ML18155A396

Lacrosse - Transmittal of Decommissioning Plan and Post-Shutdown Decommissioning Activities Report (D-Plan/PSDAR) Changes

ML18155A396
Person / Time
Site:	La Crosse  File:Dairyland Power Cooperative icon.png
Issue date:	05/23/2018
From:	Gerard van Noordennen LaCrosseSolutions
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards
Shared Package
ML15155A395	List: ML18155A396
References
LC-2018-0036
Download: ML18155A396 (43)

Text

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LACROSSESOLUTIONS 10 CFR 50.4(b) 10 CFR 50.71(e)(4)

May 23, 2018 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 La Crosse Boiling Water Reactor Facility Operating License No. DPR-45 NRC Docket Nos. 50-409 and 72-046 LC-2018-0036

Subject:

La Crosse Boiling Water Reactor (LACBWR) Submittal of Decommissioning Plan and Post-Shutdown Decommissioning Activities Report (D-Plan/PSDAR) Changes

References:

(1)

LaCrosseSolutions, LLC letter, "Notification of Amended Decommissioning Plan and Post-Shutdown Decommissioning Activities Report for La Crosse Boiling water Reactor," dated June 27, 2016 (2)

NRC Regulatory Issue Summary 2015-17, "Review and Submission of Updates to Final Safety Analysis Reports, Emergency Preparedness Documents, awl Fire Protection Documents" The LACBWR Decommissioning Plan / Post Shutdown Decommissioning Activities Report (D-Plan/PSDAR) is recognized as the LACBWR Safety Analysis Report (SAR) equivalent as it contains the plant post-fuel accident analysis. Therefore it is subject to revisions being submitted to the NRC on a 24 month interval in accordance with 10 CFR 50.71(e)(4).

In accordance with the requirements of 10 CFR 50.4, "Written Communications," paragraph (b)(6), LaCrosseSolutions, LLC is submitting the November 2017 revision of the D-Plan/PSDAR for LACBWR. In accordance with 10 CFR 50.71(e)(4), the update is being submitted within 24 months of the previous revision which was submitted in Reference 1.

This D-Plan/PSDAR update is being provided as a;total replacement on CD, satisfying the electronic submittal requirements of 10 CFR 50.4(b )( 6) as discussed in Reference 2.

The changes to the D-Plan/PSDAR reflect administrative changes (i.e., editorial and D-Plan/PSDAR text changes) and plant design changes. The November 2017 D-Plan/PSDAR revision includes changes made from June 2016 through May 2018. contains a summary of the D-Plan/PSDAR changes. Attachment 2 provides the November 2017 revision of the 0:-Plan/ PSDAR in CD format. It is submitted as a complete replacement and is not accompanied by a list of effective pages or page change instructions.

Revision bars are provided to identify changes.

S4601 State Highway 35, Genoa, WI 54632

LaCrosseSolutions LC-2018-0036 Page2 of2 This submittal contains a Solutions Proprietary Financial Information Affidavit pursuant to 10 CFR 2.390. The Affidavit sets forth the basis for which the information may be withheld from pub,lic disclosure by the Commission and addresses with specificity the considerations listed in 10 CFR 2.390(b)(4). All documents within the scope of this affidavit are marked as

. "withhold from public disclosure under 10 CPR 2.3 90."

Section 3 of the D-Plan/PSDAR tontains the proprietary financial information Solutions is providing to the NRC and seeks to have withheld from public disclosure in its entirety. of this submittal contains a redacted version ofD-Plan/PSDAR Section 3 for public disclosure. Attachment 4 contains a preflight status report identifying any noncompliance issues with the electronic submittal guidance that could not be resolved due to the source of the information being submitted. All submitted information is in legible format.

As Vice President of Regulatory Affairs, I certify that the information in this submittal accurately presents changes made since the previous submittals necessary to reflect information and analyses submitted to the NRC or prepared in accordance with NRC requirements.

There are no regulatory commitments contained in this submittal. If you have any questions regarding this letter, please contact Jim Ashley at (224) 789-4017.

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Respectfully, r,/1

. 1.

G~~v~

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Gerard van Noordennen Vice President Regulatory Affairs Attachments:

1) Summary of Changes

2) Attachment 1, LACBWR D-Plan/PSDAR, November 2017

3) Attachment 2, LACBWR D-Plan/PSDAR Section 3* Redacted Version

4) CD contains Attachment 2 & 3 PDF files and Preflight Status Report cc:

Marlayna Vaaler, U.S. NRC Project Manager Regional Administrator, U.S. NRC, Region III Service List (Attachment 2 - Section 3 not included)

LaCrosseSolutions, LLC PROPRIETARY FINANCIAL INFORMATION AFFIDAVIT Affidavit of Gerard van Noordennen, Vice President Regulatory Affairs, LaCrosseSolutions, LLC.

D-Plan/PSDAR Section 3, contained in Attachment 2 of this.submittal, consists of proprietary financial information that LaCrosseSolutions, LLC considers confidential. Release of this information would cause irreparable harm to the competitive position ofLaCrosseSolutions,

. LLC. The basis for this declaration is:

i.

This information is owned and maintained as proprietary by LaCrosseSolutions, LLC, ii.

This information is routinely held in confidence by LaCrosseSolutions, LLC and not disclosed to the public, iii.

This information is being requested to be held in confidence by the NRC by this

petition, iv.

This information is not available in public sources,

v.

This information would cause substantial harm to LaCrosseSolutions, LLC ifit were released publicly, and v1.

The information to be withheld is being transmitted to NRC in confidence.

I, Gerard van Noordennen, being duly sworn, state that I am the person who subscribes my name the foregoing statement, I am authorized to execute the Affidavit on behalf of

• LaCrosseSolutions, LLC, and that the matters and facts set forth in the statement are true to the best of my knowledge, information, and belief.

Name:

Title:

Company:

Gerard van Noordennen Vice President Regulatory Affairs LaCrosseSolutions, LLC SUBSCRIBED AND SWORN TO BEFORE ME Notary Public LINOACHOU Ofllclll SHI Notl~y Public

• Stitt ofllllnois My Commission Expires Dec 7, 2019

(

L La Crosse Boiling Water Reactor Service List

'cc:

. KenRobuck Group President Disposal and Decommissioning Energy Solutions 299 South Main Street, Suite 1700 Salt Lake City, UT 84111 John Sauger Executive VP and Chief Nuclear Officer ReactorD&D Energy Solutions

• 121 W. Trade Street, Suite 2700 Charlotte, NC 28202 Gerard van Noordennen VP Regulatory Affairs Energy Solutions 121 W. Trade Street, Suite 2700 Charlotte, NC 28202 Joseph Nowak General Manager LaCrosseSolutions S4601 State Highway 35 Genoa, WI 54632-8846 Russ Workman General Counsel Energy Solutions 299 South Main Street, Suite 1700 Salt Lake City, UT 84111 Jerome Pedretti, Clerk Town of Genoa E860 Mundsack Road Genoa, WI 54632 Jeffery Kitsembel Division of Energy Regulation Wisconsin Public Service Commission P.O. Box 7854

• Madison, WI 53707-7854 Paul Schmidt, Manager Radiation Protection Section Bureau of Environmental.and Occupational Health Divisfon of Public Health Wisconsin Department of Health Services P.O. Box 2659 Madison, WI 53701-2659 Barbara Nick President and CEO Dairyland Power Cooperative 3200 East Avenue South La Crosse, WI 54602-0817

• Cheryl Olson, ISFSI Manager La Crosse Boiling Water Reactor Dairyland Power Cooperative S4601 State.Highway 35 P.O. Box 817 Genoa, WI 54632-8846

. Lane Peters La Crosse Boiling Water Reactor Dairyland Power Cooperative S4601 State Highway 35 Genoa, WI54632-8846 Thomas Zaremba Wheeler, Van Sickle and Anderson, S.C.

44 East Mifflin Street, Suite 1000

.Madison, WI 53703 John E. Matthews Morgan, Lewis & Bockius LLP 1111 Pennsylvania A venue, NW Washington, DC 20004

Revised Sections Section 1.3 Section 1.4 Section 1.5 Section 2.1 L

ATTACHMENT 1 Summary of Changes Description of Change This section was revised to eliminate discussion of systems and components no longer in use. The dismantlement discussion for abandoned systems is located in the LACBWR License Termination Plan (LTP). System discussion for systems still in use has been modified to reflect current site conditions.

This Section was revised to eliminate the demolition discussion for buildings and structures being decommissioned. The demolition discussion for these buildings is located in the LACBWR LTP.

This section was revised to eliminate select decommission activity radiological impact discussion that is redundant with that contained in the LACBWR LTP.

This section was revised to make minor revisions to select decommissioning schedule milestone dates.

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WITHHOLD SECTION 3 FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390, ATTACHMENT 2 LACBWR D-Plan/PSDAR November 2017

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WITHHOLD SECTION 3 FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390

LACBWR Site Restoration Project Decommissioning Plan and Post Shutdown De~omm.issioning Activities Report Revision: November 2017 Preparer (Print name/Sign) 11-20-17 Secondary Reviewer (Print Name/Sign)

James Ashley /~___Pate:

ArthureR.A{ams/via:::z::7 Date; --.-1--2-1--17--

~gulatory Affairs assigned program & regulatory reviews ('~new oniy*): Initials/Date NI~

Regulatory Reqwed Reviews (attach completed LC-RAPPR-001 and QTR forms, as applicable}

Part 50 License: 10 CFR 50.59 and 50.90 lg] YES ONO Fire Protection: IO CPR 50.48(t)

DYES l8J NO Conditions of License: PSP: 10 CFR 50.54{p)

DYES

~NO Conditions ofLiceiille: E-Plan: 10 CFR 50.54(q)

DYES i81NO Termination of License: IO CFR 50.S2(a)(6) and 50.82(a)(7) l8J YES ONO Part 72 License: 10 CFR 72.48 DYES jgjNO RP: 0 YES ~ NO SIGNATURE_. ___________

DATE:. ___ _

QA: 0 YES jg] NO SIGNATURE ____________ _,DATE: ____ _

QTR: l&l YES ONO SIGNATURE ArthurR.Adams/signa.tureviaemail DATE: 11-21-17 Approv:.l Settion PROJECT MANAGER:

Effective Date: ~((~n Lil_ (assigned by Page 1 ofll

I L

D!EC(Q)M~~SS~<Ou\\m\\4G PLAN AND PO$T.. SHl/JTD0WN p)!ECOMMISS~OIM~~G ACTIV~1r~1ES ~EPORT PREPAREDBYJamesAshley/r~

• DATE 11-20-17 REVIEWED BY Arthur R. Adams/ (signature via email}

Qualified Technical Reviewer -

DATE 11-21-17

TABLE OF CONTENTS Section Page

1.0 DESCRIPTION

OF PLANNED DECOMMISSIONING ACTIVITI.ES............................ 1-1 1.1. INTRODUCTION.......................................................................................................... 1-1 1.1.1. Historical Selection Of Decommissioning Method**************************************;*********** 1-1 1.1.2. Current Selection Of Decommissioning Method.............................................. :****** 1-3 1.2. SIGNIFICANT POST-SHUTDOWN UC.ENSING ACTIONS......................................,. 1-3 1.3. DISMANTLEMENT OF SYSTEMS AND COMPONENTS........................................... 1-4 1.3.1. Well Water System....................................*........................................................... 1-5 1.3.2. Liquid Waste Collection Systems..................... :..................................................... 1-5 1.3.3. Airborne Release Monitoring.......... *.......................................... ***********................... 1-5 1.4.. BUILDINGS AND STRUCTURES............................................................................... 1-5 1.5. RADIOLOGICAL IMPACTS OF.DECOMMISSIONING ACTIVITIES........................... 1-6 1.5.1.. Control Mechanisms to,Mitigate the Recontamination of Remediated Areas......... 1-6 1.5.2. Occupational Exposure.......................................................................................... 1-6 1.5.3. Exposure to the Public........................................................................................... 1-6 1.5.4. Radioactive Waste Projections......................................... :.................................... 1-7 1.6. GROUNDWATER........................................................................................................ 1-7 1.7. ISFSI DECOMMISSIONING........................................................................................ 1-8 1.8. REFERENCES.................................... :............................ :.......................................... 1-9 2.0 SCHEDULE.................................... *............................................................................. 2-1 2.1. SCHEDULE ACTIVITIES COMPLETED PRIOR TO LICENSE TRANSFER TO SOLUTIONS.................... '............................................................................................ 2-1 2.2. REVISED SOLUTIONS SCHEDULE........................................................................... 2-2 3.0 ESTIMATE OF EXPECTED DECOMMISSIONING COSTS........................................ 3-1 3.1. INTRODUCTION............................................................... :......................................... 3-1 3.2. HISTORICAL PERSPECTIVE..................................................................................... 3-1 3.3. PREVIOUS DAIRYLAND COST ESTIMATES...................... ;............................ :......... 3-1 3.4. SOLUTIONS DECOMMISSIONING COST ESTIMATE............................................... 3-3 3.5. COST ESTIMATE DESCRIPTION AND METHODOLOGY......................................... 3-3 3.6.

