ML050260338

Annual Decommissioning Plan Revision

ML050260338
Person / Time
Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	12/29/2004
From:	Berg W Dairyland Power Cooperative
To:	Document Control Desk, NRC/FSME
References
LAC-13856
Download: ML050260338 (69)

Text

II DAIRYLAND 7*j COOPERATIVE

• 3200EASTAVE. SO.

• P.O. BOX 817

• LA CROSSE, WISCONSIN 546020817 OFFICE (608) 787-1258 FAX (608) 787-1469 WEB SITE: www.dairynet.com WILLIAM L. BERG President and CEO December 29, 2004 In reply, please refer to LAC-13856 DOCKET NO. 50409 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

Dairyland Power Cooperative La Crosse Boiling Water Reactor (LACBWR)

Possession-Only License DPR-45 Annual Decommissioning Plan Revision

REFERENCES:

(1)

DPC Letter, Taylor to Document Control Desk, LAC-12460, dated December 21, 1987 (original submittal of LACBWR's Decommissioning Plan)

(2)

NRC Letter, Erickson to Berg, dated August 7, 1991, issuing Order to Authorize Decommissioning of LACBWR (3)

NRC Letter, Brown to Berg, dated September 15, 1994, modifying Decommissioning Order The annual update of the LACBWR Decommissioning Plan has been completed, and the pages with changes and their explanations are included with this letter. Each page with a change will have a bar in the right-hand margin to designate the location of the change. None of the changes was determined to require prior NRC approval, and they have been reviewed by both the plant Operations Review Committee and the independent Safety Review Committee.

The individual pages requiring revision are attached to this letter. Please substitute these revised pages in your copy(ies) of the LACBWR Decommissioning Plan. Reasons for the changes are listed on a separate attachment.

If you have any questions concerning any of these changes, please contact Jeff Mc Rill of my staff at 608-689-4202.

Sincerely, DAIRYLAND POWER COOPERATIVE William L. Berg, President CEO WLB:JBM:dh Attachments cc: Kris Banovac NRC Project Mgr.

A Touchstone Energy Cooperative f,4,rn~SslcŽ

IDARYLAND NRC Docket No. 50-409 If[

OPERATIVE LA CROSSE BOIUNG WATER REACTOR (LACBWR)

• 4601 STATE ROAD 35 GENOA, WISCONSIN 54632-8846 * (608) 689-2331 TO:

/A)p.

tA)tlt.

CONTROLLED DISTRIBUTION NO..53 FROM:

LACBWR Plant Manager 1/13/2005

SUBJECT:

Changes to LACBWR Controlling Documents I.

The following documents have been revised:

DECOMMISSIONING PLAN, revised November 2004 Remove and replace the following pages:

Title Page page 0-4 (Table of Contents) pages 1-1 & 1-3 pages 2-1 thru 2-4 pages 3-1 thru 3-3, 3-6 & 3-7, Figure 3.4 pages 4-1 thru 4-5 pages 5-3, 5-17, 5-28 thru 5-30, 5-35, 5-38, 5-41 thru 5-45 pages 6-3, 6-14 thru 6-18, Figure 6.1 pages 7-1 thru 7-5 pages 9-2 thru 9-4, 9-6 thru 9-8 SITE CHARACTERIZATION SURVEY Remove and replace the following pages:

Title page page 5 pages 24 thru 28

. The material listed above is transmitted herewith. Please verify receipt of all listed material, destroy superseded material, and sign below to acknowledge receipt.

o The material listed above has been placed in your binder.

O Please review listed material, notify your personnel of changes, and sign below to acknowledge your review and notification of personnel. [To be checked for supervisors for department specific procedures and LACBWR Technical Specifications.]

O The material listed above has been changed. [To be checked for supervisors when materials applicable to other departments are issued to them.]

/S_

DATE_

Please return this notification to the LACBWR Secretary within ten (10) working days.

• 1 2004 LACBWR Decommissioning Plan Review Cover Page Update revision date.

Page 0-4 Table of Contents. List of Figures: A change is made at Figure 3.4 to "Genoa Site Map" from "LACBWR Site Map. " Figure 3.4 is being replaced with an updated and better defined diagram of the Genoa Site. Changes to Section 3 provide further information concerning site definition issues.

Page 1-1 Section 1.

Introduction:

In the first paragraph, third sentence, the phrase "Dairyland Power Cooperative's" is substituted by the acronym "DPC's" for correction. The acronym was established previously in the section.

Section 1.1. Selection of SAFSTOR: In the first paragraph, first sentence, the phrase "Nuclear Regulatory Commission " is substituted by the acronym "NRC's" for correction. The acronym was established previously in the section.

Page 1-3 Section 1.1. Selection of SAFSTOR: Continued at top of page from previous, first paragraph, first sentence, the phrase "Nuclear Regulatory Commission " is substituted by the acronym "NRC" for correction. The acronym was established previously in the section.

Second paragraph, in the first sentence, the phrase "Dairyland Power Cooperative" is substituted by the acronym "DPC" for correction. The acronym was established previously in the section.

Page 2-1 Section 2.1.

Introduction:

The first paragraph is revised to establish the acronym and identify location as follows, "The La Crosse Boiling Water Reactor (LA CB R9) is owned and was operated by Dairyland Power Cooperative (DPC) of La Crosse, Wisconsin."

Third paragraph, changes are made to DPC in three places for consistency and grammar correction.

Section 2.2. Initial Construction and Licensing History: Acronyms for Atomic Energy Commission (AEC) and Dairyland Power Cooperative (DPC) are established earlier in the section. Section is revised in numerous places using the established acronyms for consistency and correction.

Page 2-2 Numerous changes using the established acronyms of AEC, DPC, and LA4CBWR are installed for consistency and correction. "US. Nuclear Regulatory Commission " is shortened to "Nudear Regulatory Commission " on the page also.

Page 2-3 Two changes using the established acronym DPC are installed for consistency and correction.

Page 2-4 Page is reissued with content shift from reformatting and no other changes.

Page 1 of 11

2004 LACBWR Decommissioning Plan Review Page 3-1 Section 3.1.1. Site Location and Description of Site Layout: First paragraph has LACBWR and DPC added in parentheses to identify acronyms. The second sentence is revised by substituting "DPC" and "LA CB WR" for the associated titles and restating as follows, "The site is, in the most part, owned by Dairyland Power Cooperative (DPC) and includes LACBWR and the 350-megawatt coal-fired generating facility, Genoa Unit 3. " A third sentence is added to the end of the paragraph stating, "Figure 3.4 depicts the Genoa Site."

A new third paragraph is added to this Section 3.1.1, and states, "Attached to this Decommissioning Plan is the Initial Site Characterization Surveyfor SAFSTOR Within this document (LA C-TR-138) the LACBWR Affected Area Map is presented on page 5. This area is bounded by the LACBWR Site Enclosurefence. Definitive historical site assessment will be performed in support of the eventual License Termination Plan process."

Section 3.1.2. The Authority of the Exclusion Area and Licensee Authorities: In the third sentence the phrase, "the Dairyland Power Cooperative " is replaced by

'the lcensee (DPC)" for consistency of terms. The final sentence is revised using the established acronym DPC for consistency and correction.

Section 3.2. Transportation. Industrial and Militara Facilities: The second sentence is revised using the established acronym DPC for consistency and correction.

Page 3-2 Page is reformatted and reissued with content shift and no other changes.

Page 3-3 Section 3.3.3. Local Meteorology. The parenthetical phrase, "See Figure 3.4, "is deleted at the end of the first paragraph as unnecessary. Figure 3.4 is being replaced with an updated and better defined diagram of the Genoa Site. This diagram does not include the referred to location of the 10-meter surface meteorological tower located at the eastern limit of the Owner Controlled Area.

Page 3-6 Section 3.4.5. Potential Dam Failures: First paragraph is revised by substituting "LACBW2" for the associated title. Second and third paragraphs have the phrase "La Crosse site" replaced with the phrase "LA CB WR site" to correct terminology. LACBWR is a part of the larger Genoa site, owned by Dairyland Power Cooperative.

Section 3.4.6. Flooding Protection Requirements: The phrase "La Crosse site" is replaced with the phrase 'LACB WR site" to correct terminology in the first sentence. In the third sentence the phrase "Nuclear Regulatory Commission " is substituted by the acronym "NRC" for correction. The seventh sentence states, "It was determined that the reactor containment building and reactor stack would be able to withstand this flood " This sentence is revised to state, "It was determined that the Reactor Building and ventilation stack would be able to withstand this flood" This change more appropriately describes the stack and Page 2 of 11

2004 LACBWR Decommissioning Plan Review continues a shift from containment terminology. In the next to last sentence of the page the term "offisite" is deleted as unnecessary to describe resources available.

Page 3-7 Section 3.4.6. Flooding Protection Requirements: Continued at top of page from previous, a change is made at "containment vessel " to "Reactor Building. "

Figure 3.4 LACBWR Site Map: This diagram is replaced with an updated and better defined diagram of the Genoa Site that contains LACBWR. Discussion contained in Section 3 provides further information concerning site location issues.

Page 4-1 Section 4.1. General Plant

Description:

First paragraph, in the first sentence, "The " is deleted before LACBWR and the word "utilizes " is replaced by "utilized" for correction. Last sentence of the second paragraph has "containment building" replaced by 'Reactor Building. "

Section 4.2.1. Containment Building: This title is changed to "Reactor Building. " In the first and third paragraphs of the section on page 4-1, a change to "Reactor Building" is made in four places. In the first paragraph "same as Containment Building" is added in the parentheses following the first change to Reactor Building.

Page 4-2 Continuing Section 4.2.1 from previous page, changes to "Reactor Buildings" are made in seven places on page 4-2.

Fifth paragraph of page 4-2. In the second sentence, the phrase "access to the shell will be through " is corrected to read "access to the shell is through. " In the fourth sentence, "evacuated" is replaced by exited" in describing Reactor Building egress through the emergency airlock. The last sentence of the paragraph which states, "When the doors are closed, a clamp exerts a positive force, which is transmitted through the doors to live-rubber gaskets around the doorframes to ensure gas tightness, " is deleted. This description pertained to containment integrity of the airlock doors during operation and gas tightness is no longer maintained.

Sixth paragraph of page 4-2 is revised by deleting the second and fourth sentences which state, "Nine-inch-thick concrete blocks were placed on the outside of the doorfor shielding, " and, "Two rubber gaskets between the door and doorframe ensure a pressure-tight seal. " Freight door shield blocks have been removed under an approved facility change. Description of rubber gasket seal pertained to containment integrity of the freight door during operation and a pressure-tight seal is no longer maintained.

Seventh paragraph of page 4-2. The first two sentences state, "Approximately 300 mineral insulated (MI) cables and 75 bulkhead conductors penetrate the containment shell. These are in the northwest quadrant of the shell adjacent to Page 3 of 11

2004 LACBWR Decommissioning Plan Review the electrical room. " Efforts to reduce unused wiring continue under approved facility changes. The number of cables in this referred to area have been reduced.

The two sentences are combined and revised as follows, "Cables and bulkhead conductors from the Turbine Building provide electrical service to the Reactor Building through penetrations in the northwest quadrant of the building shell. "

Page 4-3 Continuing Section 4.2.1 from previous page, changes to "Reactor Building" are made in three places on page 4-3.

First paragraph of page 4-3 is revised in the first sentence by substituting "A" for the phrase "An approximately. " The second and third sentences stated, "The piping connection to the emergency core spray system is near the bottom of the tank The connection to the building spray system supply header is a standpipe within the tank (the spray system piping and nozzles having been removed); the top of the standpipe is sufficiently above the bottom of the tank to leave 15,000 gal. of waterfor use in the emergency core spray system. " These two sentences are deleted as unnecessary to current status. The last sentence has reference to refueling removed and information added, "The storage tank provides waterfor reactor vessel and upper cavityffall normal makeup, and other operations associated with fuel handling and theffuel element storage well. "

Page 4-4 Section 4.2.2. Turbine Building: Continuing section from previous page, a change is made twice to "Reactor Building" in the first paragraph of the page.

Section 4.2.3. Waste Treatment Building: A change is made to "Reactor Building" in the first paragraph. In the fifth paragraph, describing shielded cubicle access, the word "through " is replaced in the phrase by, "to which access is gained. " The sixth paragraph is corrected by the following phrase, "ventilation is routed through a HEPA filter. "

Page 4-5 Due to page content adjustment from changes described previously, page 4-5 is reissued with no other changes.

Page 5-3 Section 5.2.1. Reactor Vessel and Internals: First sentence of "System Status" is revised to state, 4AU1fud assemblies and startup sources have been removed from the reactor core. " The sentence that follows is revised by adding, "The 29 control rods and other core components remain in the reactor vessel. These changes improve descriptive information provided.

Page 5-17 Section 5.2.15. Hydraulic Valve Accumulator System: "System Status" is revised to state, "This system has been drained. The air compressors, water pumps, and other equipment have been electrically disconnected and are not maintained operational" This change is due to work completed under an approved facility change.

Page 4 of 1I

2004 LACBWR Decommissioning Plan Review Page 5-28 Section 5.2.26. Steam Turbine: Descriptive information is changed to past tense.

"System Status" is revised to state, "Steam piping in the Turbine Building, turbine inlet valves, and other components and instrumentation have been removed for reprocessing and disposal. Complete removal of the Steam Turbine system is in progress. I This change is due to work completed or ongoing under an approved facility change.

Page 5-29 Section 5.2.27. 60-Megawatt Generator: Descriptive information is changed to past tense. "System Status" is revised by adding, "Complete removal of the 60-Megawatt Generator system is in progress. " This change is due to work completed or ongoing under an approved facility change.

Page 5-30 Section 5.2.28. Turbine Oil and Hydrogen Seal Oil System: Descriptive information is changed to past tense. "System Status" is revised to state, "Hydrogen Seal Oil cooler piping, cooling water outlet and inlet manifolds, and instrumentation have been removed. Turbine Oil system pumps and other equipment have been electrically disconnected and will be removed with the remainder of the system. " This change is due to work completed or ongoing under an approved facility change.

Page 5-35 Section 5.2.33.1. Normal AC Distribution: First paragraph is revised to state, "69-KVpower is supplied to the reserve auxiliary transformer located in the LACBW switchyard through a three-phase air disconnect switch and three, 30-amp, 69-KVfuses. " This change is due to work completed under an approved facility change.

Page 5-38 Section 5.2.34.1. Containment Atmosphere PASS System

Description:

A change is made to 'Reactor Building Atmosphere" in the title and paragraph.

Section 5.2.34.3. Reactor Coolant PASS Sntem

Description:

In the "System Status," a change is made to "Reactor Building Atmosphere "

Page 5-41 Section 5.4.2. System Radiation Levels: In the listing of radiation levels, current Through contact dose rates are added to existing initial dose rates. Changes due to Page 5-43 equipment removal are indicated, and explanatory notes are added to each of the three pages the listing appears on.

Page 5-44 Due to page content adjustment from changes described previously, page 5-44 is reissued with no other changes.

