspections of cranes. In this training, components are visually inspected and mechanically inspected using a number of techniques and tools. While being pre-job briefed, the team heard 'Go and perform a normal mechanical inspection you have been trained to perform."

The actual 6

intent for the inspection approved by the Event Response Team was to perform a visual inspection on the mechanical components to determine if anything was broken. With this disconnect, the team performed a normal mechanical inspection which was outside the intended inspection approved by the Event Response Team-Although we all thought the information we were gathering was within the steps of the Work Order, we failed to consider the severity of the consequences if our troubleshooting caused the load to slip or fall into the Spent Fuel Pool.

This is why we set up an Event Response Organization during problems like this - to allow an open forum to recommend tests and troubleshooting activities with full consideration of how these activities will affect the plant/health and safety of the public." [emphases added]Despite the intention of Palisades management that "they wanted a visual inspection with no components touched," Palisades employee Chad Main wrote in a memo dated 10/11/05 entitled "Troubleshooting on Spent Fuel Pool Crane L-3" that, per crane manufacturer Ederer's instructions over the phone, "[t]o verify the emergency brake was set, Ederer recommended the nitrogen bottle valve be opened and the brake release moved very slowly to remove a small amount of tension on the brake mechanism.

This was done and the actuator moved approximately 2 mm." Thus, Palisades workers partially overrode the emergency brake on the crane from which dangled a 107 ton cask fully loaded with high-level radioactive waste, which, if dropped, could have caused a

radioactive inferno killing tens of thousands downwind. The NMC document lists "Vague and Incorrect Guidance-IneffectiveCommunication-[and] Over Confidence" caused the human error despite the "sensitivity of the suspended load." The document concludes by saying "The team supervisory oversight was given two days off without pay due to not following the EventResponse Team instruction."(16)

See on the next page a photo of the Palisades cask during its dangle over the irradiated nuclear fuel storage pool, dated 10.11.2005 at 15:03 Eastern Daylight Time.(17) "-the cask was approximately four feet out of the water,"(18) meaning that 11 feet remained underwater. The crane remained stuck and the cask dangled above the irradiated fuel storage pool for 43 hours4.976852e-4 days <br />0.0119 hours <br />7.109788e-5 weeks <br />1.63615e-5 months <br />, from 5:30 amon 10/11/2005 till 12:28 am on 10/13/2005.(see

footnote (26) below, p. 6) Although Palisades site leadership, NRC, NMC's reactor fleet, and the Institute for Nuclear Power Operations (an industry self-policing regulatory body) received word about the cask dangle, the public was kept in the dark for over two months, and documentation was not made available to the public by the NRC till five months later.(19) The cask, weighing about 107 tons, contained 32 irradiated nuclear fuel assemblies, nearly 13 tons worth of high-level radioactive waste.(20) The inspection team must not have read the work order carefully, because it clearly states "[r]emote visual to be performed"and "perform a visual inspection of mechanical equipment associated with the main hoist on l-3 crane." Further on, the Work Order makes explicit "Do not perform any movement of the L-3 crane" and a "WARNING" about the resetting of the emergency brake potentially causing "uncontrolled movement of the Main Hoist drum-" Further warnings that "caution and 7

conservatism during the test or evolution, particularly when uncertainties are encountered-[and] Verification that adequate margins of safety are to be maintained when interlocks and protection systems are bypassed," was also discussed, but went unheeded and were in fact violated. Despite admitting that "a MAE (maintenance avoidable error)" had occurred, Palisades inspection team answered "No" to the question "Are any Action Requests or Lessons Learned Warranted?"(21)So what caused the emergency brake to engage in the first place? Palisades reports that "Prior to 10/11/2005 the EATL [energy absorbing torque limiter] was last set

in August 2005-The as-left set-point was 175 ft-lbs. Review of the work order could not validate that Ederer Procedure 260 was utilized or the EATL was tested for repeatability after its setting was applied. The target set point was 186 +/- 18 ft-lbs