SUMMARY

OF THE SITE-SPECIFIC DECOMMISSIONING"COST ESTIMATE......... 3-4 3.7. ISFSI MANAGEMENT......................................................................... :........................ 3-4 3.8.

• SITE RESTORATION COSTS.................................................................. :................. 3-4 3.9. DECOMMISSIONING FUND..................................................................... ;................. 3-5 3.10. REFERENCES..........................................................,................................................. 3-5 4.0 PLANT POST-FUEL ACCIDENT ANALYSIS.................... '......................................... 4-1 4.1. OVERVIEW............................................................... :................................................. 4-1 4.2. RADIONUCLIDE RELEASE LIMITS APPLIED IN ANALYSIS................................'..... 4-1 D-PLAN / PSDAR 1-i November 2017 I L_

TABLE OF CONTENTS 4.2.1. Limits Applied to Postulated Airborne Release............... :...................................... 4-1 4.2.2. Limits Applied to Postulated Liquid Releases......................................................... 4-3 4.3. POST-FUEL ACCIDENT ANALYSIS ASSUMPTIONS................................................ 4-3 4.3.1. Assumptions - Remaining Non-lSFSI-Related Radioactive Source Term.............. 4-3 4.3.1.1. Portion of Total Radioactivity Assumed Releasable Via the Airborne Pathway 4-4 I

4.3.1.2. Portion of Total Radioactivity Assumed ReJeasable Via the Liquid Pathway.... 4-5 4.3.2. Additional Assumptions - Postulated Airborne Release......................................... 4-6 4.3.2.1. Genoa-3 (G-3) Office Building Occupancy....................................................... 4-6 4.3.2.2. Terrain Height Above Grade............................................................................ 4-6 4.3.2.3. crv and cr2 at Distances Less Than 100 Meters........................................... :..... 4-6 4.3.2.4. Pasquill Stability Class..................................................................................... 4-6 4.3.2.5. Rem vs Rad..................................................................................................... 4-6 4.3.2.6. Thyroid Dose................................................................................................... 4-6 4.3.2.7. Correction Factor (CF) for G-3 Office Building................................................. 4-6 4.3.2.8. Radionuclide Data............................................................................................ 4-7 4.3.2.9. Atmospheric Release Inputs............................................................................ 4-7 4.3.3. Additional Assumptions - Postulated Liquid Release......,....................................... 4-7 4.3.3.1. Retention Tank Release, Dilution, and Mixing.......,.......................................... 4-7 4.3.3.2. Duration of Retention Tank Rupture Release to Thief Slough.......................... 4-8

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4.3.3.3. Thief Slough and G-3 Outfall Flow................................................................... 4-8 4.3.3.4.- Liquid Release Inputs...................................................................................... 4-8 4.4.

SUMMARY

OF ANALYSIS RESULTS.......,................................................................. 4-8 4.4.1. Postulated Airborne Release................................................................................. 4-8 4.4.2. Postulated Liquid Release..................................................................................... 4-9 4.5. RADIOLOGICAL OCCUPATIONAL SAFETY..................... :........................................ 4-9 4.6. OFFSITE RADIOLOGICAL EVENTS......................................................................... 4-10 4.7. NON-RADIOLOGICAL EVENTS............................................................................... 4-10 4.8. REFERENCES.......................................................................................................... 4-10 5.0 ENVIRONMENTAL IMPACT........ *............................................................................... 5-1

• D-PLAN / PSDAR 1-ii November 2017 I

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1.0 DESCRIPTION

OF PLANNED DECOMMISSIONING ACTIVITIES 1.1.

INTRODUCTION The La Crosse Boiling Water Reactor (LACBWR) was,a 50 Megawatt Electric (MWe) BWR that is owned by Dairyland Power Cooperative (Dairyland). This unit, also known as Genoa 2, is.

located on the Dairyland Genoa site on the east shore of the Mississippi River south of the Village of Genoa, Vernon County, Wisconsin.

The site is licensed under Possession Only License No. DPR-45 with Docket Numbers of 50-409 for LACBWR and 72-046 for the Independent Spent Fuel Storage Installation (ISFSI).

LACBWR has been shut down since 1987 and is currently undergoing decommissioning. The spent nuclear fuel stored in the LACBWR ISFSI will be maintained under an amended Part 50 license.

There are 333 spent fuel assemblies stored in five NAC-MPC dry cask storage systems at the onsite Independent Spent Fuel Storage Installation (ISFSI). DPC currently expects the fuel to remain onsite until a federal repository, offsite interim storage facility, or licensed temporary monitored retrievable storage facility is established and ready to receive LACBWR fuel.

1.1.1.

Historical Selection Of Decommissioning Method The "Generic Environmental Impact Statement (GEIS) on Decommissioning of Nuclear Facilities," NUREG-0586, Supplement 1, evaluates the environmental impact of three methods for decommissioning. The Supplement updates information in the 1988 GEIS and discusses the three decommissioning methods; a short summary of each follows:

DECON is the alternative in which the equipment, structures, and portions of a facility and site containing radioactive contaminants are removed or decontaminated to a level that permits the property to be released for unrestricted use shortly after cessation of operations.

SAFSTOR is the alternative in which the nuclear facility is placed and maintained in such condition that the nuclear facility can be safely stored and subsequently decontaminated (deferred decontamination) to levels that.permit release for unrestricted use.

ENTOMB is the alternative in which radioactive contaminants are encased in a structurally long-lived material, such as concrete. The entombed structure is appropriately maintained and continued surveillance is carried out until the radioactivity decays to a level permitting unrestricted release of the property~ This alternative would be allowable for nuclear facilities contaminated with relatively short-lived radionuclides such that all contaminants would decay to levels permissible for unrestricted use within a period on the order of 100 years. For a power reactor, the choice was either DE CON or SAFSTOR. Due to some of the long-lived isotopes in the reactor vessel and internals, ENTOMB alone was not an allowable alternative. under the original proposed rule.

Following plant shutdown, the choice between SAFSTOR and DEGON was based on a variety of factors including availability of fuel and waste disposal, land use, radiation exposure, waste volumes, economics, safety, and availability of experienced personnel, Each alternative had advantages and disadvantages. The best option for a specific plant was chosen based on an evaluation of the factors involved.

D-PLAN / PSDAR 1-1 November 2017 I

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1.0 DESCRIPTION

OF PLANNED DECOMMISSIONING ACTIVITIES (cont'd)

The overriding factor affecting the decommissioning decision for LACBWR at the time of shutdown was that a federal repository was not expected to be available for fuel storage in the foreseeable future. With the fuel in the Fuel Element Storage Well, the only possible decommissioning option was SAFSTOR. Limited decontamination and dismantling of unused systems could be performed during this period.

There were other reasons to choose the SAFSTOR alternative. The majority of piping radioactive contamination was Co-60 (5.27 year half-life) and Fe-55 (2. 7 year half-life). If the plant was placed in SAFSTOR for 50 years, essentially all the Co-60 and Fe-55 would have decayed to stable elements. Less waste volume would be generated and radiation doses to personnel performing the decontamination and dismantling activities would be significantly lower. Therefore, delayed dismantling supported the ALARA (As Low As Reasonably Achievable) goal. The reduction iri dismantling dose would exceed the dose the monitoring crew received during the SAFSTOR period.

The shutdown of LACBWR occurred before the full funding for DEGON was acquired. The SAFSTOR period has permitted the accumulation of the full DEGON funding. The majority of studies showed that while the total cost of SAFSTOR with delayed DEGON was greater than immediate DEGON, the present value was less for the SAFSTOR with delayed DEGON option.

The main disadvantage of delayed DEGON was that the plant would continue to occupy the land during the SAFSTOR period. The land could not be released for other purposes. DPC also operates a 350 MWe coal-fired power plant on the site. Due to the presence of the coal-fired facility, DPC would continue to occupy and control the site, regardless of the nuclear plant's status. Therefore, the continued commitment of the land to LACBWR during the SAFSTOR period was not a significant disadvantage.

J A second disadvantage of delaying the final decommissioning was that the people who operated the plant would not be available for the DEGON period. When immediate DEGON is selected, some of the experienced plant staff would be available for decommissioning and dismantlement activities. When SAFSTOR is chosen, efforts must be made to maintain excellent records to compensate for the lack of staff continuity.

The remaining factor was safety.. As of October 2009, 24 power reactors have been shut down in the United States, 11 of which have been fully dismantled and decommissjoned. Experience has shown that the process can be performed safely.

The NRC issued its Waste Confidence Decision in August 1984 as codified in 10 CFR 51.23.

Amended in December 2010, the NRC has found "reasonable assurance that, if necessary, spent fuel generated in any reactor can be stored safely and without significant environmental impacts for at least 60 years beyond the expiration of that reactor's operating license at that reactor's spent fuel storage basin, or at either onsite or offsite independent spent fuel storage installations." Therefore, DPC's plan to maintain the spent fuel at LACBWR, until a federal repository, interim storage facility, or licensed temporary monitored retrievable storage facility is ready to accept the fuel, is acceptable from the safety standp*oint, as well as necessary from the practical standpoint.

After evaluating the factors involved in selecting a decommissioning alternative, DPC decided to choose an approximate 30-50 year SAFSTOR period, followed by DEGON. After 25 years in SAFSTOR and all spent fuel in dry cask storage at the ISFSI,* LACBWR,is now beginning the final decommissioning and dismantlement phase.

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D-PLAN / PSDAR 1-2 November 2017 I

1.0 DESCRIPTION

OF PLANNED DECOMMISSIONING ACTIVITIES (cont'd) 1.1.2.

Current Selection Of Decommissioning Method In a letter dated October 8, 2015 (Reference 1 ), Dairyland and LaCrosseSolutions, LLC.

(Solutions) requested Nuclear Regulatory Commission (NRC) consent to transfer Dairyland's possession, maintenance and decommissioning authorities, under Possession Only License No.

DPR-45, from Dairyland to Solutions. Following NRC approval, as documented in a letter dated May 20, 2016 (Reference 2), the LACBWR site is being decontaminated and dismantled in accordance with the DEGON alternative.