Page 545 Section 5.7.1.2. Containment Building Air Exhaust Gaseous and Particulate Monitor: The title is changed to "Reactor Building Air Exhaust Gaseous and Particulate Monitor. "

Page 5 of 1I

2004 LACBWR Decommissioning Plan Review Page 6-3 Section 6.2. Organization and Responsibilities: The section continues from previous pages; in the fourth and fifth paragraphs of page 6-3 changes to correct administrative staff titles and availability are added. "Administrative Supervisor" is changed to "Administrative Assistant. " The fourth paragraph has the phrase "involved in facility shutdown and establishment " deleted after occurring many, many years ago. The fifth paragraph is revised to state, "Additional administra-tive personnel will be made available to the Administrative Assistant as needed, and will assist in the clerical tasks at LACBWR. Such additionalpersonnel will be qualified to perform required communication functions while assisting and will be assigned other tasks as necessary by the AdministrativeAssistant. "

Page 6-14 Section 6.9.1. Fire Protection Plan: The entire section is revised following the And first paragraph sentence to be consistent with Fire Protection Program content.

Page 6-15 Changes in bold follow and are a result of corrective actions in response to a Quality Assurance audit that cited radiological concerns as lacking:

"LACBWR can safely maintain and control the Fuel Element Storage Well in the case of the worst postulatedfire in each area of the plant The fire protection plan at LA CB WR is to preventfire, effectively respond to fire, and to minimize the risk to the publicfromfire emergencies. The goals of thefire protection plan arefire prevention andfire protection. This fire protection plan, implemented through the fire protection program, provides defense-in-depth to fire emergencies and addresses the following objectives:

• Prevent fires. By administratively controlling ignition sources, flammable liquid inventory, and combustible material accumulation,fire risk is reduced. Welding and other hot work shal be performed only under Special Work Permit conditions and the use of afire watch shall be required. Routine fire and safety inspections by LACBWR staff shall be conducted to ensure flammable liquids are properly stored and combustible material is removed. These inspections shall also require identification offire hazards and result in action to reduce those hazards. General cleanliness and good housekeeping shall continue as an established practice and shall be checked during inspection.

• Rapidly detect, control, and extinguish fires that do occur and could result in a radiological hazard. Fire detection systems are installed to detect heat and smoke in spaces and areas of the protectedpremises of LACBWR. Iffire detection systems or components are unavailable, increased monitoring of affected areas by personnel shall compensate for any loss of automatic detection. Fire barriers provide containment against the spread offire between areas and provide protection to personnel responding tofire emergencies. Areas of high fire loading are provided with automatic reaction-typefire suppression systems or manually initiatedfire suppression systems. These installed systems Page 6 of 11

2004 LACBWR Decommissioning Plan Review provide immediate fire suppression automatically or provide the means to extinguish fires withoutfire exposure to personnel manually initiating them. Manualfire extinguishing equipment is installed in all areas of the LACBWRfaciity. Al fire protection equipment and systems are maintained, inspected, and tested in accordance with established guidelines. Compensatory actions and procedures for the impairment or unavailability offire protection equipment are provided. A trainedfire brigade, available at all times shall respond immediately to all fire emergencies. The function of the response by the fire brigade shall be to evaluate fire situations, to extinguish incipient stagefires, and to quickly realize the need for, and then summon, outside assistance. For any situation where afire should progress beyond the incipient stage, qualified outside fire services shallprovide assistance.

• Minimize the risk to the public, environment and plant personnel resulting from fire that could result in a release of radioactive materials.

Surface contamination has been reduced to minimal levels in most areas of the facility by cleanup efforts and the effects of long-term decay.

Contamination surveys are performed routinely and areas identifledfor attention are deconnedfurther. Good radiological work practices and contamination control are maintained. Radioactive waste generated is containerized and shipped for processing in accordance with approved procedures. Liquid effluents are collected in plant drain systems, processed, and monitored during discharge. Plant personnel are alerted to elevated radioactivity levels by area radiation monitors and air monitoring systems that are in operation at all times in buildings of the radiological controlled area. Gaseous and particulate air activities are continuously monitored prior to their release to the environment.

Procedures and protocols exist to ensure risk is minimized to the public and members of the outside fire service. "

Page 6-15 Section 6.9.2. Fire Protection Program: The introductory paragraph is revised to be consistent with other Fire Protection Program content. Unnecessary wording is deleted and statements are made succinct.

"The fire protection program for the LA CB WRfacility is based on sound engineering practices and established standards. The function of thefire protection program is to provide the specific mechanisms by which the fire protection plan is Implemented. The fire protection program utilizes an integrated system of administrative controls, equipment, personnel, tests, and inspections. Components of thefire protection program are:"

Section 6.9.2.1. Administrative Controls: The word "objective" is replaced by the word "goal" in the first sentence to be consistent with other Fire Protection Program content.

Page 7 of I1

2004 LACBWR Decommissionine Plan Review Page 6-16 Due to page content adjustment from changes described previously, page 6-16 is reissued with no other changes.

Page 6-17 Section 6.9.2.8. Outside Fire Service Assistance: The last paragraph of the page is revised to remove reference to the Fire Services Director for Vernon County Emergency Management. Local emergency response communications are directed through the Vernon County 9-1-1 Center and Vernon County Emergency Management.

Page 6-18 Due to page content adjustment from changes described previously, page 6-18 is reissued with no other changes.

Figure 6.1 La Crosse Boiling Water Reactor SAFSTOR Staff: Changes to correct administrative staff titles and availability are added. "Administrative Supervisor" is changed to "Administrative Assistant. "

Page 7-1 Section 7.1. Preparation for SAFSTOR: Minor wording changes are made to content of the section's two paragraphs to describeSAFSTOR preparations that were completed several years ago. Changes appear in bold in the following:

"The plant was shut down on April 30, 1987. Reactor defueling was completed on June 11, 1987. Since the plant shutdown, some systems were secured.

Additional systems were shut down following determination of layup methodology. Others awaited changes to plant Technical Specifications before operational status could be evaluated. Section 5.2 discusses the plant systems and their current status.

In addition to preparation of this Decommissioning Plan, proposed revisions to Technical Specifications, the Security Plan, the Emergency Plan, and the Quality Assurance Program Description were completed. An addendum to the Environmental Report and a preliminary DECONplan were also submitted."

Page 7-2 Section 7.3.1. Flushing Systems and Decontamination During SAFSTOR and, Section 7.3.1.1. Internal System Flushing: These whole sections are deleted due to the fact that natural decay has reduced radionuclide inventories to the point that flushing and decontamination of systems are not warranted.

New content is added at Section 7.3.1 to describe licensing actions that have occurred since permanent shutdown. Information added is as published by the NRC in the January 2004 Fact Sheet on Decommissioning Nuclear Power Plants or contained in other documents. Section 7.3.1 is renamed and contains the following:

Page 8 of II

2004 LACBWR Decommissioning Plan Review Section 7.3.1. Significant SAFSTOR Licensing Actions "The licensee's authority to operate Facility License No. DPR-45, pursuant to 10 CFR Part 50, was terminated by license Amendment No. 56, dated August 4, 1987, and a possession only status was granted. The decommissioning alternative ofSAFSTOR was chosen.

The NRC directed the licensee to decommission the facility in its Decommissioning Order ofAugust 7,1991. The Decommissioning Order was modified September 15, 1994, by Confirmatory Order to allow the licensee 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, ifaplant-specific safety and environ-mental review procedure containing similar requirements as specified In 10 CFR 50.59 was applied.

License Amendment No. 66, issued with the Decommissioning Order and also dated August 7, 1991, provided evaluation and approval of the proposed Decommissioning Plan, proposed SAFSTOR Technical Specifications, and license renewal to accommodate the proposed SAFSTOR period until March 29,2031.

The Initial Site Characterization Survey for SAFSTOR was completed and.L published October 1995 and is attached to this Decommissioning Plan.

License Amendment No. 69, containing the current Technical Specifications, was issued April11, 1997. This amendment revised the body of the license and the Appendix A Technical Specifications. The changes to the license and Technical Specifications were structured to reflect the permanently defueled and shutdown status of the plant. These changes deleted all requirementsfor emergency electricalpower systems and maintenance of containment integrity.

The SAFSTOR Decommissioning Plan is considered the post-shutdown decommissioning activities report (PSDAR). The PSDAR public meeting was held on May 13, 1998.

Review of and revisions to this Decommissioning Plan, the Security Plan, the Emergency Plan, the Quality Assurance Program Description, the Offsite Dose Calculation Manual, and other material continue at intervals as required. "

Page 7-2 Section 7.3.1.2. Area and System Decontamination: With deletions and additions (cont'd) described preceding, section is renumbered 7.3.2.

Page 7-3 Section 7.3.1.2. Area and System Decontamination: The section now numbered 7.3.2 continues from previous page. In second paragraph of page 7-3, the phrases "steam cleaning, abrasive blasting, " and "electropolishing, " are deleted as Page 9 of 11

2004 LACBWR Decommissioning Plan Review decontamination methods. Contamination levels have decreased negating the need for such cleaning methods.

Page 7-3 Section 7.3.2. Removal of Unused Eguipment During SAFSTOR, is renumbered (cont'd) 7.3.3.

Page 7-4 Section 7.3.3. Research, is renumbered 7.3.4.

Section 7.3.4. Testing and Maintenance Program to Maintain Systems in Use, is renumbered 7.3.5. In the first sentence the word "previously" is deleted as unnecessary.

Section 7.4. Plant Monitoring Program: In the third sentence the term "offsite" is replaced in the phrase by "An in-plant, as well as surrounding area, surveillance program " to better describe locations beyond this licensee's Owner Controlled Area.

Page 7-5 Section 7.4.4. Environmental Monitoring: In the first sentence the term "offsite" is replaced in the phrase by "Surrounding area dose rates " to better describe locations beyond this licensee's Owner Controlled Area.

Page 9-2 Section 9.2. Spent Fuel Handling Accident: The term "offsite" is deleted in the Through third paragraph. The curie content remaining as of October 2003 and calculated Page 9-3 values for Whole Body Dose and Skin Dose as of October 2003 are updated to October 2004.

Page 9-4 Section 9.3. Shipping Cask or Heavy Load Drop into FESW: The curie content remaining as of October 2003 and calculated values for Whole Body Dose and Skin Dose as of October 2003 are updated to October 2004. The maximum whole body dose is updated as a factor of 400 below the 10 CFR 100 dose limit as opposed to a factor of 375 below that limit. The final statement at the bottom of the page is restated to be consistent with that on page 9-3 as follows:

'As can be seen, the estimated maximum whole body dose is more than a factor of 400 below the 10 CFR 100 dose limit of 25 Rem (25,000 mRem) to the whole body within a 2-hour period."

Page 9-6 Section 9.5. FESW Pipe Break: In the fourth paragraph, the phrase "offsite release" is changed to "release of contamination. " Change better describes information presented.

Page 9-7 Section 9.8. Seismic Event: A change is made to "Reactor Building" in the second paragraph.

Page 9-8 Section 9.9. Wind and Tornado: Changes are made to "Reactor Building" in three places.

Page 10ofIl

Wit 2004 LACBWR Decommissionine Plan Review INITIAL SITE CHARACTERIZATION SURVEY FOR SAFSTOR (LAC-TR-138):

Cover Page Update revision date.

Page 5 LACBWR AFFECTED AREA MAP: Two changes are incorporated to reflect removal under an approved facility change of the 20,000-gallon external fuel oil storage tank and a change in nomenclature from "Containment Building" to the more appropriate "Reactor Building. "

Page 24 Curie content values stated in pages 24-28 are updated. These pages of Through Attachments 1, 2, 3 have been decay-corrected to October 2004, replacing pages Page 28 that had been decay-corrected to October 2003.

Page 11 of 11

LA CROSSE BOILING WATER REACTOR (LACBWR)

DECOMMISSIONING PLAN 4-Revised November 2004 DAIRYLAND POWER COOPERATIVE LA CROSSE BOILING WATER REACTOR (LACBWR) 4601 State Road 35 Genoa, WI 54632-8846

.I

TABLE OF CONTENTS LIST OF FIGURES Figure 3.1 Figure 3.2 Figure 3.3 Figure 3.4 Figure 3.5 Figure 3.6 Figure 3.7 Figure 3.8 Figure 4.1 Figure 4.2 Figure 4.3 Figure 4.4 Figure 4.5 Figure 6.1 Figure 6.2 General Site Location Map Population Dispersion Effluent Release Boundary Genoa Site Map Monthly Average Meteorological Data Wind Speed Frequency Distribution East Bank River Slope Generalized Soil Profile Containment Building Elevation Containment Building General Arrangement Main Floor of Turbine Building, El. 668'0" Mezzanine Floor of Turbine Building, El. 654'0" Grade Floor of Turbine Building, El. 640'0" LACBWR Organization Chart Tentative Schedule for LACBWR Decommissioning I

4--

Table 3-1 Table 5-1 LIST OF TABLES Wind Direction Frequency Distribution at LACBWR Site and La Crosse NWS Spent Fuel Radioactivity Inventory Pane No.

3-3 5-1 D-PLAN 04 November 2004

1.

INTRODUCTION The Decommissioning Plan describes Dairyland Power Cooperative's (DPC) plans for the future disposition of the La Crosse Boiling Water Reactor (LACBWR). DPC has chosen to place LACBWR in the SAFSTOR mode, so this plan describes the plant's status and provides a safety analysis for the SAFSTOR period. A separate preliminary DECON Plan has been submitted to outline DPC's intention to ultimately decommission the plant and site to radiologically releasable l levels and terminate the license in accordance with Nuclear Regulatory Commission (NRC) requirements.

This Decommissioning Plan concentrates on the status of LACBWR while the reactor fuel remains in the Fuel Element Storage Well. There are 333 activated fuel assemblies onsite. The plan at this time is to continue to store the fuel in the existing Fuel Element Storage Well. DPC currently expects the fuel to remain onsite until a federal repository, interim storage facility, or licensed temporary monitored retrievable storage facility is established and ready to receive LACBWR fuel.

1.1 SELECTION OF SAFSTOR The NRC's proposed rule on Decommissioning Criteria for Nuclear Facilities identifies 3 major classifications of decommissioning alternatives. They are DECON, SAFSTOR, and ENTOMB.

The proposed rule defines the alternatives as 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 decon-taminated (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 is either DECON or SAFSTOR. Due to some of the long-lived isotopes in the reactor vessel and internals, ENTOMB, by itself, is not an allowable alternative under the proposed rule.

D-PLAN 1-1 November 2004

1. INTRODUCTION - (cont'd)

The NRC issued its Waste Confidence Decision in the Federal Register on August 31, 1984. In it, the NRC found "reasonable assurance that, if necessary, spent fuel generated in any reactor can be stored safely and without significant environmental impacts for at least 30 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 activated 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 standpoint, as well as necessary from the practical standpoint.

After evaluating the factors involved in selecting a decommissioning alternative, DPC decided to choose an approximately 30-50 year SAFSTOR period, followed by DECON. The exact duration of the SAFSTOR period will be dependent on the availability of a high-level waste storage facility, availability of waste disposal, economics, personnel exposure, and various institutional factors. If any major changes are made in DPC's decommissioning plans, a revision to this plan will be prepared.

1.2 REFERENCES

1) Nuclear Regulatory Commission, proposed rule on Decommissioning Criteria for Nuclear Facilities, Federal Register, Vol. 50, No. 28, February 11, 1985.

2) Nuclear Regulatory Commission, Waste Confidence Decision, Federal Register, Vol.

49, No. 171, August 31, 1984.

3) "Decommissioning - Demonstrating the Solution to a Problem for the Next Century,"

Nuclear Engineering International, Vol. 32, No. 399, October 1987, p. 48.

