ML110190592

Decommissioning Plan Revision, November 2010

ML110190592
Person / Time
Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	12/28/2010
From:	Berg W Dairyland Power Cooperative
To:	Document Control Desk, NRC/FSME
References
LAC-14154
Download: ML110190592 (190)

Text

WILLIAM L. BERG, President and CEO DAIRYLAND POWER COO PE RATIVE December 28, 2010 In reply, please refer to LAC-14154 DOCKET NO. 50-409 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 Decommissioning Plan Revision - November 2010

REFERENCES:

(1)

DPC Letter (LAC-12460), Taylor to Document Control Desk, dated December 21, 1987, submitting original LACBWR 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, issuing Confirmatory Order modifying Decommissioning Order A revision of the LACBWR Decommissioning Plan (D-Plan) has been completed. The D-Plan functions as a Final Safety Analysis Report (FSAR) at LACBWR and submittal of revisions to this FSAR fulfills requirements found in 10 CFR 50.71(e)(4). The main objective of this comprehensive revision was to accommodate construction and implementation of an Independent Spent Fuel Storage Installation at the LACBWR site. As determined in accordance with 10 CFR 50.59 requirements, these changes to the LACBWR D-Plan do not require prior NRC approval. All changes have been reviewed by the plant Operations Review Committee and the independent Safety Review Committee; both groups have given approval of these changes.

The D-Plan is also considered a Post-Shutdown Decommissioning Activities Report (PSDAR).

NRC notification under 10 CFR 50.82(a)(7) is required if changes are inconsistent with or make significant schedule changes to those described in the PSDAR. Schedule changes in this revision of the D-Plan required notification to the NRC. This notification of a change to the LACBWR decommissioning schedule was made to the State of Wisconsin and the NRC by letter dated December 7, 2010.

A Touchstone Energy Cooperative fZ M

3200 East Ave. S.

• PO Box 81]7 La Crosse, WI 54602-0817

• 608-787-1258 ° 608-787-1469 fax

• www.dairynet.com

Document Control Desk Page 2 December 28, 2010 All pages of the LACBWR D-Plan have been updated with changes in formatting and numbering, and are enclosed with this letter. The location of specific changes in content are designated with a change bar in the right-hand margin. Justifications for the changes are provided in a separate enclosure. Please replace all pages of the LACBWR D-Plan with the pages enclosed showing a revision date of November 2010. Also enclosed for replacement is an update to the Initial Site Characterization Survey for SAFSTOR (TR-138).

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

Sincerely, DAIRYLAND POWER COOPERATIVE William L. Berg, President and CEO WLB:JBM:two Enclosures cc:

Kristina Banovac Project Manager U.S. Nuclear Regulatory Commission

NRC Docket No. 50-409 "TRANSMITTAL/ NOTIFICATION/ ACKNOWLEDGMENT" TO:

NRC Washington - Doc Control CONTROLLED DISTRIBUTION NO. 53 (TWO COPIES)

FROM:

LACBWR Plant Manager 12/22/2010

SUBJECT:

Changes to LACBWR Controlling Documents I.

The following documents have been revised or issued new.

DECOMMISSIONING PLAN, revised November, 2010 Remove and replace all pages SITE CHARACTERIZATION SURVEY Remove and replace all pages LxI I have received and properly filed the material(s) listed above. I have destroyed superseded material, if necessary.

CK I have placed the material(s) listed above in the appropriate "controlled" procedures binder.

E[

I have reviewed the material(s) listed above, and if necessary I have notified my reporting personnel of the changes noted above. The signatures on the back of this form serve as acknowledgment of understanding and Read and Heed Training.

FK I have updated the index or indices with pen and ink changes, if needed.

II.

The following procedure(s) has been CANCELLED. Please destroy all copies and update the index or indices with pen and ink changes.

/S/

DATE Please return this notification to the LACBWR Administrative Assistant within ten (10) working days.

ACP-06.4 Issue 5 ATTACHMENT I 10 CFR 50.59 SCREEN FORM Document No.

Not Applicable 50.59 Screen Rev. No.

0 Activity Title LACBWR Decommissioning Plan Activity Description 2010 annual review and update of the LACBWR Decommissioning Plan.

Document Listing

1. LACBWR Decommissioning Plan, December 2009.

2. LACBWR Possession-Only License, Docket No. 50-409, Amendment No. 69, date of issuance April 11, 1997.

3. LACBWR Possession-Only License, Appendix A, Technical Specifications, Amendment No. 70, date of issuance April 3, 2006.

4. NRC to DPC, Confirmatory Order Modifying NRC Order Authorizing Decommissioning of Facility, dated September 15, 1994.

Design Functions The Decommissioning Plan functions as a Final Safety Analysis Report (FSAR) at LACBWR and submittal of revisions to this FSAR fulfill requirements found in 10 CFR 50.71 (e)(4). The Decommissioning Plan (D-Plan) is also considered a Post-Shutdown Decommissioning Activities Report (PSDAR).

Page 1 of 2

ACP-06.4 Issue 5 ATTACHMENT I 10 CFR 50.59 SCREEN FORM 10 CFR 50.59 Screening Questions Yes No

1. Does the proposed activity involve a change to a structure, system, or component (SSC) that adversely affects a design function described in the D

Decommissioning Plan?

2. Does the proposed activity involve a change to a procedure that adversely affects how SSC design functions, described in the Decommissioning Plan, E]L are performed or controlled?

3. Does the proposed activity revise or replace evaluation methodology described in the Decommissioning Plan that is used in the safety analyses, or 0

which establishes the design bases?

4. Does the proposed activity involve a test or experiment not described in the Decommissioning Plan, where a SSC is utilized or controlled in a manner that El H

is outside the reference bounds of the design for that SSC or is inconsistent with analyses or descriptions in the Decommissioning Plan?

5. Does the proposed activity require a change to LACBWR Possession-Only Ei License, Appendix A, Technical Specifications?

][

6.

Will the proposed change result in a significant environmental impact not previously evaluated in NUREG-0586, Supp. 1, "Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities," dated

][

November 2002?

Conclusion Any of questions 1, 2, 3, or 4 are answered YES; questions 5 and 6 are answered NO. 50.59 Evaluation shall be performed. This Screen form does not need to be retained.

Questions 5 or 6 are answered YES. NRC approval is required prior to implementation of the activity; proceed to license amendment process. This Screen form does not need to be retained.

All screening questions have been answered NO. 50.59 Evaluation or NRC approval is not required. Implement the activity per the applicable procedure for the type of activity. Attach this Screen form, as approved, to documentation for the activity. Provide justification that a 50.59 Evaluation is not required in the space below.

Justification Changes arise from annual review and update of the D-Plan. Changes provide updated dose levels and activity concentrations, decay-corrected to current values. Changes also provide description of facility modifications being performed for ISFSI implementation. NRC notification under 10 CFR 50.82(a)(7) is required if changes are inconsistent with or make significant schedule changes to those described in the PSDAR. Such notification has been made to the NRC and State of Wisconsin of the changes in schedule described in this revision to the D-Plan.

These D-Plan changes are administrative and have no adverse effect on any design bases nor create any significant environmental impact not previously evaluated.

Signatures 50.59 Screen Preparer (print name):

(Signature)

Date:

ORC Approval and Meeting No.

(Chairman Signature)

Date:

/

',- 3 Page 2 of 2

2010 LACBWR Decommissionin2 Plan Review NOTE:

Changes described following are as they appear in 2010 Decommissioning Plan pages as revised Cover Page Update Decommissioning Plan (D-Plan) revision date.

Page 0-2 Table of Contents: Listing is revised by addition of items related to Independent Through Spent Fuel Storage (ISFSI) construction and implementation. The changes are as Page 0-6 follows with additions in bold:

Section:

3.6 ISFSI Soils and Seismology 3.7 References 4.2.6 Onsite Independent Spent Fuel Storage Installation 7.8 Dry Cask Storage Project 7.8.1 Cask Handling Crane 8.4.6 ISFSI Monitoring Figures:

4.8 Onsite ISFSI Entire All Sections have bullets and numbering reformatted for consistency.

D-Plan Page 1-1 Section 1,

Introduction:

In the second and third sentences of the first paragraph, the verbs are corrected to past tense. The first paragraph is amended at the end by addition of four sentences explaining that the preliminary DECON plan, completed in 1987, will not be revised and that the D-Plan addresses issues contained in the preliminary DECON plan. Final dismantlement activities at LACBWR will be performed in accordance with the License Termination Plan (LTP), as required by 10 CFR 50.82, following approval by the NRC of the LTP and final supplement to the Environmental Report in support of the LTP. Second paragraph of the section is revised by deleting the first sentence that states, "This Decommissioning Plan concentrates on the status of LA CB WR while the reactor fuel remains in the Fuel Element Storage Well." The purpose of this and most 2010 changes to the D-Plan is to accommodate ISFSI construction and implementation at the LACBWR site. In the second sentence of the second paragraph, the phrase, "333 activated fuel assemblies onsite," is corrected to current terminology by, "333 spent fuel assemblies onsite. " This change is made to be consistent with the definition of "spent fuel" found in 10 CFR 72.3. The third sentence of the second paragraph is revised to include dry storage as an option for the spent fuel at LACBWR by stating:

"The plan at this time is to continue to store the fuel in the existing Fuel Element Storage Well while changes are made to place the spent fuel assemblies in dry cask storage containers and store the containers at the onsite Independent Spent Fuel Storage Installation (ISFSI). "

Page I of 25

2010 LACBWR Decommissionin2 Plan Review In the final sentence of the second paragraph of the section, the term offsite is added to describe a second interim storage facility if available as an option for spent fuel storage in addition to the onsite ISFSI being implemented.

Section 1.1, Selection of SAFSTOR: The first paragraph of the section is rewritten to update regulatory action on the 1996 decommissioning rule. A second paragraph is added to describe the three methods of decommissioning now established in the current guidance found in NUREG-0586.

Page 1-2 Section 1.1, Selection of SAFSTOR: Verbs in all of Section 1.1 are changed to past tense. Following the summary of ENTOMB, in the last sentence of the first paragraph, original is added to describe the decommissioning rule when proposed. In the first sentence of the third paragraph, the unavailability of a federal repository "for about 20 years" is replaced by "the foreseeable future."

Page 1-3 Section 1.1, Selection of SAFSTOR: In the paragraph at the top of the page, the phrase, "available for the dismantling" is corrected to "available for dismantlement." The second paragraph is updated to current decommissioning experience as applicable in the U.S. by the following:

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

In the third paragraph discussion of the Waste Confidence Decision of 1984, is updated by adding "which is codified as amended in 10 CFR 51.23."

Section 1.2,

References:

An updated reference for information in the section is added as:

1.2.5 Regulatory Guide 1.184, "Decommissioning of Nuclear Power Reactors," July 2000.

Page 2-2 Section 2.2, Initial Construction and Licensing History: In the last paragraph of the section, the last sentence is changed to past tense.

Section 2.3, Operating Record: Changed verb in second paragraph to past tense.

Page 2-3 Section 2.4, Decision for Shutdown: In last paragraph of section, "333 irradiated fuel assemblies" is changed to "333 spent fuel assemblies" consistent with the definition in 10 CFR 72.3. The acronym (FESW) is identified.

Section 2.5.1, Failed Fuel: In the first paragraph, second sentence the acronym FESW replaces Fuel Element Storage Well.

Page 2-4 Section 2.5.1, Failed Fuel: In the last paragraph of the section the last sentence Page 2 of 25

2010 LACBWR Decommissioning Plan Review stating, "The majority of the material is located on horizontal surfaces in the Fuel Element Storage Well, " is deleted and replaced by, "The FESW will be emptied of all spent fuel assemblies and fuel debris which will be placed in dry cask storage containers and moved to storage at the ISFSI." This change states the purpose of ISFSI implementation.

Page 2-3 Section 2.5.2, Fuel Element Storage Well Leakage: The acronym FESW replaces And Fuel Element Storage Well in the first sentence. A new sentence is added to the Page 2-4 end of the section to state that, "Following removal of all spent fuel assemblies and fuel debris for dry storage at the ISFSI, the storage racks will be removed and disposed of, the FESW will be drained and decontaminated to eliminate leakage." Purpose of this change is to state DPC's plan to drain and dismantle FESW structures, systems, and components (SSCs).

Page 3-1 Section 3.1.1, Site Location and Description of Site Layout: After addition of the acronym (G-3) at the end of the second sentence in the first paragraph, a general description of the ISFSI at the Genoa site is given by addition of three sentences stating:

"The Independent Spent Fuel Storage Installation (ISFSI) is located south of the G-3 plant, on land which was previously used for an access road to an abandoned boat landing on the Mississippi River. The ISFSI site is about midway between the Mississippi River on the west and Highway 35 on the east and is between the two closed ash landfills of the Genoa site. Access to the new boat landing follows the property's east and south boundaries. "

Page 3-1 Section 3.1.2, The Authority of the Exclusion Area and Licensee Authorities: A And second paragraph is added to the section to provide information that the entire Page 3-2 Genoa site is included in the Part 50 license. A third paragraph is added that provides a description of the ISFSI location in relation to LACBWR, a description of the ISFSI boundaries, and a statement of the licensee's ISFSI access control authority. The new paragraphs state:

"By letter dated May 8, 2008, in response to request by DPC, the NRC agreed that the geographical area included within the LACBWR Part 50 license is the entire 163.5 acres owned or otherwise controlled by DPC. Further, the NRC found that the site where the ISFSI is being constructed is part of the NRC-licensed site under License DPR-45.

The ISFSI is located 2,232feet south-southwest of the Reactor Building center.

The ISFSI will be surrounded by a protected area fence, an isolation zone fence, and vehicle barrier system. These protective barriers will be within the ISFSI Controlled Area Boundary fence established at the perimeter of the 38.9 acre ISFSI site (See Figure 3.4). The ISFSI Controlled Area Boundary is established to limit dose to the public during normal operations and design basis accidents in accordance with the requirements of 10 CFR 72.104 and 10 Page 3 of 25

2010 LACBWR Decommissionin2 Plan Review CFR 72.106. The controlled area, as defined in 10 CFR 72.3, means the area immediately surrounding an ISFSIfor which the licensee exercises authority over its use and within which ISFSI operations are performed. DPC shall likewise exclude access to the ISFSI Controlled Area Boundary if adverse radiological conditions require."

Page 3-2 Section 3.2, Transportation, Industrial and Military Facilities within Proximity to the Plant: In first paragraph correction is made to the Burlington Northern Santa Fe Railway and the acronym (BNSF) is identified and later used to replace Burlington Northern.

Page 3-3 Section 3.3.3, Local Meteorology: In first sentence of second paragraph, clarification is added to the description of the data contained in Table 3-1 by adding 'from 1982-1984."

Page 3-6 Section 3.4.4, Flooding and Probable Maximum Flood: A description of ISFSI elevations in relation to flood elevation is added to the end of the section by stating, "The ISFSI area is constructed to a grade elevation of 640.8 feet MSL with a top of pad elevation of 643.5 feet to raise the pad above the standard project flood elevation of 643.2feet."

Page 3-6 Section 3.4.6, Flooding Protection Requirements: In the first paragraph, the And acronym (USGS) is added after United States Geological Survey. Discussion is Page 3-7 added to the end of the section explaining that the NAC-MPC storage system flood design basis bounds flood conditions of the ISFSI site as follows:

"For flood conditions at the ISFSI, the NAC Multi-Purpose Canister (NAC-MPC) storage system is evaluated for afully-immersing design basis flood with a water depth of SO feet and a steady-stateflow velocity of 15feet per second.

The analysis shows that the NAC-MPC storage system performance is not affected by the design basis flood, and demonstrates that the concrete cask will not slide and will not overturn in the design-basis flood. The hydrostatic pressure exerted by the 50foot depth of water does not produce significant stress in the canister. The NAC-MPC design basis bounds the LACBWR site probable maximum flood elevation of 658feet MSL with flow velocity of 1.91 feet per second."

Page 3-8 Section 3.4.7, Ultimate Heat Sink and Low Flow Conditions: A sentence is added to end of section to indicate that dry storage of spent fuel at the ISFSI does not require cooling water, stating, "With all spent fuel assemblies in dry storage at the ISFSI, spent fuel cooling is no longer dependent on river flow. "

Section 3.5.1, Basic Geologic and Seismic Information: The spelling of Paleozoic is corrected in the first paragraph.

Page 4 of 25

2010 LACBWR Decommissioning Plan Review Page 3-14 Section 3.6, ISFSI Soils and Seismologv: A new section is added that provides And description of the soil improvements and seismological design of the ISFSI site:

Page 3-15 "The soil parameters were analyzed from borings made in the area of the ISFSL A site dimension of 122 feet by 148feet, including the 32feet by 48feet ISFSI pad area, was improved by vibrocompaction to a depth of 35 feet. The 11 feet of in-situ soils above elevation 621feet were removed. Drain rock was placed and compacted in 8-inch lifts to an elevation of 625 feet at which point a layer of geotextile fabric was installed. Imported structural backfill was then placed in controlled compacted lifts to raise the improved area to a final elevation of 640.8feet. All soils work was completed per project earth work specifications.

The liquefaction analysis for the post-improvement site soils used the cone penetration testing (CPT) based method and considered a design groundwater elevation of 639feet and a final grade elevation of 640.8feet with and without ISFSI pad loading. The results of the liquefaction evaluation indicated that factors ofsafety for all sandy soil layers below elevation 621feet were acceptable. The soils improvement above elevation 621 feet resulted in all soil densities in the improved area to be considered non-liquefiable.

The soil structure interaction analysis of the ISFSI site soils improvement resulted in acceptable ISFSI pad accelerations of 0.402g horizontal and 0.18g vertical. For the center of gravity of the stored Vertical Concrete Cask (VCC) the resulting accelerations were 0.442g horizontal and 0.18g vertical The NA C-MPC Final Safety Analysis Report (FSAR) acceptance criteria for the VCC are 0.45g horizontal and 0.30g vertical accelerations."

Page 3-16 Section 3.7,

References:

Three updated references for information in the section are added as:

3.7.15 NA C International, Inc., NA C Multi-Purpose Canister Final Safety Analysis Report, Revision 7, Section 11.A.2.6.

3.7.16 Sargent & Lundy Report No. SL-010167, ISFSI Soil Remediation Summary.

3.7.17 NRC Letter, Banovac to Berg, dated May 8, 2008.

Figures Figure 3.4, Genoa Site Map is revised to show the ISFSI location and Controlled Area Boundary fence.

Page 4-1 Section 4.1, General Plant

Description:

A short paragraph is added to the end of the section to describe the ISFSI on the Genoa site by stating, "An Independent Spent Fuel Storage Installation for dry storage of the LACBWR spent fuel inventory is being established on the Genoa site south of the G-3 coal pile. "

Page 5 of 25

2010 LACBWR Decommissioning Plan Review Section 4.2.1, Reactor Building: In the second paragraph, the acronym (FESW) is added and then used later in replacing storage well cooling system with the FESW cooling system.

Page 4-3 Section 4.2.1, Reactor Building: In the second paragraph the acronym FESW replacesfuel element storage well. In fourth paragraph of page, the word wet is added to describe spent fuel storage in the FESW with the acronym used to replace forms of storage well in three instances. At the end of the paragraph, DPC's intent to dismantle FESW SSCs is stated by, "Following removal of all spent fuel assemblies and fuel debris for dry storage at the ISMSI, the storage racks will be removed and disposed of, the FESW will be drained and decontaminated."

Page 4-3 Section 4.2.1, Reactor Building: Dry cask loading operations required And modifications to be made to the Reactor Building to return watertight integrity to Page 4-4 the biological shield upper cavity and provide support for cask loading, cask preparation, and cask transfer operations. These modifications were designed to be Seismic Category I structures. The design of the modifications was reviewed under the LACBWR design review and acceptance process. The installations were reviewed under the LACBWR 10 CFR 50.59 process and performed in accordance with the LACBWR work control process and requirements of the LACBWR Quality Assurance Program Description (QAPD) and Dry Cask Storage Quality Assurance Project Plan (QAPP). New information describing these modifications is added to the end of the section as follows:

"The Reactor Building is being modified to facilitate movement of spentfuel assemblies from the FESW to the NA C-MPC System Transportable Storage Canister (TSC). The TSC is located inside the Transfer Cask (TFR) during fuel loading operations. The TFR/TSC assemblage will reside in the cask pool within the water-filled upper cavity during spent fuel assembly transfer from the FESW to the TSC.

The 10-6" wide opening from the mezzanine floor elevation 667feet to the fuel handling floor elevation 701 feet was previously created in the northern section of the upper cavity bio-shield and liner to permit removal of the LACBWR reactor pressure vessel. This opening will be used to facilitate movement of the TFR/TSC between the cask pool and the mezzanine floor to the north where TSC preparation operations will take place (e.g., welding, drying, etc.). The Reactor Building mezzanine floor will be reinforced in that location by adding steel struts beneath a cantilevered section of the floor. In order to provide sufficient water coverage over the spent fuel assemblies during movement into the TSCfrom the FESW, a water-tight removable gate, 16 '-9" high by 9 '-4" wide, will be installed in the bio-shield opening above approximate elevation 679'-3" extending to elevation 696feet. The cask pool gate will be supported by a 12' high structure installed at elevation 667feet. The caskpool gate is Page 6 of 25

2010 LACBWR Decommissioning Plan Review designed with inflatable pneumatic seals having a defined acceptable leakage rate. Appropriate interfacing modifications to the bio-shield liner at the edges of the opening will be installed to ensure water retention in the area between the upper cavity liner and the cask pool gate. The cask pool gate storage stand will support the 6-ton cask pool gate when not in use.

The 10' high by 10' inner diameter cask pool will be installed at elevation 669' 3" atop a 20'-10" high support structure attached to the reactor support cylinder at elevation 648'-5". The cask pool will have a 16Y2" wide horizontal flange welded to the top of the shell, the outer circumference of which will be tied into the existing upper cavity liner using L-shaped stainless steel angle at approximate elevation 679'-37. This arrangement will provide a barrier to prevent water in the upper cavity area above the cask poolfrom leaking around the outside of the cask pool into the cavity below.

The upper cavity liner and bio-shield contained a number of penetrations from reactor operation that will be sealed using steel plates welded to the upper cavity liner to maintain the pressure boundary of the upper cavity. The cask pool will also include two penetrations, one at the bottom and one on the side approximately 7"6" above the tank bottom. These penetrations will connect to piping and valves for filling, draining, and processing water from the pool and upper cavity to permit removal of the cask pool gate, and to perform other activities such as water clean-up, cask annulus flushing, and inventory control.

Each 2-inch diameter pipe will include two manual valves in series just outboard of the cask pool to ensure redundant isolation."

Page 4-5 Section 4.2.3, Waste Treatment Building and LSA Storage Building: In the last paragraph of the section, Low Specific Activity is added and LSA is parenthesized as an acronym to provide definition of what the LSA Building means.

Page 4-5 Section 4.2.5. Onsite Independent Spent Fuel Storage Installation: New section Through is added to provide an overview and description of the ISFSI by the following:

Page 4-7 "The LA CB WR Dry Cask Storage Project establishes an Independent Spent Fuel Storage Installation (ISFSI) under general license provisions of 10 CFR 72, Subpart K on the Genoa site. The ISFSI site is about midway between the Mississippi River on the west and Highway 35 on the east. The ISFSI is located 2,232feet south-southwest of the Reactor Building center on land which was previously used for an access road between the two closed ash landfills of the Genoa site.