_1.2.

SIGNIFICANT POST-SHUTDOWN LICENSING ACTIONS DPC's authority to operate LACBWR under Provisional Operating License DPR:-45, pursuant to

,

• 10 CFR Part 50, was terminated by License Amendment No. 56, dated August 4, 1987 (Reference 3), and a possess but not operate status was granted. The Decommissioning Plan was submitted December 1987 (Reference 4) with a chosen decommissioning alternative of SAFSTOR. License Amendment No. 63, dated August 18, 1988 (Reference 5), amended the Provisional Operating License to Possession-Only License DPR-45 with a term to expire March 29, 2003.

The NRC directed the licensee to decommission the facility in its Decommissioning Order of August 7, 1991 (Reference 6). License Amendment No. 66, issued with the Decommissioning Order provided evaluation and approval of the proposed Decommissioning Plan, post-operating Technical Specifications, and license renewal to accommodate the SAFSTOR period for a term to expire March 29, 2031.

The Decommissioning Order was modified September 15, 1994 (Reference 7), by Confirmatory Order to allow DPC to make changes in the facility or procedures as described in the Safety Analysis Report, and to conduct tests or experiments not described in the Safety Analysis Report, without prior NRC approval, if a plant-specific safety and environmental review procedure cqntaining similar requirements as specified in 1 O _CFR 50.59 was applied'.

The Initial Site Characterization Survey for SAFSTOR was completed and published October 1995 (Reference 8).

License Amendment No. 69, containing the SAFSTOR Technical Specifications, was issued April 11, 1997 (Reference 9). This amendment revised the body of the license and the Appendix A, Technical Specifications. The _changes to _the license and.Technical Specificati_ons were structured to reflect the permanently defueled and shutdown status of the plant. These changes deleted all requirements for emergency electrical power systems and maintenance of containment integrity.

The LACBWR Decommissioning Plan was considered the PSDAR. The PSDAR public meeting was held on May 13, 1998.

License Amendment No. 71 was issued January 25, 2011 (Reference10), making changes to the LACBWR license Appendix A, Technical Specifications in support of Jhe Dry Cask Storage Project. The amendment revised the definition of FUEL HANDLING, reduced the minimum water coverage over stored spent fuel from 16 feet to 11 feet, 6% inches, and made a _small

• number of editorial changes to clarify heavy load controls and reflect inclusion of the cask pool as part of an "extended" Fuel Element Storage Well. The intent of these changes was to facilitate efficient dry cask storage system loading operations and reduce overall occupational dose to personnel during these operations. All spent fuel assemblies and fuel debris were placed in dry cask storage in the ISFSI in September 201.2.

D-PLAN / PSDAR 1-3 November 2017. I

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1.0 DESCRIPTION

• oF PLANNED DECOMMISSIONING ACTIVITIES (cont'd) \\

The November 2012. revision of the Decommissioning Plan establish~d information in a Post-Shutdown Activities Report (PSDAR) format to provide stakeholders a better understanding of the current status of the decommissioning effort at LACBWR and the planned dismantlement activities.

License Amendment No. 72 was issued July 31, 2013 (Reference 11 ), revising certain license conditions and removing Technical Specification definitions, operational requirements, and specific design requirements for wet spent fuel storage that are no longer applicable with all spent fuel in dry cask storage. The changes removed administrative control requirements that have been relocated to the LACBWR Quality Assurance Program Description or have been superseded by regulation or other guidance. Changes to the body of the license reflected revision of and exemptions granted for the Physical Security Plan described in License Condition 2.C.(3). The body of the license was also.revised to describe ongoing changes to the Fire Protection Program described in-License Condition 2.C.(4). The amendment reduced Technical Specifications to simply two design feature items: one a description of the licensed facility, and the other a declaration that a maximum of 333 spent fuel assemblies are stored in 5 dry casks within an Independent Spent Fuel Storage Installation (ISFSI). The declaration of fuel storage at the ISFSI includes the commitment that spent fuel assemblies shall not be placed in the Fuel Element Storage Well.

On May 20, 2016, The NRC approved the transfer of Dairyland's possession, maintenance and decommissioning authorities, under Posse.ssion Only License No. DPR-45, from Dairyland to Solutions (Reference 2).

In a letter dated March 31, 2016, the NRC withdrew the September 15, 1994 Confirmatory Order modifying the NRC Order Authorizing Decommissioning of facility for LACBWR (Reference 14)

The License Termination Plan (L TP) for LACBWR (Reference 12) was submitted for NRC revi1aw in June 2016. It details final decommissioning and dismantlement activities including site remediation and survey of residual contamination.

The June 2016 revision of the Decommissioning Plan and Post-Shutdown Activities Report (DP/PSDAR) identifies DECON activities, a revised decommissioning schedule, and an updated cost estimate provided by Solutions.

The November 2017 revision of the Decommissioning Plan and Post-Shutdown Activities Report (DP/PSDAR) eliminates discussion of systems no longer in use and eliminates redundant demolition discussion that is contained in the L TP.

1.3.

DISMANTLEMENT OF SYSTEMS AND COMPONENTS Significant dismantlement had already been accomplished at the time of license transfer *to Solutions. In excess of 2 million pounds of metallic waste had been removed, shipped, and disposed of in addition to the Reactor Pressure Vessel (RPV) and spent fuel storage racks.

Removal and disposal of the RPV included disposition of irradiated hardware and all other Class B and C waste.

Waste stored in the Fuel Element Storage Well (FESW) was processed and collected with other Class B/C waste (i.e., resins, filters, and waste barrel contents) and packaged in three liners that were shipped for disposal in June 2007. The RPV containing the reactor internals and 29 control rod blades was filled with low-density cellular concrete with the reactor head installed.

Attachments to the RPV were removed and all other appurtenances were cut. The RPV was D-PLAN / PSDAR 1-4 November 2017

1.0 DESCRIPTION

OF PLANNED DECOMMISSIONING ACTIVITIES (cont'd) removed from the Reactor Building and was shipped for disposal in June 2007. After all spent fuel assemblies and fuel debris were placed in dry cask storage in the ISFSI in September 2012, the storage racks and installed components were removed from the FESW.

(

The remaining systems being decont_aminated and dismantled by Solutions in accordance with the DE CON method are described in the LACBWR L TP. Systems still in use are described below. Systems still in use will only be used until no longer necessary as decommissioning progresses. Their removal from service will be evaluated and controllec;j via'lhe appropriate review process.

1.3.1.

Well Water System Water for this system was supplied from two sealed submersible deep well pumps that took suction through stainless steel strainers, and discharged into integrated pressure tanks. The Well Water System has been isolated from the Reactor/Turbine building. Well water pump #3 supplies potable water to the Administration Building. This system is maintained in continuous operation.

-1.3.2.

Liquid Waste Collection Systems The only remaining liquid effluent releases consist of water collected during decommissioning activities in radiological controlled areas. This water is collected and processed through a temporary filtration and demineralizer system prior to storage in a monitor tank. The monitor tank is batch released to the Mississippi River via the G-3 circulating water discharge pathway following sampling in accordance with ODCM and procedural requirements.

1.3.3.

Airborne Release Monitoring Air samplers will be used to monitor airborne activity during decommissioning activities in radiologically controlled areas. Temporary HEPA ventilation systems utilized in radiological controlled areas and discharged to the outside environment in support of decommissioning activities_ will also be monitored via air samplers. Samples will be collected and analyzed in accordance with ODCM and procedural requirements.

1.4.

BUILDINGS AND STRUCTURES Located within the radiological controlled area of LACBWR are/ were the following buildings and structures. These buildings and structures, with the exception of the Cribhouse, will be or have already been demolished and disposed of.

Reactor Building

• Turbine Building and Turbine Office Building

• Waste Treatment Building Low Specific Activity (LSA) Storage Building

• Cribhouse Maintenance Eat Shack Underg'round Gas Storage Tank Vault 1 B Diesel Generator Building

• Ventilation Stack Solutions plans for the decontamination, dismantlemeQt and anticipated end-state condition(s) for the identified site structures are presented in the LACBWR LTP. The methods to remediate

• D-PLAN / PSDAR 1-5 November 2017

1.0 DESCRIPTION

OF PLANNED DECOMMISSIONING ACTIVITIES (cont'd) contaminated structures, systems, and equipment do not involve any unique safety or remediation issues.

1.5.

RADIOLOGICAL IMPACTS OF DECOMMISSIONING ACTIVITIES The decommissioning activities described in the LACBWR L TP are and will be conducted under the provisions of the Solutions Radiation Protection Program and Radioactive Waste Management Program. The Radiation Protection Program and written site procedures are intended to provide sufficient information to demonstrate that decommissioning activities will be performed in accordance with 10 CFR 19, Notices, Instructions And Reports To Workers, and 10 CFR 20, Standards For Protection Against Radiation and to maintain radiation exposures As Low As Reasonably Achievable (ALARA). The Radioactive Waste Management Program controls the generation, characterization, processing, handling, shipping, and disposal of radioactive waste in accordance with the approved Radiation Protection Program, Process Control Program, and written plant procedures.

The current Radiation Protection Program, Waste Management Program, and Radiological Effluent Monitoring and Offsite Dose Calculation Manual (ODCM) will be used to protect the workers and the public, as applicable, during the various decontamination and decommissioning activities. These well-established programs are routinely inspected by the Nuclear Regulatory Commission (NRC) to ensure that workers, the public, and the environment are protected during facility decommissioning activities.

Continued application of the current and future Radiation Protection and Radiological Effluent Monitoring Programs at LACBWR ensures public protection in accordance with 10 CFR 20 and 10 CFR 50, Appendix I. Radiological Environmental Monitoring Program (REMP) reports for LACBWR to date conclude that the public exposure as a result of decommissioning activities is bounded by the evaluation in NUREG-0586 (Reference 13), which concludes the impact is minimal.

1.5.1.

Control Mechanisms to Mitigate the Recontamination of Remediated Areas Due to the large scope of remaining structures and systems that will be decontaminated and

• dismantled, FRS of areas may be performed in parallel with decommissioning activities.

Consequently, a systematic approach will be employed to ensure that areas are adequately remediated prior to performing FRS and ongoing decommissioning activities do not impact the radiological condition of areas where compliance with the unrestricted release criteria as specified in 10 CFR 20.1402 has been demonstrated. These measures and mechanisms are described in L TP Chapter 5.

1.5.2.

Occupational Exposure The total radiation exposure estimate for remaining decommissioning activities is discussed in L TP Chapter 3.

1.5.3.

Exposure to the Public Continued application of Radiation Protection, Radioactive Waste, Radiological Effluent Technical Specification and Radiological Environmental Monitoring Programs assures public protection in accordance with 10 CFR 20 and 10 CFR 50, Appendix I.

D-PLAN / PSDAR 1-6 November 2017 I

1.0 DESCRIPTION

OF PLANNED DECOMMISSIONING ACTIVITIES (cont'd) 1.5.4.

Radioactive Waste Projections The Radioactive Waste Management Program is used to control the characterization, generation, processing, handling, shipping, arid disposal of radioactive waste during decommissioning. Activated and contaminated systems, structures, and components represent the largest volume of low ievel radioactive waste expected to be generated during decommissioning. Other forms of waste generated during decommissioning include:.

Contaminated water; Used disposable protective clothing; Expended abrasive and abso.rbent materials; Expended resins and filters; Contamination control materials (e.g., strippable coatings, plastic enclosures); and Contaminated equipment used in the decommissioning process.

Details of the radioactive wa~te disposal.activities are discussed in L TP Chapter 3.