4) Proceedings from the 1987 International Decommissioning Symposium, Conf-871018, October 4-8, 1987.

D-PLAN l-3 November 2004

2.

LA CROSSE BOILING WATER REACTOR OPERATING HISTORY

2.1 INTRODUCTION

The La Crosse Boiling Water Reactor (LACBWR) is owned and was operated by Dairyland Power Cooperative (DPC) of La Crosse, Wisconsin.

LACBWR was a nuclear power plant of nominal 50 Mw electrical output, which utilized a forced-circulation, direct-cycle boiling-water reactor as its heat source. The plant is located on the east bank of the Mississippi River in Vernon County, Wisconsin, approximately 1 mile south of the village of Genoa, Wisconsin, and approximately 19 miles south of the city of La Crosse, Wisconsin.

The plant was one of a series of demonstration plants funded in part by the U.S. Atomic Energy Commission (AEC). The nuclear steam supply system and its auxiliaries were funded by the AEC, and the balance of the plant was funded by DPC. The Allis-Chalmers Company was the original licensee; the AEC later sold the plant to DPC and provided DPC with a provisional operating license.

2.2 INITIAL CONSTRUCTION AND LICENSING HISTORY Allis-Chalmers, under a contract with the AEC, had the responsibility for the design, fabrication, construction, and startup of the reactor. Allis-Chalmers retained Sargent & Lundy Engineers as architect-engineers for the project and the Maxon Construction Company as constructors. DPC furnished the plant site and all equipment, facilities, and services necessary for a complete and operable nuclear plant.

Allis-Chalmers Atomic Energy Division and the AEC entered into a contract, AT(1 1-1)-850, on June 6, 1962, to construct a second round demonstration nuclear power plant. The last modification to the contract was No. 8, dated June 16, 1967.

DPC and the AEC entered into a contract, AT(11-1)-851 on June 6, 1962, to buy steam from the nuclear power plant to operate a turbine-generator for production of electricity.

On November 5, 1962, Allis-Chalmers applied for a Construction Authorization.

The AEC issued Construction Authorization, CAPR-5 on March 29, 1963.

On August 3, 1965, Allis-Chalmers applied for an Operating Authorization; amendments to the application continued through March 8, 1967.

The AEC issued Provisional Operating Authorization No. DPRA-5 to Allis-Chalmers on July 3, 1967.

DPC applied for an Operating Authorization on October 4, 1967.

D-PLAN 2-1 November 2004

2. LA CROSSE BOILING WATER REACTOR OPERATING HISTORY - (cont'd)

Provisional Operating Authorization No. DPRA-6 was issued to DPC on October 31, 1969, under Docket No. 115-5.

DPC applied to the AEC to convert POA No. DPRA-6 to a 10 CFR Part 50 provisional operating license on May 22, 1972.

The AEC issued Provisional Operating License No. DPR-45 under Docket 50-409 to DPC on August 28, 1973.

DPC applied to the AEC to convert POL No. DPR-45 to a full-term facility operating license on October 9, 1974. The 40-year term would expire on March 28, 2003.

In 1977, the Systematic Evaluation Program (SEP) was initiated by the Nuclear Regulatory Commission (NRC) to review the designs of older operating nuclear power plants, including LACBWR, in order to reconfirm and document their safety. The purpose of the review was to provide (1) an assessment of the significance of differences between current technical positions on safety issues and those that existed when a particular plant was licensed, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. The conversion of the provisional operating license to a full-term operating license was tied to the completion of the SEP safety assessment. The Integrated Plant Safety Assessment for LACBWR was issued as NUREG-0827 in June 1983.

Addendum 1 to NUREG-0827 was released in August 1986. DPC performed a consequence study to evaluate wind, tornado and seismic events. The study was accepted by the NRC in letters dated September 9, 1986 and April 6, 1987. DPC provided a schedule for completion of items necessary for safe shutdown during a seismic, wind or tornado event on December 11, 1986. Work on scheduled items has been terminated due to the plant shutdown.

2.3 OPERATING RECORD LACBWR achieved initial criticality on July 11, 1967, and the low power testing program was completed by September 1967. In November 1967, the power testing program began. The power testing program culminated in a 28-day power run between August 14 and September 13, 1969.

DPC has operated the facility as a base-load plant on its system since November 1, 1969, when the AEC accepted the facility from Allis-Chalmers.

LACBWR was permanently shut down on April 30,1987.

During this time the reactor was critical for a total of 103,287.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The 50 MW generator was on the line for 96,274.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Total gross electrical energy generated (MWH) was 4,046,923. The unit availability factor was 62.9%.

D-PLAN 2-2 November 2004

2. LA CROSSE BOILING WATER REACTOR OPERATING HISTORY - (cont'd) 2.4 DECISION FOR SHUTDOWN On April 24, 1987, the decision was made by the DPC Board of Directors to permanently shut down LACBWR. The official announcement to the public of this decision was made on April 27, 1987.

The major reason for the decision was projected cost savings associated with the operation of the cooperative's coal-fired plants because of recently renegotiated coal and coal transportation contracts.

Other factors include the low growth rate in electrical demand forecast for DPC's service area and the current regional surplus of generating capacity.

Final reactor shutdown was completed at 0905 hours0.0105 days <br />0.251 hours <br />0.0015 weeks <br />3.443525e-4 months <br /> on April 30, 1987. The availability factor for LACBWR in 1987 had been 96.4%.

Final reactor defueling was completed on June 11, 1987. Eleven fuel cycles over the 20 years of operation have resulted in a total of 333 irradiated fuel assemblies being stored in the LACBWR Fuel Element Storage Well.

2.5 OPERATING EVENTS WHICH AFFECT DECOMMISSIONING 2.5.1 Failed Fuel During refueling operations following the first few fuel cycles, several fuel elements were observed to have failed fuel rods. These fuel failures were severe enough to have allowed fission products to escape into the Fuel Element Storage Well and reactor coolant. These fission product particles then entered, or had the potential to enter and lodge in or plate out in, the following systems:

(1)

Forced Circulation (2) Purification (3) Decay Heat (4) Main Condenser (5) Fuel Element Storage Well (6) Overhead Storage Tank (7) Emergency Core Spray (8) Condensate system between main condenser and condensate demineralizer resin beds (9) Reactor Vessel (10) Seal Injection (11) Waste Water (12) Reactor Coolant Post-Accident Sampling System (13) Control Rod Drive System D-PLAN 2-3 November 2004

2. LA CROSSE BOILING WATER REACTOR OPERATING HISTORY - (cont'd)

Therefore, extra precautions will be taken in monitoring for and containing any fission product and transuranic radionuclide contaminants during the eventual disassembly of the above listed systems. The majority of this material is located on horizontal surfaces in the Fuel Element Storage Well and the Reactor Vessel.

2.5.2 Fuel Element Storage Well Leakage The stainless steel liner in the Fuel Element Storage Well (FESW) has had a history of leakage.

From the date of initial service until 1980, the leakage increased from approximately 2 gallons per hour (gph) to just over 14 gph. In 1980, epoxy was injected behind the liner and leakage was reduced to approximately 2 gph. In 1993, the FESW pump seals were discovered to be defective and were replaced, which reduced the leak rate to approximately I gph. FESW water level is continuously monitored in the control room and verified periodically by local inspection. The control room level instrument(s) generate an audible alarm when FESW level decreases to a selected level which is significantly above the minimum allowable level as specified in the technical specifications.

2.5.3 References 1

1) DPC Letter, LAC-4935, Madgett to Director of NRR, dated October 5, 1977.

2) DPC Letter, LAC-6274, Linder to Director of NRR, dated May 9, 1979.

3) DPC Letter, LAC-8553, Linder to Director of NRR, dated September 7, 1982.

D-PLAN 2-4 November 2004

3.

FACILITY SITE CHARACTERISTICS 3.1 GEOGRAPHY AND DEMOGRAPHY CHARACTERISTICS 3.1.1 Site Location and Description of Site Layout The La Crosse Boiling Water Reactor (LACBWR) is located on the east bank of the Mississippi River approximately 19 miles south of the city of La Crosse, Wisconsin, and 1 mile south of the populated portion of the village of Genoa, Wisconsin. The site is, in the most part, owned by the Dairyland Power Cooperative (DPC) and includes LACBWR and the 350-megawatt coal-fired generating facility, Genoa Unit 3. Figure 3.4 depicts the Genoa Site.

The site is on fill in the river-bottom area of the east bank of the Mississippi River and includes a portion of a wooded hillside to the east of the nuclear unit. The site also contains DPC's 161-KV and 69-KV transmission switching center.

Attached to this Decommissioning Plan is the Initial Site Characterization Survey for SAFSTOR.

Within this document (LAC-TR-138) the LACBWR Affected Areqa Map is presented on page 5.

This area is bounded by the LACBWR Site Enclosure fence. Definitive historical site assessment will be performed in support of the eventual License Termination Plan process.

The municipalities, including villages, towns and cities within a 25-mile radius of the facility, are shown in Figure 3.1. The population dispersion out to 5 miles is shown on Figure 3.2.

3.1.2 The Authority of the Exclusion Area and Licensee Authorities The site exclusion area referenced in 10 CFR 100, Section 3(a) was initially established as approximately 1,109 feet in radius from the center of the Reactor Building. The area to which access will potentially be excluded during a postulated accident while in SAFSTOR is the area within the Effluent Release Boundary (ERB). The ERB is the licensee (DPC) property line within the former 1,109 ft. radius exclusion area. (See Figure 3.3.) DPC exercises direct control over its own employees and visitors on the site to exclude them if adverse radiological conditions require. Additionally, DPC maintains a letter of agreement with the Vernon County Sheriffs Department for them to provide any necessary assistance.

3.2 TRANSPORTATION. INDUSTRIAL AND MILITARY FACILITIES WITHIN PROXIMITY TO THE PLANT There are no military facilities located within a 5-mile radius of the nuclear plant site. The only industrial facility of any significant size is the DPC Genoa Unit 3 coal facility located approximately 200 feet from the nuclear plant and sharing the same site. There are no major manufacturing facilities of any type in this area; it is principally used for agriculture.

Transportation routes include the Burlington Northern Railway line from Chicago, Illinois, to the D-PLAN 3-1 November 2004

3. FACILITY SITE CHARACTERISTICS - (cont'd)

West Coast which crosses through the original exclusion area. The Burlington Northern Railway line is a twin-track line of welded steel track and constitutes a major rail corridor for the railroad.

Wisconsin State Trunk Highway 35 also crosses through the original exclusion area. The Mississippi River main channel which is used for barge transportation crosses through the orig-inal exclusion area. The Milwaukee Railroad single track line from Minneapolis, Minnesota, to St. Louis, Missouri, is on the opposite side of the Mississippi River from the plant and was abandoned from 1980 to 1981. The line has since been restored to service but is not frequently used. State Trunk Highway 56 originates in the village of Genoa and runs East towards Viroqua, the county seat. The origin point for Highway 56 is approximately 1-1/2 miles north of the reactor plant.

On the Iowa and Minnesota side of the river, State Trunk Highway 26 runs within 4 miles of the original exclusion area. All the mentioned highway facilities are two-lane paved roadways with unlimited access.

The car count on the road (Highway 35) passing through the nuclear facility original exclusion area is 2,950 cars per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, as determined by the Vernon County Wisconsin Highway Department in 1984.

1 There does exist north of the plant, approximately.9 mile, a U.S. Army Corps of Engineers Lock and Dam on the Mississippi River. This lock is not classified as an industrial facility, although it employs approximately 11 individuals.

3.3 METEOROLOGY 3.3.1 Meteorological Measurement Program The LACBWR meteorological measurement program consists of onsite equipment located within the Mississippi River valley. Meteorological parameters monitored are wind speed, wind direction, and temperature. Data is also available fiom the National Weather Service (NWS),

approximately 35 km (21.7 mi.) north of LACBWR.

3.3.2 General Climatology The plant site area exhibits a typical continental type of climate. Temperature extremes in the La Crosse/LACBWR region are more marked because of the river-valley location. Average temperatures vary from -7. IC (19.2"F) in the three months of winter to 21.9"C (71.4"F) in the summer months., A maximum temperature of 42.2"C (108.0°) was recorded in July 1936, with a minimum low of -41.7C (-43.0"F) recorded in January 1873, both in La Crosse. Monthly precipitation in the area averages between 5.1 cm (2.0 in.) and 10.7 cm (4.2 in.) from March through October and 2.5 cm (1 in.) and 5.1 cm (2 in.) for the rest of the year. Average annual precipitation is 79.2 cm (31.2 in.). Monthly snow and sleet averages between 12.7 cm (5 in.) and 35.6 cm (14 in.) from November through March, the largest amount normally occurring during March. The normal annual amount of snow and sleet is 110.5 cm (43.5 in.).

D-PLAN 3-2 November 2004

3. FACILITY SITE CHARACTERISTICS - (cont'd)

The prevailing winds are subjected to the channeling effects of the river valley. This channeling directs almost all of the regional scale cross-valley winds into the north-south orientation of the valley. Wind speeds are also lower as a result of vertical decoupling caused by the river valley.

In summer, low wind speeds cause air stagnation in the valley during periods of hot humid weather, and in winter some deepening of inversion conditions can cause a stagnant layer of cold air on the valley floor.

3.3.3 Local Meteorology Onsite meteorological data was collected at two levels (top of stack and 10-meter surface tower) from 1976 to 1994 and provides the data that follows.

Wind direction frequency distributions for the surface and stack levels are shown in Table 3-1.

The distributions demonstrate the strong predominance of wind directions from the SSE and NNW sectors for the surface and S and N sectors for the stack. The Mississippi River valley has a north-south orientation at the plant site, and it would be expected that winds should be predominantly from the north and south because of the river valleys channeling effect. It is suspected that the layout of the buildings on site reduces the frequency of winds observed from the southwest to west.

TABLE 3-1 WIND DIRECTION FREQUENCY DISTRIBUTION AT LACBWR SITE AND LA CROSSE NWS (Percentage 1982-1984)

East La Crosse Surface Stack Bluff NWS N

10.4 15.9 5.6 5.5 NNE 2.1 5.1 4.7 1.9 NE 1.8 2.5 3.7 1.1 ENE 2

1.8 4.9 1.3 E

3 1.9 5.5 6.7 ESE 3.1 2.3 5.9 9.1 SE 6.1 5.2 6.2 10.6 SSE 20.9 10.2 7.5 10.9 S

14.1 22.5 10.8 12.1 SSW 3.4 6.6 9.5 3.4 SW 1.4 2.8 5.6 2.3 WSW 1.0 2.2 4.2 1.6 W

1.3 2.9 5.6 7.7 WNW 2.7 3.7 6.7 6.1 NW 9.3 7.2 6.8 10.9 NNW 17.7 7.3 6.8 8.6 D-PLAN 3-3 November 2004

3. FACILITY SITE CHARACTERISTICS - (cont'd) 3.4.5 Potential Dam Failures The U. S. Army Corps of Engineers maintains a lock and dam less than a mile upstream from the facility. The lock and darn was constructed in the 1930's. The NRC technical reviewers noted that Lock and Dam No. 8, upstream from LACBWR, has its right bank earth-bermed to control water and direct flow to the dam spillway, which is located in the main river channel. Locks are located on the east bank, adjacent to which is the U. S. Army Corps of Engineers' Field Office.