The ISFSI will be used for interim storage of LA CB WR spent fuel assemblies in the NAC International, Inc. Multi-Purpose Canister (NAC-MPC) System. The NRC issued 10 CFR 72 Certificate of Compliance (CoC) No. 1025 which confers approval of the NAC-MPC storage system design. The design basis for the NA C-MPC System is provided in the NAC-MPC Final Safety Analysis Page 7 of 25

2010 LACBWR Decommissioning Plan Review Report (FSAR). The NRC approved Amendment 6 to the NAC-MPC CoC and Technical Specifications to incorporate LA CB WR spent fuel assemblies as approved contents for storage in the NAC-MPC System. The effective date of the license amendment was October 4, 2010.

The NA C-MPC System is comprised of the Transportable Storage Canister (TSC), the Vertical Concrete Cask (VCC), and the Transfer Cask (TFR). The TSC is designed to be transported in the NAC Transport Cask (STC) licensed by the NRC pursuant to 10 CFR 71 CoC No. 71-9235for shipment of the NAC-MPC canister. The NA C-MPC System designed for and to be used at LACBWR is designated MPC-LACBWR. The LACBWR spent fuel inventory will be placed into five MPC-LACBWR dry storage casks. Each MPC-LACBWR TSC can accommodate up to 68fuel assemblies for a total of 340 fuel storage cells among the five TSCs. Thirty-two of the 68fuel cell locations in each TSC are designed for a damaged fuel canister. Seven spare locations are available in the fifth TSC.

The spent fuel assemblies will be loaded underwater into a TSC within a TFR located in the cask pool. The TSC/TFR will be removed from the cask pool and the TSC will be prepared for storage by draining, helium fill, vacuum drying, and closure lid seal welding. The TSC/TFR will be moved outdoors and the TSC will be placed in the VCC by positioning the TFR on top of the VCC and lowering the TSC within the TFR through the transfer adapter to the VCC below. The VCC containing the TSC will then be placed in storage on the ISESI pad.

The TSC assembly consists of a right circular cylindrical shell with a welded bottom plate, afuel basket, a closure lid with closure ring, and two sets of redundant port covers. The cylindrical shell, plus the bottom plate and lid, constitutes the confinement boundary. The stainless steelfuel basket is a right circular cylinder with 68fuel tubes including 32 oversized tubes designed to accommodate LA CB WR damaged fuel cans, laterally supported by a series of stainless steel support disks. The support disks are retained by spacers on radially located tie rods. The spent fuel assemblies will be contained in square stainless steelfuel tubes that include Boral on up to four sides for criticality control.

The VCC is the storage overpack for the TSC and provides structural support, shielding, protection from environmental conditions, and natural convection cooling of the TSC during long term storage. The VCC is a reinforced concrete structure with a carbon steel inner liner. The VCC has an annular air passage to allow the natural circulation of air around the TSC. The air inlet and outlet vents take non-planar paths to the VCC cavity to minimize radiation streaming.

The decay heat is transferred from the spent fuel assemblies to the tubes in the fuel basket, through the heat transfer disks, to the TSC wall. Heat flows by convection from the TSC wall to the circulating air, as well as by radiation from Page 8 of 25

2010 LACBWR Decommissionin2 Plan Review the TSC wall to the VCC inner liner. The heat flow to the circulating air from the TSC wall and the VCC liner is exhausted through the air outlet vents. The top of the VCC is closed by a single shielded lid incorporating a carbon steel plate for gamma shielding and concrete for neutron shielding. The VCC lid will be bolted in place.

The TFR, with its lifting yoke, is a qualified heavy lifting device designed, fabricated, and proof load tested to the requirements of NUREG-0612 and ANSI N14. 6. The TFR provides shielding during TSC movements between work stations, the VCC, or the transport cask. It is a multi-wall (steel/leadINS-4-FR/steel) design and has a bolted top retaining ring to prevent a fuel-loaded canister from being inadvertently removed through the top of the TFR.

Retractable, hydraulically operated, bottom shield doors on the TFR are used during TSC transfer operations. To minimize contamination of the TSC, clean water will be circulated in the gap between the TFR and the TSC during cask pool loading operations.

The ISFSI pad made of reinforced concrete will be 48feet in length, and 32feet in width, and 3feet thick. The empty TFR and five VCCs containing fuel-loaded TSCs will be placed in a storage array on the ISFSI pad. The ISFSI will be surrounded by a protected area fence, an isolation zone fence, and vehicle barrier system. These protective barriers will be within the ISFSI Controlled Area Boundary fence. The ISFSI pad will be supplied with lighting, electronic surveillance and security systems. The ISFSI Administration Building is a commercial structure approximately 60feet by 30feet located 400feet northeast of the ISFSI pad. The ISFSIAdministration Building will provide space for security monitoring and ISFSI operations support."

Figures Figure 4.1, Containment Building Elevation: The figure is updated to depict modifications within the reactor cavity.

Figure 4.8, Onsite ISESI: A new figure is added depicting the ISFSI pad, storage array, and immediate vicinity.

Page 5-1 Section 5.1, Spent Fuel Inventory: First paragraph is updated to include ISFSI implementation by stating, "All spent fuel will be placed in dry storage casks and moved to the onsite Independent Spent Fuel Storage Installation."

Page 5-3 Section 5.2.2, Forced Circulation System, Section 5.2.3, Seal Injection System, Through and 5.2.4, Decay Heat System: Minor changes in capitalization, verb tense, Page 5-5 equipment titles, acronyms, and other edits are made in the sections for correction and consistency.

Page 5-5 Section 5.2.5, Emergency Core Spray System: In first paragraph, verbs are changed to past tense and final sentence of paragraph is deleted as unnecessary.

Second and third paragraphs are also deleted as being extraneous detailed Page 9 of 25

2010 LACBWR Decommissionina Plan Review information for a system that has been removed. System Status is shortened by simply stating, "This system has been removed."

Section 5.2.6, Primary Purification System: In first paragraph, verb is changed to past tense.

Page 5-6 Section 5.2.8, Alternate Core Spray System: Minor changes in capitalization, verb tense, equipment titles, acronyms, and other edits are made in the section for correction and consistency. System Status is shortened by removing unnecessary information about the closed isolation valve, 6-inch supply line removal, and designation as HPSW. The 6-inch supply penetration to the upper cavity will be sealed during cask pool installation as part of the cask pool installation. The system designation will not be changed.

Section 5.2.9, Control Rod Drive Auxiliaries: System Status is shortened by simply stating, "This system has been removed."

Page 5-7 Section 5.2.11, Fuel Element Storage Well System: In the paragraph at the top of the page, the phrase, "Spent fuel elements" is changed to, "Spent fuel assemblies" for consistency. A similar change is made to the System Status section. Minor changes in capitalization, verb tense, equipment titles, acronyms, and other edits are made in the section for correction and consistency. Plans for the FESW are added in the System Status section by stating, "Following removal of all spent fuel assemblies and fuel debris for dry storage at the ISFSI, the storage racks will be removed and disposed of; the FESW will be drained and decontaminated."

Page 5-7 Section 5.2.12, Component Cooling Water System: Minor changes in And capitalization, verb tense, equipment titles, acronyms, and other edits are made in Page 5-8 the section for correction and consistency. System Status is shortened to, "The system remains operable," with components presently being cooled by CCW moved to the first paragraph.

Page 5-8 Section 5.2.14, Shutdown Condenser System: Second paragraph has sentences And deleted as being extraneous detailed information for a system that has been Page 5-9 removed. Last paragraph of the section on page 5-9 has a verb changed to past tense and an undefined acronym replaced by the proper title.

Page 5-9 Section 5.2.15, Hydraulic Valve Accumulator System: System description is reorganized and minor changes in capitalization, verb tense, and equipment titles are made in the section for consistency.

Page 5-10 Section 5.2.17, Demineralized Water System: The title of Genoa Unit 3 is corrected as is statement concerning level indication which was removed under an approved facility change. System Status of Condensate Storage Tank being Page 10 of 25

2010 LACBWR Decommissionin2 Plan Review drained is added while statement of tank being covered in Condensate System is deleted.

Section 5.2.18, Overhead Storage Tank: Minor changes in capitalization, verb tense, equipment titles, acronyms, and other edits are made in the section for correction and consistency. Use of the system for cask loading operations is added. System Status is updated by stating, "After the transfer of all spent fuel to dry storage at the onsite ISFSI, the OHST system will be drained and dismantled."

Page 5-11 Section 5.2.21, High Pressure Service Water System: In first paragraph, second sentence, system pressure is maintained by the Genoa Unit 3jockey pump instead of the LPSW system. Minor changes in capitalization, equipment titles, and acronyms are made.

Page 5-12 Section 5.2.23, Condensate System and Feedwater Heaters, through Section Through 5.2.30, Waste Collection System: Minor changes in capitalization, verb tense, Page 5-15 equipment titles, acronyms, and other edits are made in the sections for correction and consistency.

Page 5-15 Section 5.2.3 1, Fuel Transfer Bridge: System Status is updated for ISFSI implementation by stating, "The fuel transfer bridge will be used to transfer individual spent fuel assemblies from the Fuel Element Storage Well to the dry cask storage container. After the transfer of all spent fuel to dry storage at the onsite ISFSI, the fuel transfer bridge will no longer be required."

Page 5-16 Section 5.2.33.1, Normal AC Distribution, through Section 5.2.33.3, Emergency And Diesel Generators 1A and 1B: Minor changes in capitalization, verb tense, Page 5-17 equipment titles, acronyms, and other edits are made in the sections for correction and consistency.

Page 5-17 Section 5.2.33.4, 120-V Non-Interruptible Buses: Section is shortened to a single paragraph for update and states:

"The 120-V Non-Interruptible Buses maintained a continuous non-interruptible power supply using static inverters to a portion of the essential plant control circuitry, communications equipment and radiological monitoring equipment.

The static inverters used 125-VDC input and supplied 120-VAC output to their respective distribution panels. The static inverters have been removed and the distribution panels have been renamed ]A 120-VAC Essential Power, Regulated Bus Auxiliary Panel, and 1C 120-VA C Essential Power."

Page 5-18 Section 5.2.33.5, 125-V DC Distribution: System description is updated to reflect installation and use of smaller capacity battery charger in Diesel Building by completion of an approved Facility Change. Minor changes in capitalization and equipment titles are made in the section for consistency. System Status of Page 11 of 25

2010 LACBWR Decommissionin2 Plan Review Electrical Power Distribution System is updated to reflect how configuration changes, made possible by approved procedures, permit a single diesel generator to supply the entire plant 480-V AC system during station power loss.

Page 5-18 Section 5.2.34, Post Accident Sampling Systems: Minor changes in capitalization, equipment titles, and acronyms are made in the section and its subsections for consistency. System Status is updated for ISFSI implementation by stating, "After the transfer of all spent fuel to dry storage at the onsite ISFSI, the Post-Accident Sampling Systems will no longer be required."

Page 5-19 Section 5.3, Radionuclide Inventory Estimates: Third sentence in first paragraph is changed to present tense.

Page 5-21 Section 5.4.2, System Radiation Levels: In the listing of survey points, current Through dose rates are updated to 2010 measurements.

Page 5-23 Page 6-1 Section 6.1, Objectives: First bulleted item stating, "safely store irradiatedfuiel,"

is changed to "safely store spent fuel" consistent with the definition in 10 CFR 72.3.

Section 6.2, Organization and Responsibilities: In the first sentence of the second paragraph, plant manager's responsibility is revised to include the ISFSI by stating, "is responsible for the safety of the facility and ISFSI, their daily operation and surveillance. " In the same paragraph, second sentence quality assurance and security programs has been made plural Page 6-2 Section 6.2, Organization and Responsibilities: In the second paragraph of the page, Operator responsibility to tour the facility is clarified to "when spent fuel is stored in the fuel element storage well." Following spent fuel removal, 24-hour surveillance and monitoring of spent fuel storage conditions will be performed by ISFSI personnel. In the fourth and fifth paragraphs, Health Physics and Instrument Technician responsibilities are expanded to include the ISFSI. In the fifth paragraph irradiated fuel is changed for consistency to spent fuel.

Page 6-4 Section 6.4.1, Training Program

Description:

In subsection 6.4.1.2, a statement is added concerning the change in training requirements following ISFSI implementation:

"These programs and requirements will change when all spent fuel is at the ISFSI. CFH training and proficiency will no longer be required. Training in ISFSI administration, security, monitoring, and maintenance will be fully implemented."

Page 6-8 Section 6.4.3.4, Certified Fuel Handler Training Program: In first paragraph, the need for this training program is clarified to "while spent fuel is stored in the Page 12 of 25

2010 LACBWR Decommissionin2 Plan Review Fuel Element Storage Well. " The acronym CFH has been deleted as unnecessary. Following spent fuel removal, certified and proficient fuel handlers will no longer be required. The second paragraph of the section has been moved as more appropriate to Section 10, "SAFSTOR Operator Training and Certification Program."

Page 6-8 Section 6.5, Quality Assurance: The four existing paragraphs describing QA And requirements in the D-Plan have been revised, re-organized, and expanded to Page 6-9 include Dry Cask Storage Project activities as reflected in changes to the Quality Assurance Program Description. The majority of the changes to the LACBWR QAPD relate to the inclusion of 10 CFR 72, Subpart G as holding applicable requirements that will be met to assure dry cask storage activities. The LACBWR QAPD, which has been approved by the NRC, meets the requirements of 10 CFR 50, Appendix B. As such, the LACBWR QAPD is acceptable as stated in 10 CFR 72.140(d), and establishes a quality assurance program satisfying each of the applicable criteria of 10 CFR 72 Subpart G. The intent to apply LACBWR's previously approved quality assurance program to ISFSI activities was declared to the NRC by letter dated August 24, 2009. The section with changed content in bold reads as follows:

"Decommissioning and SAFSTOR activities will be performed in accordance with the NRC-approved Quality Assurance Program Description (QAPD) for LACBWR. The QAPD has been developed to assure safe and reliable operation of LACBWR in a SAFSTOR condition and transition to an Independent Spent Fuel Storage Installation. The program is designed to meet the requirements of 10 CFR 50, Appendix B as applicable to the possession-only license condition and 10 CFR 72, Subpart G as applicable to the onsite dry cask storage of spent nuclear fuel.

The QAPD addresses all 18 criteria of 10 CFR 50, Appendix B, and 10 CFR 72, Subpart G, and applies to all activities affecting the functions of the structures, systems, and components that are associated with a possession-only license condition using a graded approach, and the Dry Cask Storage Project, respectively. These activities include design, installation, operations, maintenance, repair, fuel handling, testing, modifications, and radioactive waste shipments. Design and fabrication of storage and shipping casks for radioactive material are addressed by the Dry Cask Storage System Certificate of Compliance holder's quality assurance program and will not be conducted under the QAPD. Safety Related as defined in 10 CFR 50.2 is no longer applicable in the possession-only license condition. A graded approach is used to implement the QAPD by establishing managerial and administrative controls commensurate with the complexity and regulatory requirements of the activities undertaken.

Scheduled activities during SAFSTOR shall be performed within schedule intervals. A schedule interval is a time frame within which each scheduled Page 13 of 25

2010 LACBWR Decommissioning Plan Review activity shall be performed, with a maximum allowable extension not to exceed 25 percent of the schedule interval.

For Important to Safety (ITS) activities, as defined in 10 CFA 72, Subpart G, a Quality Assurance Project Plan (QAPP) was established to define the quality assurance requirements to be implemented during the Dry Cask Storage Project at the LACBWR site. The QAPP does not supplant the QAPD which is used to assure safe and reliable operation of the LA CB WR plant in a SAFSTOR condition. The QAPP utilizes the QAPD as the base document of the Dry Cask Storage Project's overall quality assurance program. The QAPP references and provides clarification to each applicable section of the QAPD that collectively meet the quality assurance requirements of 10 CFR 50 Appendix B, 10 CFR 71 Subpart H, and 10 CFR 72 Subpart G.

The QAPP applies to all activities associated with the design, fabrication, installation, and preparation for operation of an Independent Spent Fuel Storage Installation and any related plant modifications and other site activities as designated by the Plant Manager. Design and fabrication of the Dry Cask Storage System (DCSS) will not be performed under the QAPP; however, selection, qualification, and performance-based overview of the selected DCSS designer and DCSS fabrication will be conducted in accordance with the QAPP. "

Page 6-9 Section 6.6, Schedule: In the first paragraph, a sentence is added from a paragraph that has been deleted at the end of the section stating, "At the time of the original Decommissioning Plan in 1987, DPC anticipated the plant would be in SAFSTOR for a 30-50 year period."

The third paragraph of the section that described DECON being planned for 2019 has been deleted as no longer being operative. DPC management has revised the schedule for final decommissioning of LACBWR; more details are provided in changes following in the section.

Page 6-10 Section 6.6, Schedule: In the first paragraph at the top of page 6-10, the third sentence is amended in describing efforts to place an ISFSI on site "by commencing the Dry Cask Storage Project." The sentence describing the duration of the project as being 3 years has been deleted as ISFSI implementation approaches completion. Discussion of transport of spent fuel to Yucca Mountain has been changed to offsite due to uncertainty in the federal repository issue.

After the paragraph that discusses Private Fuel Storage, new information is added concerning acceleration in the final decommissioning of the LACBWR facility.

The information is self-explanatory and is added as follows:

"DPC Staff completed an extensive review and analysis of the comparative costs and benefits of the current decommissioning schedule and various accelerated Page 14 of 25

2010 LACBWR Decommissioning Plan Review schedules. From this analysis, the DPC Board of Directors approved accelerating the removal of radioactive metalfrom the LA CB WR facility. By letter dated December 7, 2010, DPC gave notification to the NRC of a change in schedule that would accelerate the decommissioning of the LACBWR facility starting with a 4-year period of systems removal beginning in 2012. This activity will include the removalfor shipment of large bore (16 and 20-inch) reactor coolant piping and pumps of the Forced Circulation system and other equipment once connected to the reactor pressure vessel or primary system such as Control Rod Drive Mechanisms, Decay Heat, Primary Purification, Seal Injection, and Main Steam.

This phase of decommissioning activity does not result in significant environmental impacts and has been reviewed as documented in the "Generic Environmental Impact Statement (GEIS) on Decommissioning of Nuclear Facilities," NUREG-0586, Supplement 1, November 2002. As stated in the GEIS, licensees can rely on information in this Supplement as a basis for meeting the requirements in 10 CFR 50.82(a)(6)(ii). The GEIS characterizes the environmental impacts resulting from this decommissioning activity as generic and small. Potential site-specific environmental impacts not determined in the GEIS will be addressed in the License Termination Plan (LTP)for LA CBWR.

DPC's review and analysis found that the Nuclear Decommissioning Trust (NDT) was sufficiently funded to allow dismantlement to begin in 2012, immediately after spent fuel removal is completed. Costs of the metal removal project will befundedfrom the NDT. DPC's approved strategy requires continuing evaluation of the costs of the decommissioning activity as it progresses. During this time the LTP will be formulated determining the disposition of concrete structures and site end use. The LTP will include an updated site-specific estimate of remaining decommissioning costs. DPC's decommissioning strategy for LA CB WR with accelerated systems removal provides flexibility in that provisions are afforded to evaluate the costs and benefits of alternative methodologies for concrete removal, and delay LTP implementation if necessary to assure adequate NDTfunds are available for the final decommissioning process. Figure 6.2 depicts the revised schedule."

Page 6-11 Section 6.7.1, SAFSTOR Funding: In the first paragraph of the section, the DPC spent fuel management and funding plan applicability, pursuant to 10 CFR 50.54(bb), is expanded to include spent fuel storage costs at the ISFSI by addition of the statement, "This plan applies also to ISFSI operations."

In the second paragraph of the section, the title, Decommissioning Trust Fund, is corrected to the legal title it bears, Nuclear Decommissioning Trust and the acronym (NDT) is identified. NDT or NDTfunds is then used in three places in the paragraph replacing Decommissioning Trust Fund NDT is also used in the sixth paragraph of the section.

Page 15 of 25

2010 LACBWR Decommissionin2 Plan Review Page 6-12 Section 6.7.2, Decommissioning Cost Financing: In the first paragraph of the section, correction is made again to the title and acronym of the Nuclear Decommissioning Trust (NDT). The same correction is made to NDT in the fourth paragraph of the section.

A new paragraph describing the 2010 decommissioning cost study update is added following description of the 2007 cost study update. The paragraph states the following:

"A cost study update was completed in November 2010 to more accurately assess future costs of the remaining dismantlement needed and tofacilitate DPC decommissioning and license termination planning. This update placed the cost to complete decommissioning at $67.8 million in Year 2010 dollars.

During this process, ISFSI decommissioning costs were identified uniquely as a specific item within the cost study and estimated to be $1.6 million in Year 2010 dollars. The DPC Board of Directors will establish an externalfunding mechanism for ISFSI decommissioning costs in accordance with 10 CFR 72.30 to assure adequate funds will be available for the final decommissioning cost of the LACBWR ISFS."

Page 6-13 Section 6.7.2, Decommissioning Cost Financing: In the second paragraph of Section 6.7.2 on page 6-13, it is added that the DPC Board of Directors remain committed to assuring that adequate funding will be available for the final decommissioning of the LACBWR facility and ISFS.

Section 6.8, Special Nuclear Material (SNM) Accountability: In second paragraph it is clarified that LACBWR spent fuel inventory is stored underwater in the FESW or in dry storage casks located at the onsite ISFSI. Purpose of change is to accommodate ISFSI implementation.

Section 6.9.1, Fire Protection Plan: Information is added to accommodate ISFSI implementation in LACBWR fire protection planning. In the first paragraph it is stated that the FESW can be safely maintained and controlled during fire in each area of the plant while spent fuel is stored wet. Description is added to end of same paragraph stating the intent of LACBWR fire protection with ISFSI operations by stating:

"With implementation of ISFSI operations fire protection planning for the LACBWR facility will adapt to the absence offuel stored wet and will focus on ISFSIfire protection. As long as radiological hazards remain at the LA CB WR facility, fire protection planning will be commensurate with the risks associated with the reduction in those radiological hazards."

Page 6-14 Section 6.9.1, Fire Protection Plan: In third bulleted item of page, the word deconned is corrected to decontaminated. A new paragraph is added to the end Page 16 of 25

2010 LACBWR Decommissioning Plan Review of the section to provide description of the ISFSI fire hazards analysis and ISFSI compliance with 10 CFR 72.122(c) general design criteria. The new paragraph reads as follows:

"The fire protection plan at the ISFSI is based on the fire hazards analysis performed in support of the Dry Cask Storage Project. The ISFSI fire hazards analysis demonstrated that the explosion and heat effects of crediblefire and explosion hazards at the Genoa site will not significantly increase the risk of radioactivity release to the environment. Therefore, storage of spent nuclear fuel at the LA CB WR ISFSI is in accordance with 10 CFR 72.122(c) general design criteria. The ISFSIfire protection features and administrative controls will ensure that the Genoa site fire and explosion hazards are acceptable and within the cask system design basis for fuel-loaded Vertical Concrete Casks (VCCs) located at or in route to the ISFSI."