1.6.

GROUNDWATER Groundwater characterization is a requirement of decommissioning nuclear power plants and although each station may have varying degrees of investigations, techniques, and modeling efforts, the process is similar. Haley & Aldrich, Inc. was contracted to build the conceptual site model (CSM) for the LACBWR site with respect to the potential release of radiological and chemical materials to the environment. The Hydrogeological CSM was the first step to better' understand both groundwater flow regimes.as well as groundwater quality, with respect to radionuclides associated with LACBWR.

Data gathered through investigation were used to obtain a better understanding of the site's history and hydrogeological setting and were used to design a sampling program. This data was also used to support license termination by providing hydrogeological information for RES RAD and to develop site Derived Concentration Guideline Levels (DCGLs ).

In November 2012, five pairs of groundwater monitoring wells (10 wells total) were installed within the LACBWR radiological controlled area in the most-likely areas of potential release to determine if groundwater quality has been impacted. The paired wells were installed downgradient of the most likely areas where potential releases occurred and have sufficient spatial distribution so that groundwater flow rates and direction may be estimated. The paired wells were installed so that the shallow well intersects the water table and the deeper well is installed at depths approximately 20 to 30 feet below the shallow well. Soil samples were collected during the well installation to provide additional data points that support RESRAD, and DCGLs needed during decommissioning actions.

Two rounds of groundwater samples were collected as part of the hydrogeological investigation.

Samples were collected during the seasonal high water in June 2013 and then again during a seasonal low groundwater level in November 2013. Groundwater samples were collected using low flow methods at monitoring wells equipped with dedicated tubing.

Results of the initial groundwater sampling performed in-2013 indicate:

The shallow aquifer has slower velocities and groundwater movement below the Turbine Building and faster groundwater movement outside and around the Turbine Building, suggesting some interference of the subsurface pilings associated with the building.

D-PLAN / PSDAR 1-7 November 2017

1.0 DESCRIPTION

OF PLANNED DECOMMISSIONING ACTIVITIES (cont'd)

Groundwater velocity data for the deep aquifer indicate less variability and lack the influence of subsurface disturbances.

The most likely areas of interest (AO ls) where radionuclides could have been released to soils and groundwater include the Turbine Building waste collection system and the Underground Gas Storage Tank Vault and piping. No radionuclides were detected above background from the. groundwater monitoring wells suggesting that these AOls

• did not impact downgradient conditions. However, soils and groundwater directly below these areas have not yet been characterized.

Groundwater analytical results did not report radionuclides at activities above background in any of the samples; historic site operations did not significantly impact groundwater quality downgradient of the potential AOls.

The focus of the investigation is to characterize groundwater quality in support of NRC decommissioning requirements. However, monitoring wells are located and constructed i!Jsing methods, such that if needed, they meet the data quality objectives of other programs that may come into play during the regulatory closure of the site.

1.7.

ISFSI DECOMMISSIONING The decommissioning plan for the ISFSI is based on information contained in the NAC-MPC FSAR, Section 2.A.4, "Decommissioning C.onsiderations." The ISFSI will be decommissioned after the stored spent fuel and GTCC waste are removed and transferred to the Department of Energy. NAC-MPC dry cask storage systems in use at the ISFSI are designated as MPC-LACBWR.

The principal elements of the MPC-LACBWR storage system are the vertical concrete cask (VCC) and the transportable storage canister (TSC). The VCC provides-biological shielding and physical protection for the contents of the TSC during long-term storage. The VCC is not expected to become surface contaminated during use, except through incidental contact with other contaminated surfaces. Incidental contact could occur at the interior liner surface of the VCC, the top surface that supports the transfer cask during loading and unloading operations, and the pedestal of the VCC that supports the TSC. All of these surfaces are carbon steel, and could be decontaminated as necessary for decommissioning. A Y4-inch stainless steel plate is placed on the carbon steel pedestal of the MPC-LACBWR VCC to separate it from the stainless steel TSC bottom. Contamination of these surfaces is expected to be minimal, since the TSC was isolated from spent fuel pool water during loading in the pool and the transfer cask was decontaminated prior to transfer of the TSC to the VCC. Activation of the VCC carbon steel liner, concrete, support plates, and reinforcing bar could occur due to neutron flux from the stored fuel. Since the neutron flux rate is low, only minimal activation of carbon steel in the VCC is expected to occur.

Decommissioning of the VCC would involve the re~oval of the TSC and the subsequent disassembly of the VCC. It is expected that the concrete would be broken up, and steel components segmented to reduce volume. Any contaminated or activated items are expected to qualify for near-surface disposal as low specific activity material.

The TSC is designed and fabricated to be suitable for use as a waste package for permanent disposal in a deep Mined Geological Disposal System, in that it meets the requirements of the DOE MPC Design Procurement Specification. The TSC is fabricated from materials having high long-term corrosion resistance, a_nd the TSC contains no paints or coatings that could adversely affect the permanent disposal of the TSC. As a result, decommissioning of the TSC would occur only if the spent fuel contained in the TSC had to be removed. Decommissioning would D-PLAN / PSDAR 1-~

November 2017

1.0

. DESCRIPTION OF PLANNED DECOMMISSIONING ACTIVITIES (cont'd) require that the closure welds at tlie TSC closure lid and port covers be cut, so that the spent fuel could be removed. Removal of the contents of the TSC would require that the TSC be returned to a spent fuel popl or dry unloading facility; such as a hot cell. Closure welds can be cut either manually or with automated equipment, with the procedure being essentially the reverse of that used to initially close the TSC.

The LACBWR ISFSI storage pad, fence, and supporting utility fixtures are not expected to require decontamination as a result of use of the MPC-LACBWR system*. The design of the VCC andTSC precludes the release of contamination from the contents over the period of use of the system. Consequently, these items may be reused or disposed of as locally generated clean waste.

The decommissioning plan for the ISFSI is to dispose of the five VCCs and the 32' x 48' x 3' concrete storage pad. ISFSI decommissioning is not within the scope of the Solutions project.

1.8.

REFERENCES

1.

Letter from Dairyland Power Cooperative to the Nuclear Regulatory Commission, for Order Approving License Transfer and Conforming Administrative License Amendments,

• dated October 8, 2015.

2.

Marlayna Vaaler, U.S. Nuclear Regulatory Commission, Letter to Barbara Nick, Dairyland Power Cooperative, "Order Approving Transfer of License for the La Crosse Boiling Water Reactor from the Dairyland Power Cooperative to LaCrosseSolutions, LLC and Conforming Administrative License Amendment" dated May 20, 2016

3.

Letter from the Nuclear.Regulatory Commission to Dairyland Power Cooperative, Issuance of License Amendment No. 56, dated August 4, 1987

4.

Letter from Dairyland Power Cooperative to the Nuclear Regulatory Commission, Submittal of Decommissioning Plan, Preliminary Decon Plan; and Supplement to the Enyiron Report, dated December 21, 2015.

5..

Letter from the Nuclear Regulatory Commission to Dairyland Power Cooperative, Issuance of License Amendment No. 63, dated August 18, 1988

6.

Letter from the Nuclear Regulatory Commission to Dairyland Power Cooperative, Order

• to Authorize Decommissioning and Amendment No. 66 to Possession Only License No..

DPR-45 for La Crosse Boiling Water Reactor, dated August 7, 1991

7.

Letter from the Nuclear Regulatory Commission to Dairyland Power Cooperative, Confirmatory Order Modifying the August 7, 1991, Decommissioning Order for the La Crosse Boiling Water Reactor, dated September 15, 1994

8.

LACBWRlnitial Site Characterization Survey for SAFSTOR, dated October 1995

9. -

Letter from the Nuclear Regulatory Commission to Dairyland Power Cooperative, I

Issuance of License Amendment No. 69, dated April 11, 1997

10.

Letter from the Nuclear Regulatory Commission to Dairyland Power Cooperative, I

Issuance of License Amendment No.' 71, dated January 25, 2011

11.

Letter from the Nuclear Regulatory Commission to Dairyland Power Cooperative, I.

Issuance of License Amendment No. 72, dated July 31, 2013

12.

La Crosse Boiling Water Reactor License Termination Plan, Revision 0

13.

U.S. Nuclear Regulatory Commission NUREG-0586, Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities, Supplement 1, Volume 1 -

November 2002.

14.

Letter from the Nuclear Regulatory Commission to Dairyland Power Cooperative, Withdrawal of Confirmatory-Order, dated.March 31, 2016 D-PLAN / PSDAR 1-9 November 2017 I

2.0 SCHEDULE 2.1.

SCHEDULE ACTIVITIES COMPLETED PRIOR TO LICENSE TRANSFER TO SOLUTIONS Following final reactor shutdown in April 1987, the transition from operating plant to possession-only facility required numerous admini~trative changes. Staff' level was reduced, license required plans were revised, and operating procedures were curtailed or simplified as conditions and NRC approval allowed. The LACBWR Decommissioning Plan was approved in August

- 1991, and the facility entered SAFSTOR. License renewal granted at the same time accommodated the proposed SAFSTOR period for a term to expire March 29, 2031. At the time of the original Decommissioning Plan in 1987, DPC anticipated the plant would be in SAFSTOR for a 30-:-50 year period.

To make better use of resources during the SAFSTOR period, some incremental decommissioning and dismantlement activities were desirable. By Confirmatory Order from the NRC in 1994, changes in the facility meeting 10 CFR 50.59 requirements were permitted and limited gradual dismantlement progressed. Approximately 2 million pounds of material related to the removal of unused components or whole systems, completed in over 100 specific approved changes to the facility, has been shipped for processing and disposal. This total does not include reactor vessel and B/C waste disposal.

The 2-year Reactor Pressure Vessel Removal (RPV) Project was completed in Jun*e 2007 with disposal of the intact RPV at the Barnwell Waste Management Facility (BWMF). Disposal of the RPV was completed at this time prior to the planned closing of BWMF to out-of-compact waste in July 2008. RPV removal was* not specifically addressed in the original decommissioning schedule. The removal of this large component, as defined in 10 CFR 50.2, was an activity requiring notice be made pursuant to 10 CFR 50.82, Termination of License, (a)(7); This notice was made by submittal to the NRC on August 18, 2005.

In 2007, DPC began efforts to place an ISFSI on site by commencing the Dry Cask Storage Project. An on-site ISFSI was the available option that provided flexibility for license termination

• of the LACBWR facility.

DPC Staff completed an extensive review and analysis of the comparative costs and benefits of the current decommissioning schedule and various accelerated schedules. From this analysis, the DPC Board of Directors approved accelerating the removal of radioactive metal from the LACBWR facility. By letter dated December 7, 2010, DPC gave notification to the NRC of a change in schedule that would accelerate the decc;>mmissioning of the LACBWR facility starting with a 4-year period of systems removal beginning in 2012. This activity included the removal for shipment of large bore (16 and 20-inch) reactor coolant piping and pumps of the Forced Circulation system and other equipment once connected to the reactor pressure vessel or primary system such as Control Rod Drive Mechanisms, Decay Heat, Primary Purification, Seal Injection, and Main Steam.

This metal removal phase of decommissioning activity did not result in significant environmental impacts compared to the "Generic Environmental lnipact Statement (GEIS) on Decommissioning of Nuclear Facilities," NUREG-0586, Supplement 1, November 2002: The GEIS characterizes the environmental impacts resulting from metal removal as generic and small.