The failure of the main darn or adjacent earth-berms will have a variable effect on the water surface elevations at the LACBWR site. Barges depend on the river discharge for adequate channel depth. The nominal operating pool elevations of Lock and Dam No. 8 are 631 feet MSL (upstream) and 620 feet MSL (downstream). The difference in elevation between head and tail waters of the dam is 0.8 feet at the five-year discharge flow rate of 134,000 cfs. The elevation difference decreases with increasing discharge, so that at a 500-year discharge flow rate (321,000 cfs), the difference is reduced to 0.4 feet. Additional increases in discharge result in a smaller difference in elevation up to the elevation at which the dam is submerged.

Should the dam fail with discharges ranging from 100,000 cfs to 300,000 cfs, the increase in dam tail water elevations will be attenuated as water reaches the LACBWR site. Consequent increase of water elevation will certainly be less than 1 foot of elevation at the site. It was concluded that the effect of a catastrophic failure of Lock and Dam No. 8, during high flow conditions, would have negligible effect on water surface elevations measured at the LACBWR site.

3.4.6 Flooding Protection Requirements Historical flooding protection at the LACBWR site was consistent with the initial design criteria of meeting the passive protection needs of a 100-year return frequency flood. These design criteria were based on the return frequencies established by the United States Geological Survey.

During the Systematic Evaluation Program, the NRC reviewed, through a consultant, current criteria which establish a return frequency more in the approximate range of one in a million years. In this particular review, it was determined that in order to comply with criteria for a probable maximum flood certain evaluations would have to be completed. The existing reactor site did not comply with passive protection requirements to the level of one-in-a-million-year return frequency (probable maximum flood). The site was required to review the stability of certain structures at this flood level. It was determined that the Reactor Building and ventilation stack would be able to withstand this flood. Procedures have been established which require certain actions to be taken at various water levels. A river elevation of 630 feet MSL activates the flood-alert stage in which management is alerted of the condition and monitoring is increased. A flood warning condition is declared at 635 feet MSL. At this point, flood control operations are coordinated with any resources needed. Any anticipated temporary dike construction is commenced depending on estimated final flood level.

D-PLA 3-6 November 2004

3. FACILITY SITE CHARACTERISTICS - (cont'd)

At 639 feet MSL, a flood emergency is declared. At 643 feet MSL a flood crisis level is declared. At this point, actions are taken to minimize the differential pressure on the Reactor Building. The warning available to the facility of flood cresting is 4-5 days following crest at Minneapolis, Minnesota.

3.4.7 Ultimate Heat Sink and Low Flow Conditions The ultimate heat sink of the La Crosse Boiling Water Reactor is the Mississippi River. Low flow to the site occurs in the fall and winter and the most frequently recorded lowest monthly average flow occurs in February. Minimum flows have also been recorded in August and September during periods of drought. Records of minimum and average flows maintained over the period of 1930 to 1955 at the United States Geological Survey Station at La Crosse were reviewed and are summarized as follows. These low flows should vary only slightly from those at the site.

Summary Flow Data for the Mississippi River at La Crosse Station 1930-1955:

Condition Discharging Cu. Ft./Sec.

All Time Low Flow Rate 3,200 December 30 and 31, 1933 Median of Annual Minimum Flow Rates 8,100 (Averaged over 1 day)

Overall Average Flow Rate 1930-1955 27,970 As of January 1, 1988, the fuel pool could be without active cooling for up to five days before boiling commenced. A test (LACBWR Technical Report 137) performed in September of 1993, after six years of decay, indicates that water temperatures would stabilize at a temperature less than boiling. This allows ample time to arrange for alternate cooling methods.

3.4.8 Ground Water As the site has valley sand overlaying a layer of Eau Claire sandstone of the Cambrian Age which is underlaid by a Mount Simon sandstone, wells have been driven in areas closest to the site but not in valleys characterized by sub-layers of Mount Simon sandstone. Deep wells penetrating the Mount Simon layer flow to the surface indicating an artesian head above the level of the river valley floor. Use of water from these artesian aquifers has been limited because the chemical quality of this deep water is poorer than that from shallow aquifers. As a result, there has been no extensive withdrawal of water and no serious decrease in the artesian head.

Therefore, an accidental release of contaminants cannot enter the artesian aquifer.

D-PLAN 3-7 November 2004

4-II Genoa Site Map FIGURE 3.4 I

D-PLAN

4.

FACILITY DESCRIPTION 4.1 GENERAL PLANT DESCRIPTION LACBWR was a nuclear power plant of nominal 50 Mw electrical output, which utilized a forced-circulation, direct-cycle boiling-water reactor as its heat source.

The reactor and its auxiliary systems were within a steel containment building. The turbine-generator and associated equipment, the control room for both turbine and reactor controls, and plant shops and offices were in a conventional building adjacent to the Reactor Building.

Miscellaneous structures which were associated with the power plant, and were located adjacent to the Turbine Building, include the electrical switchyard, Cribhouse, Waste Treatment Building, LSA Storage Building, oil pump house, stack, warehouses, administration building, annex building, guard house, outdoor fuel oil tanks, underground septic tanks, gas storage tank vaults, underground oil tanks and the condenser circulating water discharge seal well at Genoa Unit 3.

Miscellaneous onsite improvements included roads, walks, parking areas, yard lighting, fire hydrants required for plant protection, access to and use of rail siding facilities, fencing, landscaping, and communication services.

4.2 BUILDING AND STRUCTURES 4.2.1 Reactor Building The Reactor Building (same as Containment Building, Figs. 4.1 and 4.2) is a right circular cylinder with a hemispherical dome and semi-ellipsoidal bottom. It has an overall internal height of 144 ft. and an inside diameter of 60 ft., and it extends 26 ft. 6 in. below grade level. The shell thickness is 1.16 in., except for the upper hemispherical dome which is 0.60 in. thick.

The building contained most of the equipment associated with the nuclear steam supply system, including the reactor vessel and biological shielding, the fuel element storage well, the forced circulation pumps, the shutdown condenser, and process equipment for the reactor water purification system, decay heat cooling system, shield cooling system, seal injection system, emergency core spray system, boron injection system, and storage well cooling system.

The Reactor Building was designed to withstand the instantaneous release of all the energy of the l primary system to the Reactor Building atmosphere at an initial ambient temperature of 80'F, neglecting the heat losses from the building and heat absorption by internal structures. Its design pressure is 52 psig, compared to a calculated maximum pressure buildup of 48.5 psig following the maximum credible accident while in operation. The Reactor Building shell is designed and constructed according to the ASME Boiler and Pressure Vessel Code, Sections II, VIII, and IX, and Nuclear Code Cases 1270N, 1271N, and 1272N.

D-PLAN 4-1 November 2004

4. FACILITY DESCRIPTION - (cont'd)

The interior of the shell is lined with a 9-inch-thick layer of concrete, to an elevation of 727 ft.

10 in., to limit direct radiation doses in the event of a fission-product release within the Reactor Building.

The Reactor Building is supported on a foundation consisting of concrete-steel piles and a pile capping of concrete approximately 3 ft. thick. This support runs from the bottom of the semi-ellipsoidal head at about el. 612 ft. 4 in. to an elevation of 621 ft. 6 in. The 232 piles that support the containment structure are driven deep enough to support over 50 tons per pile.

The containment bottom head above el. 621 ft. 6 in. and the shell cylinder from the bottom head to approximately 9 in. above grade elevation (639 ft. 9 in.) are enveloped by reinforced concrete laid over a 1/2 in. thickness of premolded expansion joint filler. The reinforced concrete consists of a lower ring, mating with the pile capping concrete. The ring is approximately 414 ft. thick at its bottom and 212 ft. thick at a point I I/2 ft. below its top (due to inner surface concavity). The ring then tapers externally to a thickness of 9 in. at the top (el. 627 ft. 6 in.) and the 9 in.

thickness of concrete extends up the wall of the shell cylinder to 639 ft. 9 in. The filler and concrete are not used, however, where cavities containing piping and process equipment are immediately adjacent to the shell.

I Except for areas of the shell adjacent to other enclosures, the exterior surface of the shell above el. 639 ft. 9 in. is covered with 1 YIa-inch-thick siliceous fiber insulation, faced with aluminum.

The insulation of the dome is Johns-Manville Spintex of 9 lb/ft3 density, faced with embossed aluminum sheet approximately 0.032 in. thick. The insulation of the vertical walls is Johns-Manville Spintex of 6 lb/ft3 density, faced with corrugated embossed aluminum sheet approximately 0.016 in. thick. The insulation minimizes heat losses from the building and maintains the required metal temperature during cold weather, and reduces the summer air-conditioning load.

The shell includes two airlocks. The principal access to the shell is through the personnel airlock that connects the Reactor Building to the Turbine Building. The airlock is 21 ft. 6 in. long between its two doors, which are 5 ft. 6 in. by 7 ft. and are large enough to permit passage of a spent fuel element shipping cask. The Reactor Building can also be exited, if necessary, through the emergency airlock, which is 7 ft. long and 5 ft. in diameter, with two circular doors of 32Y/2 in. diameter (with a 30-in. opening). Both airlocks are at el. 642 ft. 9 in. and lead to platform structures from which descent to grade level can be made.

An 8 ft. by 10 ft. freight door opening in the Reactor Building accommodates large pieces of equipment. The door is bolted internally to the door frame in the shell.

Cables and bulkhead conductors from the Turbine Building provide electrical service to the Reactor Building through penetrations in the northwest quadrant of the building shell. The majority of pipe penetrations leave the Reactor Building I to 10 ft. below grade level and enter either at the northwest quadrant into the pipe tunnel that runs to the Turbine Building, or on the northeast side into the tunnel connecting the Turbine Building, Reactor Building, stack, and the water treatment and waste gas storage areas.

D-PLAN 4-2 November 2004

4. FACILITY DESCRIPTION - (cont'd)

A 45,000-gal. storage tank in the dome of the Reactor Building supplied water for the emergency core spray system and the building spray system. The storage tank provides water for reactor vessel and upper cavity fill, normal makeup, and other operations associated with fuel handling and the fuel element storage well.

A 50-ton traveling bridge crane with a 5-ton auxiliary hoist is located in the upper part of the Reactor Building. The bridge completely spans the building and travels on circular tracks supported by columns around the inside of the building just below the hemispherical upper head.

A trolley containing all the lifting mechanisms travels on the bridge to near the crane rail, and it permits crane access to any position on the main floor under the trolley travel-diameter. The lifting cables of both the 50-ton and the 5-ton hoists are also long enough to reach down through hatchways into the basement area. Hatches at several positions in the main and intermediate floors may be opened to allow passage of the cables and equipment.

The spent fuel is stored in racks in the bottom of the spent fuel storage well located adjacent to the reactor biological shielding in the Reactor Building. The storage rack system is a two-tier configuration such that each storage location is capable of storing two (2) fuel assemblies, one above the other. Fuel assemblies stored in the lower tier are always accessible (e.g., for periodic inspection) by moving, at most, one other assembly. Each storage rack consists of a welded assembly of fuel storage cells spaced 7 inches on center. A neutron absorbing B4CIPolymer Composite plate is incorporated between each adjacent fuel storage cell in each orthogonal direction. Horizontal seismic loads are transmitted from the rack structures to the fuel storage well walls at three elevations (the top grid of the upper tier rack section, the top grid of the lower tier rack section and the bottom grid of the lower tier rack section) through adjustable pads attached to the rack structures. The vertical dead-weight and seismic loads are transmitted to the storage well floor by the rack support feet. The fuel storage racks and associated seismic bracing are fabricated from Type 304 stainless steel.

4.2.2 Turbine Building The general location of the Reactor and Turbine Buildings is shown in Figure 4.3. The Turbine Building contained a major part of the power plant equipment. The turbine-generator was on the main floor. Other equipment was located below the main floor. This equipment included the feedwater heaters, reactor feedwater pumps, air ejector, vacuum pump, full-flow demineralizers, condensate pumps, air compressors, air dryer, oil purifier, service water pumps, component cooling water coolers and pumps, demineralized water system, domestic water heater, turbine oil reservoir, oil tanks and pumps, turbine condenser, unit auxiliary transformer, 2400-volt and 480-volt switchgear, motor control centers, diesel engine-generator sets, emergency storage batteries, inverters and other electrical, pneumatic, mechanical and hydraulic systems and equipment required for a complete power plant. A 30/5-ton capacity, pendant-operated overhead electrical traveling crane spanned the Turbine Building. The crane has access to major equip-ment items located below the floor through numerous hatches in the main floor. A 40-ton capacity, pendant-operated overhead electric crane spanned the space between turbine building loading dock and Waste Treatment Building.

D-PLAN 4-3 November 2004

4. FACILITY DESCRIPTION - (cont'd)

The Turbine Building also contained the main offices, the Control Room (for both turbine-generator and reactor), locker room facilities, laboratory, shops, counting room, personnel change room, and decontamination facilities, heating, ventilating and air conditioning equipment, rest rooms, storeroom, and space for other plant services. In general, these areas were separated from power plant equipment spaces. The Control Room is on the main floor on the side of the Turbine Building that is adjacent to the Reactor Building. The general arrangement of the Reactor and Turbine Buildings is shown in Figures 4.3 through 4.5.

4.2.3 Waste Treatment Building and LSA Storage Building The Waste Treatment Building (WTB) is located to the northeast of the Reactor Building. The building contains facilities and equipment for decontamination and the collection, processing, storage, and disposal of low level solid radioactive waste materials in accordance with the Process Control Program.

The grade floor of the Waste Treatment Building contains a shielded compartment which encloses a 320 ft3 stainless steel spent resin receiving tank with associated resin receiving and transfer equipment. A high integrity disposal liner can be located in the adjacent shielded cubicle.

Adjacent to these shielded resin handling cubicles are two open cubicles, one of which is about 3' above grade. The grade level area contains two back-washable radioactive liquid waste fiqters, the spent resin liner level indication panel and the spent resin liner final dewatering piping, container, and pumps. The second above-grade area is a decontamination facility, consisting of a steam cleaning booth, a decontamination sink, and heating/ventilation/air conditioning units.

The remaining grade or above-grade areas contain a shower/wash/frisking area, and the dry active waste (DAW) compactor unit and temporary storage space for processed DAW containers.

Beneath the grade floor are two shielded cubicles. One cubicle, to which access is gained by removal of floor shield plugs, is available for the storage of up to nine higher activity solid waste drums. The other area, to which access is gained by a stairway, contains the dewatering ion exchanger, the WTB sump and pump, and additional waste storage space.

The Waste Treatment Building ventilation is routed through a HEPA filter to the stack plenum.

The building is normally maintained at a negative pressure. The general arrangement of the WTB is shown on Figure 4.5.

The LSA Storage Building is southwest of the Turbine Building. It is used to store processed, packaged and sealed low level dry active waste materials, and sealed low level activity components for a period of approximately 5 years. The building has the capacity for 500 DOT17H-55 gallon drums of waste. No liquids are stored in this building. There are no effluent releases from this building during normal use.

D-PLAN 44 November 2004

4. FACILITY DESCRIPTION - (cont'd) 4.2.4 Cribhouse The Cribhouse is located on the bank of the Mississippi River to the west of the plant and through its intake structure, provides the source of river water to the various pumps supplying river water to the plant. The Cribhouse contains the diesel-driven high pressure service water pumps, traveling screens, low pressure service water pumps and the circulating water pumps.

a-D-PLAN 4-5 November 2004

5. PLANT STATUS - (cont'd) 5.2 PLANT SYSTEMS AND THEIR STATUS 5.2.1 Reactor Vessel and Internals The reactor vessel consists of a cylindrical shell section with a formed integral hemispherical bottom head and a removable hemispherical top head which is bolted to a mating flange on the vessel shell to provide for vessel closure. The vessel has an overall inside height of 37 feet, an inside diameter of 99 inches, and a nominal wall thickness of 4 inches (including 3/16-inch of integrally bonded stainless steel cladding).