Page 6-15 Section 6.9.2, Fire Protection Program: A new paragraph is added at the beginning of the section that provides a general description of the ISFSI fire protection program and states:

"The fire protection program for the ISFSI consists mainly of administrative controls to limit flammable liquids and combustible materials in the area of the fuel-loaded VCCs and is implemented by ISFSI procedures. There will be no organized fire brigade at the ISFSI. Personnel monitoring dry cask storage conditions may extinguish incipient fires, but Genoa Fire Department will be summoned for fire emergencies at the ISFSI site."

Section 6.9.2.2, Fire Detection System: A short description of ISFSI Administration Building fire detectors is added.

Page 6-16 Section 6.9.2.4, Fire Suppression Water System: After completion of an approved facility change, pressure maintenance of the High Pressure Water System is changed from the Low Pressure Service Water System (LPSW) to the Genoa Unit 3 jockey pump.

Section 6.9.2.6, Portable Fire Extinguishers and Other Fire Protection Equipment:

In the first paragraph Halon is removed as a type of fire extinguisher because Halon units have been retired at LACBWR. A short mention is added that "The ISFSI is supplied with its own complement of portable fire extinguishers." In the final paragraph of the section a correction is made at "sprinkler heads and sprinkler equipment are located in the Maintenance Shop emergency locker."

Page 6-17 Section 6.9.2.7, The Fire Brigade: A correction is made to change Duty Shift Supervisor to Operations Shift Supervisor to avoid confusion with the future addition of the ISFSI Security Shift Supervisor.

Page 17 of 25

2010 LACBWR Decommissionin2 Plan Review Section 6.9.2.8, Outside Fire Service Assistance: Genoa Fire Department support during emergencies will include LACBWR or the ISFSI.

Page 6-17 Section 6.9.2.9, Reporting: An addition is made to provide reporting requirements for ISFSI fire emergencies by the following:

3) Any incident requiring outside fire service assistance within the ISFSI Controlled Area Boundary shall be reported by the ISFSI Security Shift Supervisor using an ISFSI Security Incident Report.

Page 6-18 Section 6.9.2.10, Training: In the first paragraph, Security badged visitors is corrected to Security badged personnel. The second paragraph is revised to clarify that Fire Brigade training is applicable to only Operations and Security personnel assigned duties at the LACBWR plant. The second paragraph reads as follows with additions in bold:

"Operations and Security personnel have Fire Brigade responsibilities and are given basic practicalfire fighting and specific fire protection program instruction annually. Fire Brigade members shall also participate in at least one drill annually."

A new paragraph is added at the end of the section to clarify the response role of ISFSI personnel during fire and to indicate that there will not be an organized fire brigade at the ISFSI. The new paragraph states:

"ISFSI personnel will be termed designated employees, and as such will not be members of an organized fire brigade. These personnel will be properly trained to use portable fire extinguishers to fight incipient fires in the employee's immediate work area."

Section 6.10, Security during SAFSTOR and/or Decommissioning: In the first paragraph, maintenance of security is expanded to the LACBWR facility and ISESI. The last paragraph of the section has a sentence added stating, "ISESI security requirements are addressed and implemented as applicable."

Section 6.11, Testing and Maintenance of SAFSTOR Systems: In the second paragraph a statement is added indicating that 10 CFR 50.65 is no longer applicable during dry cask storage. The sentence states:

"When all spent fuel is at the ISFSI, Maintenance Rule Program requirements will no longer be applicable."

Page 6-19 Section 6.12.3.1, Stack: The discussion of stack effluents is clarified by indicating that with dry cask storage krypton-85 will no longer be a component of any effluent release. A new sentence is added stating:

Page 18 of 25

2010 LACBWR Decommissioning Plan Review "With all spent fuel stored at the ISFSI, no concentration of Kr-85 will be available as a source of radioactivity in effluent releases."

Page 6-20 Section 6.13, Records: The section is revised to include record requirements for ISFSI operation. Four bulleted items are changed or added in bold as follows:

0 Baseline surveys performed in and around the LACBWRfacility and ISFSI.

0 Analysis and evaluations of total radioactivity concentrations at the LACBWR facility and ISFSI.

0 Records for ISFSI construction, dry cask storage system fabrication, and dry cask loading.

N Any other records or documents, which would be needed to facilitate decontamination and dismantlement of the LACBWR facility or ISFSI and are not controlled by other means.

Figure 6.1 La Crosse Boiling Water Reactor SAFSTOR Staff: The organization chart has a block signifying ISFSI Operations and Security added under Plant Manager. A note is added at bottom clarifying who will be performing ISFSI duties and states,

" *Duties to be performed by existing LA CB WR staff and security force."

Figure 6.2 LACBWR Schedule: Figure is updated to reflect decommissioning planning and schedule changes as discussed previously in Section 6.6.

Page 7-1 Section 7.2, SAFSTOR Modifications: In the final paragraph of the section, clarification is added to indicate that with dry cask storage krypton-85 will no longer be a component of any effluent release. A new sentence states:

"With all spent fuel stored at the Independent Spent Fuel Storage Installation (ISFSI) in seal-welded canisters protected by concrete overpacks, there is no concentration of the isotope Kr-85 available for release from the LACBWR plant."

Section 7.3, Significant SAFSTOR Licensing Actions: In the second paragraph of the section, "and also of the same date," and "proposed" are deleted in the description of License Amendment No. 66 as being unnecessary. On page 7-2, after the description of License Amendment No. 69, a new paragraph is added to provide information of License Amendment No. 71 which has been submitted for Dry Cask Storage Project needs. The amendment remains in the NRC review and approval process at this time. The new paragraph reads as follows:

"License Amendment No. 71 was submitted July 28, 2009, requesting changes to the LACBWR Appendix A, Technical Specifications in support of the LACBWR Dry Cask Storage Project. The request seeks approval of a revised Page 19 of 25

2010 LACBWR Decommissioning Plan Review definition of FUEL HANDLING, approval of a reduction of the minimum water coverage over stored spent fuel from 16feet to llfeet, 6Y2 inches, and a small number of editorial changes to clarify heavy load controls and reflect inclusion of the cask pool as part of an "extended" Fuel Element Storage Well.

These changes were requested to accommodate efficient dry cask storage system loading operations and reduce overall occupational dose to personnel during these operations."

Page 7-3 Section 7.5, Removal of Unused Equipment during SAFSTOR: The fourth paragraph is clarified to commit to the requirements of 10 CFR 50.59 by stating, "Removal of plant equipment will be performed in accordance with the requirements of 10 CFR 50.59."

Page 7-4 Section 7.6.1, Temporary Lifting Device: In conjunction with reformatting, the And section is re-organized and shortened by the deletion of overly detailed Page 7-5 information about the Temporary Lifting Device (TLD) runway construction used in removal of the reactor pressure vessel. A reasonable amount of historical information about the TLD, basically unchanged from the previous revision, is presented by the following:

"A Temporary Lifting Device/Gantry Rail System (TLD) was erected and installed inside and outside the RB. The TLD system consisted of a temporary runway structure and rolling trolley which incorporated hydraulic strand jacks for lifting the RPV The runway structure consisted of 37-feet girders inside the RB and 74-feet girders outside the RB. The runway structure design inside the RB met NUREG-0612 criteria. The runway structure design outside the RB was not required to meet NUREG-0612 criteria, as NUREG-0612 pertains to lifts and equipment inside buildings where spent fuel is stored.

The trolley was a moveable platform with four two-wheeled bogie end trucks (8 total double flanged wheels) designed to run on the box girder rails. Two of the trucks had electric mechanical drives. Each drive consisted of a gearbox, motor, and brake. There were two driven/braked wheels in the 8 wheel set. The brake was automatically set when the momentary directional motion switch was released to the neutral position. The trolley had two travel speeds; 1. 62feet per minute (FPM) and 6.80 FPM. Both travel speeds were very slow. At the slower speed it would have taken over an hour to traverse the runway from south to north. Two hydraulic strandjack hoisting systems were mounted on top of the trolley platform. The strandjack systems were independent from each other and were specially fabricated to meet the specifications for the LACB WR RPV lift and transport. Hoisting speed was 0.5 FPM. The strand jacks were comprised of 36 strands per jack; failure of any given strand would not result in loss of control of the suspended load. Failure of over 75% of the strands would have had to occur before the remaining strands could not carry the load. Two separate electrical sources were used to power the two strandjack power packs and one trolley drive system through three dedicated load disconnect switches.

Page 20 of 25

2010 LACBWR Decommissionins Plan Review The strand jack system was designed such that the load would remain secured at the height lifted upon loss of power or hydraulic pressure. The trolley assembly was designed to meet NUREG-0612 criteria.

The TLD was constructed of components within the Bigge equipment inventory along with new fabricated assemblies. Prior to TLD use for the RPVlift, a load test of 110% of the load lifted outside the RB (service load 639,000 lbs/test load 703,000 lbs) was conducted. Since a load test of 150% of the load lifted inside the RB (service load 380,000 lbs/test load 570,000 lbs) was less than the outside load test weight, the inside load test was not performed. The percent increases above static weight or service load were consistent with NQA-1 and ANSI N14.6.

The custom built RPV attachment/hlandlingfixture used inside the RB was load tested in accordance with ANSI N14.6-1993, Section 7, "Special lifting devices for critical loads." Section 7.3.1(a) required the test load to be three times (3x) the weight the fixture would support. The handling fixture load test was documented for record.

All TLD equipment was removed following RPV removal with the exception of two rocker bearing assemblies installed on the bio-shield at elevation 701 'and two bearing assemblies mounted at the RB wall opening. "

Page 7-6 Section 7.6.3, 50.59 Evaluations: The numbered listing of the eight criteria examined in the LACBWR 50.59 review process is deleted as unnecessary.

Page 7-8 Section 7.8, Dry Cask Storage Project: A new section is added giving description of the Dry Cask Storage Project and related ISFSI licensing information and reference:

"The LACBWR Dry Cask Storage Project establishes an ISFSI under general license provisions of 10 CFR 72, Subpart K, on the Genoa site. The ISFSI is located 2,232feet south-southwest of the Reactor Building center on land which was previously used for the access road between the two closed ash landfills of the Genoa site. The ISFSI will be used for interim storage of LA CB WR spent fuel assemblies in the NA C International, Inc. (NA C) Multi-Purpose Canister (MPC) System. 10 CFR 72.212 requires a general licensee to conduct and document an array of reviews to confirm that the physical ISFSI site and the site organization are prepared to implement dry spent fuel storage and that the generically designed dry spent fuel storage cask chosen for use bounds applicable site-specific design criteria and conditions. This evaluation is documented in the LACBWR ISFSI 10 CFR 212 Report and meets the requirements set forth in 10 CFR 72.212(b)(2)(i), (b)(3), and (b)(4) that mandate such written evaluations prior to use of the cask system under a Part 72 general license.

Page 21 of 25

2010 LACBWR Decommissioning Plan Review Refer to the NA C-MPC Storage System Certificate of Compliance No. 1025 and Final Safety Analysis Report for details of the MPC-LACBWR design, operation, and safety analyses. Decommissioning Plan Section 4.2.1 discusses modifications made to the Reactor Building for dry cask loading operations and Section 4.2.5 discusses the onsite ISESI."

Page 7-8 Section 7.8.1, Cask Handling Crane: A new section is added providing And description of the single failure proof lift system that will be used during dry Page 7-9 storage cask transfer operations and states the following:

"American Crane and Equipment Corporation (ACECO) was contracted to supply an 85-ton capacity temporary cask handling system for the Dry Cask Storage Project. The temporary cask handling system consists of the cask handling crane being supplied by ACECO; and a temporary runway system being supplied by Rigging International. The cask handling crane consists of a refurbished single failure proof trolley and hoist previously utilized at Maine Yankee designed, manufactured, and tested in accordance with ASME NOG 1998 and NUREG 0554. The temporary runway system is a new structure designed, manufactured, and tested in accordance with the requirements of ASME NOG-1-2004 for a Type I Crane (i.e. single failure proof crane).

Features are included in the crane design to assure that any credible failure of a single component will not result in the loss of capability to stop and hold the critical load.

Dry cask storage system component lifts will be made using the single failure proof cask handling crane. All other lifts of equipment in the Reactor Building are performed in accordance with the requirements of NUREG-0612, LA CB WR heavy load control procedures, and activity-specific rigging plans."

Page 7-9 Section 7.9, Environmental Impact: In the second paragraph limited is deleted before dismantlement activities due to planned acceleration of LACBWR decommissioning. ISFSI activity not within the scope of the GEIS is addressed in the licensing process for the ISFSI by completion of the LACBWR ISFSI 10 CFR 72.212 Report rather than will be addressed Page 8-1 Section 8, Health Physics: At end of paragraph it is added that exposures will be maintained ALARA "at LA CB WR and the onsite ISFSI."

Page 8-2 Section 8.2.3, Radiation Exposure Limits: An introductory statement is added for explanation stating, "Radiation exposure to individuals at LACBWR will be controlled and limited in accordance with the following:" and the following limits are renumbered. The description of "Daily Administrative Limit" is deleted after being removed from the Operating Manual earlier under a process that included a 50.59 review. The daily limit was a holdover from operations when higher doses were routine. Today, doses are much lower and it is not appropriate Page 22 of 25

2010 LACBWR Decommissioning Plan Review to have such a high daily limit. ALARA planning and review control dose distribution during work; annual limits are not exceeded.

Page 8-3 Section 8.2.3.3, The NRC establishes the annual occupational dose limits, has been renumbered, annual replaces following in the initial passage which has also been underlined as the section topic. At the end of the introduction, the phrase "annual limits" is deleted as redundant.

Page 8-8 Section 8.4.6, ISFSI Monitoring: A new section is added to provide information And about ISFSI radiological conditions and monitoring:

Page 8-9 "Spent fuel will be placed in dry storage at the ISFSI in the MPC-LACB WR system which will provide an inert environment; passive shielding, cooling and criticality control; and a confinement boundary closed by welding. The structural integrity of the system precludes the release of contents in any of the design basis normal conditions and off-normal or accident events, thereby assuring public health and safety during use of the system.

The MPC-LACBWR 5-cask array is evaluated to determine the minimum distance necessary to achieve a controlled area boundary dose of 25 mrem/year as required by 10 CFR 72.104(a). In the NAC-MPC FSAR, Section IO.A, annual exposures, based on an 8 760-hour residence year, were determined from the center of a single cask and a 5-cask array. The NAC-MPC FSAR includes a plot of the 25 mrem/year footprint and the boundary required. A rectangular boundary a minimum of 300feetfrom the pad center around the ISFSI ensures compliance with the requirements of 10 CFR 72.104(a) that dose rate will not exceed 25 mrem/year at the Controlled Area Boundary.

Prior to ISFSI construction, baseline radiation sampling and surveys were performed at the ISFSI site. With implementation of ISFSI operations, the fuel-loaded Vertical Concrete Cask (VCC) dose rates will be verified to be compliant with limits specified in Technical Specifications for the NAC-MPC System to maintain dose rates ALARA at locations on the VCCs where surveillance is performed and to reduce offsite exposures. Radiological conditions at the ISFSI will be monitored routinely to evaluate the continued effectiveness of the dry storage cask confinement boundary."

Page 8-11 Section 8.6.3, Dismantlement (Metallic): It is added that metallic waste will be sent to a reprocessor or disposal site as an option.

Page 9-1 Section 9, SAFSTOR Accident Analysis: Throughout section, reference citations are changed due to renumbering.

Section 9.1,

Introduction:

First paragraph is revised to clarify that SAFSTOR accident analyses apply to wet storage conditions in the FESW, and that the NAC-MPC FSAR provides dry cask storage accident discussion which will not be Page 23 of 25

2010 LACBWR Decommissioning Plan Review addressed in the Decommissioning Plan. The paragraph states the following with changes in bold:

"While spent fuel is in wet storage during SAFSTOR, the only major concern is protecting the fuel in the Fuel Element Storage Well (FESW). Once all spent fuel is placed in dry storage, accidents associated with the onsite ISFSI are discussed in the NAC-MPC Final Safety Analysis Report and the LACBWR ISFS1 10 CFR 72.212 Report. Accidents associated with dry cask storage at the ISFSI will not be addressed in this Decommissioning Plan."

In the second paragraph of the same section, changes are made for consistency to clarify spent fuel stored in the FESW.

Page 9-2 Section 9.2, Spent Fuel Handling Accident: In the first paragraph the acronym and FESW is used replacing Fuel Element Storage Well.

Page 9-3 The curie content remaining as of October 2009 and calculated values for Whole Body Dose and Skin Dose as of October 2009 are updated to October 2010.

Page 9-3 Section 9.3, Shipping Cask or Heavy Load Drop into FESW: In the first And paragraph the acronym FESW is used replacing Fuel Element Storage Well. The Page 9-4 curie content remaining as of October 2009 and calculated values for Whole Body Dose and Skin Dose as of October 2009 are updated to October 2010.

Page 9-5 Section 9.5, FESW Pipe Break: In first paragraph, discharge to the FESW is revised to state, "underwater in the FESW." The discharge pipe is being extended for reduced water level during dry cask loading operations. In second paragraph an abbreviated drain level is clarified by elevation 679feet.

Page 9-7 Section 9.7, Loss of Offsite Power: The first sentence of the fourth paragraph is corrected grammatically by stating, "If neither EDG can be started, FESW and CCW pumps cannot run."

Section 9.8, Seismic Event: In last sentence of first paragraph, spent fuel is corrected and the acronym FESW is used.

Page 9-8 Section 9.9, Wind and Tornado: In last sentence of first paragraph, spentfiuel is corrected and the acronym FESW is used.

Page 10-1 Section 10.1,

Introduction:

In first paragraph, clarification is added at the end that certification of Operations personnel is only necessary "while spent fuel is in wet storage in the FESW." A second paragraph, removed from Section 6, is added with a statement that CFH training is no longer needed following transfer of all spent fuel to the ISFSI. The new paragraph states, Page 24 of 25

2010 LACBWR Decommissionin2 Plan Review "The Operator Training and Certification Programs ensure that people trained and qualified to operate LACBWR will be available during the SAFSTOR period. Licensee certification of personnel makes it unnecessary for the NRC to periodically conduct license examinations for persons involved in infrequent activities and prevents delays due to obtaining NRC Fuel Handler Licenses for any evolutions that may require fuel movements. When all spent fuel is in dry storage at the ISFSI, CFH training and proficiency will no longer be required."

Section 10.2, Applicability: The acronym CFH is identified and used.

Page 10-1 Section 10.3, Initial Certification: The acronyms CFH and FESW are used.

And With renumbering, the entire section is reorganized with minor wording changes.

Page 10-2 INITIAL SITE CHARACTERIZATION SURVEY FOR SAFSTOR (LAC-TR-138):

Cover Page Update revision date.

Entire Where used, Containment Building and abbreviation CB have been changed to Report Reactor Building and abbreviated RB for consistency with previous updates to the D-Plan and relevant material.

Page 24 Attachment I curie content values decay-corrected to October 2009 are updated to October 2010 values.

Page 26 curie content values decay-corrected to October 2009 are updated Through to October 2010 values.

Page 28 Page 25 of 25

LA CROSSE BOILING WATER REACTOR (LACBWR)

DECOMMISSIONING PLAN I

Revised November 2010 DAIRYLAND POWER COOPERATIVE LA CROSSE BOILING WATER REACTOR (LACBWR) 4601 State Road 35 Genoa, WI 54632-8846 I

TABLE OF CONTENTS Page No.

1.

Introduction 1-1 1.1 Selection of SAFSTOR 1-1 1.2 References 1-3

2.

La Crosse Boiling Water Reactor Operating History 2-1 2.1 Introduction 2-1 2.2 Initial Construction and Licensing History 2-1 2.3 Operating Record 2-2 2.4 Decision for Shutdown 2-2 2.5 Operating Events which Affect Decommissioning 2-3 2.5.1 Failed Fuel 2-3 2.5.2 Fuel Element Storage Well Leakage 2-3 2.6 References 2-4

3.

Facility Site Characteristics 3-1 3.1 Geography and Demography Characteristics 3-1 3.1.1 Site Location and Description of Site Layout 3-1 3.1.2 The Authority of the Exclusion Area and Licensee Authorities 3-1 3.2 Transportation, Industrial and Military Facilities within Proximity to the Plant 3-2 3.3 Meteorology 3-2 3.3.1 Meteorological Measurement Program 3-2 3.3.2 General Climatology 3-3 3.3.3 Local Meteorology 3-3 3.4 Hydrology 3-5 3.4.1 Hydrologic Description 3-5 3.4.2 Drainage 3-5 3.4.3 Downstream Water Use 3-5 3.4.4 Flooding and Probable Maximum Flood 3-6 3.4.5 Potential Dam Failures 3-6 3.4.6 Flooding Protection Requirements 3-6 3.4.7 Ultimate Heat Sink and Low Flow Conditions 3-7 3.4.8 Ground Water 3-8 3.5 Geology, Seismology and Geotechnical Engineering 3-8 3.5.1 Basic Geologic Seismic Information 3-8 3.5.2 Proximity of Capable Tectonic Structures in the Plant Vicinity 3-9 3.5.3 Surface Faulting at the Site 3-12 D-PLAN 0-1 November2010 I

TABLE OF CONTENTS Page No.

3.5.4 Stability of Slopes and Subsurface Materials 3-12 3.5.5 Stability of Subsurface Materials and Foundations 3-13 3.6 ISFSI Soils and Seismology 3-14 3.7 References 3-15

4.

Facility Description 4-1 4.1 General Plant Description 4-1 4.2 Buildings and Structures 4-1 4.2.1 Reactor Building 4-1 4.2.2 Turbine Building 4-4 4.2.3 Waste Treatment Building 4-5 4.2.4 Cribhouse 4-5 4.2.5 Onsite Independent Spent Fuel Storage Installation 4-5

5.

Plant Status 5-1 5.1 Spent Fuel Inventory 5-1 5.2 Plant Systems and their Status 5-3 5.2.1 Reactor Vessel and Internals 5-3 5.2.2 Forced Circulation System 5-3 5.2.3 Seal Injection System 5-4 5.2.4 Decay Heat Cooling System 5-5 5.2.5 Emergency Core Spray System 5-5 5.2.6 Boron Injection System 5-5 5.2.7 Primary Purification System 5-5 5.2.8 Alternate Core Spray System 5-6 5.2.9 Control Rod Drive Auxiliaries 5-6 5.2.10 Gaseous Waste Disposal System 5-6 5.2.11 Fuel Element Storage Well System 5-6 5.2.12 Component Cooling Water System 5-7 5.2.13 Shield Cooling System 5-8 5.2.14 Shutdown Condenser System 5-8 5.2.15 Hydraulic Valve Accumulator System 5-9 5.2.16 Well Water System 5-9 5.2.17 Demineralized Water System 5-9 5.2.18 Overhead Storage Tank 5-10 5.2.19 Station and Control Air System 5-10 5.2.20 Low Pressure Service Water System 5-11 5.2.21 High Pressure Service Water System 5-11 5.2.22 Circulating Water System 5-11 5.2.23 Condensate System and Feedwater Heaters 5-12 5.2.24 Reactor Feedwater Pumps 5-12 5.2.25 Full-Flow Condensate Demineralizer System 5-12 5.2.26 Steam Turbine 5-13 D-PLAN 0-2 November 2010

TABLE OF CONTENTS Page No.