-The low as-found EATL break away torque value confirms that the EATL is a credible cause for the emergency brake actuation. Discussions with the crane vendor Ederer indicated that EATL slippage from a 93 ton load would be equivalent to an EATL set-point of 121 ft-

lbs, which is sufficiently close to the as-found set-point of 140 ft-lbs when consideration is given to the additional loading experienced due to the dynamic affects of a moving

load."(22)"Why did this occur?" NMC asks itself. Its answer: "Review of the work order from August 2005 shows that the break-away torque setting may not have been done correctly. It appears that they may have only adjusted the setting once, yet the procedure in the vendor's manual requires multiple evolutions."(23) Despite the Ederer crane representative's improper instructions to manipulate the emergency brake during the cask dangle, an internal NMC document reveals that "[thecrane] Vendor does not recommend adjusting the EATL with a load suspended" and suggested precautions and "conservative measures consistent with the nature of the load being handled." This document again mentioned that "Big Rock [Point nuclear power plant in Charlevoix, MI] experienced several emergency brake set incidences...The conclusion was that even when the EATL was set within the vendor recommended range (186+/-ft-lbs) the EATL caused emergency brake sets with the crane heavily loaded,"

apparently during emptying of the storage pool of high-level wastes into dry casks as part of Big Rock Point's decommissioning. However, news of those incidents was not, to this author's knowledge, ever made readily accessible to the public or the media by NRC or the company. And it appears the company did not learn lessons from Big Rock Point, at least not sufficiently enough to prevent a repeat of a cask dangle at Palisades. An important question regarding the Big Rock Point cask dangles is, was the Ederer crane there single-failure-proof, and were safety precautions such as the emergency brake

overridden improperly as occurred at Palisades in Oct., 2006?(24) NMC admitted that "The former L-3 crane-main hoist was not designed as single-failure-proof." But "In 2002/2003, NMC modified the L-3 crane to increase the rated load capacity to 110 tons and incorporate single-failure-proof technology." Luckily, this 2006 cask dangle had not occurred several years earlier, for a cask drop would have been much more likely then. Another important question to answer about the Big Rock 8

Point cask dangles in Nov. 2002 - before or after the crane at Big Rock was made single-failure-proof? Did Palisades upgrade its Palisades crane to single-failure-proof because of

its cask dangles at Big Rock Point?(25) In its "Root Cause Analysis Report: Crane operator heard loud noise during lift with L-3 crane," Nuclear Management Company admits that "[t]he EATL [Energy Absorbing Torque Limiter] is the last-line-of-defense for overload." It goes on: "Completion of the annual PM crane inspection activities in August 2005 resulted in the EATL being adjusted. The 'as-found' condition was not recorded at that time. In a telephone interview the vendor representative, who was here in August, indicated the as

found condition on August 5 th was '-well over 200 ft-lbs.' The acceptable setting range of the EATL is 168-204 ft-lbs. With the acceptance criteria not met, the vendor, with the assistance of an inexperienced plant repair person, reset the EATL.The plant PM procedure and the referenced section of the vendor procedure procedure did not contain steps to reset the EATL. The vendor considered the activity to be routine as he had done it several times. Once the EATL had been set within the acceptance criteria, at 175 ft-lbs, the vendor did not proceed to recheck the torque setting as the procedure known to him did not require it, and in his experience, it was not required to verify the setting.According to the vendor, due to the torque imparted on the reduction gearing from themain hoist motor shaft to the hoist drum it requires two to three workers, working in unison on the torque wrench and motor hoist brake, to prevent kickback of the wrench in accomplishing this task."[emphases added](26) Note that only two workers, not three, each with significant gaps in their knowledgeof proper procedure, were assigned to the

job. NMC concluded:"Investigation into the cause for the EATL not being set at 175 ft-lbs, as was recorded in August, identified that plant procedure direction for checking the EATL setting was inadequate and that no direction existed for adjusting the EATL setting. The Plant's Administrative Service Management procedure had not been implemented to ensure a plant staff member understood the activities that the vendor was performing, and that the vendor was made aware of the plants (sic) process and expectations. Other factors also influenced the cause of the EATL being set incorrectly including: The word orders contain deficiencies. Work proceeded beyond what was detailed in the procedure. The plant staff was not knowledgable of the crane components and has relied on the vendor to complete the annual inspection activities. Additionally, error precursors including high heat and humidity and time pressure influenced the outcome."[emphasis added](27)