The reactor vessel is a ferritic steel (ASTM A-302-Gr-B) plate with integrally bonded Type 304L stainless steel cladding. The flanges and large nozzles are ferritic steel (ASTM A-336) forgings.

The small nozzles are made of Inconel pipe.

The reactor internals consist of the following: a thermal shield, a core support skirt, a plenum separator plate, a bottom grid assembly, steam separators, a thermal shock shield, a baffle plate structure with a peripheral lip, a steam dryer with support structure, an emergency core spray tube bundle structure combined with fuel holddown mechanism, control rods and the reactor core.

System Status All fuel assemblies and startup sources have been removed from the reactor core. The 29 control l rods and other core components remain in the reactor vessel. The reactor vessel head is installed l and partially bolted in place. The reactor vessel and primary systems have been drained to the maximum extent practical. The reactor vessel is capable of being refilled. The control rods may be removed to the Fuel Element Storage Well or a licensed facility during SAFSTOR.

D-PLAN 5-3 November 2004

5. PLANT STATUS - (cont'd) 5.2.15 Hydraulic Valve Accumulator System The major components of the Hydraulic Valve Accumulator System are mounted on a common bed plate on the grade floor of the Reactor Building. The system consists of a water accumulator tank, a water return sump tank, two air compressors, two water pumps, piping, valves, and the necessary instrumentation and controls.

Approximately 300 gallons of demineralized water is maintained in the Water Accumulator Tank by pumps which take a suction from the Water Return Sump Tank. The level is automatically maintained by a float switch which operates the pumps as required. The water is stored in the Accumulator Tank under 140 psi air pressure which is supplied by air compressors. The air compressors are automatically controlled by a pressure switch and are interlocked with the level float switch to prevent the compressors from running while a pump is running.

The function of the Hydraulic Valve Accumulator System is to supply the necessary hydraulic force to operate the five piston-type valve actuators, which operate the five Rotoport valves in the Forced Circulation and Main Steam Systems.

System Status This system has been drained. The air compressors, water pumps, and other equipment have been electrically disconnected and are not maintained operational.

D-PLAN 5-17 November 2004

5. PLANT STATUS - (cont'd) 5.2.26 Steam Turbine The turbine was a high pressure, condensing, reaction, tandem compound, reheat 3600 rpm unit rated at 60,000 KW with the following steam conditions: 1250 psig, 5470F, exhausting at 1.0" Hg Absolute. The turbine consisted of a high pressure and intermediate pressure and a low pressure element.

System Status Steam piping in the Turbine Building, turbine inlet valves, and other components and instru-mentation have been removed for reprocessing and disposal. Complete removal of the Steam Turbine system is in progress.

D-PLAN 5-28 November 2004

5. PLANT STATUS - (cont'd) 5.2.27 60-Megawatt Generator The 60-Mw generator was a high-speed turbine-driven wound-rotor machine that was rated at 76,800 kva, 85 percent P.F., 3600 rpm, 60 cycle, 3 phase, 13,800v A-C, and 3213 amp. The generator was cooled by a hydrogen system, lubricated by a forced-flow lubricating system, and excited by a separate exciter attached to the end of the generator shaft through a reduction gear.

A reserve exciter was provided.

System Status The main and reserve exciters have been disposed of The generator rotor has been removed and unconditionally released for reuse. Complete removal of the 60-megawatt Generator system is in progress.

D-PLAN 5-29 November 2004

5. PLANT STATUS - (cont'd) 5.2.28 Turbine Oil and Hydrogen Seal Oil System The Turbine Bearing Oil System received cooled oil from the lube oil coolers to supply the necessary lubricating and cooling oil (via a bearing oil pressure regulator) to the turbine and generator bearings, exciter bearings, and exciter reduction gear. During normal operation, the necessary oil pressures were provided by the attached lube oil pump. During startup and shutdown, an ac motor-driven auxiliary lube oil pump provided oil pressure. Backup protection consisted of the ac turbine bearing oil pump and the dc emergency bearing oil pump.

The Hydrogen Seal Oil System received cooled oil from the lube oil coolers and supplied this oil, via a pressure regulator, to the inboard and outboard hydrogen seals of the generator.

Backup protection was provided in the event the normal supply pressure dropped or was lost, with an ac hydrogen seal oil pump and a dc emergency hydrogen seal oil pump.

Flexibility of the Turbine Oil Transfer System was brought about by the piping arrangement that allowed the lubricating oil to be transferred or purified from several sources. With the lube oil transfer pump, turbine oil could be transferred from the lube oil reservoir to either the clean oil or dirty oil tanks located in the oil storage room.

System Status This system is not required to be operational. Turbine oil reservoir, clean and dirty oil tanksare drained. Hydrogen Seal Oil cooler piping, cooling water outlet and inlet manifolds, and instrumentation have been removed. Turbine Oil system pumps and other equipment have been electrically disconnected and will be removed with the remainder of the system.

D-PLAN 5-30 November 2004

5. PLANT STATUS - (cont'd) 5.2.33 Electrical Power Distribution 5.2.33.1 Normal AC Distribution 69-KV power is supplied to the reserve auxiliary transformer located in the LACBWR switchyard through a three-phase air disconnect switch and three 30-amp, 69-KV fuses.

Air Circuit Breakers 252RIA and 252R1B supply the 2.4-kv Bus IA and Bus lB from the 69/2.4-KV reserve transformer.

The 2400/480-volt Auxiliary Transformers IA and lB receive their power from the 2400-volt Buses 1A and lB through Air Circuit Breaker 252AT1A from Bus IA to Transformer 1A, and through Air Circuit Breaker 252AT1B from Bus lB to Transformer lB. The auxiliary transformers supply the 480-volt Buses IA and lB through Air Circuit Breaker 452M1A for Bus IA and through Air Circuit Breaker 452M1B for Bus IB.

The 480-volt buses supply larger equipment directly. They also supply motor control centers which furnish power to motors and other associated equipment connected to them through their respective breakers, including Motor Control Center (MCC) 120-volt ac Distribution Panels which supply 120-volt ac to equipment and instrumentation.

The regular lighting cabinets are supplied from 480-volt buses IA and lB.

5.2.33.2 480-V Essential Buses 1A and IB The 480-v Essential Bus IA Switchgear is normally supplied with electrical power from the 480-v Bus IA through Breaker 452-52A. In the event of a loss of station power, the 480-v Essential Bus IA is supplied with electrical power from Emergency Diesel Generator IA through Breaker 452 EGA. Breakers 452-52A and 452 EGA are electrically interlocked to prevent both sources from supplying the bus.

The 480-v Essential Bus lB Switchgear is normally supplied with electrical power from 480-v Bus IB through Breaker 452-52B. In the event of a loss of station power, the 480-v Essential Bus lB is supplied with electrical power fiom Diesel Generator IB through Breaker 452 EGB.

Breakers 452-52B and 452 EGB are electrically interlocked to prevent both from supplying the bus.

The 480-v Essential Buses IA and lB may be cross-connected through the 480-v Essential Bus Tie Breakers 452 TBA and 452 TBB.

5.2.33.3 Emerzency Diesel Generators IA and lB The IA Diesel Generator set system consists of a 250-kw diesel generator, a day tank fuel supply, a fuel transfer pump, a remote radiator and fan, a 100-kw test load, a local engine instrument panel, a local generator panel, and a remote selector switch and alarms in the Control Room. The Diesel Generator set is located in the emergency generator cubicle which is on the grade floor level adjacent to the Machine Shop.

D-PLAN 5-35 November 2004

5. PLANT STATUS - (cont'd) 5.2.34.1 Reactor Building Atmosphere PASS System Description The Reactor Building Atmosphere Post-Accident Sampling System consists of a vacuum pump which takes a suction on the Reactor Building atmosphere at the 714' level. The atmosphere sample is drawn through two solenoid operated isolation valves, a heat exchanger, and moisture trap. Then the sample is discharged to the two in-parallel hydrogen analyzers with preset flowmeters; then either through a bypass line or a remote sample cylinder and back to the Reactor Building at the 676' level through two solenoid operated isolation valves.

5.2.34.2 Stack Gas PASS System Description The Stack Gas Post-Accident Sampling System makes use of the same equipment that provides the normal stack gas sample flow. The vacuum pump for stack gas sampling draws the extra flow, above what the stack monitors draw, to make the total flow isokinetic to the stack discharge. This flow can be diverted through the post-accident sample canister by opening manual isolation valves. The sample canister is connected to the system by two quick disconnects and, therefore, can be easily removed from the system and taken to the laboratory for analysis. The sample canister diversion valve is controlled from the local control panel in the No. 3 Feedwater Heater area.

5.2.34.3 Reactor Coolant PASS System Description The Reactor Coolant Post-Accident Sampling System took primary coolant from an incore flux monitoring flushing connection, through 2 solenoid-operated isolation valves with a heat exchanger between them, to a motor-operated pressure reducing valve. Downstream of the pressure reducing valve, the coolant sample could be diluted with demineralized water which then flowed through the sample cylinder or its bypass valve, through another solenoid isolation valve, and back to the Reactor Building basement or to the waste water tanks.

System Status The Stack Gas PASS System is maintained in continuous operation. The Reactor Coolant PASS System has been removed. The Reactor Building Atmosphere PASS System is retained in place.

5.2.35 Containment Integrity Systems With the plant in the SAFSTOR condition, there is no longer a postulated accident that would result in containment pressurization or that takes credit for Containment integrity.

System Status Containment integrity systems are not required to be operable.

D-PLAN 5-38 November 2004

5. PLANT STATUS - (cont'd) 5.4.2 System Radiation Levels During SAFSTOR the major radioactively contaminated systems at LACBWR will be monitored in order to trend system cleanups and radioactivity decay. A program consisting of 100 survey points located throughout the plant has been established. Initial system contact readings have been taken and will be monitored on a frequency determined to adequately trend any radiation level changes. The individual survey locations may change during the SAFSTOR period as plant parameters change.

The following is a list of the initial survey points, their initial dose rates, and the current surveyl point dose rates.

I Note: All readings are contact dose rates.

Survey Point #

Survey Point Location I

Condensate Line to and from OHST 2

Condensate Line to and from OHST 3

Condensate Line to and from OHST 4

1A Condensate Pump Discharge Line 5

Emergency Overflow Line 6

Emergency Overflow Bypass Line 7

Ice Melt Line 8

1A Reactor Feed Pump 9

Near lB Reactor Feed Pump Discharge Valve 10 Side of #3 Feedwater Heater 11 Reheater Level Control Chamber 12 South End of Reheater 13 Gland Exhaust Condenser Loop Seal 14 Main Steam Line 15 Main Steam Line 16 Offgas System Flame Arrestor 17 lB Waste Water Pump 18 IA Waste Water Pump 19 End of 3000 Gallon Waste Tank 20 End of 4500 Gallon Waste Tank 21 Side of Gland Seal Steam Generator 22 Side of Gland Seal Steam Generator 23 Main Steam Bypass Line 24 Turbine Inlet Valve Body 25 Main Steam Line 26 Reheat to Flash Tank Line Initial Dose Rate (mRem/hr) 25 24 33 12 27 33 3

16 11 26 26 13 35 48 50 8

26 60 170 120 1100 160 17 23 24 11 Current I

Dose Rate (mRemhr) 1 9

<1 4

9 14 18

• Survey Point removed due to dismantlement activities.

I D-PLAN 5-41 November 2004

5. PLANT STATUS - (cont'd)

Survey Point #

Survey Point Location 27 Flash Tank 28 Seal Injection Heater 29

1. 2 Feedwater Heater Bypass Line 30 Feedwater Heater Bypass Line 31 Bottom of Gland Exhaust Condenser 32 Top of Gland Exhaust Condenser 33 Condensate into Air Ejector Line 34 Air Ejector 35 Low Pressure Turbine Manhole Cover 36 End of High Pressure Turbine 37 Primary Purification IA Filter Inlet Line 38 Primary Purification Pump 39 Exhaust Ventilation Duct 40 Reactor Bldg. Grade Level N Shield Wall 41 IA Fuel Element Storage Well Pump 42 lB Fuel Element Storage Well Pump 43 FESW Filter Discharge Line 44 FESW System Cooler 45 Hydraulic Valve Actuation System Header 46 Base of Hydraulic Valve Accumulator 47 Wall at Electrical Penetration 48 Handrail on NW Nuclear Instrumentation (NI)

Platform 49 Shield Wall on N NI Platform 50 Primary Purification to OHST Line 51 Above Primary Purification Cooler Inlet Valve 52 Cold Leg of Reactor High Level Transmitter Line 53 Seal Injection Reservoir 54 Reactor Cavity Drain Line 55 IA Core Spray Pump Discharge Line 56 Reactor Water Level Sightglass Line 57 Reactor Water Level Sightglass Line 58 Reactor Bldg. Mezzanine Level N Shield Wall 59 Steam Trap Reactor Bldg. Mezzanine Level NW Wall 60 Fuel Element Storage Well Line 61 Fuel Element Storage Well Line 62 Fuel Element Storage Well Line Initial Dose Rate (mRemlhr) 5 31 100 24 170 20 7

8 6

2 38 140 9

6 70 80 180 1000 60 24 30 100 4

6 25 46 30 44 10 180 100 4

23 400 420 60 Current I

Dose Rate (mRem/hr)

<1

<1

<1 4

10 1

<1 5

7 11 30-3 1

1 4

<1

<1 5

12 5

<1 2

9 3

• Survey Point removed due to dismantlement activities.

I D-PLAN 5-42 November 2004

5. PLANT STATUS - (cont'd)

Survey Point #

63 64 65 66 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 85 86 087 88 89 90 91 92 93 94 95 96 97 98 99 100 Survey Point Location Fuel Element Storage Well Skimmer Line Wall near Fuel Transfer Canal Drain Relief Valve Platform at Level Transmitter Shutdwn Condenser Shutdown Condenser Condensate Line lB Retention Tank IA Retention Tank By Primary Purification Cation Tank Decay Heat Cooler Decay Heat Cooler Decay Heat Cooler Bypass Valve Decay Heat Pump Suction Line Handrail at Shutdown Condenser Condensate Valves Seal Injection DP Transmitter Top of Upper Control Rod Drive Mechanism Top of Upper Control Rod Drive Mechanism Wire mesh screen on N Upper Control Rod Platform Bottom of Upper Control Rod Drive Mechanism Top of Upper Control Rod Drive Mechanism Bottom of Upper Control Rod Drive Mechanism Effluent Lines on Upper Control Rod Platform Sump Pump Discharge Line to Retention Tank At Forced Circulation Pump Filters Retention Tank Pump Under Lower Control Rod Drive Mechanism Control Rod Drive Hydraulic System Header Decay Heat Pump lB Forced Circulation Pump Suction Line lB Forced Circulation Pump Suction Line IA Forced Circulation Pump Suction Line IA Forced Circulation Pump Suction Line IA Forced Circulation Pump Discharge Line Feedwater Line in Forced Circulation Cubicle IA Forced Circulation Pump Handrail at IA Forced Circ. Pump Suction Line IA Forced Circulation Pump Discharge Line IA Forced Circulation Pump Discharge Line IA Forced Circulation Pump Suction Line Initial Dose Rate (mRemlhr) 90 35 80 11 6

300 130 24 25 18 70 32 28 44 370 200 22 1000 500 800 390 260 33 60 246 190 150 1000 1100 500 600 700 130 130 250 800 600 700 Current Dose Rate (mRem/hr) 6 4

9 36 10 1

4 5

13 27 6

50 33 3

90-50 70 19 4

6 22 19 130 200 130 90 130 35 22 24 90 60 60 I

• Survey Point removed due to dismantlement activities.