5.2.27 60-Megawatt Generator 5-13 5.2.28 Turbine Oil and Hydrogen Seal Oil System 5-13 5.2.29 Heating, Ventilation, and Air-Conditioning Systems 5-14 5.2.30 Waste Collection Systems 5-15 5.2.31 Fuel Transfer Bridge 5-15 5.2.32 Communications Systems 5-16 5.2.33 Electrical Power Distribution 5-16 5.2.34 Post-Accident Sampling Systems 5-18 5.2.35 Containment Integrity Systems 5-19 5.3 Radionuclide Inventory Estimates 5-19 5.4 Radiation Levels 5-19 5.4.1 Plant Radiation Levels 5-19 5.4.2 System Radiation Levels 5-20 5.5 Plant Personnel Dose Estimate 5-24 5.6 Sources 5-24 5.7 Radiation Monitoring Instrumentation 5-24 5.7.1 Fixed Plant Monitors 5-24 5.7.2 Portable Monitors 5-25 5.7.3 Laboratory-Type Monitors 5-25

6.

Decommissioning Program 6-1 6.1 Objectives 6-1 6.2 Organization and Responsibilities 6-1 6.3 Contractor Use 6-3 6.4 Training Program 6-4 6.4.1 Training Program Description 6-4 6.4.2 General Employee Training (GET) 6-4 6.4.3 Technical Training 6-5 6.4.4 Other Decommissioning Training 6-8 6.4.5 Training Program Administration and Records 6-8 6.5 Quality Assurance 6-8 6.6 Schedule 6-9 6.7 SAFSTOR Funding and Decommissioning Cost Financing 6-11 6.7.1 SAFSTOR Funding 6-11 6.7.2 Decommissioning Cost Financing 6-12 6.8 Special Nuclear Material (SNM) Accountability 6-13 6.9 SAFSTOR Fire Protection Program 6-13 6.9.1 Fire Protection Plan 6-13 6.9.2 Fire Protection Program 6-15 D-PLAN 0-3 November 2010 1

TABLE OF CONTENTS Page No.

6.10 Security during SAFSTOR and/or Decommissioning 6-18 6.11 Testing and Maintenance of SAFSTOR Systems 6-18 6.12 Plant Monitoring Program 6-19 6.12.1 Baseline Radiation Surveys 6-19 6.12.2 In-Plant Monitoring 6-19 6.12.3 Release Point / Effluent Monitoring 6-19 6.12.4 Environmental Monitoring 6-20 6.13 Records 6-20

7.

Decommissioningz Activities 7-1 7.1 Preparation for SAFSTOR 7-1 7.2 SAFSTOR Modifications 7-1 7.3 Significant SAFSTOR Licensing Actions 7-1 7.4 Area and System Decontamination 7-2 7.5 Removal of Unused Equipment During SAFSTOR 7-3 7.6 Reactor Pressure Vessel Removal 7-3 7.6.1 Temporary Lifting Device 7-4 7.6.2 NUREG-0612 Compliance 7-5 7.6.3 50.59 Evaluations 7-6 7.6.4 References 7-7 7.7 B/C Waste Removal 7-8 7.8 Dry Cask Storage Project 7-8 7.8.1 Cask Handling Crane 7-8 7.9 Environmental Impact 7-19

8.

Health Physics 8-1 8.1 Organization and Responsibilities 8-1 8.2 ALARA Program 8-1 8.2.1 Basic Philosophy 8-1 8.2.2 Application of ALARA 8-2 8.2.3 Radiation Exposure Limits 8-2 8.3 Radiation Protection Program 8-4 8.3.1 Personnel Monitoring 8-4 8.3.2 Respiratory Protection Program 8-5 8.3.3 Protective Clothing 8-6 8.3.4 Access Control 8-6 8.3.5 Postings 8-6 8.4 Radiation Monitoring 8-6 D-PLAN 0-4 November 2010 1

TABLE OF CONTENTS Page No.

8.4.1 Airborne Radioactivity Surveys 8-7 8.4.2 Radiation Surveys 8-7 8.4.3 Contamination Surveys 8-8 8.4.4 Liquid Activity Surveys 8-8 8.4.5 Environmental Surveys 8-8 8.4.6 ISFSI Radiation Monitoring 8-8 8.5 Radiation Protection Equipment and Instrumentation 8-9 8.5.1 Portable Instruments 8-9 8.5.2 Installed Instrumentation 8-9 8.5.3 Personnel Monitoring Instrumentation 8-10 8.5.4 Counting Room Instrumentation 8-10 8.6 Radioactive Waste Handling and Disposal 8-10 8.6.1 Resin 8-10 8.6.2 Dry Active Waste (DAW) 8-10 8.6.3 Dismantlement (Metallic) 8-11 8.7 Records 8-11 8.8 Industrial Health and Safety 8-11

9.

SAFSTOR Accident Analysis 9-1 9.1 Introduction 9-1 9.2 Spent Fuel Handling Accident 9-1 9.3 Shipping Cask or Heavy Load Drop into FESW 9-3 9.4 Loss of FESW Cooling 9-4 9.5 FESW Pipe Break 9-5 9.6 Uncontrolled Waste Water Discharge 9-6 9.7 Loss of Offsite Power 9-6 9.8 Seismic Event 9-6 9.9 Wind and Tornado 9-7 9.10 References 9-8

10.

SAFSTOR Operator Traininz and Certification Program 10-1 10.1 Introduction 10-1 10.2 Applicability 10-1 10.3 Initial Certification 10-1 10.4 Proficiency Training and Testing 10-2 10.5 Certification 10-2 10.6 Physical Requirements 10-3 10.7 Documentation 10-3 D-PLAN 0-5 November 2010 1

TABLE OF CONTENTS 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 4.6 Figure 4.7 Figure 4.8 Figure 6.1 Figure 6.2 Table 3-1 Table 5-1 LIST OF FIGURES 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" Reactor Building Opening Reactor Building Bi-Parting Door Onsite ISFSI LACBWR Organization Chart LACBWR Schedule LIST OF TABLES Wind Direction Frequency Distribution at LACBWR Site and La Crosse NWS Spent Fuel Radioactivity Inventory Page No.

3-3 5-1 D-PLAN 0-6 November2010 I

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 chose 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 was submitted to outline DPC's intention to ultimately decommission the plant and site to radiologically releasable levels and terminate the license in accordance with Nuclear Regulatory Commission (NRC) requirements.

The preliminary DECON Plan was completed in December 1987 and there are currently no plans to revise the preliminary DECON Plan. This Decommissioning Plan addresses the issues contained in the preliminary DECON Plan. The License Termination Plan (LTP) for LACBWR will detail final dismantlement activities, processes for demolition of structures, site remediation, survey of residual contamination, and determination of site end-use. A final supplement to the Environmental Report in support of the LTP will address all environmental impacts of the license termination stage.

There are 333 spent fuel assemblies onsite. The plan at this time is to continue to store the fuel in the existing Fuel Element Storage Well while changes are made to place the spent fuel assemblies in dry cask storage containers and store the containers at the onsite Independent Spent Fuel Storage Installation (ISFSI). DPC currently expects the fuel to remain onsite until a federal repository, offsite interim storage facility, or licensed temporary monitored retrievable storage facility is established and ready to receive LACBWR fuel.

1.1 SELECTION OF SAFSTOR Effective August 28, 1996, the NRC's final decommissioning rule amended the regulations on decommissioning procedures. The rule clarified ambiguities in previous regulation, reduced unnecessary requirements, provided additional flexibility, and codified procedures and terminology that had been used on a case-by-case basis. The 1996 rule extended the use of the process described in 10 CFR 50.59, "Changes, Tests, and Experiments," to allow licensees to make changes to facilities undergoing decommissioning if determined that prior NRC approval was not required.

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

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

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

D-PLAN 1-1 November2010

1. INTRODUCTION - (cont'd)

ENTOMB is the alternative in which radioactive contaminants are encased in a structurally long-lived material, such as concrete. The entombed structure is appropriately maintained and continued surveillance is carried out until the radioactivity decays to a level permitting unrestricted release of the property. This alternative would be allowable for nuclear facilities contaminated with relatively short-lived radionuclides such that all contaminants would decay to levels permissible for unrestricted use within a period on the order of 100 years.

For a power reactor, the choice was either DECON or SAFSTOR. Due to some of the long-lived isotopes in the reactor vessel and internals, ENTOMB alone was not an allowable alternative under the original proposed rule.

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

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

There were other reasons to choose the SAFSTOR alternative. The majority of piping radioactive contamination was Co-60 (5.27 yr half-life) and Fe-55 (2.7 yr half-life). If the plant was placed in SAFSTOR for 50 years, essentially all the Co-60 and Fe-55 would have decayed to stable elements. Less waste volume would be generated and radiation doses to personnel performing the decontamination and dismantling activities would be significantly lower.

Therefore, delayed dismantling supported the ALARA (As Low As Reasonably Achievable) goal. The reduction in dismantling dose would exceed the dose the monitoring crew received during the SAFSTOR period.

The shutdown of LACBWR occurred before the full funding for DECON was acquired. The SAFSTOR period has permitted the accumulation of the full DECON funding. SAFSTOR funding and decommissioning cost financing are discussed in Section 6.7. The majority of studies showed that while the total cost of SAFSTOR with delayed DECON was greater than immediate DECON, the present value was less for the SAFSTOR with delayed DECON option.

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

A second disadvantage of delaying the final decommissioning was that the people who operated the plant would not be available for the DECON period. When immediate DECON is selected, D-PLAN 1-2 November 2010

1. INTRODUCTION - (cont'd) some of the experienced plant staff would be available for dismantlement. Their knowledge of plant characteristics and events would be extremely helpful. In the absence of these knowledgeable people, all information would have to be obtained from plant records. When SAFSTOR is chosen, efforts must be made to maintain excellent records to compensate for the lack of staff continuity.

The remaining factor to be discussed is safety. As of October 2009, 24 power reactors have been shut down in the United States, 11 of which have been fully dismantled and decommissioned.

Experience has shown that the process can be performed safely.

The NRC issued its Waste Confidence Decision on August 31, 1984, which is codified as amended in 10 CFR 51.23. 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 spent fuel at LACBWR, until a federal repository, interim storage facility, or licensed temporary monitored retrievable storage facility is ready to accept the fuel, is acceptable from the safety 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

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

1.2.2 Nuclear Regulatory Commission, Waste Confidence Decision, Federal Register, Vol. 49, No. 171, August 31, 1984.

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

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

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

1.2.5 Regulatory Guide 1.184, "Decommissioning of Nuclear Power Reactors," July 2000.

D-PLAN 1-3 November 2010 1

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( 11-1)-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(l 1-1)-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.

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

D-PLAN 2-1 November 2010 1

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

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 was 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 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%.

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.

D-PLAN 2-2 November 2010

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

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 included 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 spent fuel assemblies being stored in the LACBWR Fuel Element Storage Well (FESW).

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 FESW 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 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 FESW will be emptied of all spent fuel assemblies and fuel debris which will be placed in dry cask storage containers and moved to storage at the ISFSI.

2.5.2 Fuel Element Storage Well Leakage The stainless steel liner in the 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 D-PLAN 2-3 November 2010

2. LA CROSSE BOILING WATER REACTOR OPERATING HISTORY - (cont'd) replaced, which reduced the leak rate to approximately I gph. FESW leakage has stabilized over the years to an average of approximately 21 gallons per day. 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. Following removal of all spent fuel assemblies and fuel debris for dry storage at the ISFSI, the storage racks will be removed and disposed of; the FESW will be drained and decontaminated to eliminate leakage.

2.6 REFERENCES

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

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

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

D-PLAN 2-4 November 2010 1

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 I 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 (G-3). The Independent Spent Fuel Storage Installation (ISFSI) is located south of the G-3 plant, on land which was previously used for an access road to an abandoned boat landing on the Mississippi River. The ISFSI site is about midway between the Mississippi River on the west and Highway 35 on the east and is between the two closed ash landfills of the Genoa site. Access to the new boat landing follows the property's east and south boundaries. 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 Area 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.

By letter dated May 8, 2008, in response to request by DPC, the NRC agreed that the geographical area included within the LACBWR Part 50 license is the entire 163.5 acres owned or otherwise controlled by DPC. Further, the NRC found that the site where the ISFSI is being constructed is part of the NRC-licensed site under License DPR-45.

The ISFSI is located 2,232 feet south-southwest of the Reactor Building center. The ISFSI will be surrounded by a protected area fence, an isolation zone fence, and vehicle barrier system. These protective barriers will be within the ISFSI Controlled Area Boundary fence established at the perimeter of the 38.9 acre ISFSI site (See Figure 3.4). The ISFSI Controlled Area Boundary is D-PLAN 3-1 November 2010

3. FACILITY SITE CHARACTERISTICS - (cont'd) established to limit dose to the public during normal operations and design basis accidents in accordance with the requirements of 10 CFR 72.104 and 10 CFR 72.106. The controlled area, as defined in 10 CFR 72.3, means the area immediately surrounding an ISFSI for which the licensee exercises authority over its use and within which ISFSI operations are performed. DPC shall likewise exclude access to the ISFSI Controlled Area Boundary if adverse radiological conditions require.

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 LACBWR site. The only industrial facility of any significant size is the G-3 plant 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 Santa Fe (BNSF) Railway line from Chicago, Illinois, to the west coast which crosses through the original exclusion area. The BNSF 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 original 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 V2 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.

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 from the National Weather Service (NWS), approximately 35 km (21.7 mi.) north of LACBWR.

D-PLAN 3-2 November 2010 1

3. FACILITY SITE CHARACTERISTICS - (cont'd) 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 -7.1 °C (19.2°F) in the three months of winter to 21.9'C (71.4 0F) in the summer months. A maximum temperature of 42.21C (108.00) was recorded in July 1936, with a minimum low of-41.7°C (-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.).

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 from 1982-1984 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 valley's channeling effect. It is suspected that the layout of the buildings on site reduces the frequency of winds observed from the southwest to west.

D-PLAN 3-3 November 2010

3. FACILITY SITE CHARACTERISTICS - (cont'd)

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 The stack wind directions demonstrate a higher percentage of winds coming from the S to N directions than the surface distribution, due to the better exposure of the stack sensor from those directions. It is obvious that winds from the eastern and western sectors (at either level) are very infrequent because of the bluffs approximately 305m (1,000 feet) east of the site and the interference of buildings on the site to the west. The similarity between topographical features of the site and the La Crosse NWS station are shown in the wind direction frequency distribution in Table 3-1. The major differences in the La Crosse distribution are due to the better exposure of the instrumentation at the airport, and LACBWR's proximity to the eastern bluff.

Monthly average temperature and wind speed data is presented for both the La Crosse and LACBWR sites in Figure 3.5. There is excellent correlation between the temperature records for both sites, with some small differences in the winter period. The differences for the months of December through February are most likely the result of a micrometeorological heat island effect at the LACBWR site. The warmer winter temperatures result from the influence that the large concrete buildings and roads have on local environment. Wind speeds for both the stack and the surface follow each other quite closely, with the stack wind speeds on the average approximately 65 percent higher than the surface wind speeds at the LACBWR site. La Crosse NWS wind speeds average 42 percent higher than site surface wind speeds. All locations also exhibit the typical case of higher wind speeds in the spring and fall than in the summer and winter months.

High wind speeds, in excess of 11.2 m/s (25 mph) are not prevalent at either the surface or stack locations, equaling 0.0 and 1.3 percent, respectively. Low wind speeds (Figure 3.6) are very prevalent at the surface site due to the sheltering effects of the nearby bluffs and buildings.

D-PLAN 3-4 November 2010 1

3. FACILITY SITE CHARACTERISTICS - (cont'd)

Overall, there is very good agreement between LACBWR and La Crosse NWS wind speed, direction, and temperature data. The minor differences that do exist are due to differences in river valley influences and sheltering effects.

3.4 HYDROLOGY 3.4.1 Hydrologic Description The LACBWR site is in the Mississippi River valley. In the vicinity of the site, the valley is deeply cut into highly dissected uplands. From La Crosse to Prairie du Chien, approximately 40 miles south, the valley varies between 2/2 and 42 miles in width. The valley walls rise sharply 500 to 600 feet from river level.

There is little or no agricultural use of the river valley floor which consists primarily of marshy land, islands between river channels and extensions of low lying flood plain cut by ponds, sloughs and meandering stream channels. Numerous short, steep-sided valleys that have been cut into the uplands by tributary streams intercept the main river valley. Both walls of the main channel are wooded. The flat upland areas and some of the tributary valleys are cultivated and grazed.

The main channel of the river varies greatly in width above and below the site. A series of dams are operated by the U.S. Army Corps of Engineers for navigational purposes. Above Dam No. 8 (about /4 mile north of the site) the river is nearly four miles wide. Below the site, the river is relatively narrow for a distance of 20 miles, then gradually widens as the river approaches Dam No. 9, 33 miles south of the site.

3.4.2 Drainage The site is on a filled-in area south of the original Genoa-I steam plant. Therefore, drainage at the site must be provided. There is allowance for runoff from the high valley walls to the east. The site is favorably located with respect to this runoff, however, because of two short valleys east of the bluffs bordering the site. One valley drains to the north and one to the south, so that only precipitation that falls on the bluff adjacent to the site and on a small portion of the upland area contributes to runoff directly across the site. This runoff is presently channeled along the highway and railroad to prevent interference with traffic. No problems of flash floods have occurred at the site.

3.4.3 Downstream Water Use For a distance of 40 miles downstream of the site, virtually all municipal water supplies for cities and towns along the river are obtained from ground water. On the basis of readily available published records, the nearest major city using the river water for direct human consumption is Davenport, Iowa, about 195 miles downstream. The nearest user of river water for industrial purposes, excluding the adjacent fossil plant, is the steam-power plant in Lansing, Iowa, about 15 miles downstream. River water is used at this plant for condenser cooling. There is no other known user of river water for industrial purposes between the reactor site and Prairie du Chien, 40 miles down-river.

D-PLAN 3-5 November 2010 1

3. FACILITY SITE CHARACTERISTICS - (cont'd) 3.4.4 Flooding and Probable Maximum Flood The flood profile at the LACBWR site has a return frequency (as described by the U.S. Army Corps of Engineers) of 635.2 feet above mean sea level (MSL) for a 50-year flood, 637.2 feet MSL for 100-year, and 640.0 for a 500-year. The site fill is at 639 feet. The NRC, during the Systematic Evaluation Program, determined that the maximum historic flood (1965) was 638.2 feet MSL. The standard project flood is 643.2 feet MSL and the probable maximum flood is 658 feet MSL. The period of record keeping for evaluation of flooding, in the region of the La Crosse plant, goes back to records kept by the United States Weather Bureau in La Crosse, Wisconsin, from approximately 1873 on. Site surface run-off flooding for a local probable maximum precipitation was determined to meet NRC criteria as the total run-off would be approximately 6.4 inches above grade and the equipment is protected to a level of approximately 1 foot or more above grade. This was in compliance with the applicable Regulatory Guide criteria based on a local run-off area of the 35-acre water shed to the east of the facility.

The ISFSI area is constructed to a grade elevation of 640.8 feet MSL with a top of pad elevation of 643.5 feet to raise the pad above the standard project flood elevation of 643.2 feet.

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 dam 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 dam 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 (USGS). 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 D-PLAN 3-6 November 2010 1

3. FACILITY SITE CHARACTERISTICS - (cont'd) 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.

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.

For flood conditions at the ISFSI, the NAC-MPC storage system is evaluated for a fully-immersing design basis flood with a water depth of 50 feet and a steady-state flow velocity of 15 feet per second. The analysis shows that the NAC-MPC storage system performance is not affected by the design basis flood, and demonstrates that the concrete cask will not slide and will not overturn in the design-basis flood. The hydrostatic pressure exerted by the 50 foot depth of water does not produce significant stress in the canister. The NAC-MPC design basis bounds the LACBWR site probable maximum flood elevation of 658 feet MSL with flow velocity of 1.91 feet per second.

3.4.7 Ultimate Heat Sink and Low Flow Conditions The ultimate heat sink of LACBWR 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 USGS 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 described in Section 9, "SAFSTOR Accident Analysis," substantial time is available for restoration of spent fuel cooling. If water is added to the FESW, any consequences of water heat D-PLAN 3-7 November 2010 1

3. FACILITY SITE CHARACTERISTICS - (cont'd) up can be delayed or prevented. With all spent fuel assemblies in dry storage at the ISFSI, spent fuel cooling is no longer dependent on river flow.

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 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.

3.5 GEOLOGY, SEISMOLOGY, AND GEOTECHNICAL ENGINEERING 3.5.1 Basic Geologic and Seismic Information LACBWR is located within the Wisconsin Driftless section of the Central Lowland physiographic province. The Wisconsin Driftless section was not glaciated during the Pleistocene Epoch and is characterized by flat lying naturally dissected sedimentary rocks of early Paleozoic age.

Moderate to strong relief has been produced on the unglaciated landscape which has been modified only slightly by a mantle of loess and glacial outwash in the larger valleys of the area.

Maximum relief in the region is about 1000 feet.

Bedrock in the site region consists of Pre-Cambrian crystalline rocks exposed at the crest of the Wisconsin Dome by early Paleozoic (Cambrian and Ordovician, 572 million years before present

[mybp] to 435 mybp) sedimentary strata. Basement rocks in the site vicinity are of granitic composition. The Paleozoic rocks are 1200-1300 feet thick in the site vicinity and consist of dolomites, sandstones and shales. About 600 feet of this sequence is exposed along the bluffs on both sides of the Mississippi River in the plant vicinity. Prior to the Pleistocene Epoch (more than 2 mybp) the river had carved a gorge as much as 150 to 210 feet deeper than can be seen today. It was buried by post-glacial sediment.

The site is located within the Central Stable Region tectonic province. The Central Stable Region consists of a vast area of large circular uplifts and sedimentary basins, and broad synclines and arches. Major structural features include the Wisconsin Dome and Arch, Lake Superior syncline, Forest City Basin, Michigan Basin, and Illinois Basin. These structures were formed during the Late Pre-Cambrian and Early Paleozoic (more than 435 mybp).

Major uplift and down-warping also occurred during Late Paleozoic (330 mybp to 240 mybp).

Some minor tilting occurred during and following the Pleistocene glaciation (2 mybp to 10,000 ybp). The site is located on the southwest flank of the Wisconsin Dome and the western flank of the Wisconsin Arch, a southward projection of the Wisconsin Dome. For this reason, sedimentary strata in the site vicinity dip less than 20 feet per mile to the southwest.

Many faults have been mapped in the site region. None of these faults are considered to be capable according to 10 CFR 100, Appendix A. These faults are discussed in Section 3.5.2. A "capable fault" is a fault which has exhibited one or more of the following characteristics:

D-PLAN 3-8 November 2010

3. FACILITY SITE CHARACTERISTICS - (cont'd)

1) Movement at or near the ground surface at least once within the past 35,000 years or movement of a recurring nature within the past 500,000 years.