NMC reports that "A crane vendor representative noted that the EATL is the last-line-of-defense for overload." And the explanation for the 43 hour4.976852e-4 days <br />0.0119 hours <br />7.109788e-5 weeks <br />1.63615e-5 months <br /> dangle before the cask was lowered back down to the pool floor included: "Several factors contributed to thislonger than expected time including a general lack of knowledge related to L-3 operation and components related to its single-failure proof design-The system engineer and backup system engineer, who were involved with the shop testingt and acceptance testing 9

of the crane in 2002 and 2003, are no longer employed by NMC. Ownership of the crane in Outage Management is with its second owner since the beginning of 2004. Maintenance support is dependent on vendor support from Ederer Inc."[emphasis added](28)A number of follow-on mistakes were made. The as-found EATL setting of "well over 200-pounds (sic)-was not recorded in the [Work Order], not was it entered into the corrective action system-Additionally, another Ederer representative brought to the plant on 10/12/05 agreed, based on his experience that repeating the EATL check is needed to verify its setting. A post maintenance test (PMT) of the EATL, as was specified in the work order, requireda verification the EATL works properly. To complete the PMT a validation test should have included a retest of the EATL torque setting. The target set point for the EATL is 186 +/- 18 ft-lbs (168-204 ft-lbs). The work order summary does not maintain a PMT. There is no indication an adequate PMT was done." In addition, "there is no indication in the [Work Order] that the torque wrenches"

used to check the EATL setting were properly calibrated.(29) NMC goes on "The implication in the CAP [corrective action plan] is there is a relationship between setting the EATL and the failure of the diaphragm of the air

canister. This is unclear communication that appears to be due to inadequate knowledge of the crane components-It is apparent that the repair of the air canister was believed to

be, in part, repair of the EATL. This is not the case. This knowledge deficiency resulted in vague direction in the [Work Order]."(30)Again referring to the faulty setting of the EATL in August 2005, NMC reports "No provisions were made fro placekeeping or step signoffs on the vendor procedure and neither were used to denote steps were completed-Since there is no specific direction in either of the [Work Order] documents the conclusion is that the EATL was set only by the experience of the vendor. Work outside the direction of a procedure or work order is

not within our processes-the FIN repair-worker assigned to assist the Ederer

representative on August 5 identified that he was assigned the-task-but had no

knowledge of the crane brake operation as he had not previously worked on the equipment. This was a first time task for the FIN repair-worker. He sat down with the vendor and questioned what the task was, and got some understanding of the job, but was

dependent on the vendor for direction.(31) Citing more crane vendor errors both before and during the cask dangle incident, NMC reports "Deficiencies are evident in meeting all these service manager administrative requirements in August [2005] and additionally in March and during the latest vendor assistance fromEderer on October 12 and 13."(32) The crane vendor and Palisades communicated poorly prior to the August 5, 2005 EATL setting. "This PMT step may not have provided clear direction to the vendor who was unfamiliar with the plant process requirements-This lack of clear direction points to a knowledge deficiency of the EATL operation-There is no indication the checklist was

reviewed on August 5 th when the EATL was reset. Interviews have indicated the PJB [pre-job briefts] was minimal on August 5-The as found condition of the EATL, per 10 interview with the vendor, was the setting '-was well over 200 ft-lbs.' This exceeded the acceptance criteria-Work activities did not stop at this point. The 'as-found' condition was not documented in the work order nor was an action request initiated.

Work proceeded under the direction of the vendor representative to reset the EATL-this was outside the [Work Order] and procedure guidance that was directing the work activity."(33)Heat, humidity, stress, and the desire to finish the job as soon as possible in order to leave for vacation contributed to the errors: "The temperature and humidity up on the crane trolley near the ceiling was said to be very hot and it was humid-The vendor indicated it was hot and that he perspired heavily

while working on the crane trolley that day. There were no ice-vests worn. The FIN repair-worker indicated it was extremely hot and humid and protective clothing was an issue from a heat stress standpoint. The personnel involved did not initiate any actions to address heat stress that would have addressed ice vest requirements or stay time