I D-PLAN 5-43 November 2004

5. PLANT STATUS - (cont'd) 5.5 PLANT PERSONNEL DOSE ESTIMATE During normal/routine SAFSTOR operations at LACBWR, average whole body radiation dose received by plant personnel should be no more than 0.600 Rem per individual per year. This average dose is expected to decrease during the SAFSTOR period due to isotopic decay.

Individual doses will be dependent upon work being performed. Plant personnel will not be allowed to exceed 5.0 Rem/year.

5.6 SOURCES As authorized by the facility license, sealed sources for radiation monitoring equipment calibration, reactor instrumentation, reactor startup, and fission detectors will continue to be possessed and/or used. Additionally, sources will be used as authorized without restriction to chemical or physical form for sample analysis, instrument calibration and/or as associated with radioactive apparatus and components.

5.7 RADIATION MONITORING INSTRUMENTATION Radiation monitoring instrumentation for the LACBWR consists of fixed plant surveillance equipment, portable survey meters, laboratory-type counting instrumentation, and personnel monitoring equipment.

b The Radiation Monitoring System performs the following functions:

(I)

Provides a permanent record of radioactivity levels of plant effluents.

(2)

Provides alarms and automatic valve closures to prevent excessive radioactive releases to environment.

(3)

Provides warning of leakage of radioactive gas, liquid, or particulate matter within the plant.

(4)

Provides continuous radiation surveillance in normally accessible plant areas.

(5)

Provides portable instrumentation for use in conducting radiation surveys.

(6)

Provides instrumentation for personnel and material contamination surveillance, including that necessary for control of egress from restricted areas.

(7)

Provides pocket dosimeters and necessary charging and readout equipment for personnel radiation exposure control and estimates.

D-PLAN 544 November 2004

5. PLANT STATUS - (cont'd) 5.7.1 Fixed Plant Monitors The plant fixed surveillance monitoring equipment consists of liquid monitors, air monitors, and area monitors.

5.7.1.1 Liquid Monitors. The liquid monitors consist of a modular nim bin electronic system in the Control Room coupled to a NaI scintillation detector. The Nal scintillation detector is coupled to a photomultiplier tube base-preamplifier.

5.7.1.2 Reactor Building Air Exhaust Gaseous and Particulate Monitor. A monitor is located on the Reactor Building mezzanine level. This monitor has a fixed filter particulate detector and a gaseous detector. It takes suction from the outlet of the Reactor Building ventilation filters.

5.7.1.3 Stack Monitor. A monitor is installed to sample the stack emissions. This monitor draws air from the stack through an isokinetic nozzle. This monitor detects particulate and gaseous activity released to the stack. This monitor alarms locally and in the control room.

5.7.1.4 Fixed Location Monitors. Area radiation monitors are used to detect and measure gamma radiation fields at various remote locations. There are fifteen remote units located throughout the plant. The measured dose rate is displayed on meters located in the Control Room.

5.7.2 Portable Monitors Portable instruments are located throughout the plant. Instruments are available to detect various levels of beta, gamma, and alpha radiation.

5.7.3 Laboratorv-TTpe Monitors Laboratory instruments are available to determine contamination levels and radioisotope concentrations. These instruments consist of internal proportional counters, gamma analyzers, and liquid scintillation counters.

D-PLAN 5-45 November 2004

6. DECOMMISSIONING PROGRAM - (cont'd) stationed at or visiting LACBWR comply with it in spirit as well as regulation. This supervisor will also assign the day-to-day duties of the health physics technicians.

The Health Physics Technicians will be responsible for the radiation protection and chemistry programs at LACBWR. They will perform all tasks required for surveillance and will provide all work coverage required by special work permits. They will maintain as required the exposure records of personnel, take all the readings necessary to guard against the spread of contamination and provide input to the long-term radionuclide inventory program. They will report, as directed by the Health and Safety Supervisor, to the Duty Shift Supervisor as required.

The Mechanical Maintenance Lead Mechanic is responsible for the assignment of mechanical maintenance duties and will direct the completion of all maintenance requests and surveillance tests of a mechanical nature. He is responsible for the preventive maintenance program established on those systems necessary to maintain the SAFSTOR condition. The Lead Mechanic is responsible for overall maintenance on all of the plant equipment which may serve as backups to the required systems or backup supplies to the rest of the Dairyland system.

Maintenance Mechanics are responsible for the completion of all mechanical maintenance tasks.

These tasks include all surveillance requirements and work requests defined in maintenance orders as well as general duties as assigned by the Lead Mechanic.

The Administrative Assistant is responsible for overall administration of LACBWR. She will l

maintain all records required under technical specifications for plant operation and will maintain a record of all activities of the SAFSTOR mode. The Administrative Assistant will ensure that all clerical functions are performed adequately. She will maintain all budget expense and project accounts and will coordinate preparation of the LACBWR budget. Duties will also include assigning to staff personnel required responses to regulatory agencies, other Dairyland departments, etc., and ensuring that these tasks are completed by the established deadline.

Additional administrative personnel will be made available to the Administrative Assistant as needed, and will assist in the clerical tasks at LACBWR. Such additional personnel will be qualified to perform required communication functions and will be assigned other tasks, as necessary, by the Administrative Assistant.

The Licensing Engineer will be responsible for all facility licensing. This will include steps preparatory to eventual shipment of SAFSTOR fuel and proceeding into the DECON mode. The Licensing Engineer will be the principal liaison on behalf of the Plant Manager for the contact with the Nuclear Regulatory Commission and other regulatory agencies. This engineer will be D-PLAN 6-3 November 2004

6. DECOMMISSIONING PROGRAM - (cont'd)

The LACBWR Spent Fuel (333 assemblies) is stored under water in the high density spent fuel storage racks in the LACBWR Fuel Storage Well which is located adjacent to the reactor in the LACBWR Reactor Building.

Additional small quantities of SNM are contained in neutron and calibration sources, which are appropriately stored at various locations in the LACBWR plant.

All fuel handling and all shipment and receipt of SNM is accomplished according to approved written procedures. Appropriate accounting records will be maintained and appropriate inventories, reports and documentation will be accomplished by or under the direction of the LACBWR Accountability Representative in accordance with the requirements set forth in 10 CFR 70, 10 CFR 73 and 10 CFR 74.

6.9 SAFSTOR FIRE PROTECTION 6.9.1 Fire Protection Plan LACBWR can safely maintain and control the Fuel Element Storage Well in the case of the worst postulated fire in each area of the plant.

The fire protection plan at LACBWR is to prevent fire, effectively respond to fire, and to minimize the risk to the public from fire emergencies. The goals of the fire protection plan are fire prevention and fire protection. This fire protection plan, implemented through the fire protection program, provides defense-in-depth to fire emergencies and addresses the following objectives:

• Prevent fires. By administratively controlling ignition sources, flammable liquid inventory, and combustible material accumulation, fire risk is reduced. Welding and other hot work shall be performed only under Special Work Permit conditions and the use of a fire watch shall be required. Routine fire and safety inspections by LACBWR staff shall be conducted to ensure flammable liquids are properly stored and combustible material is removed. These inspections shall also require identification of fire hazards and result in action to reduce those hazards. General cleanliness and good housekeeping shall continue as an established practice and shall be checked during inspection.

• Rapidly detect, control, and extinguish fires that do occur and could result in a radiological hazard. Fire detection systems are installed to detect heat and smoke in spaces and areas of the protected premises of LACBWR. If fire detection systems or components are unavailable, increased monitoring of affected areas by personnel shall compensate for any loss of automatic detection. Fire barriers provide containment against the spread of fire between areas and provide protection to personnel responding to fire emergencies. Areas of high fire loading are provided with automatic reaction-type fire suppression systems or manually initiated fire suppression systems. These installed systems provide immediate fire D-PLAN 6-14 November 2004

6. DECOMMISSIONING PROGRAM - (cont'd) suppression automatically or provide the means to extinguish fires without fire exposure to personnel manually initiating them. Manual fire extinguishing equipment is installed in all areas of the LACBWR facility. All fire protection equipment and systems are maintained, inspected, and tested in accordance with established guidelines. Compensatory actions and procedures for the impairment or unavailability of fire protection equipment are provided. A trained fire brigade, available at all times shall respond immediately to all fire emergencies.

The function of the response by the fire brigade shall be to evaluate fire situations, to extinguish incipient stage fires, and to quickly realize the need for, and then summon, outside assistance. For any situation where a fire should progress beyond the incipient stage, qualified outside fire services shall provide assistance.

Minimize the risk to the public, environment, and plant personnel resulting from fire that could result In a release of radioactive materials. Surface contamination has been reduced to minimal levels in most areas of the facility by cleanup efforts and the effects of long-term decay. Contamination surveys are performed routinely and areas identified for attention are deconned further. Good radiological work practices and contamination control are maintained. Radioactive waste generated is containerized and shipped for processing in accordance with approved procedures. Liquid effluents are collected in plant drain systems, processed, and monitored during discharge. Plant personnel are alerted to elevated radioactivity levels by area radiation monitors and air monitoring systems that are in operation at all times in buildings of the radiological controlled area. Gaseous and particulate air activities are continuously monitored prior to their release to the environment.

Procedures and protocols exist to ensure risk is minimized to the public and members oftthe outside fire service.

6.9.2 Fire Protection Program The fire protection program for the LACBWR facility is based on sound engineering practices and established standards. The function of the fire protection program is to provide the specific mechanisms by which the fire protection plan is implemented. The fire protection program utilizes an integrated system of administrative controls, equipment, personnel, tests, and inspections. Components of the fire protection program are:

6.9.2.1 Administrative Controls are the primary means by which the goal of fire prevention is accomplished. Administrative controls also ensure that fire protection program document content is maintained relevant to its fire protection function. By controlling ignition sources, combustible materials, and flammable liquids, and by maintaining good housekeeping practices, the probability of fire emergency is reduced. Procedures are routinely reviewed for adequacy and are revised as conditions warrant.

6.9.2.2 Fire Detection System. The LACBWR plant fire detection system is designed to provide heat and smoke detection. A Class B protected premises fire alarm system is installed which uses ionization or thermal-type fire detectors. Detectors cover areas throughout the plant and outlying buildings. The plant fire alarm system control panel is located in the Control Room. Alarms as a result of operation of a protection system or equipment, such as water D-PLAN 6-15 November 2004

6. DECOMMISSIONING PROGRAM - (cont'd) flowing in a sprinkler system, the detection of smoke, or the detection of heat, are sounded in the Control Room. Alarm response is initiated from the Control Room.

The Administration Building fire detection system provides alarm functions using a combination of thermal detectors ionization detectors, and manual pull stations. Audible alarms are sounded throughout the building and provide immediate notice to occupants of fire emergency. The control panel for the Administration Building fire detection system is located within the Security Electrical Equipment Room.

6.9.2.3 Fire Barriers are those components of construction (walls, floors, and doors) that are rated in hours of resistance to fire by approving laboratories. Any openings or penetrations in these fire barriers shall be protected with seals or closures having a fire resistance rating equal to that of the barrier. The breaching of fire barriers is administratively controlled to ensure their fire safety function is maintained.

6.9.2.4 Fire Suppression Water System. The fire suppression water system is designed to provide a reliable supply of water for fire extinguishing purposes in quantities sufficient to satisfy the maximum possible demand. Fire suppression water is supplied by the High Pressure Service Water System (HPSW) which is normally pressurized from the Low Pressure Service Water (LPSW) system. Two HPSW diesel pumps provide fire suppression water when started manually or when started automatically by a decrease in HPSW pressure to <60 PSIG. Fire suppression water can be supplied from Genoa Station No. 3 (G-3) as a backup system to the HPSW system.

Fire suppression water is available from an external underground main at five 6-inch fire hydrants spaced at 200-foot intervals around the plant. Four outside hose cabinets contain the necessary hoses and equipment for hydrant operation.

Fire suppression water is available at five hose cabinets in the Turbine Building, one hose reel in the IB Diesel Generator Building, and one hose cabinet in the Waste Treatment Building. Fire suppression water is available from hose reels located on each of four levels in the Reactor Building.

Fire suppression water is also supplied to sprinkler systems in areas with high fire loads.

Sprinkler systems suppress fire in these areas without exposure to personnel. Automatic sprinkler systems are installed in the Oil Storage Room and in the Crib House HPSW diesel pump and fuel tank area. A manually initiated sprinkler system is installed in IA Diesel Generator Room. An automatic reaction-type deluge system protects the Reserve Auxiliary Transformer located in the LACBWR switchyard.

6.9.2.5 Automatic Chemical Extinguishing Systems are installed in two areas of LACBWR containing high fire loads. The IB Diesel Generator Room is protected by a CO2 Flooding system. The Administration Building Records Storage Room is protected by a Halon system.

These systems automatically extinguish fire using chemical agents, upon detection by their associated fire protection circuits. Fire in these areas is extinguished without exposure to personnel.

D-PLAN 6-16 November 2004

6. DECOMMISSIONING PROGRAM - (cont'd) 6.9.2.6 Portable Fire Extinguishers and Other Fire Protection Equipment. An assortment of dry chemical, CO2, and Halon portable fire extinguishers rated for Class A, B, and C fires are located throughout all areas of the LACBWR facility. These extinguishers provide the means to immediately respond to incipient stage fires. Spare fire extinguishers are located on the Turbine Building grade floor.

Portable smoke ejectors are provided for the removal of smoke and ventilation of spaces. Smoke ejectors are located in the Change Room, on the Turbine Building mezzanine floor, and in the Maintenance Shop.

Four outside hose cabinets contain necessary lengths and sizes of fire hose for use with the yard fire hydrants. These hose cabinets also contain hose spanner and hydrant wrenches, nozzles, gate valves, coupling gaskets, and ball-valve wye reducers.

Tool kits are located in the Crib House outside fire cabinet and in the Maintenance Shop. Spare sprinkler heads and other sprinkler equipment is located in the Change Room locker. Recharge-able flashlights are wall-mounted in various locations and at entries to spaces. Portable radios are available at various locations and used for Fire Brigade communication.

6.9.2.7 The Fire Brigade is an integral part of the fire protection program. The Fire Brigade at LACBWR shall be organized and trained to perform incipient fire fighting duties. Personnel qualified to perform Operations Department duties and all LACBWR Security personnel sh4il be designated as Fire Brigade members and trained as such. Fire Brigade responsibilities shall be assigned to members of these groups while on duty.

The Fire Brigade shall be a minimum of two people at all times. The Duty Shift Supervisor (or his designee) shall respond to the fire scene as the Fire Brigade Leader. One member of the Security detail shall respond, as directed by the Fire Brigade Leader, and perform duties as the second Fire Brigade member.