2) Macro-seismicity instrumentally determined with records of sufficient precision to demonstrate a direct relationship with the fault.

3) A structural relationship to a capable fault according to characteristics (1) or (2) of this paragraph such that movement on one could be reasonably expected to be accompanied by movement on the other.

The site lies on the east bank of the Mississippi River. Local drainage has dissected the upland areas into a pinnate pattern. The site is north of one of the drainages (Italian Hollow) which opens into the Mississippi Valley perpendicularly from the east. The Mississippi Valley is broad (2.6 miles at the site) and is bounded on both sides by vertical bluffs several hundred feet high composed of relatively flat lying early Paleozoic strata (570 mybp to 410 mybp).

The plant is founded on about 15 to 20 feet of hydraulic fill on the river flood plain. Beneath the fill are from 115 to 135 feet of fine sand that was deposited as alluvium and glacial outwash. The bedrock below these glacial fluvial deposits is sandstone of the Dresbachian Group of Cambrian age (570 mybp to 500 mybp).

3.5.2 Proximity of Capable Tectonic Structures in the Plant Site Vicinity The NRC has been sponsoring research programs since the early 1970's in an attempt to determine if there are bases for delineating seismotectonic provinces and earthquake source structures in the eastern and central United States. Much new geologic and seismic information has been developed as a direct result of these studies, but positive evidence substantiating province boundaries and specific earthquake generating tectonic structures has not been found.

The most successful results have been attained in the Mississippi Embayment region where a reactivated Precambrian rift zone (Reelfoot Rift) is indicated to be responsible for the relatively high seismicity there. Some success has also been achieved with respect to the Central Stable Region (site province) in developing seismic and geologic evidence that suggests the presence of seismic source zones. The study, based on an independent examination of large geologic structures that appear to be spatially related to seismicity, proposes ten seismic source zones.

These zones are related to either basement rifts or to major uplifts and basins. The nearest of these major zones to LACBWR is the Cincinnati Arch, the northwest segment of which (Kankakee Arch) trends to within 200 miles southeast of the site. There is no geologic evidence, however, that any of these major tectonic structures (except for the Reelfoot Rift, also called the New Madrid Fault zone) are capable or that they are associated with capable structures.

Numerous faults have been mapped in the site region. Many investigations have been carried out concerning these faults by the state geological surveys, oil companies, mining companies, and consultants to utilities and agencies constructing nuclear facilities. Geologic history interpreted from all of these studies indicates that the last major tectonic activity took place during the period between post-Pennsylvanian (290 mybp) and pre-Cretaceous (138 mybp). The general absence of middle to late Paleozoic (410 mybp to 330 mybp), Mesozoic (240 mybp to 36 mybp) and Cenozoic (63 mybp to 2 mybp) strata make it extremely difficult to pinpoint the age of the last movement along these faults. However, Pleistocene (younger than 2 mybp) deposits are D-PLAN 3-9 November 2010 1

3. FACILITY SITE CHARACTERISTICS - (cont'd) relatively common in the region except in the Wisconsin Driftless Area. Investigations in the region have not found evidence that Pleistocene deposits are offset nor that accumulations of Pleistocene deposits are thicker on the down thrown side of faults than on the up thrown side, even though such characteristics have been extensively searched for. Additionally, drainage patterns in the region that trend across mapped faults show no offset along the fault traces.

Therefore, the faults are at least pre-Pleistocene and not capable according to Appendix A of 10 CFR 100.

The following is a brief discussion of each of the more significant faults in the site region within a radius of 200 miles.

The closest mapped fault of any size to the LACBWR site is the Mifflin fault, which is located on the northeast flank of an anticline, the Mineral Point anticline. It is located in Iowa and La Fayette Counties, Wisconsin, about 45 miles southeast of the site. The fault strikes N 400 W for about 10 miles and is offset with the northeast side down at least 65 feet. About 1000 feet of strike-slip displacement has also been indicated. Last movement on the fault is believed to have occurred in Late Paleozoic (330 mybp to 240 mybp).

The Sandwich fault is a major regional fault in northern Illinois about 150 miles southwest of the site. The Sandwich fault is an 85-mile long vertical fault that strikes northwest-southeast, and is down to the northeast a maximum of 900 feet. A subsidiary fault is present near the north end of the Sandwich fault with 150 feet displacement down to the south, forming a graben between the two faults. The nearest age of last movement that can be determined is post-Silurian and pre-Pleistocene as no intervening age rocks are present in association with the fault. During subsurface investigations for an expansion of the General Electric Company's Fuel Recovery Operation near the Dresden nuclear site, a complex fault zone was found. This fault zone was considered to be related to the same tectonic events that caused deformation on the Sandwich fault zone. These investigations showed that the fault zone at the GE facility did not offset the Pennsylvanian Spoon formation, thus demonstrating that last movement occurred more than 280 million years ago.

The Plum River fault zone (formerly the Savanna fault) is located 120 miles south-southeast of the site in northwestern Illinois and eastern Iowa. It strikes east-west and is due west of the northern end of the Sandwich fault. The fault zone consists of a series of echelon faults with south sides up from 100 to 400 feet. This fault zone is associated with the Savanna anticline, a major fold in the region. Last movement on this fault zone took place between post-middle Silurian (425 mybp) and pre-middle Illinoian (700 thousand years bp).

The Madison fault and the Janesville fault are about 50 miles southeast of the site. They trend parallel in a general east-west strike with the Madison fault being the northernmost. The north side of both faults is down relative to the south side. The Janesville is the most well known of the two and has been mapped in the subsurface for a distance of 75 miles. It is composed of two branches, the predominant east-west one, and another that strikes northeast-southwest. Both faults are interpreted to have last moved sometime between post-S ilurian and pre-Cretaceous (410 mybp to 138 mybp).

O The Appleton and Green Bay faults are located 180 and 130 miles northeast of the site. Both faults have been postulated based on abrupt changes in elevation (south sides down) on the D-PLAN 3-10 November 2010 1

3. FACILITY SITE CHARACTERISTICS - (cont'd)

Precambrian basement. The faults are dated as post-Silurian to pre-Cretaceous (410 mybp to 138 mybp).

Several faults are mapped in the Precambrian basement in Minnesota and northern Wisconsin based on geophysical evidence. These faults, in general, are oriented in a northeast to north-northeast direction. Two of these faults are the Douglas and Lake Owen faults which bound the north and south flanks of the Lake Superior syncline respectively. A southwest extension of the Lake Owen fault is the Hastings fault which approaches to about 110 miles from the site. Due to the lack of post-Precambrian rocks over these faults, it is impossible to determine an upper limit of last movement. They probably experienced activity during late Precambrian (600 mybp) and throughout the Paleozoic (570 mybp to 240 mybp). There is no evidence that they are capable.

Faulting has been identified in the Chicago area. A fault zone is inferred in the basement rocks from gravity and seismic data, north of, and parallel to, the Sandwich fault zone. The south side is down relative to the north side. It is not known whether or not the fault extends into the overlying Paleozoic rocks. Twenty-five minor faults are mapped in the Chicago area based on seismic data. These faults are dated as post-Silurian (410 mybp) and pre-Pleistocene (2 mybp).

There is no surface evidence for these faults, but from the seismic data, displacements of up to 55 feet are recognized. The fault zone which trends west-northwest consists of blocks that have been down-dropped both to the north and to the south within the zone.

The LaSalle anticlinal belt trends along the eastern flank of the Illinois Basin to within 200 miles of the site. Faults have been postulated on the west flank of the LaSalle anticlinal belt, the Oglesby and Tuscola faults. There is no direct evidence for faulting but they are based on the presence of changes in dip of the rock strata and as much as 1200 feet of stratigraphic difference on the west side of the anticline. Movement on these faults, if they exist, is interpreted to be pre-Cretaceous (more than 138 mybp).

The nearest region containing possible known capable faults is the Mississippi Embayment or New Madrid fault systems which is about 370 miles from the site. Extensive investigations in this area by the USGS, local universities, and state geological surveys, funded in part by the NRC, have indicated that recent faulting and earthquake activity are related to the reactivation of a north-south striking Precambrian and Paleozoic rift zone. Although seismicity of the New Madrid fault system must be considered in the seismic analysis of the LACBWR plant, it is not significant, by virtue of its distance from the site, in a consideration of potential surface faulting at the site.

Little is known about faulting in the rock beneath the site, but no faults have been mapped in the 400 to 600 feet high bluffs immediately east of the site. It is possible that minor faults mapped in the lead-zinc mining district southeast of the site are representative of faulting in the immediate site vicinity, although the lead-zinc district is in an area of relatively strong tectonic deformation.

It is possible that minor faults are present in rock beneath the site, but based on low seismicity, a lack of any indication of fault displacements in outcrops in the area, and the lack of evidence for recent fault displacement in the region, it is concluded that faults beneath the site, if they exist, are not capable.

D-PLAN 3-11 November 2010 I

3. FACILITY SITE CHARACTERISTICS - (cont'd) 3.5.2.1 Conclusions of the NRC "There are no geologic conditions in the site vicinity that represent hazards to the facility.

Numerous faults are mapped in the site region, but investigations of all of these faults during the course of validating several nuclear power plant sites in the region, in addition to studies for the LACBWR, have not found any evidence of capable faulting. Additionally, the area is one of relatively low seismicity. Therefore, capable faulting does not need to be considered in the analysis of this site."

3.5.3 Surface Faulting at the Site LACBWR is located one mile south of Genoa, Vernon County, Wisconsin, on the east bank of the Mississippi River. The LACBWR facilities are situated on about 15 feet of hydraulic fill which overlies approximately 100-130 feet of glacial outwash and fluvial deposits. Due to the absence of bedrock exposures at the plant site, the geologic investigation was restricted to an examination of the rock-bluffs in the site vicinity. The following observations were compiled from several vantage points:

" Bluffs on both sides of the Mississippi River are heavily vegetated;

" Scattered outcrops are visible;

" Bluff tops appear to be sub-horizontal at a relatively constant elevation;

" Valleys on one side of the Mississippi River do not appear to be linearly continuous with valleys on the opposite side;

" Several widely spaced joints are visible;

" A lithologic contact (bluff to light gray sandy dolomite to dolomitic sandstone overlying yellow-brown sandstone with gray siltstone interbeds) could be discontinuously observed in the bluff face immediately east of LACBWR. This contact appeared to be sub-horizontal and was not observed.to be off-set by folding or faulting; and

" No closely-spaced joints, shear zones, or faults were observed in the bluffs east of LACBWR or in either valley immediately north and south of the plant.

3.5.3.1 Conclusions of the NRC Based upon the preceding observations, there is no apparent evidence to indicate that faulting has affected the Late Cambrian-Ordovician age rocks exposed in the bluffs east of LACBWR. It is therefore apparent, based upon the available evidence, that there are no faults in the vicinity of the LACBWR site that have the potential to represent a seismic hazard to site safety.

3.5.4 Stability of Slopes and Properties of Subsurface Materials 3.5.4.1 Properties of Subsurface Materials. The initial soil investigations at the LACBWR site were conducted in 1962. Between 1962 and 1980 soil test borings were made at 36 locations in the site vicinity, Of this number, five were associated with subsurface investigations in the power D-PLAN 3-12 November 2010 1

3. FACILITY SITE CHARACTERISTICS - (cont'd) station area, four were associated with the switchyard area, and one was drilled to locate an offsite borrow area for construction fill materials. The remaining 23 were associated with subsurface investigations in the main plant facility area. DPC has boring logs depicting the soil conditions encountered in these investigations. Field investigation effort included standard penetration tests (SPT) and split-barrel sampling in accordance with ASTM D-1586-67 procedures. Relatively undisturbed samples were also obtained at several locations in thin-walled tubes using an Osterberg piston sampler. Laboratory testing of soil samples was accomplished to determine index properties and to establish soil strength parameters. Testing included specific gravity determinations in accordance with ASTM D-854-58, particle size analysis testing in accordance with ASTM D-422-63, relative density determinations in accordance with ASTM D-2049-69 and cyclic triaxial testing in accordance with the procedures of NUREG-003 1.

3.5.4.2 Plant Facilities. The reactor containment structure, turbine building, diesel generator building, stack, waste disposal building and the gas vault are supported on cast-in-place concrete piles driven to develop a 50-ton capacity and filled with concrete specified to have a minimum 28-day compressive strength of 3500 psi. Using the data presented, the NRC staff independently estimated the bearing capacity of the in-place piles. Results indicate that the piles can be expected to safely carry a loading of greater than 50 tons per pile without significant settlement under static loading conditions. The cribhouse and associated water intake and discharge piping are not designated as Seismic Category I structures.

3.5.4.3 Slope Stability. Review of available onsite and offsite topographic data indicates there are no onsite slopes whose failure could cause radiological consequences adversely affecting the public health and safety. One offsite slope, the east bank of the Mississippi River adjacent to the plant cribhouse site, was identified from topographic data and evaluated for safety in conjunction with this topic. A generalized typical section for this slope is presented in Figure 3.7.

A stability analysis of this slope was performed using conservative solid strength parameters developed from the results of the site subsurface investigation. In the analysis an angle of internal friction of 270 was assigned each soil layer and the slope stabilizing influence of the slope protection riprap material was conservatively ignored. Results of the analysis indicate the factor of safety against failure under static loading conditions is greater than 1.5. The NRC concluded that an adequate margin of safety exists under static loading conditions.

Due to the fact that the cribhouse and the associated water intake and discharge piping are the only plant structures in the vicinity which will potentially be affected by a postulated failure of the east river bank slope, and these structures are not designated as Seismic Category I structures, a dynamic (pseudostatic) slope stability analysis is not appropriate and was not performed.

3.5.4.4 Conclusions of the NRC Based on the review of available site data and on information obtained during an NRC staff visit to the site, it was concluded that the stability of slopes associated with the LACBWR site does not pose a safety concern for this plant.

3.5.5 Stability of Slopes and Properties of Subsurface Materials 3.5.5.1 Liquefaction and Seismic Settlement. The safe shutdown earthquake (SSE) peak ground acceleration postulated for the La Crosse site is 0.11 g with an equivalent duration (NEQ) of 5 D-PLAN 3-13 November2010 I

3. FACILITY SITE CHARACTERISTICS - (cont'd) cycles. Results of standard penetration tests (SPT) undertaken by DPC in 1980 show a range in N-values in clean sand below the water table beneath the turbine building of from 12-34 blows/ft.

SPT N-values taken beneath the stack ranged from 23 to over 50 blows/ft. Based on the NRC staffs review of the site foundation conditions, the borings under the turbine and stack foundations are considered representative for other adjacent structures that are pile supported including the reactor containment building.

Results of an NRC staff safety evaluation concerning liquefaction potential at the LACBWR site were reported in August 1980. Based upon an evaluation of information provided by DPC, the staff concluded in that report that the materials under the existing turbine building, stack, and the reactor containment structure are adequately safe against liquefaction effects for an earthquake up to a magnitude 5.5 with a peak ground acceleration of 0.12g.

Based upon the information presented by DPC that all plant Seismic Category I structures are supported upon piles and the results of the NRC staffs previous studies, the staff concluded that induced settlements of Seismic Category I structures would not be significant under the postulated dynamic conditions.

3.5.5.2 Turbine Building Floor Support Grouting. In July 1980, borings drilled under the turbine building (Borings DM-12 and DM-13) encountered voids at several locations beneath the "on grade" concrete floor slab. In order to identify the lateral and vertical extent of the voids and to investigate potential for voids under other safety-related structures, DPC accomplished an exploratory drilling program at the site. Results of the program indicated voids ranging in depths up to 10 in. existed only within the turbine building area. Although voids of this relatively small size would not significantly affect lateral support to the 310 fifty to seventy ft. long piles supporting the turbine building or the integrity of the overall turbine building pile foundation under dynamic loading conditions, DPC accomplished an injection grouting program to fill the voids and restore continuous "on-grade" support to the turbine building concrete floor. About 460 cu. ft. of grout was injected under a floor area of approximately 10,000 sq. ft. The grouting program was completed on October 28, 1980.

3.5.5.3 Conclusions of the NRC "Based on the review of licensee's submittal and of other available referenced data, the NRC staff concurs with DPC's conclusion that the pile supported structures, systems and components are not expected to experience excessive settlements under static or dynamic conditions."

3.6 ISFSI SOILS AND SEISMOLOGY The soil parameters were analyzed from borings made in the area of the ISFSI. A site dimension of 122 feet by 148 feet, including the 32 feet by 48 feet ISFSI pad area, was improved by vibrocompaction to a depth of 35 feet. The 11 feet of in-situ soils above elevation 621 feet were removed. Drain rock was placed and compacted in 8-inch lifts to an elevation of 625 feet at which point a layer of geotextile fabric was installed. Imported structural backfill was then placed in controlled compacted lifts to raise the improved area to a final elevation of 640.8 feet. All soils work was completed per project earthwork specifications.

The liquefaction analysis for the post-improvement site soils used the cone penetration testing (CPT) based method and considered a design groundwater elevation of 639 feet and a final grade D-PLAN 3-14 November 2010 I

3. FACILITY SITE CHARACTERISTICS - (cont'd) elevation of 640.8 feet with and without ISFSI pad loading. The results of the liquefaction evaluation indicated that factors of safety for all sandy soil layers below elevation 621 feet were acceptable. The soils improvement above elevation 621 feet resulted in all soil densities in the improved area to be considered non-liquefiable.

The soil structure interaction analysis of the ISFSI site soils improvement resulted in acceptable ISFSI pad accelerations of 0.402g horizontal and 0.18g vertical. For the center of gravity of the stored Vertical Concrete Cask (VCC) the resulting accelerations were 0.442g horizontal and 0.1 8g vertical. The NAC-MPC Final Safety Analysis Report (FSAR) acceptance criteria for the VCC are 0.45g horizontal and 0.30g vertical accelerations.

3.7 REFERENCES

3.7.1 U. S. Department of Commerce, "Local Climatological Data with Comparative Data," La Crosse, Wisconsin, 1981.

3.7.2 Thornbury, W. D., 1965, Regional Geomorphology of the United States, John Wiley and Sons, Inc., New York.

3.7.3 Dames and Moore, 1979, Review of Liquefaction Potential, La Crosse Boiling Water Reactor (LACBWR), near Genoa, Vernon County, Wisconsin; prepared for Dairyland Power Cooperative, March 12, 1979.

3.7.4 Eardley, A. J., 1973, Tectonic Division of North America, In Gravity and Tectonics, edited by DeJong and R. Scholten, John Wiley Publishing Co.

3.7.5 King, P. B., 1969, Discussion to Accompany the Tectonic Map of North America, USGS-Department of the Interior Publication.

3.7.6 Barstow, N. L., K. G. Brill, 0. W. Nuttli, and P. W. Pormerey, 1981, An Approach to Seismic Zonation for Siting Nuclear Electric Power Generating Facilities in the Eastern United States; prepared for USNRC, NUREG/CR-1577, May 1981.

3.7.7 Heyl, A. V., A. F. Agnew, E. J. Lyons, and C. H. Behre, Jr., 1959, The Geology of the Upper Mississippi Valley, Zinc-Lead District; Professional Paper No. U. S. Geol. Survey.

3.7.8 Templeton, J. S., and H. B. Willman, 1952, Central Northern Illinois Guidebook for the 16th Annual Field Conference of the Tri-State Geological Society; Guidebook Series 2, Illinois State Geological Survey.

3.7.9 U.S. Nuclear Regulatory Commission, 1978, Safety Evaluation Report for General Electric Fuel Storage Capacity Expansion - Morris, Illinois; Geology and Seismology input transmitted by December 19, 1978, memo to L. C. Rouse from J. C. Stepp.

3.7.10 Kolata, D. P. and T. C. Buschbach, 1976, Plum River Fault Zone of Northwestern Illinois; Circular 491, Illinois State Geological Survey.

3.7.11 Thwaites, F. T., 1957, Map of Buried Precambrian of Wisconsin, Wisconsin Geol. Survey (Rev. from 1931).

D-PLAN 3-15 November 2010

3. FACILITY SITE CHARACTERISTICS - (cont'd) 3.7.12 McGinnis, L. D., 1966, Crustal Tectonics and Precambrian Basement in Northeastern Illinois, Illinois State Geological Survey, Report of Investigation 219, 29 pp.

3.7.13 Buschbach, T. C., and G. E. Heim, 1972, Preliminary Geological Investigations of Rock Tunnel Sites for Flood and Pollution Control in the Greater Chicago Area, Environ. Sed. Notes, No. 52, Illinois State Geological Survey, 1972.

3.7.14 Green, D. A., 1957, Trenton Structure in Ohio, Indiana, and Northern Illinois; Bull. Am.

Assoc. Petroleum Geol., 41, pp. 627-642.

3.7.15 NAC International, Inc., NAC Multi-Purpose Canister Final Safety Analysis Report, Revision 7, Section 11.A.2.6.

3.7.16 Sargent & Lundy Report No. SL-010167, ISFSI Soil Remediation Summary.

3.7.17 NRC Letter, Banovac to Berg, dated May 8, 2008.

D-PLAN 3-16 November2010 I

3. FACILITY SITE CHARACTERISTICS - (cont'd)

General Site Location Map FIGURE 3.1 D-PLAN 3-17 November 2010 1

3. FACILITY SITE CHARACTERISTICS - (cont'd) 62E t{HW HH3. WNW w

26 WSW EHE E*

ESE rotal Segmeant Popualation i.

0 to 5 Mile 6M-POPULATION TOTALS RINGTOTAL MILES CUM4ULAMlV1 INIES OPULArTION

-P~.

t*S OPULATION 0-2 07 -1 471 2-s

,-S I

o71 Estimated Permanent Population (6/8-2)

Is MN = 148 (13.R2%)

WI = 923 (86.18%)

IA = 0 Population Dispersion FIGURE 3.2 D-PLAN 3-18 November2010 1

3. FACILITY SITE CHARACTERISTICS - (cont'd)

Effluent Release Boundary FIGURE 3.3 3-19 November 2010 D-PLAN

3. FACILITY SITE CHARACTERISTICS - (cont'd)

I Genoa Site Map FIGURE 3.4 D-PLAN 3-20 November 2010 1

3. FACILITY SITE CHARACTERISTICS - (cont'd)

MONTHLY AVERAGE TEMPERATURE DATA LACBWR and LACROSSE NWS SITES 26 1982 -

1984 24 22 20."

18 16 14 12 to Doegrees 8-Celsius 64 2

-G-LACBWR Surface 0 X LaCrosse NWS

-6

-10

-12 JAN FEB MAR APR MAY JUN JdL AUG SEP OCT NOV DEC 20-MONTHLY AVERAGE WIND SPEED DATA 19 LACBWR and LACROSSE NWS SITES

18.

1982 -

1984 17-16 154

.& Surface Stock LaCrosse NWS 14-13-Wind 11 Speed t09 (mph) 9 4

3-JAN FEB MAR APR MAY JUN 9iL AUG SEP 0CT NOV DEC Monthly Average Meteorological Data FIGURE 3.5 D-PLAN 3-21 November 2010

3. FACILITY SITE CHARACTERISTICS - (cont'd) 12.

LACBWR SURFACE -

IOM WIND SPEED FREQUENCY DISTRIBUTION

1982 -

1984

('

Frequency of OccurrenCe (7.)