restrictions if any were needed-The vendor representative had been onsite for two days prior to commencing work on the crane. On August 3 rd the crane was not in the correctspot to access and work was stopped that day. On August 4 th the failed diaphragm, in the air canister that failed on March 17, was replaced. On August 5 th work was done on the crane that includedreplacing the rebuilt air canister and checking and resetting the EATL with the vendor representative. Additionally, on Friday August 5 th the vendor had a plane reservation to return home and it was perceived he was anxious to leave on time as his vacation was to begin."[emphases added](34) NMC admits that "Dependence on vendor experience due to plant staff lack of knowledge may be prevalent in other cases."(35) Incredibly, "Because BRP [Big Rock Point] is not an operating unit it did not submit any operating experience to this issue." This despite a cask dangle at Big Rock on an Ederer crane on Nov. 6, 2002. In addition, "On August 15, 2003 while attempting to lift the BRP Reactor Vessel and place it in a shipping container the Containment Building Crane, an Ederer X-SAM single-failure-proof crane, malfunctioned."(36)Insuring that the public could not demandthat industry learn from these repeated mistakes, NMC reports that "There was no report required to the NRC."(37) This, when false fire alarms and plant management personnel changes are required to be reported to

NRC!Internal NRC emails show that NRC officials were aware of the potential of a cask drop. Magdalena Gryglak wrote to Jamnes Cameron on 10/11/05 that "We were just briefed on the potential cask drop-Based on some older documents, before the crane was upgraded to single failure proof crane, the licensee determined that if the cask were to be dropped, there would be significant damage to the pool and flooding could result-" It seems that the removal of the impact-limiting pads from the bottom of the pool, mentioned above, would only make such pool damage worse.(38) 11 An NMC document reveals that "All unnecessary personnel were removed fromthe Spent Fuel Pool Floor," during the cask dangle.(39) NRC's Mary Jane Ross-Lee, emailing Eric Benner and Carla Roque-Cruz "Re: Palisades," on 10/11/05, shows that Palisades cover up of the incident begin immediately:

"the licensee was trying to find out if this event is reportable or not." It seems in NRC's estimation it was not, given how they helped the company keep the public in the dark for months on end. And, despite Ross-Lee assuring her NRC colleagues that "the cask has its own source of cooling," Nuclear Management Company felt the need to "Take temperature of the cask every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />," as well as to "Collect sample and determine boron concentration of SFP [spent fuel pool] and cask every 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />...[and] Develop plan to sample cask for Boron concentration." Apparently, NRC's exemption on boron concentration safeguards granted just five days earlier lowered safety margins during this

cask dangle incident. NMC also initiated evaluating "the need to whether to refill cask with Spent Fuel Pool Water," apparently to insure adequate boron concentration to

prevent nuclear criticality, and cool enough water to prevent waste fuel overheating.

Other precautions were taken as well, such as ordering that "No Continuous Work allowed in Auxiliary Building." NMC also had personnel "Analyze Worst Case Condition - Dropped Load" and prepared a "L-3 [crane] Contingency: Preparations for Potential Damage w/Heavy Load Drop."(40) In order to ensure that they could observe the lowering of the cask once the stuck crane was addressed, NRC officials made sure that all shifts at Palisades would be covered by an NRC official even throughout the wee hours of the night. However, they failed to report the incident for many months, keeping the public in the dark.(41) Internal NRC emails also expressed concern about boron concentrations. "They should continue to monitor pool boron concentrations-and maintain provisions to identify, mitigate and terminate the consequences of a boron dilution accident as requiredby our exemption-Keep us informed, especially if there are any plans to add makeup

water to the cask." A boron dilution accident could occur if unborated water were added to the pool or dry cask in order to maintain cooling. The risk, however, would be that the unborated water would provide sufficient neutron moderation that a nuclear chain reaction could occur in the still-fissile waste.(42)

It is not entirely clear why NRC reported the cask was "13 feet off the pool floor" during the dangle. Most pools are around 40 feet deep, so the bottom of the cask in that

typical situation would have been 29 feet above the pool floor. The higher above the pool

floor, the more force the cask would have delivered to the pool floor if dropped. NRC

patted itself on the back, saying "The Region-based inspectors and the resident inspectors have been working very well together to provide coverage of this issue." Did they mean "covering up" of this issue, because they kept it quiet for months.(43)On Wednesday, March 29 the Cook nuclear power plant dropped a 35 ton missile block 15 feet onto the reactor cavity floor. Again, NRC held that the incident was not 12 reportable. It would not have been reported by NRC until the next quarterly inspectionreport about Cook. But an anonymous source notified Dave Lochbaum of Union of Concerned Scientists. Lochbaum wrote to the NRC Region III Office of Public Affairs:

"Good Day: An industry colleague informed me about an incident that happened recently at DC Cook Unit 2that's making the rounds inside the industry. I find zero information about it on the NRC's website.Can you confirm any or all of the following:1) On March 29, 2006, a heavy load was dropped at Unit 2 during refueling.