D-PLAN / PSDAR

\\

2-1

• November 2017

2.0 SCHEDULE (cont'd)

The Dry Cask Storage Project established an ISFSI on the LACBWR site under the general license provisions of 10 CFR 72, Subpart K. The ISFSI is located 2,232 feet south,-southwest of the Reactor Building center. The ISFSI is used for interim storage of LACBWR spent fuel in the NAC International, Inc. (NAC) Multi-Purpose Canister (MPC) System. The ISFSI contains all LACBWR spent fuel in five NAC-MPC dry cask storage systems. Cask loading and transport operations were completed on September 19, 2012, when the fifth and final dry storage cask was placed on the ISFSI pad.

2.2.

REVISED SOLUTIONS SCHEDULE The revised schedule for decommissioning activities at LACBWR, reflecting the transfer of decommissioning responsibilities to LaCrosseSo/utions (Solutions), is depicted in Table 2.1.

Table 2.1 Decommissioning Schedule Milestones Date Milestone Q2/2016 License Transfer Complete Q2/2016 Submit L TP to NRC Q2/2016 Mobilization Complete Q3/2017 Stack Demolition Complete Q2/2018 L TP Approval by NRC Q4/2017 Component Removal Complete Q2/2018 Building Demolition Complete 04/2018 Transportation and Disposal Complete

. Q4/2018 Site Remediation Complete I

Q4/2018.

FSS Complete I

Q1/2019 Site Restoration Complete Q1/2019 Submit Remaining FSS Reports Q4/2018 Submit License Transfer to Dairyland Amendment Request to NRC Q4/2019 License Transfer to Dairyland Approved by NRC Q4/2019 LACBWR License Termination Approval by NRC Note: Circumstances can change during decommissioning. If it is determined that the decommissioning cannot be completed as outlined in this schedule, an updated schedule will be provided to the NRC.

D-PLAN / PSDAR 2-2 November 2017 I

4.0 PLANT POST-FUEL ACCIDENT ANALYSIS 4.1.

OVERVIEW This section presents the results of an analysis (Reference 4.8.1) of postulated accidents that reflect the significantly reduced non-lSFSI radiological source term as compared to the LACBWR source term during plant operations. With consideration for the current stage of LACBWR decommissioning and with spent nuclear fuel now stored in the ISFSI, this analysis confirms that the minimal radioactive material resulting from LACBWR operation and remaining on the LACBWR site is insufficient for any potential event to result in exceeding dose limits or otherwise involving a significant adverse effect on public health and safety.

The analysis considers the spontaneous release of the (non-lSFSl-related) radioactive source term that was remaining at the LACBWR site in 2012. Decommis~ioning activities and the associated decontamination subsequent to 2012 have significantly reduced that source term such that the analysis presented here remains bounding for future activities. The 2012 source term is in a form and quantity immediately releasable through the:

Airborne pathway; and Liquid discharge pathway.

The airborne release and one of the liquid release events considered in the analysis are non-mechanistic in that there are no credible phenomena that could reasonably be postulated to cause such releases. However, these events are analyzed and conservative assumptions for other credible liquid release events are selected to bound any remaining decommissioning

• events that can still be postulated considering the current stage of LACBWR decommissioning.

It should be noted that the accident analysis of the original liquid release system has been*

retained as the bounding analysis, and is not the system currently in use. The operation and malfunctions of the temporary liquid waste system are bounded by the analysis presented here.

4.2.

RADIONUCLIDE RELEASE LIMITS APPLIED.IN ANALYSIS 4.2.1.

Limits Applied to Postulated Airborne Release The following regulatory limits were considered in the analysis of a postulated airborne release:

1.

1 The limits of 10 CFR 100.11 that specify that the total radiation dose to an individual at the exclusion area boundary for two hours immediately following onset of a postulated fission product release shall not exceed 25 rem (whole body) and 300 rem (thyroid; see Section 4.3.2.6).

2.

The EPA protective action guidelines (PAGs - Reference 4.8.2) that specify the potential offsite dose levels at which actions should be taken to protect the health and safety of-*

the public. The EPA PAG limits include a total effective dose equivalent (TEDE) of 1 rem.

• The EPA PAGs are limiting values for the LACBWR post-fuel accident analysis. This conclusion is based on the sum of the effective dose equivalent resulting from exposure to external sources and from the committed effective dose equivalent incurred from the significant inhalation pathways during the early phase of an event. As detailed further in Section 4.4, this analysis demonstrates that there is insufficient releasable radioactive contamination remaining D-PLAN / PSDAR 4-1 November 2017 I

on the LACBWR site for reasonably conceivable radiological accident scenarios that could result in exceeding the EPA PAGs.

D-PLAN / PSDAR 4-2 November 2017 I

4.0 PLANT POST-FUEL ACCIDENT ANALYSIS (cont'd) 4.2.2.

Limits Applied to Postulated Liquid Releases The LACBWR analysis conservatively applies the normal effluent concentration limits of 10 CFR 20, Appendix B, Table 2, Column 2, to the event scenarios involving release of bulk radioactive liquids. As detailed further in Section 4.4, this analysis demonstrates that there is no reasonable likelihood that a postulated radioactive liquid release event could result in exceeding the normal effluent concentration limits of 10 CFR 20, Appendix B (Reference 4.8.3).

4.3.

POST-FUEL ACCIDENT ANALYSIS ASSUMPTIONS 4.3.1.

Assumptions - Remaining Non-lSFSI-Related Radioactive Source Term With the spent nuclear fuel stored in the LACBWR ISFSI, the amount of (non-lSFSl-related) radioactive contamination conservatively assumed in the analysis to remain at the LACBWR site bounds the decreasing amounts present as decommissioning progresses and is completed.

Potential sources of non-lSFSI radioactivity that remain at LACBWR include the following:

1.

Radioactivity on surfaces of plant structures, systems, and components (SSCs);

2.

Sealed and unsealed sources used for instrument calibration;

3.

Filters used for liquid rad-waste cleanup;

4.

Assorted tools and equipment used to perform decommissioning activities; and,

5.

Radioactive waste containers stored awaiting shipment.

For purposes of the LACBWR post-fuel accident analysis, the radioactivity on plant surfaces is assumed to reasonably represent the non-lSFSI radioactive source term remaining at the LACBWR site (i.e., the other identified potential sources are negligible or are already accounted for as part of plant surface contamination). Specifically, sealed sources are designed to prevent the release of the contents and are not considered in this analysis to be a potential source of releasable radioactive material. Unsealed sources remaining at LACBWR are of extremely low radioactivity levels, such that they do not contribute significantly to the total releasable source term considered in the analysis. Filters are used to remove radioactive material from radioactive liquids generated from decommissioning activities. The radioactive material in these filters is material that is already accounted for above when considering the contamination contained on plant surfaces. Thus, liquid radioactive waste filters do not result in additional releasable source term beyond that already considered.

Radioactive material on or within tools and equipment used at LACBWR is of extremely low radioactivity levels, such that this material constitutes only a small fraction of the radioactivity on plant surfaces. Thus, tools and equipment do not contribute significantly to the total releasable source term considered in the analysis. Finally, radioactive waste containers are used to* hold radioactive materials as they are being removed from the planf during decommissioning. The radioactive material in/on these containers is material,that is already accounted for above when considering the contamination contained on plant surfaces. Thus, radioactive waste containers do not result in significant additional releasable source term beyond that already considered..

The assumed radioactive material on plant surfaces is derived from the results of the LACBWR initial site cha_racterization performed in 1998 following permanent shutdown and decay-corrected to December 2012 (Reference 4.8.4). Specifically, the radioactivity on plant surfaces is conservatively estimated by assuming that the surface contamination present is at levels twice those determined from the LACBWR site characterization. Doubling the site characterization results is intended to provide sufficient margin for the unexpected but potential discovery of localized radiological contamination that could exceed amounts estimated by site D-PLAN / PSDAR 4-3 November 2017

4.0 PLANT POST-FUEL ACCIDENT. ANALYSIS (cont'd) characterization measurements. Radioactive decay since December 2012 is ignored in the analysis. Since much of the remaining radionuclide inventory is of relatively long half-life, *this assumption ensures reasonably conservative values for the remaining source term.

Using the above-described assumptions, approximately 1.175 Ci.of radioactive material is conservatively estimated in the analysis to be present on plant surfaces, and as such represents the assumed total non-lSFSI radioactive source term remaining at the LACBWR site. The

• LACBWR analysis of postulated release events separately considers the portion of this remaining radioactive.contamination that is immediately releasable as airborne contamination and that immediately releasable as contaminated liquid.

4.3.1.1.

Portion of Total Radioactivity Assumed Releasable Via the Airborne Pathway A conservative fraction of 30 percent of the total remaining source term is assumed in the analysis to be immediately available for airborne release. This assumption is reasonably conservative while ensuring that the analysis results well bound the consequences of a postulated airborne release during the LACBWR decommissioning. Specifically, the vast majority of radioactive material remaining at LACBWR is in the form of fixed surface

• contamination on plant SSCs.1 The removal and/or decontamination of these SSCs inherently involves the potential generation of airborne radioactive particulates (e.g., grinding, chemical decontamination, or thermal cutting of contaminated components).2 However, radioactive contamination is distributed throughout numerous SSCs and over relatively large areas. Industry experience at previously decommissioned nuclear reactor plants demonstrates that dismantlement/decontamination is done in distinct manageable "pieces." For example, a system or several small systems, and/or portions thereof, may be designated for removal and/or decontamination at any one time. After that effort is completed, the next system or systems is addressed. The radioactive material collected during each effort is processed, packaged, and shipped. on an ongoing basis, such that its accumulation on site is limited. This "piece-by-piece" process inherently ensures that there is no reasonable likelihood that a significant fraction of the total remaining radioactive. material could be simultaneously disturbed and released as airborne particulate.

Based on the above, it is determined that an assumed fraction of 30 percent of the total remaining source term represents a cons~rvative bounding value for the LACBWR post-fuel accident analysis. Additional assumptions used in the analysis of a postulated airborne release event are described in Section 4.3.2 below.

• 2 Airborne contamination is minimized by minimizing loose contamination levels and their sources. The use of installed and temporary ventilation systems prevents the build-up of air contamination concentrations.

Airborne radioactive particulate emissions will continue to be filtered, as applicable, and effluent discharges monitored and quantified. This includes ( 1) the operation of appropriate portions of building ventilation systems, or approved alternate systems, as necessary during decontamination and dismantlement activities; and (2) use of local high efficiency particulate air (HEPA) filtration systems for activities expected to result in the generation of airborne raQioactive particulates D-PLAN / PSDAR 4-4 November 2017

4.0 PLANT POST-FUEL ACCIDENT ANALYSIS (cont'd) 4.3.1.2.

Portion of Total Radioactivity Assumed Releasable Via the Liquid Pathway Potential sources of radioactive liquid that remain at LACBWR include water generated during decommissioning/decontamination activities (e.g., draining, decontamination, and cutting processes). The portion of the total remaining source term conservatively assumed in the analysis to be available for liquid release at any one time is radioactively contaminated liquid of the following volume, radionuclide concentration, and release flow rate associated with the retention tank contents:

1. 80 percent of the total 6000 gallon volume of the retention tank, which is 4800 gallons.

2. Maximum total radionuclide concentration of 3.9E-03 µCi/cc, which based on the LACBWR-specific radionuclide mix corresponds to a Co-60 concentration of 3.6E-03 µCi/cc.

3. Maximum flow rate from the retention tank of 20 gpm.

This assumption is reasonably conservative while ensuring that the analysis results well bound the consequences of a postulated liquid release during the LACBWR decommissioning.