Frequency of Occurrence

-t0 9

7 6

5 4

3 2

i 0

0 t2 it to.

9 8

7 6

5 4

3 2

91 AD 16 5il 20 22 24 26i 28 30 32 34 36 3 @40 2

4 5 5 10 12 14 LACBWR STACK -lOOM WIND SPEED FREQUENCY DISTRIBUTION 1982 -

1984 2 4 6 8 10 12 14 16 18 20 22 2425 26'i8'303i23i4"368'i Hourly Wind Speeds (mph)

Wind Speed Freauencv Distribution FIGURE 3.6 D-PLAN 3-22 November 2010 1

3. FACILITY SITE CHARACTERISTICS - (cont'd)

't, "*

irse S O"t 16 Mississippi River W..........~

Gray Miedu Send 7

,ý,

East Bank River Slope FIGURE 3.7 D-PLAN 3-23 November 2010

3. FACILITY SITE CHARACTERISTICS - (cont'd)

AVERAGE MSL ELEV.

+639

/

/

/

+619 -<

+614 y-

//

<.599 -4 Depth Ft.

Plant Grade 0

• 0 20 25 40

-130

+509 Generalized Soil Profile FIGURE 3.8 D-PLAN 3-24 November 2010

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.

An Independent Spent Fuel Storage Installation for dry storage of the LACBWR spent fuel inventory is being established on the Genoa site south of the G-3 coal pile.

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 feet and an inside diameter of 60 feet, and it extends 26'-6" below grade level. The shell thickness is 1.16 inch, except for the upper hemispherical dome which is 0.60 inch 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 (FESW), 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 the FESW cooling system.

The Reactor Building was designed to withstand the instantaneous release of all the energy of the primary system to the Reactor Building atmosphere at an initial ambient temperature of 800 F, neglecting the heat losses from the building and heat absorption by internal structures. The design pressure was 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 was 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 2010

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'-10" 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 feet thick. This support runs from the bottom of the semi-ellipsoidal head at about elevation 612'-4" to an elevation of 621 '-6". The 232 piles that support the containment structure are driven deep enough to support over 50 tons per pile.

The containment bottom head above elevation 621 '-6" and the shell cylinder from the bottom head to approximately 9 inches above grade elevation 639 feet are enveloped by reinforced concrete laid over a '/2 inch thickness of pre-molded expansion joint filler. The reinforced concrete consists of a lower ring, mating with the pile capping concrete. The ring is approximately 41 feet thick at its bottom and 22 feet thick at a point 1 /2 feet below the top due to inner surface concavity. The ring then tapers externally to a thickness of 9 inches at the top (elevation 627'-6") and extends up the wall of the shell cylinder to elevation 639'-9". The filler and concrete are not used, however, where cavities containing piping and process equipment are immediately adjacent to the shell.

Except for areas of the shell adjacent to other enclosures, the exterior surface of the shell above elevation 639'-9" is covered with 1 V2 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 inch thick. The insulation of the vertical walls is Johns-Manville Spintex of 6 lb/ ft3 density, faced with corrugated embossed aluminum sheet approximately 0.0 16 inch 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'-6" long between its two rectangular doors that measure 5'-6" by 7'. The Reactor Building can also be exited, if necessary, through the emergency airlock, which is 7 feet long and 5 feet in diameter, with two circular doors of 32V2 inch diameter (with a 30-inch opening). Both airlocks are at elevation 642'-9" and lead to platform structures from which descent to grade level can be made.

There is an 8 feet by 10 feet freight door opening in the Reactor Building that was intended to accommodate large pieces of equipment. The door is bolted internally to the door frame in the shell.

To facilitate reactor pressure vessel removal and dry cask storage, an opening was created in the Reactor Building. The opening has a total length of 58'-8". The width of the upper 24'-8" of the opening is 16'-9/4" and the width of the lower 34' of the opening is 10'-6". The opening is closed by a weather tight, insulated, roll-up, bi-parting door. The opening and door are depicted in Figures 4.6 and 4.7.

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 1 to 10 feet below grade level and enter either at the northwest quadrant into the pipe tunnel that runs to the Turbine Building, or on the D-PLAN 4-2 November 2010 1

4. FACILITY DESCRIPTION - (cont'd) northeast side into the tunnel connecting the Turbine Building, Reactor Building, stack, and the water treatment and waste gas storage areas.

A 45,000-gallon 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 a source of water inventory for fuel handling operations and the FESW.

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 wet in racks in the bottom of the FESW 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 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 B4C Polymer 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 FESW 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 FESW floor by the rack support feet. The fuel storage racks and associated seismic bracing are fabricated from Type 304 stainless steel. Following removal of all spent fuel assemblies and fuel debris for dry storage at the ISFSI, the storage racks will be removed and disposed of; the FESW will be drained and decontaminated.

The Reactor Building is being modified to facilitate movement of spent fuel assemblies from the FESW to the NAC-MPC System Transportable Storage Canister (TSC). The TSC is located inside the Transfer Cask (TFR) during fuel loading operations. The TFR/TSC assemblage will reside in the cask pool within the water-filled upper cavity during spent fuel assembly transfer from the FESW to the TSC.

The 10'-6" wide opening from the mezzanine floor elevation 667 feet to the fuel handling floor elevation 701 feet was previously created in the northern section of the upper cavity bio-shield and liner to permit removal of the LACBWR reactor pressure vessel. This opening will be used to facilitate movement of the TFR/TSC between the cask pool and the mezzanine floor to the north where TSC preparation operations will take place (e.g., welding, drying, etc.). The Reactor Building mezzanine floor will be reinforced in that location by adding steel struts beneath a cantilevered section of the floor. In order to provide sufficient water coverage over the spent fuel assemblies during movement into the TSC from the FESW, a water-tight removable gate, 16'-9" high by 9'-4" wide, will be installed in the bio-shield opening above approximate elevation 679'-3" extending to elevation 696 feet. The cask pool gate'will be supported by a 12' high structure installed at elevation 667 feet. The cask pool gate is designed with inflatable D-PLAN 4-3 November 2010 1

4. FACILITY DESCRIPTION - (cont'd) pneumatic seals having a defined acceptable leakage rate. Appropriate interfacing modifications to the bio-shield liner at the edges of the opening will be installed to ensure water retention in the area between the upper cavity liner and the cask pool gate. The cask pool gate storage stand will support the 6-ton cask pool gate when not in use.

The 10' high by 10' inner diameter cask pool will be installed at elevation 669'-3" atop a 20'-

10" high support structure attached to the reactor support cylinder at elevation 648'-5". The cask pool will have a 16V2" wide horizontal flange welded to the top of the shell, the outer circumference of which will be tied into the existing upper cavity liner using L-shaped stainless steel angle at approximate elevation 679'-3". This arrangement will provide a barrier to prevent water in the upper cavity area above the cask pool from leaking around the outside of the cask pool into the cavity below.

The upper cavity liner and bio-shield contained a number of penetrations from reactor operation that will be sealed using steel plates welded to the upper cavity liner to maintain the pressure boundary of the upper cavity. The cask pool will also include two penetrations, one at the bottom and one on the side approximately 7'-6" above the tank bottom. These penetrations will connect to piping and valves for filling, draining, and processing water from the pool and upper cavity to permit removal of the cask pool gate, and to perform other activities such as water clean-up, cask annulus flushing, and inventory control. Each 2-inch diameter pipe will include two manual valves in series just outboard of the cask pool to ensure redundant isolation.

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 equipment 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.

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.

D-PLAN 4-4 November 2010 I

4. FACILITY DESCRIPTION - (cont'd) 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 WTB 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 filters, 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 WTB 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 Low Specific Activity (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.

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.

4.2.5 Onsite Independent Spent Fuel Storage Installation The LACBWR Dry Cask Storage Project establishes an Independent Spent Fuel Storage Installation (ISFSI) under general license provisions of 10 CFR 72, Subpart K on the Genoa site.

The ISFSI site is about midway between the Mississippi River on the west and Highway 35 on the east. The ISFSI is located 2,232 feet south-southwest of the Reactor Building center on land D-PLAN 4-5 November 2010 1

4. FACILITY DESCRIPTION - (cont'd) which was previously used for an access road between the two closed ash landfills of the Genoa site.

The ISFSI will be used for interim storage of LACBWR spent fuel assemblies in the NAC International, Inc. Multi-Purpose Canister (NAC-MPC) System. The NRC issued 10 CFR 72 Certificate of Compliance (CoC) No. 1025 which confers approval of the NAC-MPC storage system design. The design basis for the NAC-MPC System is provided in the NAC-MPC Final Safety Analysis Report (FSAR). The NRC approved Amendment 6 to the NAC-MPC CoC and Technical Specifications to incorporate LACBWR spent fuel assemblies as approved contents for storage in the NAC-MPC System. The effective date of the license amendment was October 4,2010.

The NAC-MPC System is comprised of the Transportable Storage Canister (TSC), the Vertical Concrete Cask (VCC), and the Transfer Cask (TFR). The TSC is designed to be transported in the NAC Transport Cask (STC) licensed by the NRC pursuant to 10 CFR 71 CoC No. 71-9235 for shipment of the NAC-MPC canister. The NAC-MPC System designed for and to be used at LACBWR is designated MPC-LACBWR. The LACBWR spent fuel inventory will be placed into five MPC-LACBWR dry storage casks. Each MPC-LACBWR TSC can accommodate up to 68 fuel assemblies for a total of 340 fuel storage cells among the five TSCs. Thirty-two of the 68 fuel cell locations in each TSC are designed for a damaged fuel canister. Seven spare locations are available in the fifth TSC.

The spent fuel assemblies will be loaded underwater into a TSC within a TFR located in the cask pool. The TSC/TFR will be removed from the cask pool and the TSC will be prepared for storage by draining, helium fill, vacuum drying, and closure lid seal welding. The TSC/TFR will be moved outdoors and the TSC will be placed in the VCC by positioning the TFR on top of the VCC and lowering the TSC within the TFR through the transfer adapter to the VCC below. The VCC containing the TSC will then be placed in storage on the ISFSI pad.

The TSC assembly consists of a right circular cylindrical shell with a welded bottom plate, a fuel basket, a closure lid with closure ring, and two sets of redundant port covers. The cylindrical shell, plus the bottom plate and lid, constitutes the confinement boundary. The stainless steel fuel basket is a right circular cylinder with 68 fuel tubes including 32 oversized tubes designed to accommodate LACBWR damaged fuel cans, laterally supported by a series of stainless steel support disks. The support disks are retained by spacers on radially located tie rods. The spent fuel assemblies will be contained in square stainless steel fuel tubes that include Boral on up to four sides for criticality control.

The VCC is the storage overpack for the TSC and provides structural support, shielding, protection from environmental conditions, and natural convection cooling of the TSC during long term storage. The VCC is a reinforced concrete structure with a carbon steel inner liner.

The VCC has an annular air passage to allow the natural circulation of air around the TSC. The air inlet and outlet vents take non-planar paths to the VCC cavity to minimize radiation streaming. The decay heat is transferred from the spent fuel assemblies to the tubes in the fuel basket, through the heat transfer disks, to the TSC wall. Heat flows by convection from the TSC wall to the circulating air, as well as by radiation from the TSC wall to the VCC inner liner. The heat flow to the circulating air from the TSC wall and the VCC liner is exhausted through the air outlet vents. The top of the VCC is closed by a single shielded lid incorporating a carbon steel D-PLAN 4-6 November 2010 1

4. FACILITY DESCRIPTION - (cont'd) plate for gamma shielding and concrete for neutron shielding. The VCC lid will be bolted in place.

The TFR, with its lifting yoke, is a qualified heavy lifting device designed, fabricated, and proof load tested to the requirements of NUREG-0612 and ANSI N14.6. The TFR provides shielding during TSC movements between work stations, the VCC, or the transport cask. It is a multi-wall (steel/lead/NS-4-FR/steel) design and has a bolted top retaining ring to prevent a fuel-loaded canister from being inadvertently removed through the top of the TFR. Retractable, hydraulically operated, bottom shield doors on the TFR are used during TSC transfer operations.

To minimize contamination of the TSC, clean water will be circulated in the gap between the TFR and the TSC during cask pool loading operations.

The ISFSI pad made of reinforced concrete will be 48 feet in length, and 32 feet in width, and 3 feet thick. The empty TFR and five VCCs containing fuel-loaded TSCs will be placed in a storage array on the ISFSI pad. The ISFSI will be surrounded by a protected area fence, an isolation zone fence, and vehicle barrier system. These protective barriers will be within the ISFSI Controlled Area Boundary fence. The ISFSI pad will be supplied with lighting, electronic surveillance and security systems. The ISFSI Administration Building is a commercial structure approximately 60 feet by 30 feet located 400 feet northeast of the ISFSI pad. The ISFSI Administration Building will provide space for security monitoring and ISFSI operations support.

D-PLAN 4-7 November 2010 1

4. FACILITY DESCRIPTION - (cont'd)

REMOVABLE GATE CASK POOL CASK POOL SUPPORT Containment Buildini Elevation Figure 4.1 D-PLAN 4-8 November 2010 1

4. FACILITY DESCRIPTION - (cont'd)

I

'I Containiment Buildingz General Arrangzement Figure 4.2 D-PLAN 4-9 November 2010

4. FACILITY DESCRIPTION - (cont'd)

SPENT FUEL STORAGE WELL Main Floor of Turbine and Containment Building, El. 668'0" Figure 4.3 D-PLAN 4-10 November 2010 1

4. FACILITY DESCRIPTION - (cont'd)

SPENT FUEL STORAGE WELL Mezzanine Floor of Turbine and Containment Building, El. 654'0" Figure 4.4 D-PLAN 4-11 November 2010 1

4. FACILITY DESCRIPTION - (cont'd)

FORCED CIRCULATIOIS TURBINE BUILDING CED PUMP IB, I.p FULL F HEATER NO. 3 DIMINI p

BLDG.

4 t64 iC Z--47J 0

CONDENSER M6CIIZI IA. DIESEL MAIN LABORATORYI[

WASTS FILTER6 ELECTRICA.L OIL MAINTENANCE CuSTORM 3uj ROOK DECONTAM-INATION AREA LUZ COUNTING HOWER &

ROOM fASH AREA JILDING Grade Floor of Turbine, Containment, and Waste Treatment Building, El. 640'0" Figure 4.5 D-PLAN 4-12 November 2010 1

4. FACILITY DESCRIPTION - (cont'd)

HIS rafUOM Of" AACTOR..

SUILOING WALL.

REMMOW Reactor Building Opening Figure 4.6 4-13 November 2010 D-PLAN

4. FACILITY DESCRIPTION - (cont'd) 0 Reactor Building Bi-Parting Door Figure 4.7 D-PLAN 4-14 November 2010 I

4. FACILITY DESCRIPTION - (cont'd)

Onsite ISFSI Figure 4.8 D-PLAN 4-15 November 2010 1

5.

PLANT STATUS 5.1 SPENT FUEL INVENTORY During June 1987 all fuel assemblies were removed from the reactor vessel. Currently there are 333 spent fuel assemblies stored in the spent fuel pool. All spent fuel will be placed in dry storage casks and moved to the onsite Independent Spent Fuel Storage Installation.

This spent fuel consists of three different types of fuel assemblies. Type I (82 assemblies) and Type 11 (73 assemblies) were fabricated by Allis-Chalmers (A-C) and Type Il (178 assemblies) by EXXON. All of the fuel assemblies are 10xl0 arrays of Type 348 stainless steel clad rods with stainless steel and Inconel spacers and fittings. The initial enrichment of the uranium in the Type I and Type II fuel was 3.63% and 3.92% respectively and the nominal average initial enrichment of the Type III fuel was 3.69%. The Type III assemblies contain 96 fueled rods and 4 inert Zircaloy-filled rods.

The 72 fuel assemblies removed from the reactor in June 1987 have assembly average exposures ranging from 4,678 to 19,259 megawatt-days per metric ton of uranium (MWD/MTU). The exposures of the 261 fuel assemblies discharged during previous refuelings range from 7,575 to 21,532 MWD/MTU. The oldest fuel stored was discharged from the reactor in August 1972.

Forty-nine of the A-C fuel assemblies discharged prior to May 1982 contain one or more fuel rods with visible cladding defects and 54 additional A-C fuel assemblies discharged prior to December 1980 contain one or more leaking fuel rods as indicated by higher than normal fission product activity observed during dry sipping tests.

The estimated radioactivity inventory in the 333 spent fuel assemblies is tabulated in Table 5-1.

TABLE 5-1 SPENT FUEL RADIOACTIVITY INVENTORY January 1988 (a)

Radio-Half Life Activity Radio-Half Life nuclide (Years) (b)

(Curies) nuclide (Years) (b)

(Curies) 144Ce 7.801 E-1 2.636 E+6 90Sr 2.770 E+I 1.147 E+6 137CS 3.014 E+1 1.666 E+6 241pu 1.440 E+1 1.138 E+6 106Ru 1.008 E+0 1.524 E+6 51Fe

• 2.700 E+0 5.254 E+5 95Zr(Nb) 1.754E-1 3.555 E+5 95Zr

• 1.750 E-1 3.52 E+2 (9.58E-2) 134CS 2.070 E+0 3.291 E+5 59Ni

• 8.000 E+4 2.87 E+2 (a)

Computer Program, FACT-I, DPC, July 1987, and hand calculations.

(b)

Computer Program, TPASGAM, Nuclide Identification Package, J. Keller, Analytical Chemistry Division, ORNL, June 1986.

Activity in fuel assembly hardware based on neutron activation analysis.

D-PLAN 5-1 November 2010

5. PLANT STATUS - (cont'd)

TABLE 5 (cont'd)

Radio-Half Life Activity Radio-Half Life nuclide (Years) (b)

(Curies) nuclide (Years) (b)

(Curies) 85Kr I 'OmAg 89Sr 12 7rnTe 60Co

• 103Ru 147pm 63Ni

• 14tCe 242Cm 241Am 238pu 239pu 240pu 154Eu 244Cm 51Cr

• 1291nTe 3H 59Fe

• 152Eu 242mAm 1.072 E+I 6.990 E-1 1.385 E-1 2.990 E-1 5.270 E+O 1.075 E-1 2.620 E+0 1.000 E+2 8.890 E-2 4.459 E-1 4.329 E+2 8.774 E+1 2.410 E+4 6.550 E+3 8.750 E+0 1.812 E+1 7.590 E-2 9.340 E-2 1.226 E+I 1.220 E-1 1.360 E+I 1.505 E+2 1.160 E+5 1.018 E+5 1.009 E+5 8.238 E+4 6.395 E+4 6.334 E+4 4.129 E+4 3.540 E+4 2.638 E+4 1.858 E+4 1.474 E+4 1.262 E+4 8.837 E+3 7.165 E+3 4.020 E+3 3.603 E+3 3.002 E+3 1.170 E+3 5.510 E+2 5.120 E+2 5.110 E+2 4.900 E+2 9 9Tc 125 Sb 155Eu 234U 24 3Am 1131nCd 94Nb

• 135Cs 238U 156Eu 242pu 236U 12h1nSn 237Np 235U 15 1Sm 1268n 79Se 1291 9 3Zr 1311 2.120 E+5 2.760 E+0 4.960 E+0 2.440 E+5 7.380 E+3 1.359 E+I 2.000 E+4 3.000 E+6 4.470 E+9 4.160 E-2 3.760 E+5 2.340 E+7 7.600 E+1 2.140 E+6 7.040 E+8 9.316 E+I 1.000 E+5 6.500 E+4 1.570 E+7 1.500 E+6 2.200 E-2 2.76 E+2 2.73 E+2 1.68 E+2 6.37 E+I 6.31 E+I 1.78 E+I 1.59 E+I 1.40 E+I 1.22 E+I 8.63 E+0 8.58 E+0 6.32 E+0 4.44 E+0 2.19 E+0 1.89 E+0 1.51 E+0 7.01 E-i 5.52 E-i 3.90 E-i 1.11 E-1 2.00 E-3 (a)

Computer Program, FACT-I, DPC, July 1987, and hand calculations.

(b)

Computer Program, TPASGAM, Nuclide Identification Package, J. Keller, Analytical Chemistry Division, ORNL, June 1986.

Activity in fuel assembly hardware based on neutron activation analysis.

D-PLAN 5-2 November 2010 1

5. PLANT STATUS - (cont'd) 5.2 PLANT SYSTEMS AND THEIR STATUS 5.2.1 Reactor Vessel and Internals The reactor vessel consisted of a cylindrical shell section with a formed integral hemispherical bottom head and a removable hemispherical top head bolted to a mating flange on the vessel shell to provide for vessel closure. The vessel had 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 was a ferritic steel (ASTM A-302-Gr-B) plate with integrally bonded Type 304L stainless steel cladding. The flanges and large nozzles were ferritic steel (ASTM A-336) forgings. The small nozzles were made of Inconel pipe.

The reactor internals consisted 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.

5.2.1.1 System Status: All fuel assemblies have been removed from the reactor vessel. Startup sources have been disposed of. The reactor vessel with head installed, internals intact, and 29 control rods in place was filled with low density cellular concrete. Attachments to the reactor vessel flange were removed to a diameter of 119 inches. All other nozzles and appurtenances were cut to within the diameter of the flange. Under-vessel nozzles and appurtenances were removed from an envelope of within 6 inches of bottom dead center of the reactor vessel shell bottom. The reactor pressure vessel was removed from the Reactor Building and shipped to the Barnwell Waste Management Facility in June 2007.

5.2.2 Forced Circulation System The Forced Circulation system was designed to circulate sufficient water through the reactor to cool the core and to control reactor power from 60 to 100 percent. Primary water passed upward through the core, and then down through the steam separators to the re-circulating water outlet plenum. The water then flowed to the 16-inch Forced Circulation pump suction manifold through four 16-inch nozzles and was mixed with reactor feedwater that entered the manifold through four 4-inch connections. From the manifold, the water flowed through 20-inch suction lines to the two 15,000 gpm variable-speed Forced Circulation pumps. Hydraulically-operated rotoport valves were installed at the suction and discharge of each pump. The 20-inch pump discharge lines returned the water to the 16-inch Forced Circulation pump discharge manifold.

From the manifold, the water flowed through four equally spaced 16-inch reactor inlet nozzles to the annular inlet plenum, and then downward along the bottom vessel head to the core inlet plenum.

The system piping was designed for a maximum working pressure of 1450 psig at 650'F (a pressure above the maximum reactor working pressure to allow for the static head and the pump head). Since the piping from the reactor to the rotoport valves was within the biological shield and not accessible, the valves and piping were clad with stainless steel. The piping between the rotoport valves and the pumps was low-alloy steel.

D-PLAN 5-3 November 2010 I

5.

PLANT STATUS - (cont'd)

Each Forced Circulation pump had an auxiliary oil system and a hydraulic coupling oil system.

Each auxiliary oil system supplied oil to cool and lubricate the three (1 radial and 2 thrust) pump coupling bearings. Each hydraulic coupling oil system supplied cooled oil at a constant flow rate to the hydraulic coupling.