2) The heavy load was a 35-ton missile shield. 3) The load dropped onto the reactor cavity floor. 4) The load was dropped either because of a rigging problem or a crane failure.

5) A "stop work" was issued by the company in response to the incident.6) NRC Region III has had more heavy load drops than any other NRC region in the past 12 months."(44)It is still not clear how much damage was done to the Cook nuclear power plant by this heavy load drop.

In conclusion, with the 20 th anniversary of the Chernobyl nuclear catastrophe approaching on April 26 th, 2006, it is very sobering to realize that Palisades came all too close to a catastrophic radiation release due to near-drop of a 107 ton cask onto its waste storage pool floor. And that the Cook nuclear power plant, just 30 miles south of Palisades on the Lake Michigan shoreline, actually did drop a heavy load near its reactor vessel, with as-yet incomplete damage assessments. When combined with the near melt down at Davis-Besse nuclear power plant near Toledo in 2002, it seems that by the grace of God, or by sheer luck, the Great Lakes

region has dodged a Chernobyl-scale catastrophe on its very shores.

References:

(1)The NRC quarterly inspection report, covering October 1 through December 31, 2005, is entitled "NRC Inspection Report 05000255/2005012" and is available upon request from Kevin Kamps, NIRS, 301.270.6477x14. (2)Document available upon request from Kevin Kamps, NIRS, 301.270.6477x14.

(3)NRC NUREG-1738, "Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants," published February, 2001, Executive Summary, page Roman numeral x.

13 (4)DOE's Feb. 2002 Final Environmental Impact Statement for Yucca Mountain is viewable online. See Table A-7, "Proposed Action spent nuclear fuel inventory,"

and Table A-8, "Inventory Module 1 and 2 spent nuclear fuel inventory," at http://www.ocrwm.doe.gov/documents/feis_2/vol_2/apndx_a/index2_a.htm Calculations, based upon the above tables, showing Palisades' current waste pool inventory available upon request from Kevin Kamps, NIRS, 301.270.6477x14. (5)Robert Alvarez, Jan Beyea, Klaus Janberg, Jungmin Kang, Ed Lyman, Allison Macfarlane, Gordon Thompson, and Frank N. von Hippel, "Reducing the hazards from stored spent power-reactor fuel in the United States, Science & Global Security , Vol. 11, No. 1, 2003, page 1. See the report at: http://www.princeton.edu/%7Eglobsec/publications/pdf/11_1Alvarez.pdfThey report, at page 6: "Inventories of Cs-137 in spent-fuel storage pools. The spent-fuel pools adjacent to most power reactors contain much larger inventories of [Cs-137] than the 2 MegaCuries (MCi) that were released from the core of Chernobyl 1000-Megawatt electric (MWe) unit #4 or the approximately 5 MCi in the core of a 1000-MWe light-water reactor. A typical 1000-MWe pressurized water reactor (PWR) core contains about 80 metric tons of uranium in its fuel, while a typical U.S. spent fuel pool today contains about 400 tons of spent fuel-Furthermore, since the concentration of [Cesium-137] builds up almost linearly with burnup, there is on average about twice as much in a ton of spent fuel as in a ton of fuel in the reactor core." (6)Alvarez et al., page 6, citing "Exposures and effects of the Chernobyl accident,"

Annex J in Sources and Effects of Ionizing Radiation, United Nations, 2000, p.472-475. See http://www.unscear.org/pdffiles/annexj.pdf

.(7)Alvarez et al., p. 6, see immediately above at footnote (6).

(8)NRC NUREG-1738, "Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants," published February, 2001, Appendix 4, Consequence Assessment from ZirconiumFire; Table A4-7, p. A4-9; Table A4-

15, p.A4-15. (9)NRC NUREG-1437, Supplement 27, Generic Environmental Impact Statementfor License Renewal of Nuclear Plants, Regarding Palisades Nuclear Plant, Draft Report for Comment, Feb. 2006, Section 2.2.8.5, Demography, p.2-56, citing Nuclear Management Company, LLC, Applicant's Environmental Report-Operating License Renewal Stage, Palisades Nuclear Plant. Docket No. 50-255.