Specifically, the selection of "80 percent" of the total tank volume is an NRG-accepted conservative assumption, based on the Staff guidance of Branch Technical Position (BTP) 11-6, as further clarified in DC/COL-ISG-013. The assumption that the total radionuclide concentration of the retention tank contents is less than or equal to 3.9E-03 µCi/cc is also conservatively bounding. The value of 3.9E-03 µCi/cc is sufficiently above minimum detectable levels for the monitoring instrumentation used at LACBWR, while also allowing for operational flexibility considering the radionuclide concentrations anticipated to be generated by decommissioning activities.

The vast majority of radioactive material remaining at LACBWR is in the form of fixed surface contamination on plant SSCs. The removal and/or decontamination of these SSCs inherently involves the potential generation of liquid radioactive waste (e.g., as a result of draining, decontamination, and cutting processes during plant decommissioning). The "piece-by-piece" decommissioning process discussed in Section 4.3.1.1 above inherently ensures that there is no reasonable likelihood that a significant fraction of the total remaining radioactive material could be released as radioactively contaminated liquid. Any contaminated liquids that are generated during decommissioning are contained within existing or supplemental barriers and processed (i.e., recirculated, sampled, and diluted) to ensure the radionuclide concentration of the retention tank contents does not exceed an appropriate operational limit established in LACBWR procedures. This operational limit incorporates sufficient margin to the 3.9E-03 µCi/cc limit to ensure that, with allowance for instrumentation uncertainty, the design-basis 3.9E-03 µCi/cc limit will not be exceeded.

Finally, the post-fuel accident analysis demonstrates that, in the unlikely event that 80 percent of the retention tank volume at a total radionuclide concentration of 3.9E-03 µCi/cc were to be released from the retention tank at a flow rate of 20 gpm, the normal effluent concentration limits of 10 CFR 20, Appendix B, Table 2, would not be exceeded (see Section 4.4). Thus, the 20 gpm maximum flow rate from the retention tank is a reasonable value to be established as a design-basis limit. An appropriate operational limit is established in LACBWR procedures that incorporates sufficient margin to the 20 gpm limit. This margin ensures that, with allowance for instrumentation uncertainty, the design-basis 20 gpm limit will not be exceeded.

Based on the justification documented above, this assumption represents a reasonably conservative bounding input to the analysis. Additional assumptions used in the analysis of a postulated liquid release event are described in Section 4.3.3 below.

D-PLAN / PSDAR 4-.5 November 2017 I

4.0 PLANT p*osT-FUEL ACCIDENT ANALYSIS (cont'd) 4.3.2.

Additional Assumptions - Postulated Airborne Release The following assumptions were used in the LACBWR analysis of a postulated airborne release scenario:

4.3.2.1.

Genoa-3 (G-3) Office Building Occupancy For the LACBWR post-fuel accident analysis, it is assumed that an individual working in the G-3 office building stays in the building for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. This is reasonably conservative since it exceeds by two hours the typical work day duration of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

4.3.2.2.

Terrain Height Above Grade The X/Q methodology of Regulatory Guide 1.145 (Reference 4.8.7)] uses the terrain height above grade to calculate the effective stack height. The terrain height difference over the LACBWR site is negligible. Therefore, for purposes of the post-fuel accident analysis, it is assumed that the terrain height is the same as plant grade.

4.3.2.3.

av and al at Distances Less Than 100 Meters The NRC regulatory guidance governing development of av and az do not provide av and Oz values at distances less than 100 meters. Thus for the LACBWR post-fuel accident analysis, the methodology used to obtain av and Oz at distances less than 100 meters is derived from the equations and figures in "Meteorology and Atomic Energy" (M&AE - Reference 4.8.5) and linearly extrapolated to distances less than 100 meters. It is assumed that the av and Oz values used in the analysis at 60 m and 70 m are reasonably representative because they are extrapolated from a region of the curve that is essentially linear.

4.3.2.4.

Pasquill Stability Class It is conservatively assumed that the meteorological category is Pasquill Stability Class F.

4.3.2.5.

Rem vs Rad For the purposes of the LACBWR post-fuel accident analysis, 1 rad is assumed to be equivalent to 1 rem. This is acceptable because the calculated exposures (in rad) are a small fraction of the total dose.

4.3.2.6.

Thyroid Dose The dose to the thyroid is not considered in determining if the dose criteria are met. This has no significant effect on the arn~lysis results since:

1.

There is no radioiodine present in the LACBWR site (non-lSFSI) radionuclide inventory; and

2.

The CEDE dose conversion,factor (DCF) for the only other thyroid significant nuclide, Co-60, is approximately 3.5 times greater than the thyroid DCF. Since the CEDE DCF is larger, and the CEDE acceptance criterion is lower, the limiting dose is the CEDE dose rather than the thyroid dose.

4.3.2.7.

Correction Factor (CF) for G-3 Office Building -

Radioactivity inside the G-3 office building is a function_ of time. The analysis considers two time periods:

D-PLAN / PSDAR 4-6 November 2017 I

4.0 PLANT POST-FUEL ACCIDENT ANALYSIS (cont'd).

1.

0 to1800 seconds - Radioactivity builds up over the first 30 minutes when the fumiga.tion X/Q is used. During this period the inlet concentration is determined by the fumigation X/Q.

2.

1800 to 36,000 seconds - Radioactivity is exhausted over the remaining 9-Y'2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when the non-fumigation X/Q is used. During this period the inlet concentration is 0.0 because the elevated release X/Q is 0.0.

4.3.2.8.

Radionuclide Data For the accident doses and doses for alpha emitting radionuclides (Pu, Am, Cm), the dose conversion factors were taken from Federal Guidance Report No. 11 (Reference 4.8.6). Doses are early phase projections during the first two hours or less.

4.3.2.9.

Atmospheric Release Inputs The following values were used in the analysis.

Input Parameter Value Distance; Release Point to Road 50m Distance; Release point to G-3 Office Buildina 70m Distance; Release Point to Front Gate 120 m Stack Height, hs 350 ft-0 in Breathing Rate 3.47E-04 m3/sec Fumigation Condition Duration One-half hour Elevated Wind Speed, Uhe 2 m/sec Release Duration 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 4.3.3.

Additional Assumptions - Postulated Liquid Release The following assumptions were used in the LACBWR analysis of a postulated contaminated liquid release:

4.3.3.1.

Retention Tank Release, Dilution, and Mixing It is assumed that the release is fully diluted and mixed at the Thief Slough outlet (which empties into the Mississippi River). This is a reasonable location for the analyses because the nearest drinking water intake is 195 miles downstream and the Thief Slough outlet is the nearest sport fishing location. Also, the river is not used for irrigation, and the shoreline deposits pathway is insignificant for the Mississippi River. It is reasonable to assume complete mixing because the transit time to the Thief Slough outlet is 1.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and the annual average dilution factor is 107.

D-PLAN / PSDAR 4-7 November 2017 I

4.0 PLANT POST-FUEL ACCIDENT ANALYSIS (cont'd) 4.3.3.2.

• Duration of Retention Tank Rupture Release to Thief Slough The 6000-gal,lon retention tank is below grade in the containment building. For the postulated non-mechanistic tank rupture scenario, it is conservatively assumed that contaminated water enters the slough at a uniform rate over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This would require the tank water to leak out of the containment building and travel underground from the containment building to the slough at 3.33 gpm. There are no credible phenomena that could reasonably be postulated to cause such a release.

4.3.3.3.

Thief Slough and G-3 Outfall Flow The G-3 Outfall Circulating Water flow is withdrawn from and returned to Thief Slough.

Reflecting this configuration, it is assumed that the G-3 Outfall Circulating Water flow has no net effect on total flow in or out of the slough.

4.3.3.4.

Liquid Release Inputs The following values were used in the analysis.

Input Parameter Value Retention Tank Volume 6000 gal Minimum Mississippi River Flow 2250 cfs Conversion from gal to cc 3785.4 ml/gal Minimum G-3 Circulating Water Flow 43,840 gpm Fraction of Flow Through Thief Slough 25 percent Annual Average Dilution Factor for Thief Slough 107 4.4.

SUMMARY

OF ANALYSIS RESULTS 4A.1.

Postulated Airborne Release The results of the LACBWR post-fuel accident analysis involving a postulated airborne release are summarized in Table 4-1. As indicated in !able 4-1, the following four doses are calculated:

1.

The dose to a person at the edge cif the access road; I

2.

The dose to a person located in the G-3 parking lot;

3.

The dose to a person working inside the G-3 office building; and

4.

The dose to a person at the G-3 entry gate.

The analysis results summarized in Table 4-1 demonstrate that the consequences of releasing 30 percent of the non-lSFSI radioactive source term remaining at the LACBWR site to the atmosphere are well within the applicable 10 CFR 100.11 and EPA PAG limits.

D-PLAN / PSDAR 4-8 November 2017

4.0 PLANT POST-FUEL ACCIDENT ANALYSIS (cont'd) 4.4.2.

Postulated Liquid Release The results of the LACBWR post~fuel accident analysis involving a postulated liquid release are summarized in Table 4-2. As indicated in Table 4-2, the following three postulated release scenarios were evaluated:

1.

A (non-mechanistic) retention tank rupture with a direct release to Thief Slough;

2.

A 20 gpm release rate directly to Thief Slough; and

3.

A 20 gpm release rate into the minimum Genoa-3 Circulating Water flow, which empties into Thief Slough.

The analysis results are summarized in Table 4-2. These results demonstrate that the consequences of releasing 4800 gallons of water containing a radionuclide concentration of 3.90E-03 µCi/cc are less than the normal effluent concentration limit (1 E-3 µCi/ml) of 10 CFR 20, Appendix B, Table 2, Column 2, for all three liquid release *scenarios. It is noted that the release consequences for all three scenarios also are less than the 1 O CFR 20.2003 annual release limits for disposal into sanitary sewerage systems. Although the 10 CFR 20.2003 limits are not directly applicable to these scenarios, the fact that the liquid release results are less than those limits further demonstrates the conclusion that the postulated releases would not have an adverse impact on the health and safety of the public or the environment.

4.5.

RADIOLOGICAL OCCUPATIONAL SAFETY

\\

Radiological events could occur that result in increased exposure of decommissioning workers to radiation. However, the occurrences of these events are minimized or the consequences are mitigated through the implementation of the LACBWR Radiation Protection Program. The Radiation Protection Program is applied to activities performed onsite involving radioactive materials. A primary objective of the Radiation Protection Program is to protect workers and visitors to the site from radiological hazards during decommissioning. The program requires LACBWR and its contractors to provide sufficient qualified staff, facilities, and equipment to perform decommissioning activities in a radiologically safe manner.

Activities conducted during decommissioning that have the potential for exposure of personnel to either radiation or radioactive materials will be managed by qualified individuals who will implement program requirements in accordance with established procedures. Radiological hazards will be monitored. The Radiation Protection Program at LACBWR implements administrative dose guidelines for TEDE to ensure personnel do not exceed federal 10 CFR 20 dose limits for occupational exposure to ionizing radiation.

LACBWR work control procedures will ensure that work specifications, designs, work packages, and radiation work permits involving potential radiation exposure or handling of radioactive materials incorporate effective radiological controls.

D-PLAN / PSDAR 4-9 November 2017

4.0 PLANT POST-FUEL ACCIDENT ANALYSIS (cont'd) 4.6.

OFFSITE RADIOLOGICAL EVENTS Offsite radiological events related to decommissioning activities are limited to those associated with the shipment of radioactive materials. Radioactive shipments will be made in accordance.

with applicable regulatory requirements. The LACBWR Radiation Protection Program, Process Control Program, and the Decommissioning Quality Plan assure compliance with these requirements such that both the probability of occurrence and the consequences of an offsite event do not significantly affect the public health and safety.