5.2.2.1 System Status: The Forced Circulation system and attendant oil systems have been drained. The Forced Circulation pumps, auxiliary oil pumps, and hydraulic coupling oil pumps have been electrically disconnected. All 16-inch and 20-inch Forced Circulation system piping was filled with low density cellular concrete. Four 16-inch Forced Circulation inlet nozzles and four 16-inch outlet nozzles were cut to allow removal of the reactor pressure vessel. Piping located within the reactor cavity was also cut at the biological shield, segmented into manageable pieces, and disposed of. Pumps and piping in the shielded cubicles remain.

5.2.3 Seal Injection System The Seal Injection system provided cooling and sealing water for the seals on the two Forced Circulation pumps and the 29 control rod drive units.

The Seal Injection system had two positive-displacement pumps, supplied with water from the Seal Injection reservoir. The reservoir was supplied from the Condensate Demineralizer system with a backup supply from the overhead storage tank. One pump was required for operation with the other on standby.

A cartridge-type filter was provided in the pump suction header. A deaerator, vented to the Seal Injection reservoir, was located on the suction of the pumps to remove entrained air from the system in the event air was introduced during makeup to the system.

Bladder-type accumulators were connected to the suction and discharge of each Seal Injection pump. The suction accumulators reduced pump and piping vibrations. The discharge accumulators dampened the pulsating flow from the pumps to provide the constant flow rate required for stable system conditions.

The water to the Forced Circulation pump seals passed through a full-flow filter. There were two filters arranged in parallel with one normally in service and the other in standby. Each seal supply line contained a flow control valve, check valve, and a three-way valve for switching injection points on the seal. Both valves in each supply line could be operated from the Control Room.

The water to the control rod drive seals was filtered in the same manner. The water passed through a flow control valve which maintained a constant flow rate to the seals. The individual supply lines to the 29 control rod drive units had a throttle valve and flow indicator for setting the required flow rate to each seal. The normal leakoff from the seals (0.15 gpm or less) was drained to the reactor basement floor sump.

Continuous blowdown of water from the control rod drive upper housing was removed through a connection on each control rod drive upper housing flange. Each line contained a throttle valve, and all lines joined at a common manifold, from which suction was taken by the control rod nozzle effluent pumps. The pumps discharged into the inlet of the Decay Heat system, which was directly connected to the Forced Circulation system.

D-PLAN 5-4 November2010 I

5. PLANT STATUS - (cont'd) 5.2.3.1 System Status: This system is drained and not maintained operational.

5.2.4 Decay Heat Cooling System The Decay Heat Cooling system was a single high pressure closed loop containing a pump, cooler, interconnecting piping, and the necessary instrumentation.

The decay heat loop was connected across the 20-inch Forced Circulation pump 1 A supply and return lines on the reactor side of the Forced Circulation pump isolation valves. Both connections to the forced circulation piping were located outside the reactor biological shield.

Reactor water from the Forced Circulation pump supply line passed through an 8-inch line to the suction side of the Decay Heat pump. The water was then discharged through a 6-inch line into the Decay Heat cooler where it was cooled and returned to the Forced Circulation pump return line.

A 2-inch blowdown line to the Main Condenser was located downstream of the Decay Heat cooler. This was used to remove the excess water due to seal inleakage and thermal expansion of the water. Another 2-inch line downstream of the Decay Heat cooler connected to the reactor vessel head vent line which could be used to promote better circulation of the water within the reactor vessel head.

5.2.4.1 System Status: This system is drained and not maintained operational.

5.2.5 Emergency Core Spray System The Emergency Core Spray system consisted of a spray header with individual spray lines for each fuel assembly mounted inside the reactor vessel. The low pressure supply line allowed demineralized water from the Overhead Storage Tank, or the service water from the High Pressure Service Water supply line, to flow directly to the core spray header.

5.2.5.1 System Status: This system has been removed.

5.2.6 Boron Injection System The purpose of the Boron Injection system was to inject enough sodium pentaborate solution into the primary system to make the reactor subcritical in the event of stuck control rods.

5.2.6.1 System Status: This system has been removed.

5.2.7 Primary Purification System The Primary Purification system was a high pressure, closed loop system consisting of a regenerative cooler, purification cooler, pump, two ion exchangers and filters.

The functions of the Primary Purification system were:

N To maintain optimum reactor water quality (pH and conductivity) to minimize corrosion, D-PLAN 5-5 November 2010 I

5. PLANT STATUS - (cont'd)

" To remove dissolved and suspended solids in order to minimize fouling of heat transfer surfaces, pipes and vessels, and maintain primary coolant activity level low, and

" To provide an alternate means of removing excess reactor water.

5.2.7.1 System Status: The ion exchanger resins have been removed and the system has been drained.

5.2.8 Alternate Core Spray System The Alternate Core Spray system consisted of two diesel-driven High Pressure Service Water (HPSW) pumps which took suction from the river and discharged to the reactor vessel through duplex strainers and two motor-operated valves installed in parallel.

The Alternate Core Spray system was installed to provide backup for the High Pressure Core Spray system. It provided further assurance that melting of fuel-element cladding would not occur following a major recirculation line rupture. It had a secondary function of providing backup to the HPSW system and Fire Suppression system. The Emergency Service Water Supply system (ESWSS) pumps were portable pumps which served as backups to the diesel-driven HPSW pumps in the event the Cribhouse or underground piping was damaged. The ESWSS has been removed.

5.2.8.1 System Status: The Alternate Core Spray system is not required to be operational in SAFSTOR. Motor-operated valves and instrumentation in the Turbine Building have been electrically removed. System components continue to serve requirements of the HPSW system.

5.2.9 Control Rod Drive Auxiliaries The function of this system was to provide hydraulic fluid under pressure for the purpose of hydraulically inserting the control rods when an immediate plant shutdown was necessary.

5.2.9.1 System Status: This system has been removed.

5.2.10 Gaseous Waste Disposal System This system routed main condenser gasses through various components for drying, filtering, recombining, monitoring and holdup for decay.

5.2.10.1 System Status: This system, except for the storage tanks, has been removed.

5.2.11 Fuel Element Storage Well System The storage well is a stainless lined concrete structure 11 feet by 11 feet by approximately 42 feet deep. When full, it contains approximately 38,000 gallons. It is completely lined with Type 316 stainless steel. The walls are 16-gauge sheet and the bottom a 3/8-inch plate. All joints are full penetration welds. Vertical and horizontal expansion joints in the storage well allow for thermal expansion. A three-section aluminum cover, with two viewing windows per section, has been manufactured to cover the pool. The floor of the storage well has a design safe uniform live load of 5,000 lb/ft2.

D-PLAN 5-6 November 2010 1

5. PLANT STATUS - (cont'd)

Spent fuel assemblies are stored in two-tiered racks in the Fuel Element Storage Well (FESW) until removed to cask storage. A transfer canal connects the upper portion of the well to the upper cavity and is closed with a water-tight gate and a concrete shield plug. The water level in the well is normally maintained at an elevation of > 695 feet with fuel in upper rack.

FESW cooling is accomplished by drawing water through a 6-inch penetration at elevation 679 feet, or a 4-inch line at elevation 679 feet 11 inches, and pumping it through the FESW cooler and returning it to the well, with either of two FESW pumps. The return line enters the top of the storage well and extends down to discharge at elevation 695 feet. The bottom inlet line ends at the biological shield wall and is sealed with a welded plug.

Cleanup is provided by the FESW ion exchanger. A 4-inch line from the Overhead Storage Tank is used to flood the well or pump water back to the Overhead Storage Tank. Overflow and drain pipes from the well and cavity are routed to the retention tanks. Normal makeup to the storage well is provided by demineralized water through one of two fill valves which are remotely operated from Benchboard E in the Control Room.

The FESW cooling system is conservatively designed to remove the decay heat of a full core one week after shutdown, with the storage well water at 120'F and the ultimate heat sink, the river, at 850F.

5.2.11.1 System Status: The Fuel Element Storage Well contains 333 spent fuel assemblies and will remain in operation as part of the SAFSTOR Program as long as wet fuel storage or wet fuel handling is necessary. Following removal of all spent fuel assemblies and fuel debris for dry storage at the ISFSI, the storage racks will be removed and disposed of; the FESW will be drained and decontaminated.

5.2.12 Component Cooling Water System The Component Cooling Water (CCW) system provided controlled quality cooling water to the various heat exchangers and pumps in the Reactor Building during plant operation. It also served as an additional barrier between radioactive systems and the river. Currently, the system provides cooling water to the FESW heat exchanger and the two Reactor Building Air Conditioner compressors.

The CCW system is a closed system consisting of two pumps, two heat exchangers, a surge tank, and the necessary piping, valves, controls, and instrumentation to distribute the cooling water.

The CCW pumps, coolers, and the surge tank are located in the Turbine Building. Water flows from the pumps, to the cooler, and then to the component cooling water supply header in the Reactor Building. The flow requirements of the components cooled by the CCW system were as follows during plant operation:

D-PLAN 5-7 November 2010

5. PLANT STATUS - (cont'd)

Design Nominal (GPM)

(GPM)

1) FCP Hydraulic Coupling Coolers ----------------.............................

60 60

2) FCP Lube Oil Coolers 30 30

3) Shield Cooler 75 75

4) Control Rod Nozzle Effluent Pumps........................................

30 30

5) P urifi cation Pum p ---------------------------------------------.............

15 15

6) Purification Cooler 260 200

7) Reactor Building Air Conditioners --------------------.......................

60 each 120

8) D ecay H eat Pum p......................................................................

20 20

9) D ecay H eat C ooler 5

570 100

10) Fuel Elem ent Storage W ell Cooler --------------------.......................

260 100

11) Sam ple Coolers 5a..................................................................

5-10 each 40

12) Failed Fuel Element Location System Cooler -------------------------

40 0

13) Station A ir C om pressors ------------------------------------

20 each 0

14) PASS a) R eactor C oolant Sam ple ------------------------.......................

10 5

b) Containment Atmosphere Sample...............................

40 0

TOTAL 1560 855 Water from each of the components, listed above, flows to the CCW return header. This header leaves the Reactor Building and connects to the suction of the CCW pumps. A sample stream from the supply header is monitored for radioactivity and returned to the suction header. The temperature of the water in the supply header is automatically controlled by varying the Low Pressure Service Water flow to the tube side of the CCW coolers.

5.2.12.1 System Status: This system remains operable.

5.2.13 Shield Cooling System The Shield Cooling system was designed to maintain the temperature of the thermal shield and biological shield concrete below 140'F and 150TF, respectively.

5.2.13.1 System Status: All system components external to the biological shield have been removed, but because they are inaccessible the cooling coils have been abandoned in place.

5.2.14 Shutdown Condenser System The primary function of the Shutdown Condenser was to provide a backup heat sink for the reactor, in the event the reactor was isolated from the main condenser, by the closure of either the Reactor Building Steam Isolation valve or the Turbine Building Steam Isolation valve. In addition, the Shutdown Condenser acted as an over-pressure relief system in limiting over-pressure transients.

The Shutdown Condenser was located on a platform 10 feet above the main floor in the Reactor Building. Steam from the 10-inch main steam line passed through a 6-inch line, two parallel inlet steam control valves, back to a 6-inch line and into the tube side of the condenser where it was condensed by evaporating cooling water on the shell side. The steam generated in the shell D-PLAN 5-8 November 2010 1

5.

PLANT STATUS - (cont'd) was exhausted to the atmosphere through a 14-inch line which penetrates the Reactor Building.

The main steam condensate was collected in the lower section and returned to the reactor vessel by gravity flow.

A vent line containing two parallel control valves was connected to the 6-inch condensate return line. The valves discharged directly to the Reactor Building atmosphere and were capable of remote manual operation to vent the primary system directly to the Reactor Building atmosphere under emergency conditions. They performed the function of Reactor Emergency Flooding Vent Valves to equalize water level in the building with that in the reactor vessel, for a below-core break, and Manual Depressurization System to rapidly depressurize the reactor vessel, on failure of the High Pressure Core Spray coincident with a major leak.

5.2.14.1 System Status: This system has been removed.

5.2.15 Hydraulic Valve Accumulator System The function of the Hydraulic Valve Accumulator system was to supply the necessary hydraulic force to operate the five piston-type valve actuators, which operated the five rotoport valves in the Forced Circulation and Main Steam systems. The system consisted 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 was maintained in the water accumulator tank by pumps which took suction from the water return sump tank. The water in the accumulator tank was maintained under 140 psi air pressure by the air compressors. The pressurized water was then directed to pistons for hydraulic valve operation.

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

5.2.16 Well Water System Water for this system is supplied from two deep wells. Well No. 4 is located 115 feet southeast of the containment vessel center, and Well No. 3 is located 205 feet northeast of this centerline.

The wells are 12 inches in diameter, with 8-inch pump casings and piping. The upper 40 feet of casing is set in concrete. The pumps are sealed submersible pumps. They take suction through stainless steel strainers, and they discharge into pressure tanks.

The system supplies water to the plant and office for sanitary and drinking purposes. Water supplied by the system is used at personnel and material decontamination stations. It is used as cooling water for the two Turbine Building air-conditioning units and in the heating boiler blowdown flash tank and sample cooler. The well water system is the source of supply to laundry equipment.

5.2.16.1 System Status: This system is maintained in continuous operation.

5.2.17 Demineralized Water System The Virgin Water Tank provides the supply to the Demineralized Water transfer pumps which D-PLAN 5-9 November 2010 1

5. PLANT STATUS - (cont'd) distribute demineralized water throughout the plant, including to the Overhead Storage Tank and the Fuel Element Storage Well makeup in the Reactor Building. Water is demineralized in batches at Genoa Unit 3, transferred to LACBWR where it is sampled, and, if of acceptable quality, stored in the Virgin Water Tank.

The Condensate Storage Tank and the Virgin Water Tank are actually two sections of an integral aluminum tank located on the office building roof. The lower section of this tank is the Condensate Storage Tank, and it has a capacity of 19,100 gallons. The upper, virgin-water, section will hold 29,780 gallons and has level indication in the Control Room.

5.2.17.1 System Status: The Demineralized Water system remains in service, mainly as a source of water for the Fuel Element Storage Well and the heating boiler. The Condensate Storage Tank has been drained.

5.2.18 Overhead Storage Tank The Overhead Storage Tank (OHST) is located at the top of, and is an integral part of, the Reactor Building. The OHST system consists of the approximately 45,000-gallon tank, the tank level instrumentation and controls, and the piping to the first valve of the systems served by the tank.

The OHST now serves as a reservoir for water used to flood the Fuel Element Storage Well, cask pool, and upper cavity during cask loading operations. During operation, the OHST acted as a receiver for rejecting refueling water using the Primary Purification system. The OHST also supplied the water for the Emergency Core Spray system and Reactor Building Spray system, and was a backup source for the Seal Injection system.

5.2.18.1 System Status: The OHST remains in use, primarily for a source of makeup water to the Fuel Element Storage Well and for cask loading operations. A 4-inch line from the OHST is used to flood the well or pump water back to the OHST using FESW system pumps, valves, and piping. After the transfer of all spent fuel to dry storage at the onsite ISFSI, the OHST system will be drained and dismantled 5.2.19 Station and Control Air System There are two single-stage positive displacement lubricated type compressors. The complete compressor consists of an encapsulated compressor system, inlet system, cooling system, and control system. The encapsulated compressor includes compressor unit, fluid management system, and motor section. One compressor is normally running, and the other compressor can be started when necessary. The air receivers act as a volume storage unit for the station.

The air receiver outlet lines join to form a header for supply to the station and the control air systems. Station air is provided to the Cribhouse, where it is piped to near the suction of the Low Pressure Service Water pumps; to the High Pressure Service Water tank to charge the tank; and to the generator and reactor plants at all floor levels, for station usage as needed.

Control air is supplied from the receiver discharge header through a refrigerated air dryer and coalescing filter to various instruments and valves in the reactor and generator plants. Alarms D-PLAN 5-10 November 2010 1

5. PLANT STATUS - (cont'd) are provided in the Control Room to warn of low control air header pressure and compressor failures.

5.2.19.1 System Status: This system is maintained in continuous operation.

5.2.20 Low Pressure Service Water System The system is supplied by two vertical pumps located in the Cribhouse through a duplex strainer unit. The Low Pressure Service Water (LPSW) system supplies the Component Cooling Water coolers and Circulating Water pump mechanical seals, and is the normal supply to the High Pressure Service Water (HPSW) system through the motor-driven HPSW pump. During plant operation the LPSW system also supplied the Turbine Oil coolers, generator hydrogen system coolers, Condenser Vacuum pump, and Reactor Feedwater pumps.

5.2.20.1 System Status: This system is maintained in continuous operation.

5.2.21 High Pressure Service Water System The High Pressure Service Water (HPSW) system supplies fire suppression water and is available as backup cooling water for the Component Cooling Water coolers. During normal operation, HPSW system pressure is maintained by the Genoa Unit 3 jockey pump. A motor-driven HPSW pump with suction from the LPSW system is available for periods of high demand. With the motor-driven pump cycling in automatic, HPSW system pressure is maintained 110 to 135 psig at the expansion tank pressure switch elevation, 25 feet above site grade elevation. The pump is protected by a 35-psig low suction pressure trip. Backup supply is available from two HPSW diesel pumps. IA HPSW diesel pump will start automatically if system pressure decreases to 90 psig. 1B HPSW diesel pump will start automatically if system pressure decreases to 80 psig. The HPSW diesel pumps will maintain system pressure at approximately 150 psig. System pressure swings are cushioned by the air space in the HPSW surge tank.

The HPSW system is divided into two main loops. The internal loop serves the Turbine Building, Reactor Building, and Waste Treatment Building interior hose stations and sprinkler systems. The external loop supplies outside fire hydrants and Cribhouse sprinklers. The external loop is also cross-connected with the Fire Suppression system of the adjacent coal-fired generating facility, Genoa Unit 3. This cross-connect provides excess HPSW diesel pump capacity to this operating plant.

5.2.21.1 System Status: This system is maintained in operation to provide fire protection.

5.2.22 Circulating Water System Circulating water is drawn into the Cribhouse intake flume from the river through traveling screens by circulating water pumps 1 A and 1 B, which are located in separate open suction bays.

Each pump discharges into 42-inch pipe; the pipes join a common 60-inch pipe leading to the main condenser in the Turbine Building. At the condenser, the 60-inch pipe branches into two 42-inch pipes feeding the top section of the water boxes. The main condenser is a two-pass divided water box type. Circulating water enters the top section of the condenser tube side and is D-PLAN 5-11 November 2010 1

5. PLANT STATUS - (cont'd) discharged from the bottom section tube side. The condenser tubes extend the length of the condenser and are fastened at each end to the tube sheets inserted between the water boxes and the shell.

The 42-inch condenser circulating water outlet lines tie into a common 60-inch line which discharges to the seal well from Genoa Unit 3, located approximately 600 feet downstream from the LACBWR Cribhouse.

5.2.22.1 System Status: This system is maintained operational for periodic use for dilution of liquid waste discharges.

5.2.23 Condensate System and Feedwater Heaters The Condensate system took condensed steam from the condenser hotwell and delivered it under pressure to the suction of the reactor feed pumps. Two identical full-capacity condensate pumps took suction from the hotwell, and pumped the condensate through a full-flow demineralizing system, the air ejector condensers, the gland steam exhaust condenser and two feedwater heaters before entering the feed pumps.

The Condensate system also supplied the turbine exhaust sprays, the reactor feed pump shaft sealing cooling system, the normal makeup to the seal injection system, and gland seal steam generator. Hotwell level was maintained by automatic makeup from, or overflow to, the Condensate Storage Tank.

5.2.23.1 System Status: This system has been removed, with the exception of three feedwater heaters that remain in place with piping connections removed. The Condensate Storage Tank has been drained.

5.2.24 Reactor Feedwater Pumps The Reactor Feedwater pumps took preheated condensate from No. 2 feedwater heater and delivered it through No. 3 feedwater heater to the reactor. The pumps boosted the system pressure from about 200 psi to approximately 1300 psi. The pump coupling arrangement was such that pump speed, and therefore capacity, could be varied to control reactor water level.

Each pump was a separate unit containing all the auxiliaries, controls, and other components necessary for independent operation.

5.2.24.1 System Status: The Reactor Feedwater pumps have been removed.

5.2.25 Full-Flow Condensate Demineralizer System The Full-Flow Condensate Demineralizer system consisted of three service tanks, each with one-half system capacity and arranged in parallel. Its purpose was to remove ionic impurities from the condensate system water before admitting it to the reactor. Each service tank was capable of delivering 700 gpm. With one of the three tanks on standby, the system was capable of delivering 1400 gpm to satisfy primary system requirements. The standby service tank was available for service whenever the effluent conductivity of the inservice tanks rose to an unacceptable level. Each of the three demineralizer tanks normally contained 45-50 ft3 of pre-D-PLAN 5-12 November 2010 1

5. PLANT STATUS - (cont'd) regenerated mixed resins with a cation/anion ratio of 2 to 1. The three service tanks were designed for 400 psig operation, and normal flow was supplied by the condensate pumps. A circulating pump was provided to circulate water through the standby demineralizer tank prior to placing it into service.

5.2.25.1 System Status: This system, except for six empty tanks located in the Full-Flow room, has been removed.

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, 5470 F, exhausting at 1.0" Hg Absolute. The turbine consisted of a high pressure and intermediate pressure and a low pressure element.

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

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 PF, 3600 rpm, 60 cycle, 3 phase, 13,800-V AC, 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.

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

5.2.28 Turbine Oil and Hydrogen Seal Oil System The Turbine 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 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 D-PLAN 5-13 November 2010 1

5. PLANT STATUS - (cont'd) oil tanks located in the oil storage room.

5.2.28.1 System Status: This system, with exception of the drained clean and dirty oil tanks, has been removed 5.2.29 Heating, Ventilation, and Air-Conditioning Systems The Reactor Building ventilation system utilizes two 30-ton, 12,000-cfm air conditioning units for drawing fresh air into the building and for circulating the air throughout the building. Each air-conditioning unit air inlet is provided with a filter box assembly, face and bypass dampers; and one 337,500-Btu/hr capacity steam coil that is used when heating is required. Air enters the building through two series 20-inch dampers and is exhausted from the building by action of the stack blowers. Additional exhaust flow is available using a centrifugal exhaust fan that has a capacity of 6000 cfm at 4 inches of water static pressure. The exhaust fan and building exhaust air discharge through two series 20-inch dampers to the Reactor Building ventilation outlet plenum connected to the tunnel. A 20-inch damper is also provided for recirculation of the exhaust fan discharge air. The exhaust system is provided with conventional and high-efficiency filters and with a gaseous and particulate radiation monitor system.

The Waste Treatment Building ventilation is provided by a 2000-cfm exhaust fan that draws air from the shielded vault areas of the building and exhausts the air through a duct out the floor of the building to the waste gas storage vault. The stack blowers then exhaust the air from the waste gas storage vault through the connecting tunnel and discharge the air up the stack.

The exhaust air from the Reactor Building and the Waste Treatment Building are discharged into the tunnel connecting the Waste Treatment Building, the Reactor Building, and the Turbine Building to a plenum at the base of the stack. The stack is 350 feet high and is of structural concrete with an aluminum nozzle at the top. The nozzle tapers to 4 feet 6 inches at the discharge, providing a stack exit velocity of approximately 70 fps with the two 35,000-cfm stack blowers in operation.