The Control Room Operator shall communicate the status of fire detection system alarms or specific hazard information with the Fire Brigade, shall monitor and maintain fire header water pressure, and shall expeditiously summon outside fire service assistance as directed by the Fire Brigade Leader. The Control Room Operator shall use the page system to announce reports of fire, evacuation orders, and other information as requested by the Fire Brigade Leader.

6.9.2.8 Outside Fire Service Assistance. The LACBWR Fire Brigade is organized and trained as an incipient fire brigade. Fire Brigade Leaders are responsible for recognizing fire emergencies that progress beyond the limits of incipient stage fire fighting. Fire Brigade Leaders shall then immediately request assistance from outside fire services.

The LACBWR Emergency Plan contains a letter of agreement with the Genoa Fire Department.

This letter of agreement states that the Genoa Fire Department is responsible for providing rescue and fire fighting support to LACBWR during emergencies. Upon request by the Genoa Fire Chief, all fire departments of Vernon County can be coordinated and directed to support the l

Genoa Fire Department during an emergency at LACBWR.

D-PLAN 6-17 November 2004

6. DECOMMISSIONING PROGRAM - (cont'd) 6.9.2.9 Reporting. Fire emergencies shall be documented under the following reporting guidelines:

• Any fire requiring Fire Brigade response shall be reported by the Duty Shift Supervisor using a LACBWR Incident Report.

• Any incident requiring outside fire service assistance within the LACBWR Site Enclosure (LSE fence) shall require activation of the Emergency Plan and shall require declaration of Unusual Event.

6.9.2.10 Training. Security badged visitors and contractors located at LACBWR shall receive indoctrination in the areas of fire reporting, plant evacuation routes, fire alarm response, and communications systems under General Employee Training.

Personnel who work routinely at LACBWR, and are given basic practical fire fighting instruction annually, are termed designated employees.

In addition to the annual practical fire fighting instruction, Fire Brigade members shall receive specific fire protection program instruction and participate in at least one drill annually.

Personnel not subject to Fire Brigade responsibilities shall receive training prior to performing fire watch duties.

6.9.2.11 Records. Fire Protection records shall be retained in accordance with Quality Assurance records requirements.

6.10 SECURITY DURING SAFSTOR AND/OR DECOMMISSIONING During the SAFSTOR status associated with the LACBWR facility, security will be maintained at a level commensurate with the need to insure safety is provided to the public from unreasonable risks.

Guidance and control for security program implementation are found within the LACBWR Security Plan, Safeguards Contingency Plan, Guard Force Training and Qualification Plan, and Security Control Procedures. The Security Plan for Transportation of LACBWR Hazardous Materials is found in the Process Control Program.

D-PLAN 6-18 November 2004

(

LA CROSSE BOILING WATER REACTOR SAFSTOR STAFF

(

I I

PLANT MANAGER I

ORC, SRC, QA OPERATIONS/ I&E/ MECHANICAL MAINTENANCE GROUP 2 Instrument Technicians 2 Electricians 5 Shift Supervisors 6 Operators 3 Mechanics I

HEALTH & SAFETY SUPERVISOR 3 Health Physics Technicians I

I ADMINISTRATIVE ASSISTANT Additional Administrative Personnel as Needed LICENSING ENGINEER REACTOR / RADIATION PROTECTION ENGINEER

• I Assumes Cooperative-wide Security & QA

• Duties to be performed with assistance of qualified consultants when necessary.

FIGURE 6.1 D-PLAN November 2004

7.

DECOMMISSIONING ACTIVITIES 7.1 PREPARATION FOR SAFSTOR The plant was shut down on April 30, 1987. Reactor defueling was completed June 11, 1987.

Since the plant shut down, some systems were secured. Additional systems were shut down following determination of lay-up methodology. Others awaited changes to plant Technical Specifications before operational status could be evaluated. Section 5.2 discusses the plant systems and their current status.

In addition to preparation of this Decommissioning Plan, proposed revisions to Technical Specifications, the Security Plan, the Emergency Plan, and the Quality Assurance Program Description were completed. An addendum to the Environmental Report and a preliminary DECON plan were also submitted.

7.2 SAFSTOR MODIFICATIONS The LACBWR staff reviewed the facility to determine if any modifications should be imple-mented to enhance safety or improve monitoring during the SAFSTOR period while fuel is stored onsite. Some modifications were evaluated as being beneficial and therefore have been performed.

The majority involve the Fuel Element Storage Well System (FESW). A redundant FESW level indicator has been added. A second remote manually operated FESW makeup line has been installed, which supplies water from the Overhead Storage Tank. Also, a local direct means of measuring FESW water level has been installed.

The air activity monitoring system has been replaced with new equipment. The gas activity monitors have been recalibrated to a Kr-85 equivalent. Kr-85 will be the predominant gaseous isotope during the SAFSTOR period.

D-PLAN 7-l November 2004

7. DECOMMISSIONING ACTIVITIES - (cont'd) 7.3 ACTIVITIES DURING SAFSTOR PERIOD 7.3.1 Significant SAFSTOR Licensing Actions The licensee's authority to operate Facility License No. DPR-45, pursuant to 10 CFR Part 50, was terminated by license Amendment No. 56, dated August 4, 1987, and a possession-only status was granted. The decommissioning alternative of SAFSTOR was chosen.

The NRC directed the licensee to decommission the facility in its Decommissioning Order of August 7, 1991. The Decommissioning Order was modified September 15, 1994, by Confirmatory Order to allow the licensee 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 containing similar requirements as specified in 10 CFR 50.59 was applied.

License Amendment No. 66, issued with the Decommissioning Order and also dated August 7, 1991, provided evaluation and approval of the proposed Decommissioning Plan, proposed SAFSTOR Technical Specifications, and license renewal to accommodate the proposed SAFSTOR period until March 29, 2031.

The Initial Site Characterization Survey for SAFSTOR was completed and published October 1995 and is attached to this Decommissioning Plan.

License Amendment No. 69, containing the current Technical Specifications, was issued April 11, 1997. This amendment revised the body of the license and the Appendix A Technical Specifications. The changes to the license and Technical Specifications 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 SAFSTOR Decommissioning Plan is considered the post-shutdown decommissioning activities report (PSDAR). The PSDAR public meeting was held on May 13, 1998.

Review of and revisions to this Decommissioning Plan, the Security Plan, the Emergency Plan, the Quality Assurance Program Description, the Offsite Dose Calculation Manual, and other material, continue at intervals as required.

7.3.2 Area and System Decontamination. The decontamination program during the SAFSTOR l period will be a continuation of routine decontamination work performed at LACBWR. Plant areas and component outer surfaces will be decontaminated to reduce the requirements for protective equipment use and to reduce the potential for the translocation of radioactive material.

Decontamination methods that are used are dependent upon a number of variables, such as surface texture, material type, contamination levels, and the tenacity with which the radioactive material clings to the contaminated surfaces.

D-PLAN 7-2 November 2004

7. DECOMMISSIONING ACTIVrTIES - (cont'd)

Surface areas are primarily decontaminated using hand wiping, wet mopping, and wet vacuuming techniques. Detergents and other mild chemicals may be used with any of these techniques. The residual water cleaning solutions are collected by floor drains and processed through the liquid waste system. Most areas are routinely decontaminated to levels below 2000 dpm/ft2 (about 500 dpm/lO0 cm2). Many areas are maintained below the Lower Limit of Detection (LLD). Efforts will be made to maintain all accessible areas in the plant as free of surface contamination as is reasonably achievable.

Small tools and components will be periodically decontaminated by wiping with cleaning agents, dishwasher, ultrasonic cleaning, or other methods. Some unused equipment may be decontami-nated as a prior step to removal for disposal as commercial or radioactive solid waste. Some unused equipment may be decontaminated prior to continued use in unrestricted areas.

Larger systems and components in accessible areas may be decontaminated using hydrolazers, abrasives, chemicals or other methods, after appropriate ALARA and economic evaluations are conducted.

7.3.3 Removal of Unused Equipment During SAFSTOR During the SAFSTOR period, some equipment and plant components will no longer be considered useful or necessary to maintain the plant in the SAFSTOR condition. Some equipment located in unrestricted areas may be transferred directly for use at another location or disposed of as commercial solid waste.

Some unused equipment or components located within restricted areas, which have not previously been used for applications involving radioactive materials will be thoroughly surveyed and documented as having no detectable radioactive material (less than LLD) prior to transfer to another user or disposal as commercial solid waste.

Other unused equipment or plant system components which have previously been used for applications involving radioactive materials may be removed, thoroughly surveyed and transferred to another licensed user, or disposed of as low level solid radioactive waste material.

Some equipment may be decontaminated and will be surveyed to verify that it contains no detectable radioactive material (less than LLD), prior to transfer to an unlicensed user, or for disposal as commercial solid waste.

Removal of plant equipment will be performed only after review. A 10 CFR 50.59 review will be conducted prior to dismantling any system.

Asbestos removed from plant systems will be handled in accordance with the Dairyland Power Cooperative asbestos control program.

D-PLAN 7-3 November 2004

7. DECOMMISSIONING ACTIVITIES - (cont'd) 7.3.4 Research During the SAFSTOR period, an Aging Research Program may be conducted. This program may entail records research and possible removal of unused components for testing.

7.3.5 Testing and Maintenance Program to Maintain Systems in Use During the SAFSTOR period, a testing and maintenance program will continue for those systems designated as being required for SAFSTOR. Routine preventive maintenance will be performed as before, but where the present maintenance interval is listed as "Outage," a new interval will be specified. Corrective maintenance will be performed as necessary. Instrument calibrations and other routine testing will continue as before for equipment which will be required to be operable.

LACBWR has established a program implementing the maintenance rule. This program contains key aspects of the maintenance rule. Included are those aspects specifically necessary to adequately identify structures, systems, and components (SSCs) to be monitored under the rule, establish goals and implement monitoring for those SSCs.

7.4 PLANT MONITORING PROGRAM Activities and plant conditions at LACBWR will continue to be maintained to protect the health and safety of both the public and plant workers. Baseline radiation surveys have been performed to establish the initial radiological conditions at LACBWR during SAFSTOR. An in-plant, as well as surrounding area, surveillance program will be established and maintained to assure plant l conditions are not deteriorating and environmental effects of the site are negligible.

7.4.1 Baseline Radiation Surveys Baseline surveys have been performed to establish activity levels and nuclide concentrations throughout the plant and surrounding area. These surveys included:

a)

Specific area dose rates and contamination levels.

b)

Specified system piping and component contact dose rate.

c)

Radionuclide inventory in specified plant systems.

d)

Radionuclide concentration in the soil and sediment in close proximity of the plant.

Baseline conditions will be compared with routine monitoring values to determine the plant/

system trends during SAFSTOR. Some specific monitoring points may be reassigned during the SAFSTOR period if it is determined that a better characterization can be obtained based on radiation levels measured or due to decontamination or other activities which are conducted and experience achieved.

D-PLAN 7-4 November 2004

7. DECOMMISSIONING ACTIVITIES - (cont'd) 7.4.2 In-Plant Monitoring Routine radiation dose rate and contamination surveys will be taken of plant areas along with more specific surveys needed to support activities at the site. A pre-established location contact dose rate survey will be routinely performed to assist in plant radionuclide trending. These points are located throughout the plant on systems that contained radioactive liquid/gases during plant operation.

7.4.3 Release Point/Effluent Monitoring During the SAFSTOR period, effluent release points for radionuclides will be monitored during all periods of potential discharge, as in the past. The two potential discharge points are the stack and the liquid waste line.

a)

Stack - the effluents of the stack will be continuously monitored for particulate and gaseous activity. The noble gas detector(s) have been recalibrated to an equivalent Kr-85 energy. The stack monitor will be capable of detecting the maximum Kr-85 concentra-tion postulated from any accident during the SAFSTOR period. Filters for this monitor will be changed and analyzed for radionuclides on a routine basis established in the ODCM.

b) Liquid discharge - the liquid effluents will be monitored during the time of release. Each batch release will be gamma analyzed before discharge to ensure ODCM requirements will not be exceeded.

All data collected concerning effluent releases will be maintained and will be included in the annual effluent report.

7.4.4 Environmental Monitoring Surrounding area dose rates as well as fish, air, liquid, and earth samples will continue to be taken and analyzed to ensure the plant is not adversely affecting the surrounding environment during SAFSTOR. The necessary samples and sample frequencies will be specified in the ODCM.

All data collected will be submitted in the annual environmental report.

D-PLAN 7-5 November 2004

9. SAFSTOR ACCIDENT ANALYSIS - (cont'd)

The assumptions used in evaluating this event during SAFSTOR were similar to those used in the FESW reracking analyses.12 The fuel inventory calculated for October 1987 was used. The only significant gaseous fission product available for release is Kr-85. The plenum or gap Kr-85 represents about 15% (215.7 Curies) of the total Kr-85 in the fuel assembly. However, for conservatism and commensurate with Reference 1, 30% of the total Kr-85 activity, or 431.4 Curies, is assumed to be released in this accident scenario. (Due to decay, as of October 2004 only 33.4% of the Kr-85 activity remains - 143.9 Curies.)

No credit was taken for decontamination in the FESW water or for containment integrity, so all the activity was assumed to be released into the environment. Meteorologically stable conditions at the Exclusion Area Boundary (1109 ft. 338m) were assumed, with a release duration of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> commensurate with 10 CFR 100 and Regulatory Guides 1.24 and 1.25.

A stack release would be the most probable, but a ground release is not impossible given certain conditions. Therefore, offsite doses were calculated for 3 cases. The first is at the worst receptor location for an elevated release, which is 500m E of the Reactor Building. The next case is the dose due to a ground level release at the Exclusion Area Boundary. The maximum dose at the Emergency Planning Zone boundary3 for a ground level release is also calculated. Adverse meteorology is assumed for all cases.

Elevated Release Average Kr-85 Release Rate 431.4 Curies

= 6.00 E-2 Ci/sec 2 hrs. x 3600 sec/hr X

Worst Case Q for 0-2 hours at 500m E = 2.3 E-4 sec/M3 Kr-85 average concentration at 500m E 6.00 E-2 Ci/sec x 2.3 E-4 sec/m3 = 1.38 E-5 Ci/m3 Immersion Dose Conversion at 500m E Kr-85 Gamma Whole Body Dose Factor (Regulatory Guide 1.109) 1.61 E+I mRem/vr x 106 uCi x 1.142 E-4 yr = 1,839 mRem/hr pCi/m3 Ci hr Ci/m3 Whole Body Dose at 500w E 1839 mRem/hr x 1.38 E-5 Ci/m3 x 2 hr = 0.05 mRem (as of 10/04 =0.02 mRem) l Ci/M3 D-PLAN 9-2 November 2004

9. SAFSTOR ACCIDENT ANALYSIS - (cont'd)

Kr-85 Beta/Gamma Skin Dose Factor (Regulatory Guide 1.109) 1.34 E + 3 mRem/yr x x.142E

=1.53E5 mRem/hr pCi/m3 Ci hr Ci/m3 Skin Dose at 500m E 1.53 E5 x 1.38E-5 Ci/m3 x 2hr = 4.2 mRem (as of 10/04 = 1.4 mRem)

C m3 Ground Level Release at EAB Worst Case X for 2 hrs at 338m NE or 338m SSE using Regulatory Guide 1.25 Q

2.2 E-3 sec m3 Whole Body Dose at 338m Skin Dose at 339m 10/87 = 0.49 mRem 10/04 = 0.16 mRem 10/87 = 40.4 mRem 10/04 = 13.5 mRem I

Ground Level Release at Emergency Planning Zone Boundar_

Worst Case for 2 hrs at 100m E Q

1.02 E-2 sec m3 Whole Body Dose at 100m E Skin Dose at 100m E 10/87 = 2.25 mRem 10/04 = 0.75 mRem 10/87 = 187 mRem 10/04 = 62.4 mRem I

As can be seen, the estimated maximum whole body dose is more than a factor of 30,000 below the 10 CFR 100 dose limit of 25 Rem (25,000 mRem) to the whole body within a 2-hour period.