4.7.

NON-RADIOLOGICAL EVENTS Decommissioning LACBWR may require different work activities than were typically conducted during normal plant operations. However, effective application of the LACBWR safety program to decommissioning activities will ensure worker safety. No decommissioning events were identified that would be initiated from non-radiological sources that could significantly impact

. public health and safety.

Hazardous materials handling will be controlled by the LACBWR Process Control Program and the corporate Hazardous Material Control Program using approved procedures. There are no chemicals stored onsite in quantities which, if released, could significantly threaten public health and safety.

Flammable gases stored onsite include combustible gases ~sed for cutting and welding. Safe storage and use of these gases and other flammable materials is controlled through the Fire Protection Program and plant safety procedures.

Plant safety procedures and off-normal instructions have been established which would be implemented if a non-radiological event occurred af LACBWR. Implementation of these programs and procedures ensur~s that the probability of occurrence and consequence of onsite non-radiological events do not significantly affect occupational or public health and safety. Plant safety procedures provide personnel safety rules and responsibilities. These safety procedures control both chemical and hazardous waste identification, inventory, handling, storage, use, and disposal.

4.8.

REFERENCES

\\

1.

Sargent & Lundy Calculation No. 2013-0~098,."Doses from Release of Site Non-lSFSI Radioactivity"

2.

Environmental Protection Agency (EPA) 400-R-92-001, "Manual of Protective Action Guides and Protective Actions for Nuclear Incidents," October 1991

3.

U.S. Code of Federal Regulations, "Standards for Protection Against Radiation," Part 20, Chapter I, Title* 10, "Energy" (10 CFR.20)

4.

LACBWR Technical Report No. LAC-TR-138, "Initial Site Characterization Survey for SAFSTOR," Revised December 2012

5.

Meteorology and Atomic Energy 1968, Slade, D. H., Editor, TID-24190, July, 1968, http://www.osti.gov/energycitations/product.biblio.jsp?osti_id=4492043

6.

Federal Guidance Report No: 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion,"

EPA-520/1-88-020, 1988

7.

NRC Regulatory Guiqe, 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assess*ments at Nuclear Power Plants," Rev. 1, February.1983.

8.

Accidental Radioactive Contamination of Human Food and Animal Feeds:.

Recommendations for State and Local Agencies, US Department of Health and Human Services, 08/13/1998 D-PLAN / PSDAR 4-10 November 2017 I

4.0 PLANT POST-FUEL ACCIDENT ANALYSIS (cont'd)

Table 4-1 Summary Results of 2-Hour Airborne Release Analysis Location Dose (rem)

Acceptance Criteria Meets (Note 1)

(rem)

Criterion Edge of Access Road ( 50 m)

CEDE 0.065 Immersion

<1.0E-04 TEDE 0.065 1.0 rem TEDE Yes Genoa 3 Parking Lot (70 m)

CEDE 0.046 Immersion

<1.0E-04 TEDE 0.046 1.0 rem TEDE Yes Genoa 3 Office Building (70 m)

(Note 2)

I CEDE 0.038 Immersion

<1.0E-04 TEDE 0.038 1.0 rem TEDE

• Yes Front Gate (120 m)

CEDE 0.027 Immersion

<1.0E-04 TEDE 0.027 1.0 rem TEDE Yes Notes:

1.

1 rem = 1 rad (see Section 4.3.2.5)

2.

Dose reflects assumed 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> occupancy (see Section 4.3.2.1 ).

D-PLAN / PSDAR 4-11 November 2017 I

4.0 PLANT POST-FUEL ACCIDENT ANALYSIS (cont'd)

Table 4-2 Summary Results of Liquid Release Analysis Release Description Results Acceptance Criteria Meets Criterion Tank Rupture to Thief Slough Sum of Fractions 0.02894 Sums 1.0 Yes Total Quantity Released (Ci) (Note 1) 0.07086 Total< 1.0 Ci Yes 20 gpm Discharge to Thief Slough Sum of Fractions (Note 2) 0.1736 Sums 1.0 Yes Total Quantity Released (Ci) (Note 1) 0.07086 Total< 1.0 Ci Yes 20 gpm Discharge to G-3 Outfall Sum of Fractions G-3 Circulating Water 0.9999 Sums 1.0 Yes Slough Outlet (Note 2) 0.1736 Sums 1.0 Yes Total Quantity Released (Ci) (Note 1) 0.07086 Total< 1.0 Ci Yes Notes:

3.

Total radionuclide concentration in the tank is 3.900E-03 Ci/cc and the tank volume is 1.817E+07 cc; thus, the total activity released is (3.900E-03 Ci/cc x 1.817E+07 cc x 1.0E-06 Ci/Ci=) 0.07086 Ci.

4.

The G-3 Outfall Circulating Water flow affects the sum of the fractions only at the outfall, not at the outlet of Thief Slough. Thus, the sum of the fractions for a 20 gpm release rate is 0.1736 at the slough outlet regardless of the G-3 Outfall Circulating Water flow.

D-PLAN / PSDAR 4-12 November 2017 I

5.0 ENVIRONMENTAL IMPACT Review of post-operating license stage environmental impacts was* documented in a supplement to the Environmental Report for LACBWR dated December 1987. LACBWR decommissioning and dismantlement activities have resulted in no significant environmental impact not previously evaluated in the NRC's Environmental Assessment in support of the August 7,' 1991, Decommissioning Order or the Final Environmental Statement (FES) related to operation of LACBWR, dated April 21, 1980 (NUREG-0191 ).

The environmental impact of decommissioning and dismantlement activities is defined in the "Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities (GEIS),"

NUREG-0586, Supplement 1, November 2002. For decommissioning, the NRC uses a standard of significance derived from the Council on Environmental Quality (CEQ) terminology.

The NRC has defined three significance levels: SMALL, MODERATE, and LARGE:

SMALL - Environmental impacts are not detectable or are so minor that they will neither destabilize nor noticeably alter any important attribute of the resource.

MODERATE - Environmental impacts are sufficient to alter noticeably, but not to destabilize, important attributes of the resource.

LARGE - Environmental impacts are clearly noticeable and are sufficient to destabilize important attributes of the resource.,

The environmental impact of all completed or planned LACBWR decommissioning and dismantlement activities is SMALL as determined by the GEIS. LACBWR decommissioning is specifically evaluated in the GEIS. As stated in the GEIS, licensees can rely on information in this Supplement as a basis for meeting the requirements in 10 CFR 50.82(a)(6)(ii). Site-specific potential environmental impacts not determined in the GEIS are:

Offsite land use activities Aquatic ecology as to activities beyond the operational area Terrestrial ecology as to activities beyond the operational area Threatened and endangered species Socioeconomic Environmental justice The L TP for LACBWR has been prepared and details final decommissioning activities including site remediation, survey of residual contamination, and determination of site end-use. Chapter 8 of the L TP contains a supplement to the Environmental Report that describes any new information or significant environmental change associated with the site specific decommissioning and site closure activities performed at the LACBWR site. The L TP supplement to the Environmental Report concluded the following:

As previously evaluated in the D-Plan/PSDAR, the non-radiological environmental impacts from decommissioning LACBWR are temporary and not significant. The potential issues identified as "site-specific" in NUREG-0586 have been evaluated and there is no significant impact. The potential environmental impacts associated with decommissioning LACBWR have already been predicted in and will be bounded by the previously issued environmental impact assessments (NUREG-0191, NUREG-0586, and D-Plan/PSDAR). Therefore, there are no new or significant environmental changes associated with decommissioning.

D-PLAN / PSDAR 5-1 November 2017 I

LTP SECTION 3 REDACTED VERSION FOR PUBLIC' DISCLOSURE ATTACHMENT 3 LACBWR D-Plan/PSDAR Section 3 Redacted Version LTP SECTION 3 REDACTED VERSION FOR PUBLIC DISCLOSURE

r 3.0 ESTIMATE OF EXPECTED DECOMMISSIONING COSTS 3.1.

INTRODUCTION The revised decommissioning cost estimate prepared by Solutions evaluates the following cost elements:

1.

Cost assumptions used, including contingency factors;

2.

• Major decommissioning activities and tasks;

3.

Unit cost factors;

4.

Estimated costs for decontamination and removal of equipment and structures;

5.

Estimated costs for waste disposal, including disposal site surcharges;

6.

Estimated Final Radiation Survey (FRS) costs; and

7.

Estimated total costs.

The cost estimate focuses on the remaining work, including costs of labor, materials, equipment, energy, and services. The cost estimate includes the cost of the planned_

remediation activities as well as the cost of the transportation and disposal of the waste generated by the planned work.

3.2.

HISTORICAL PERSPECTIVE The LACBWR Decommissioning Plan was approved on August 7, 1991. Because the licensing history of LACBWR spans a period that includes several decommissioning regulation changes,

- The D-Plan has been revised to the LACBWR Decommissioning Plan and Post-Shutdown Decommissioning Activities Report (D-Plan/PSDAR):

In a letter dated October 8, 2015 (Reference 1 ), Dairyland and LaCrosseSolutions, LLC (Solutions) requested Nuclear Regulatory Commission (NRG) consent to transfer Dairyland's possession, maintenance and decommissioning authorities, under Possession Only License No.

DPR-45, from Dairyland to Solutions. The NRG approved this transfer in a letter dated May 20, 2016 (Reference 2). The revised cost estimates presented reflect those. developed by Solutions.

After the balance of the site is remediated and the as-left radiological conditions are demonstrated to be below the unrestricted use criteria specified in 10 CFR 20.1402, the licensed area will be reduced to a small area around the ISFSI and Possession Only License No. DPR-45 will be transferred back to Dairyland.

3.3.

PREVIOUS DAIRYLAND COST ESTIMATES In late 1983, the DPC Board of Directors resolved,to provide resources for the final dismantlement of LACBWR DPC began making deposits to a decommissioning fund in 1984.

The Nuclear Decommissioning Trust (NDT) was established in July 1990 as an.external fund outside DPC's administrative control *holding fixed income and equity investments.

The cost of DEGON was based on the selection of unrestricted use as the criteria to be pursued for LACBWR. At the time of preparation of this plan in 1987, decommissioning cost was based on studies by Nuclear Energy Services, Inc., available generic decommissioning cost guidance, and technology as it existed. In the Safety Evaluation Report dated August 7, 1991 (Reference 3), related to the order authorizing decommissioning and approval of the Decommissioning Plan, the NRG found the estimate of $92 million in Year 2010 dollars reasonable for the final dismantling*cost of LACBWR.

D-PLAN / PSDAR 3-1

- November 2017 _ I

3.0 ESTIMATE OF EXPECTED DECOMMISSIONING COSTS (cont'd)

An improved site-specific decommissioning cost study was performed by Sargent & Lundy (S&L) in 1994 and provided basis for the updated cost estimate and funding. The S&L study determined the cost to complete decommissioning to be $83.4 million in Year 1994 dollars with commencement of decommissioning assumed to occur in 2019. A cost study revision completed in July 1998 placed the cost to complete decommissioning at $98.7 million in Year 1998 dollars. A cost study revision, prompted by significant changes in radioactive waste burial costs, as well as lessons learned on decontamination factors and methods, was prepared in November 2000 and placed the cost to complete decommissioning at $79.2 million in Year 2000 dollars. During 2003, the cost study was revisited again to include changes in escalation rates,

• progress in limited dismantlemcint, and a revised reactor vessel weight definition. This update placed the cost to complete decommissioning at $79.5 million in Year 2003 dollars.