Covert, Michigan (March, 2005). (10)The NRC quarterly inspection report, covering October 1 through December 31, 2005, entitled "NRC Inspection Report 05000255/2005012," p.9. (11)Hubert T. Bell, NRC Inspector General, "NRC's Regulation of Davis-Besse Regarding Damage to the Reactor Vessel Head," Case No.02-03S, Dec.

30, 2002, at http://www.nrc.gov/reading-rm/doc-collections/insp-gen/2003/02-03s.pdf . (12)NRC event report, Event Number 42053, Event Date 10/12/2005, Event Time 17:09 EDT. This condition persisted till 05:00 10/13/2005, or nearly 12

hours.(13)Letter from L. Mark Padovan, Project Manager, Section 1, ProjectDirectorate III, Divisionof Licensing Project Management, Office of Nuclear 14 Reactor Regulation, NRC, to Mr. Paul A. Harden, Site Vice President, Nuclear Management Company, LLC, Palisades Nuclear Plant, dated Oct. 6, 2006, see

especially p. 4. (14)Palisades "L-3 SFP [Spent Fuel Pool] Crane Operation Action Plan,"

undated.(15)FSAR Chapter 9 - Auxiliary Systems, Fuel Handling and Storage Systems, Revision 24, Page 9.11-1 of 9.11-26, undated. (16)Nuclear Management Company, A/R 01000753: Intent of WO Task Exceeded During Troubleshooting, 4 pages, undated. (17)Pictures of Palisadescask, 7 pages, undated.

(18)NMC, Work Order: L-3 Contingency Dry Fuel Storage 200, 35 pages, undated.(19)NMC, "Root Cause Evaluation Charter: CAP/RCE0100065901 Crane Operator Heard Loud Noise during Lift with L-3 Crane," 1 page, 10/17/2005. (20)See Footnotes (4) and (19) above.

(21)See footnote (18) above.

(22)NMC, Manager Sponsor Darrel Turner, "Validation of Cause/PMT Load Test Considerations w/handwritten notes," 1 page, undated. (23)NMC, A/R 01000980, "Break-away Torque Setting out of Spec (Low) for L-3 SFP Crane," 1 page, undated. (24)NMC, "L-3 Spent Fuel Pool Suspended Load Recovery," 2 pages, undated.(25)NMC, "Maintenance History of L-3, Spent Fuel Pool Crane," 2 pages, undated; also see footnote (26) below, p. 14. (26)NMC, Root Cause Analysis Report, NCE0100065901, "CAP01000659, Crane operator heard loud noise during lift with L-3 crane,"58 pages, undated. (27)See footnote (26), p. 3. (28)See footnote (26) pgs. 6-7.

(29)See footnote (26), p. 8.

(30)See footnote (26), p. 9.

(31)See footnote (26), p. 10.

(32)See footnote (26), p. 11-12.

(33)See footnote (26), p. 11.

(34)See footnote (26), p. 12.

(35)See footnote (26), p. 16.

(36)See footnote (26), p. 18.

(37)See footnotes (26), p. 20. (38)Email, M. Gryglak to J. Cameron,

Subject:

Palisades crane, 2 pages, 10/11/05.(39)Action Request Report, Number 01000626, Crane operator heard loud noise during lift with L-3 crane, 1 page, 10/11/2005. (40)NRC email, and NMC, Event Response Plan: L-3 Spent Fuel Pool Crane, 5 pages, 10/12/05. (41)Email, J. Ellegood to C. Lipa/J. Cameron/L.M. Padovan,

Subject:

Schedule for lowering load, 2 pages, 10/12/2005.

15 (42)Email from Robert Taylor to Mark Padovan regarding Palisades Dry Storage Cask Suspended in Air, 1 page, 10/13/05. (43)Email, J. Cameron to M. Phillips,

Subject:

FOR YOUR ACTION: Palisades EDO Bullet, 3 pages, 10/14/05. (44)Email from Dave Lochbaum, UCS, to NRC RIII OPA, 3/30/2006; Hugh McDiarmid, Jr., "Concrete shield falls in nuclear plant mishap," March 31, 2006.

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