D-PLAN 9-3 November 2004

9. SAFSTOR ACCIDENT ANALYSIS - (cont'd) 9.3 SHIPPING CASK OR HEAVY LOAD DROP INTO FESW This accident postulates a shipping cask or other heavy load falling into the Fuel Element Storage Well. Reference I stated that extensive local rack deformation and fuel damage would occur during a cask drop accident, but with an additional plate (installed during the reracking) in place, a dropped cask would not damage the pool liner or floor sufficiently to adversely affect the leak-tight integrity of the storage well (i.e., would not cause excessive water leakage from the FESW).

For this accident, it is postulated that all 333 spent fuel assemblies located in the FESW are damaged. The cladding of all the fuel pins ruptures. The same assumptions used in the Spent Fuel Handling Accident (Section 9.2) are used here. A total of 35,760 Curies of Kr-85 is released within the 2-hour period. The doses calculated are as follows. (Due to decay, as of Oct. 2004 only 33.4% of the Kr-85 activity remains - 11,928 Curies.)

Elevated Release I

Whole Body Dose at 500m E Skin Dose at 500m E 10/87 = 4.2 mRem 10/04= 1.4 mRem 10/87 = 350 mRem 10/04 = 116.7 mRem I

8-Ground Level Release at EAB Whole Body Dose at 338m Skin Dose at 338m 10/87 = 40.2 mRem 10/04 = 13.4 mRem 10/87 = 3.34 Rem 10/04 = 1.11 Rem I

Ground Level Release at Emereencv Planning Zone Boundary Whole Body Dose at 100m E Skin Dose at 100m E 10/87 = 186 mRem 10/04 = 62.0 mRem 10/87 = 15.6 Rem 10/04 = 5.2 Rem I

As can be seen, the estimated maximum whole body dose is more than a factor of 400 below the 10 CFR 100 dose limit of 25 Rem (25,000 mRem) to the whole body within a 2-hour period.

D-PLAN 9-4 November 2004

9. SAFSTOR ACCIDENT ANALYSIS - (cont'd)

The operator would be alerted to this accident by receipt of the FESW Level Lo/High alarm.

Any makeup water added may run out the break, depending on the size of the break.

A calculation has been performed to determine the radiation levels due to the exposed control rods. In the vicinity of most of the FESW piping and isolation valves, the radiation dose would not be substantially increased due to the loss of water.

A repair team should be able to access the break location or piping isolation valves and either isolate the break or effect temporary repairs. FESW level could then be restored to normal.

There would be no immediate urgency to restore the level. The partially uncovered control rods only create a local problem. No release of contamination is associated with this event. Active FESW cooling would be lost during this accident, but as discussed in Section 9.4, considerable time is available to take action. Due to the lesser water volume to act as the heat sink and reduced fuel coverage, less time would be available to restore cooling during this accident scenario than in just a loss of FESW cooling event, but boiling would not commence for more than one (1) day. As with the loss of FESW cooling event, if water is added to the FESW, any consequences of water heatup can be delayed or prevented. Water can be added from the Demineralized Water System or the Overhead Storage Tank.

9.6 UNCONTROLLED WASTE WATER DISCHARGE This accident postulates that an operator starts pumping a Waste Water or Retention Tank to the river which is not sampled or for which the sample was incorrectly analyzed. If the contents of the tank are of normal activity, this event will not be detected until the lineup is being secured after pumping, if then.

If the liquid in the tank is of high activity, the liquid waste monitor will alarm and the Auto Flow Control Valve (54-22-002) automatically will close, terminating the discharge. If the automatic valve does not close, an operator will try to close it from the Control Room. If it cannot be closed, an operator will close a local valve or secure the pump to terminate the discharge.

After the discharge is terminated, a sample of the tank will be taken to analyze the uncontrolled release. Waste water is diluted by LACBWR Circulating Water and Low Pressure Service Water flow, in addition to circulating water from the adjacent coal-fired plant, prior to being discharged into the river.

9.7 LOSS OF OFFSITE POWER This accident postulates a loss of offsite power. If both Emergency Diesel Generators and a High Pressure Service Water (HPSW) Diesel start, FESW cooling can be provided and adequate instrumentation is available to monitor FESW conditions fiom the Control Room. All that is needed is for an operator to cross-connect HPSW to the Component Cooling Water (CCW) coolers.

D-PLAN 9-6 November 2004

9. SAFSTOR ACCIDENT ANALYSIS - (cont'd)

If an HPSW Diesel and IB Emergency Diesel Generator start, FESW cooling can be provided.

If IA Emergency Diesel Generator (EDG) starts, but IB does not, adequate cooling can be provided only if the essential buses are tied together.

If one or more EDGs start, but neither HPSW diesel starts, no ultimate heat sink for the FESW would be available. The consequences would be the same as in the Loss of FESW Cooling Event (Section 9.4).

If neither EDG can be started, neither FESW or CCW pump can run. The consequences again are the same as a Loss of FESW Cooling Event. Some instrumentation will be lost immediately and the rest will be lost if packaged uninterruptible power supplies (UPS) are depleted. The operator would have to check the FESW locally periodically.

As discussed in Section 9.4, the fuel pool heatup test conducted in 1993 indicated that the temperature of the pool water would stabilize at less than boiling. Therefore, no immediate action needs to be taken and sufficient time is available to take corrective actions to restore power.

9.8 SEISMIC EVENT This accident postulates that a design basis earthquake occurs. The magnitude of the seismic event and damage incurred is the same as that assumed during the Systematic Evaluation Program (SEP) and the Consequence Study prepared as part of the SEP Integrated Assessment (References 4-7). The major concern of the previous evaluation was to safely shut down the plant and maintain adequate core cooling to prevent fuel damage. The focus now is to prevent damage to the fuel stored in the Fuel Element Storage Well.

Seismic analysis has shown the Reactor Building structure, LACBWR stack and Genoa Unit 3 stack are capable of withstanding the worst postulated seismic event at the LACBWR site.

Reference I documented that the storage well, itself, the racks and the bottom-entry line between the check valves and the storage well can withstand the postulated loads.

The potential consequences of most interest due to a seismic event could include loss of all offsite and onsite power and a break in the FESW System piping. This event, therefore, can be considered as a combination of a Loss of Power Event (Section 9.7) and FESW Line Break (Section 9.5). As with these individual events, considerable time is available for response to a seismic event, with the FESW System pipe break requiring the earlier response. Access to the break location may be more difficult following a seismic event due to failure of other equipment in the plant. The time available, though, should be more than sufficient to initiate mitigating actions. (Refer to Section 9.5).

D-PLAN 9-7 November 2004

9. SAFSTOR ACCIDENT ANALYSIS - (cont'd) 9.9 WIND AND TORNADO This accident postulates that design basis high wind or tornado event occurs. The magnitude of the event and damage incurred is the same as that assumed during the Systematic Evaluation Program (SEP) and the Consequence Study prepared as part of the SEP Integrated Assessment (References 4-9). The major concern of the previous analyses was to ensure that adequate cooling of the reactor core was maintained. The focus now is to prevent damage to the fuel stored in the Fuel Element Storage Well.

The previous evaluations determined that the Reactor Building would withstand this event. The Turbine Building, Diesel Building, Cribhouse and Switchyard may be damaged. The probability of the LACBWR or Genoa Unit 3 stacks failing and impacting the Reactor Building was determined to be low enough that it need not be considered. Personnel outside the Reactor Building may not survive.

The potential plant consequence of primary concern is the loss of all offsite and onsite power.

As discussed in Section 9.7, Loss of Offsite Power, considerable time is available before action must be taken to protect the fuel.

9.10 REFERENCES

1)

NRC Letter, Ziemann to Linder, dated February 4, 1980.

2)

NRC Letter, Reid to Madgett, dated October 22, 1975.

3)

DPC Letter, Taylor to Document Control Desk, LAC-12377, dated September 29, 1987.\\

4)

DPC Letter, Linder to Paulson, LAC-10251, dated October 11, 1984.

5)

NRC Letter, Zwolinski to Linder, dated January 16, 1985.

6)

DPC Letter, Linder to Zwolinski, LAC-10639, dated March 15, 1985.

7)

NRC Letter, Zwolinski to Taylor, dated September 9, 1986.

8)

DPC Letter, Taylor to Zwolinski, LAC-12052, dated January 14, 1987.

9)

NRC Letter, Bernero to Taylor, dated April 6, 1987.

D-PLAN 9-8 November 2004

LAC-TR-138 LACBWR INITIAL SITE CHARACTERIZATION SURVEY FOR SAFSTOR By:

Larry Nelson Health and Safety Supervisor a-October 1995 Revised: November 2004 Dairyland Power Cooperative 3200 East Avenue South La Crosse, WI 54601

LAC-R-138 PAGE 5 LACBWR AFFECTED AREA MAP

LAC-TR-138 PAGE 24 ATTACHMENT 1 SPENT FUEL RADIOACTIVITY INVENTORY Decay-Corrected to October 2004 I

Half Life Activity Half Life Radionuclide (Years)

(Curies)

Radionuclide (Years)

(Curies)

Ce-144 Cs-137 Ru-106 Cs-134 Kr-85 Ag-110m Co-60 Pm-147 Ni-63 Am-241 Pu-238 Pu-239 Pu-240 Eu-154 Cm-244 H-3 Eu-152 Am-242m 7.801 E-l 3.014 E+1 1.008 E+0 2.070 E+0 1.072 E+1 6.990 E-1 5.270 E+0 2.620 E+0 1.000 E+2 4.329 E+2 8.774 E+1 2.410 E+4 6.550 E+3 8.750 E+0 1.812 E+1 1.226 E+1 1.360 E+I 1.505 E+2 0.909 1.133E6 15.2 1.208E3 3.928E4 0.006 7.067E3 492 3.152E4 1.435E4 1.106E4 8.833E3 7.152E3 1.067E3 1.899E3 214 218 454 Sr-90 Pu-241 Fe-55 Ni-59 Tc-99 Sb-125 Eu-155 U-234 Am-243 Cd-1 13m Nb-94 Cs-135 U-238 Pu-242 U-236 Sn-121m Np-237 U-235 Sm-151 Sn-126 Se-79 1-129 Zr-93 2.770E+ 1 1.440 E+1 2.700 E+0 8.000 E+4 2.120 E+5 2.760 E+0 4.960 E+0

• 2.440 E+5 7.380 E+3 1.359 E+1 2.000 E+4 3.000 E+6 4.470 E+9 3.760 E+5 2.340 E+7 7.600 E+1 2.140 E+6 7.040 E+8 9.316 E+1 1.000 E+5 6.500 E+4 1.570 E+7 1.500 E+6 7.543E5 5.082E5 7.135E3 287 276 4.07 16.18 63.7 61 7.577 15.89 14.0 12.2 8.58 6.32 3.811 2.19 1.89 1.333 0.7 0.552 0.39 0.111 Total Activity = 2.53 E6 Curies I

C

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LAC-TR-138(

PAGE 25 ATTACHMENT 2 CORE INTERNAUJRX COMPONENT RADIONUCLIDE INVENTORY - OCTOBER 2004 I

Estimated Curie Content Y

i Components Co-60 Fe-55 Ni-63 Other Nuclides T, 2 > Sy Total L

J.

.4

-

In Reactor Fuel Shrouds (72 Zr, 8 SS)

Control Rods (29)

Core Vertical Posts (52)

Core Lateral Support Structure Steam Separators (16)

Thermal Shield Pressure Vessel Core Support Structure Horizontal Grid Bars (7)

Incore Monitor Guide Tubes Total In FESW Fuel Shrouds (24 SS)

Fuel Shrouds (73 Zr)

Control Rods (10)

Start-up Sources (2)

Total 2,418 534 139 996 3,657 158 38 706 19 34 8,699 1,495 100 378 347 2,320 841 64 8

286 1,049 45 14 203 5

3 2,518 199 13 32 30 2'74 1,203 727 56 685 2,515 109 9

486 13 544 6,347 2,122 85 810 139 3,156 8

8 2

4 15

' 0.5 3

3 43.5 13 2

9 2

26 4,470 1,333 205 1971 7,236 312.5 61 1398 37 584 17,607.5 3,829 200 1,229 518 5,776

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LAC-TR-138(

PAGE 26 ATTACHMENT 3 PLANT SYSTEMS INTERNAL RADIONUCLIDE INVENTORY - OCTOBER 2004 I

Nuclidce Activity, in uCi Plant System System Total uCi Content Fe-55 Alpha Co-60 Cs-137 1

t I

4 CB Ventilation Offgas -

upstream of filters Offgas -

downstream of filters TB drains CB drains TB Waste Water CB Waste Water Main Steam Turbine Primary Purification Emergency Core Spray Overhead Storage Tank Seal Inject 21 SYSTEMIREMOVED SYSTEM 226 506 48 2,794 3,459 12 1,184 SYSTEM 173 21 REMOVED 40 3

7 79 290 2

12 REMOVED 34 4

175 1,859 4,156 394 22,965 28,433 102 9,733 1,422 175 115 3,394 1,629 81 1,561 136 529 37 311 5,519 6,294 530 27,399 32,182 252 10,929 2,158 237 lI

C C

LAC-TR-138 (

PAGE 27 ATTACHMENT 3 PLANT SYSTEMS INTERNAL RADIONUCLIDE INVENTORY - OCTOBER 2004 - (cont'd)

I

.f Nuclide Activity. in uLC1 Plant System r

System Total uCi Content Fe-55 Aloha Co-60 Mn-54

+

a A

Decay Heat Boron Inject Reactor Coolant PASS Alternate Core Spray Shutdown Condenser Control Rod Drive Effluent Forced Circulation Reactor Vessel and Internals Condensate after beds & Feedwater Condensate to beds l

1,330 SYSTEM SYSTEM 266 SYSTEM 1,996 19,956 33,260 SYSTEM SYSTEM 490 REMOVED REMO VED 94 REMOVED 720 7,000 12,000 REMOVED REMOVED I

10,936 2,187 16,404 164,037 273,395 l

12,756 2,547 19,120 190,993 318,656 0.5 0.9 A

a L

II

C C

LAC-TR-138 C PAGE 28 ATTACHMENT 3 PLANT SYSTEMS INTERNAL RADIONUCLIDE INVENTORY - OCTOBER 2004

- (cont'd)

Nuclide Activity, in

_Ci_l System Total Plant System Fe-55 Alpha Co-60 Mn-54 Cs-137

• CiContent Fuel Element Storage Well System 11,308 390 92,954 104,652 Fuel Element Storage Well

- all but floor 17 5

142 3,122 3,286 Fuel Element Storage Well floor 345,901 7,600 2,843,306 0.6 27,829 3,224,637 Resin lines 1,730 100 14,217 16,047 Main Condenser 146,343 8,500 1,202,937 4.3 1,357,784