In preparation for removal of the reactor pressure vessel (RPV}, cost figures were brought current to $84.6 million in Year 2005 dollars. As of December 2006, NOT funds were approximately $83.4 million. NOT funds for B/C waste and RPV removals, approved by the Board of Directors, have been drawn in the amount of $18.2 million. Following B/C waste and RPV disposal a revis.ion to the cost estimate was performed in September 2007 that placed the cost to complete decommissioning at $62.5 million in Year 2007 dollars.

A cost study update was completed in November 2010 to more accurately assess future costs of the remaining dismantlement needed and to facilitate DPC decommissioning and license termination planning. This update placed the cost to complete decommissioning at $67.8 million in Year 2010 dollars. During this process, ISFSI decommissioning costs were identified uniquely as a specific item and estimated to be $1.6 million in Year 2010 dollars. The DPC Board of Directors established an external funding mechanism for ISFSI decommissioning costs in accordance with 10 CFR 72.30 to assure adequate funds will be available for the final decommissioning cost of the LACBWR ISFSI.

A cost study update was completed in March 2013 for the LACBWR plant. During the revision to the cost study, some potentially contaminated structures previously assumed to be decontaminated and left intact were evaluated for demolition and disposal. This change in decommissioning methodology to demolition and disposal of structures, in lieu of decontamination of structures, resulted in an increase in the LACBWR plant decommissioning cost estimate in the range of $20 million over the previous November 2010 decommissioning cost estimate of $67.8 million. The March 2013 cost study update included the cost of demolition and disposal of the LACBWR stack, Turbine and Turbine Office Buildings, Waste Treatment Building,.and Underground Gas Storage Tank Vault structure. This update placed the cost to complete plant decommissioning at $90.7 million in Year 2013 dollars. The DPC Board of Directors formally adopted the change in decommissioning methodology to demolition and disposal of potentially contaminated structures and authorized adjustments to decommissioning funding be made as necessary.

• The ISFSI Decommissioning Cost Estimate was revised in March 2013 to reflect the MPC-LACBWR as-built vertical concrete cask (VCC) dimensions. These VCC dimensions differ from those used to establish the ISFSI Decommissioning Cost Estimate in 2010. Use of the as-

• built VCC dimensions resulted in a reduction in the volume of concrete to be disposed of. The cost for ISFSI decommissioning was estimated to be $1,435,626 in Year 2013 dollars.

D-PLAN / PSDAR 3-2 November 201 t I

3.0 ESTIMATE OF EXPECTED DECOMMISSIONING COSTS (cont'd) 3.4.

SOLUTIONS DECOMMISSIONING COST ESTIMATE The decommissioning cost estimate presented herein represents the projected costs to complete the remaining decommissioning work as of October 1, 2015. This estimate was prepared by SOLUTIONS based upon an assessment of the remaining work and incorporating experience gained while performing similar decommissioning tasks including the ongoing decommissioning of the Zion Nuclear Power Station (ZNPS) through the work of its subsidiary ZionSolutions LLC.

, The decommissioning cost estimate includes application of contingency, as specific provision for unforeseeable elements of cost within the defined project scope. Contingencies are particularly important where previous experience has shown that unforeseeable events, which may increase costs, are likely to occur. The contingency, as used in this estimate, does not account for price escalation and inflation in the costs of decommissioning over the remaining project duration.

The site-specific decommissioning cost estimate presents a breakdown of all costs associated with completing the decommissioning and unrestricted release of the LACBWR site, other than the area bounded by the ISFSI. The estimate includes the costs required to accomplish

• unrestricted release and restore the site to a safe and stable condition as well as the operation of the ISFSI until the site and the remaining ISFSI are transferred back to Dairyland.

3.5.

COST ESTIMATE DESCRIPTION AND METHODOLOGY The cost estimates include consideration of regulatory requirements, contingency for unknown or uncertain conditions, and the availability of low and high-level radioactive waste disposal sites. The methodology utilized to develop the cost estimate follows the basic approach presented in "Guidelines for Producing Commercial Nuclear Power Plant Decommissioning Cost Estimates, (Reference 4)" which uses a unit cost factor approach for estimating the decommissioning activity costs. It also includes the use of site specific information when available (e.g., hourly-labor rates, and commodities).

The updated DPC estimate completed in March 2013 has been utilized to obtain site-specific commodity quantities for this estimate. The commodity weights and estimated unit cost factors were applied, which take into consideration the current decommissioning approach and schedule, to arrive at an updated cost estimated to decommission LACBWR. Dairyland and Solutions also utilized 25 years of corporate experience in planning and scheduling as well as the latest available industry experience (e.g., information from the decommissioning of ZNPS).

The estimate does not include the transfer of spent fuel, which has been previously transferred to an ISFSI facility, the security c9sts for the ISFSlfacility, or the removal of certain large components and decommissioning work previously completed.

Additionally, Dairyland and Solutions performed a contingency and risk analysis so that the potential additional costs due to expected but undefined risks and uncertainties could be addressed and included in the cost estimate.

The resulting information was then compiled into a decommissioning cost estimate. The following sections provide a summary of those results.

D-PLAN / PSDAR 3-3 November 2017 I

3.0 ESTIMATE OF EXPECTED DECOMMISSIONING COSTS (cont'd) 3.6.

SUMMARY

OF THE SITE-SPECIFIC DECOMMISSIONING COST ESTIMATE The overall remaining decommissioning cost (including scope risk contingency) is estimated to be $- Million (in current ~r dollars), with a base estimated cost of$- Million, plus a scope risk contingency of$* Million. The cost estimates include provisions for cost escalation based upon the following assumptions:

All estimated costs including labor, staff, materials, equipmel")t, professional services, waste transportation and disposal are in 2015 dollars.

Although all costs are in current year dollars, the project baseline does include*

provisions to escalate costs based on the Consumer Price Index for all Urban Customers - U.S. City Average All Items, Not Seasonally Adjusted (CPI-U NSA).

The associated Clf1SS A radioactive waste management costs are covered by existing fixed-price contracts with EnergySolutions. Therefore, the waste management costs for these items are well known and not likely to vary due to waste volume uncertainties.

No costs for Class 8/C waste are included in the estimate, as all materials classified as 8/C waste were previously removed by Dairyland.

The cost estimate includes the costs for radiological decommissioning and site restoration. A summary of the cost for each part of the decommissioning program is provided in Table 3-1.

Table 3-1 Cost Summary for Radiological Decommissioning and Site Restoration Radiological Site Restoration 1 Total Project Decommissioning Performance Baseline

$-Million

$2.6 Million

$111111 Million Contingency

$. Million

$0.3 Million

$II Million

$-Million l

$Ill Million Total

$2.9 Million Note 1: Site restoration is included for completeness, but not required for license termination funding purposes.

3.7.

ISFSI MANAGEMENT All spent nuclear fuel elements from LACBWR have been transferred from the FESW to dry cask storage in the ISFSI. Solutions will assume responsibility for the ISFSI Site, including security requirements. Solutions has entered into a "Company Services Agreement" with Dairyland, pursuant to which Dairyland will provide operations, maintenance, access control, and security services to and for the ISFSI site. Dairyland is responsible for the costs relating to the ISFSI and those costs are not included in this decommissioning estimate.

3.8.

SITE RESTORATION COSTS Solutions acknowledges that the costs to restore the LACBWR property are not considered by the NRC staff as part of decommissioning costs. Nevertheless, there is significant interest by many stakeholders in these costs and they are presented herein. The estimated cost for the D-PLAN / PSDAR 3-4 November 2017 I

3.0 ESTIMATE OF EXPECTED DECOMMISSIONING COSTS (cont'd) anticipated work scope is $2.6 Million. A contingency of $0.3 Million is estimated, bringing the total cost to $2.9 Million.

Overall, that work scope includes removal of any remaining hazardous materials, demolition of remaining structures, backfilling of any open excavations or void spaces, and final grading and stabilization against erosion. The estimated costs.include the labor, equipment, materials, professional services and fees needed to conduct the work.. In general, most of this work is

• anticipated to be performed by contractors however, the estimated cost also includes al.I of the program support activities and services necessary to manage and safeiy carry out the work.

3.9.

DECOMMISSIONING FUND Decommissioning costs will be paid for with funds from the site's Nuclear Decommissioning Tru~t (NDT) fund. The decommissioning of the LACBWR site ISFSI will be undertaken by Solutions and will be financed separately to the NDT account amount identified here for decommissioning of the LACBWR site.

The project cash balance of the NDT identified for the decommissioning of the LAC_BWR site, as a~d to by Solutions, and held in trust by the Owner trustee as of October 1, 2015 was

$- Million.

Based on a time phased cash flow analysis of the radiological decommissioning and site restoration costs, and assuming NDT returns at an annual 2% real, after tax rate of return, the required minimum funding assurance amount to fund the future radiological decommissioning costs equals $- Million, which is below the $- Million available balance described above.

This NDT position, together with the $. Million Surety Bond payable to the NDT, provides for sufficient funding and financial assurance for the completion of the decommissioning of the LACBWR site.

Additionally, although not relied upon here, Solutions parent EnergySolutions has agreed with Dairyland to provide a performance guaranty defined in the LACBWR Decommissioning Agreement submitted as part of the license transfer application.

This PSDAR will not be updated for minor changes in anticipated decommissioning costs.

However, the status of the decommissioning funding will continue to be reported to the NRC in accordance with 10 CFR 50.75(f)(1), "Reporting and recordkeeping for decom.missioning planning." Additionally, Solutions will inform the NRC in writing of any significant schedule and decommissioning cost changes per 10 CFR 50.82(a)(7), and provide an updated site-specific estimate of remaining decommissioning.costs with the license termination plan per 10 CFR 50.82(a)(9)(ii)(F).

3.10.

REFERENCES

1.

Letter from Dairyland Power Cooperative to the Nuclear Regulatory Commission, for Order Approving License Transfer and Conforming Administrative License Amendments, dated October 8, 2015.

2.

Marlayna Vaaler, U.S. Nuclear Regulatory Commission, Letter to Barbara Nick, Dairyland Power Cooperative, "Order Approving Transfer of License for the La Crosse D-PLAN / PSDAR 3-5 November 2017

• I

3.0 ESTIMATE OF EXPECTED DECOMMISSIONING COSTS (cont'd)

Boiling Water Reactor from the Dairyland Power Cooperative to LaCrosseSolutions, LLC and Conforming Administrative License Amendment" dated May 20, 2016

3.

Letter from the Nuclear Regulatory Commission to Dairyand Power Cooperative, Order to Authorize Decommissioning and Amendment No. 66 to Possession Only License No.

DPR-45 for La Crosse Boiling Water Reactor, dated August 7, 1991

4.

T.S. LaGuardia et al., Guidelines for Producing Commercial Nuclear Power Plant Decommissioning Cost Estimates, AIF/NESP-036, May 1986.

I

/

o~PLAN / PSDAR

• 3-6 November 2017 I

i 1, i j I

I l i

,I ATTACHMENT 4 CD contains Attachment 2 & 3 PDF flies and Preflight Status Report

~~::::,

-:~

!1 I

This document serves as preflight report for Attachment 2 and 3 to the letter LC-2018-0036. The following file(s) did not pass pre-flight criteria or do not meet NRC criteria, but text is word searchable with clarity/legibility of high quality.

Reference Document Name File Name Preflight Reason Status This file contains scanned LACBWR D-Plan LACBWR D-Plan Revision November 2017 Failed document (unembedded fonts),

logos, and signatures < 300 ppi, clear and legible LACBWR D-Plan, Section 3 LACBWR D-Plan Revision November 2017 Section 3 Passed Redacted