The Turbine Building heating system provides heat to the turbine and machine shop areas through unit heaters and through automatic steam heating units.

The Control Room Heating and Air-Conditioning unit serves the Control Room, Electrical Equipment Room, Shift Supervisor's area, and adjacent office.

The office area and laboratory are provided with a separate multi-zone heating and air-conditioning unit.

The heating boiler is a Cleaver-Brooks, Type 100 Model CB-189, 150-hp unit. At 150 psig, the boiler will deliver 6,275,000 Btu/hr. The boiler fuel is No. 2 fuel oil. The oil is supplied by and atomized in a Type CB-1 burner which will deliver 45 gph.

Two 14.7-kW resistance heaters with power supplied from the essential busses are available to heat the Reactor Building in the event normal heating is lost.

5.2.29.1 System Status: These systems are maintained operational D-PLAN 5-14 November 2010 1

5.

PLANT STATUS - (cont'd) 5.2.30 Waste Collection Systems The functions of the Waste Collection and Treatment System are:

" To collect and store radioactive liquid waste generated in the plant.

" To collect and transfer depleted ion exchange resins to a shipping container.

" To process the collected waste as required for safe and economical disposal.

The Turbine Building Liquid Waste system collects the liquid waste from the Turbine Building, the Waste Treatment Building, the waste gas storage vault, and the tunnel area in two storage tanks (4500 gallons and 3000 gallons) located in the tunnel between the Reactor Building and the Turbine Building.

The Reactor Building Liquid Waste system consists of two retention tanks, each with a capacity of 6000 gallons, a liquid waste transfer pump, two sump pumps, and the necessary piping to route the waste liquid to the retention tanks and from the retention tanks out of the Reactor Building.

After a tank's contents are recirculated, a sample is withdrawn from the tank and analyzed for radioactivity concentrations prior to discharge. The total amount of liquid waste discharged to the circulating water discharge line is measured with a flow totalizer water meter which can handle flow rates up to 100 gpm and a flow rate monitor with Control Room readout.

Spent resin will be transferred to the spent resin receiving tank, where it will be held until there is a sufficient quantity available for shipment to an approved processing facility. The resin will be transferred to an approved shipping container where it will be dewatered and made ready for shipment.

5.2.30.1 System Status: The Waste Collection systems are maintained operational.

5.2.31 Fuel Transfer Bridge The fuel transfer bridge is a specially-designed structure which is power-driven north and south on rails recessed in the floor at elevation 701' of the Reactor Building.

The bridge traverses over the areas where service operations are performed at the reactor cavity, transfer canal, spent fuel storage pool and the new fuel storage racks. The bridge serves as the structural support for the fuel transfer hoist, and it provides an operating platform for personnel.

5.2.31.1 System Status: The fuel transfer bridge is kept operational and is tested routinely. The fuel transfer bridge will be used to transfer individual spent fuel assemblies from the Fuel Element Storage Well to the dry cask storage container. After the transfer of all spent fuel to dry storage at the onsite ISFSI, the fuel transfer bridge will no longer be required.

D-PLAN 5-15 November 2010 1

5.

PLANT STATUS - (cont'd) 5.2.32 Communications Systems The communications systems installed or otherwise available in the plant are:

1) Central office trunk line telephone service for off-plant local and long distance calls.

2) PABX (Private Automatic Branch Exchange) for interplant and intraplant calls and for off-plant calls to or from the site.

3) Paging system for in-plant and site calls.

4) DPC ultra high-frequency radio network, for voice communications within DPC systems and headquarters, including mobile units.

5) Microwave system for calls between LACBWR, Genoa Unit 3, La Crosse, and Alma, and for calls to local numbers in La Crosse.

6) Portable transceivers (handie-talkie) for mobile interplant and site voice communication.

5.2.32.1 System Status: The various communications systems are presently maintained operational.

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. Reserve Feed Breakers 252R1A and 252R1B supply the 2400-V Bus IA and Bus lB from the 69/2.4-kV reserve auxiliary transformer.

The 2400/480-V Auxiliary Transformers IA and lB receive their power from the 2400-V Buses 1A and lB through breaker 252AT1A from Bus lA to Transformer I A, and through breaker 252AT1B from Bus 1B to Transformer lB. The auxiliary transformers supply the 480-V Buses 1A and lB through Main Feed Breakers 452M1A for Bus 1A and 452M1B for Bus lB.

The 480-V buses supply larger equipment directly. They also supply motor control centers (MCCs) which furnish power to motors and other associated equipment connected to them through their respective breakers, including the MCC 120-V AC Distribution Panels which supply equipment and instrumentation. The regular lighting cabinets are supplied from 480-V Buses lA and 1B.

5.2.33.2 480-V Essential Buses IA and 1B The 480-V Essential Bus 1A switchgear is normally supplied with electrical power from the 480-V Bus 1A 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 D-PLAN 5-16 November 2010 1

5. PLANT STATUS - (cont'd) prevent both sources from supplying the bus.

The 480-V Essential Bus lB switchgear is normally supplied with electrical power from 480-V Bus lB through breaker 452-52B. In the event of a loss of station power, the 480-V Essential Bus lB is supplied with electrical power from Diesel Generator lB 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 1A and lB may be cross-connected through the 480-V Essential Bus Tie Breakers 452 TBA and 452 TBB.

5.2.3 3.3 Emergency Diesel Generators lA and lB The 1A Diesel Generator 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.

The function of the 1A Diesel Generator is to supply emergency power to the 480-V Essential Bus 1 A which, in turn, supplies power to the Turbine Building MCC 1 A, the Turbine Building 120-V Bus, the Turbine Building 120-V Regulated Bus and the feed to the Regulated Bus Auxiliary Panel.

The 1 B Diesel Generator system consists of a 400 kW diesel driven generator, a 300-gallon fuel oil day tank, fuel oil transfer system and external remote radiator and fan, a 200 kW fan-cooled test load, a local engine control and instrument cabinet, and remote instrumentation and controls in the Control Room. The diesel generator set is located in the Generator Room of the Diesel Building which is south of the Electrical Penetration Room at elevation 641 feet.

The function of the lB Diesel Generator is to supply emergency power to the 480-V 1 B Essential Bus, which in turn supplies power to the Reactor MCC lA 480-V Bus, Diesel Building MCC 480-V Bus, and the loads supplied by these MCCs. A feed from 480-V Essential Bus 1B to Genoa Unit 3 provides an alternate source of energy to supplement their plant's batteries during emergency shutdown with subsequent plant blackout.

5.2.33.4 120-V Non-Interruptible Buses The 120-V Non-Interruptible Buses maintained a continuous non-interruptible power supply using static inverters to a portion of the essential plant control circuitry, communications equipment and radiological monitoring equipment. The static inverters used 125-V DC input and supplied 120-V AC output to their respective distribution panels. The static inverters have been removed and the distribution panels have been renamed 1 A 120-V AC Essential Power, Regulated Bus Auxiliary Panel, and 1 C 120-V AC Essential Power.

D-PLAN 5-17 November 2010 I

5. PLANT STATUS - (cont'd) 5.2.33.5 125-V DC Distribution The 125-V DC distribution systems supply DC power to all equipment requiring it. The 125-V DC distribution systems were divided into three separate and independent systems each with its own battery, battery charger, and distribution buses. The buses could be cross-connected but were normally isolated from each other. The Reactor Plant and Diesel Building batteries and chargers have been removed. A new smaller capacity battery charger has been installed in the Diesel Building to provide reliability to the system. The Diesel Building Battery Charger, Generator Plant Battery, and Generator Plant Battery Charger remain as sources of DC power to the 125-V DC distribution system. The once three separate systems have been interconnected by using installed bus tie breakers.

For the system, the Diesel Building Battery Charger provides the normal DC supply with the Generator Plant Battery as the reserve supply. The battery floats on the line maintaining a full charge, and provides emergency DC power in the event of a loss of AC power to the battery charger or failure of the charger. Due to age, the Generator Plant Battery Charger is maintained available as a standby unit.

5.2.33.6 System Status: The Electrical Power Distribution system is maintained operational and required surveillance tests are performed on the Emergency Diesel Generators and 125-V DC batteries. The Electrical Power Distribution system is configured to provide backup diesel generator power to the entire 480-V AC system. This configuration provides continued operation of normal lighting systems, air conditioning systems, and sanitary well water supply for habitability reasons and plumbing system use during loss of station power.

5.2.34 Post-Accident Sampling Systems The Post-Accident Sampling Systems (PASS) are designed to permit the removal for analysis of small samples of either Reactor Building atmosphere, reactor coolant, or stack gas when normal sample points are inaccessible following an accident. These samples will aid in determining the amount of fuel degradation and the amount of hydrogen buildup in the Reactor Building.

Samples will be removed to the laboratory for analysis.

5.2.34.1 Reactor Building Atmosphere PASS Description The Reactor Building Atmosphere PASS consists of a vacuum pump that takes suction from the Reactor Building atmosphere at the 714' elevation. The sample is drawn through a bypass line or to a remote sample cylinder and discharged back to the Reactor Building at the 671' elevation.

5.2.34.2 Stack Gas PASS Description The Stack Gas PASS 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.

D-PLAN 5-18 November2010 I

5. PLANT STATUS - (cont'd) 5.2.34.3 Reactor Coolant PASS Description The Reactor Coolant PASS 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.

5.2.34.4 System Status: The Stack Gas PASS is maintained in continuous operation. The Reactor Coolant PASS has been removed. The Reactor Building Atmosphere PASS is retained in place. After the transfer of all spent fuel to dry storage at the onsite ISFSI, the Post-Accident Sampling Systems will no longer be required.

5.2.3 5 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.

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

5.3 RADIONUCLIDE INVENTORY ESTIMATES Testing was conducted in-house, using Health Physics personnel, to determine the location and the quantities of the radionuclides present at LACBWR. Several different types of samples and sampling techniques were used to qualify/quantify the radionuclide inventory. Each method is described in the initial site characterization survey for SAFSTOR. All samples were gamma scanned using HPGe detectors coupled to a gamma spectroscopy computer system. This equipment has been calibrated to NBS traceable sources and is checked periodically to maintain this calibration.

5.4 RADIATION LEVELS 5.4.34 Plant Radiation Levels Upon entering the initial phase of LACBWR's SAFSTOR mode, base line gamma radiation surveys were performed throughout the plant. General area radiation levels are listed below.

These levels will be routinely monitored and tracked. Specific area hot spots will also be looked for and recorded on each area survey.

Area General Area Gamma Radiation Levels Reactor Building:

Shutdown Condenser Platform 10-20 mRem/hr 701' Level 6-12 mRem/hr Mezzanine Level East 5-10 mRem/hr Mezzanine Level West 20-30 mRem/hr West Nuclear Instrument Platform 40-90 mRem/hr D-PLAN 5-19 November 2010 1

5.

PLANT STATUS - (cont'd)

East Nuclear Instrument Platform Purification Cooler Platform Grade Floor North and East Grade Floor West Upper Control Rod Drive Area Basement Primary Purification Demineralizer Retention Tank Area Lower Control Rod Drive Area Forced Circulation Pump Cubicles Turbine Building:

Main Floor Mezzanine Stop Valve Area Grade Floor Feedwater Heater Area Tunnel Machine Shop lB Diesel Room Electrical Penetration Room Waste Treatment Building:

Main Floor Basement 10-20 mRem/hr 5-10 mRem/hr 7-20 mRem/hr 75-120 mRem/hr 60-120 mRem/hr 10-40 mRem/hr 7-17 mRem/hr 250-400 mRem/hr 60-150 mRem/hr 150-400 mRem/hr

<1-3 mRem/hr

<1-4 mRem/hr 10-85 mRem/hr 1-10 mRem/hr 5-20 mRem/hr 10-50 mRemihr

<1 mRem/hr

<1 mRem/hr 2-7 mRem/hr 1-20 mRem/hr 10-100 mRem/hr Building Exteriors:

Exterior of Waste Treatment Building Exterior of Reactor Building 5.4.35 System Radiation Levels

<1 except for south side where there is one spot between 3-4 mRem/hr

<1 except for one spot on south side reading 7 mRem/ hr 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 iestablished. 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 survey point dose rates.

Note: All readings are contact dose rates.

D-PLAN 5-20 November 2010 1

5. PLANT STATUS - (cont'd)

Survey Point #

1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 Survey Point Location Condensate Line to and from OHST Condensate Line to and from OHST Condensate Line to and from OHST 1 A Condensate Pump Discharge Line Emergency Overflow Line Emergency Overflow Bypass Line Ice Melt Line IA Reactor Feed Pump Near 1 B Reactor Feed Pump Discharge Valve Side of #3 Feedwater Heater Reheater Level Control Chamber South End of Reheater Gland Exhaust Condenser Loop Seal Main Steam Line Main Steam Line Offgas System Flame Arrestor 1 B Waste Water Pump 1 A Waste Water Pump End of 3000 Gallon Waste Tank End of 4500 Gallon Waste Tank Side of Gland Seal Steam Generator Side of Gland Seal Steam Generator Main Steam Bypass Line Turbine Inlet Valve Body Main Steam Line Reheat to Flash Tank Line Flash Tank Seal Injection Heater

1. 2 Feedwater Heater Bypass Line Feedwater Heater Bypass Line Bottom of Gland Exhaust Condenser Top of Gland Exhaust Condenser Condensate into Air Ejector Line Air Ejector Low Pressure Turbine Manhole Cover End of High Pressure Turbine Primary Purification IA Filter Inlet Line Primary Purification Pump 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 5

31 100 24 170 20 7

8 6

2 38 140 Current Dose Rate (mRem/hr)

<1 4

<1 2

6 11 10

<1 1

7

• Survey Point removed due to dismantlement activities.

D-PLAN 5-21 November 2010 1

5. PLANT STATUS - (cont'd)

Survey Point #

39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 72 73 74 75 Survey Point Location Exhaust Ventilation Duct Reactor Bldg. Grade Level N Shield Wall 1A Fuel Element Storage Well Pump 1B Fuel Element Storage Well Pump FESW Filter Discharge Line FESW System Cooler Hydraulic Valve Actuation System Header Base of Hydraulic Valve Accumulator Wall at Electrical Penetration Handrail on NW Nuclear Instrumentation (NI)

Platform Shield Wall on N NI Platform Primary Purification to OHST Line Above Primary Purification Cooler Inlet Valve Cold Leg of Reactor High Level Transmitter Line Seal Injection Reservoir Reactor Cavity Drain Line lA Core Spray Pump Discharge Line Reactor Water Level Sightglass Line Reactor Water Level Sightglass Line Reactor Bldg. Mezzanine Level N Shield Wall Steam Trap Reactor Bldg. Mezzanine Level NW Wall Fuel Element Storage Well Line Fuel Element Storage Well Line Fuel Element Storage Well Line Fuel Element Storage Well Skimmer Line Wall near Fuel Transfer Canal Drain Relief Valve Platform at Level Transmitter Shutdown Condenser Shutdown Condenser Condensate Line l B Retention Tank 1 A 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 Initial Dose Rate (mRem/hr) 9 6

70 80 180 1000 60 24 30 100 4

6 25 46 30 44 10 180 100 4

23 400 420 60 90 35 80 11 6

300 130 24 25 18 70 32 28 Current Dose Rate (mRem/hr)

<1

<1 9

10 24 140 11 3

2 8

<1

<1 5

7 4

5 22 10 14 3

5 24 4

1 3

3 10 18 4

• Survey Point removed due to dismantlement activities.

D-PLAN 5-22 November 2010

5. PLANT STATUS - (cont'd)

Survey Point H 76 77 78 79 80 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 100 Survey Point Location 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 1 B Forced Circulation Pump Suction Line 11B Forced Circulation Pump Suction Line 1 A Forced Circulation Pump Suction Line 1A Forced Circulation Pump Suction Line 1 A Forced Circulation Pump Discharge Line Feedwater Line in Forced Circulation Cubicle 1 A Forced Circulation Pump Handrail at 1A Forced Circ. Pump Suction Line 1 A Forced Circulation Pump Discharge Line 1 A Forced Circulation Pump Discharge Line 1 A Forced Circulation Pump Suction Line Initial Dose Rate (mRem/hr) 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) 42 25 2

50 60 34 16 2

5 11 65 140 55 43 41 12 7

7 34 38 16

• Survey Point removed due to dismantlement activities.

D-PLAN 5-23 November2010 I

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 will continue to be possessed and used. Additionally, sources will be used as authorized without restriction to chemical or physical form for sample analysis, instrument calibration and 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.

The Radiation Monitoring system performs the following functions:

" Provides a permanent record of radioactivity levels of plant effluents.

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

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

" Provides continuous radiation surveillance in normally accessible plant areas.

" Provides portable instrumentation for use in conducting radiation surveys.

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

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

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 Nal scintillation detector. The Nal scintillation detector is D-PLAN 5-24 November 2010 1

5.

PLANT STATUS - (cont'd) 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 Laboratory-Type 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 5-25 November2010

6.

DECOMMISSIONING PROGRAM 6.1 OBJECTIVES The primary objective of the Decommissioning Program at LACBWR will be to safely monitor the facility and prevent any unplanned release of radioactivity to the environment. Some of the goals during the SAFSTOR period are as follows:

" To safely store spent fuel while minimizing risk to the public health and safety.

" To maintain a radiation protection program that ensures SAFSTOR activities have minimal effect on the environment, the general public, and site personnel.

" To maintain systems required during SAFSTOR period.

" To handle radioactive waste generated during the SAFSTOR period in accordance with plant procedures and applicable requirements.

" To limit plant personnel radiation exposure to levels as low as reasonably achievable (ALARA).

" To dismantle unused systems with assurance that no unacceptable environmental impacts or other adverse effects are created from these activities.

" To maintain qualified and trained staff.

6.2 ORGANIZATION AND RESPONSIBILITIES The organization of the SAFSTOR staff at LACBWR is as indicated in Figure 6.1. The staff may change as activities being performed vary and staffing needs change. The organization is directed by a Plant Manager, who reports directly to the Dairyland Power Cooperative Vice President, Generation. The individuals who report directly to the Plant Manager each have distinct functions in ensuring the safety of the facility during the SAFSTOR mode.

The Plant Manager is responsible for the safety of the facility and ISFSI, their daily operation and surveillance, long range planning, and licensing. Quality assurance activities and security control and support are provided by a Cooperative-wide quality assurance and security programs.

The Plant Manager is responsible for operation of any onsite security required as well as ensuring compliance with the Quality Assurance Program Description (QAPD). The Plant Manager is responsible to ensure that adequate staff is present to comply with the terms of the license, training commitments and responsibilities are met, and that the personnel reporting are fit for duty.

The Operations, Training/Relief Supervisor is responsible for the day to day activities of the Shift Supervisors and Operators. This supervisor is responsible for the coordination of all Technical Specification required tests. This supervisor is responsible for proper implementation of the LACBWR Training Program.

D-PLAN 6-1 November2010 I

6. DECOMMISSIONING PROGRAM - (cont'd)

The Shift Supervisor is responsible for operating the shift and ensuring that the facility is maintained in a safe and efficient manner. The Shift Supervisor directs and is responsible for all operations and maintenance activities occurring on shift. The Shift Supervisor ensures that routine rounds are made, logs are kept and equipment maintenance requests are properly initiated.

The Operators are responsible for the operation of the facility. They ensure that all equipment is operated in a proper manner consistent with the license and procedures of the facility. The Operators perform fuel handling operations and maintain qualification in compliance with their training certification. The Operators are responsible to ensure that procedural deficiencies discovered are identified. The Operators tour the facility when spent fuel is stored in the fuel element storage well and ensure that the fuel element storage well and supporting systems are operable and the cleanliness of the facility is maintained.

The Health and Safety/Maintenance Supervisor is responsible for the radiological health and safety of the general public in the area surrounding the plant as well as the safety of the staff and all visitors to the plant. The Health and Safety/Maintenance Supervisor ensures that all radiological and environmental monitoring programs are performed. This individual ensures that all radiation exposure controls are in place and ensures that contamination and daily, monthly and annual exposure limits on personnel are complied with. The Health and Safety/Maintenance Supervisor is responsible for the ALARA program and ensures that all personnel at LACBWR comply with ALARA requirements. This supervisor also assigns the day-to-day work of the Health Physics Technicians, Instrument Technicians, Electricians, and Maintenance Mechanics through coordination with Forepersons of these groups.

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

The Instrument Technicians are responsible for maintaining the instrumentation within the facility and ISFSI necessary to safely store the spent fuel. They perform surveillance tests required as well as maintenance requests initiated on instrumentation.

The Electricians are responsible for maintaining electrical equipment in operating systems in accordance with procedures and completing maintenance requests and surveillance tests that are required. They are responsible for maintaining equipment within the plant used by other DPC facilities for system reliability. They are also responsible for electrical breaker maintenance and such other responsibilities as assigned by supervision.

The Maintenance Mechanics are responsible for the completion of mechanical maintenance tasks. They are responsible for the completion of maintenance requests and surveillance tests of a mechanical nature. They are responsible for the preventive maintenance program established on those systems necessary to maintain the SAFSTOR condition. The Maintenance Mechanics are responsible for overall maintenance on plant equipment used by other DPC facilities for system reliability.

D-PLAN 6-2 November2010 I

6. DECOMMISSIONING PROGRAM - (cont'd)

The Administrative Assistant is responsible for overall administration of LACBWR. The Administrative Assistant processes and maintains file records of all LACBWR programs, procedures, document changes, budget information, written communications, and numerous other records related to SAFSTOR activities. The Administrative Assistant ensures that clerical functions are performed adequately. This person maintains all budget expense and project accounts and coordinates preparation of the LACBWR budget. Duties 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 are made available to the Administrative Assistant as needed, and assist in the clerical tasks at LACBWR. Such additional personnel are qualified to perform required communication functions and are assigned other tasks, as necessary, by the Administrative Assistant.

The Licensing Engineer is responsible for facility licensing during the SAFSTOR condition, during preparations for dry cask storage, and eventual license termination activities. The Licensing Engineer is the principal liaison on behalf of the Plant Manager for contact with the Nuclear Regulatory Commission and other regulatory agencies.

The Radiation Protection Engineer is responsible for radiation protection, projections and trending. This engineer is responsible for working with the Health and Safety/Maintenance Supervisor in ensuring that an aggressive ALARA program is carried out and that contamination and background radiation exposure are reduced as low as reasonably achievable during the SAFSTOR period.

The Reactor Engineer performs administrative functions in support of SAFSTOR operations and assists with plans for dry cask storage. The Reactor Engineer monitors stored fuel conditions and provides nuclear safety oversight. This engineer monitors the condition of structures, systems and components important to safety. Presently, the Reactor Engineer is assigned responsibilities of the LACBWR Accountability Representative as discussed in Section 6.8.

The Safety Review Committee (SRC) is the offsite review group responsible for oversight of facility activities. A quorum of four persons, including the chairman, is required. No more than a minority of the quorum shall have line responsibility for operation of the facility. The SRC shall meet at least once per year.

The Operations Review Committee (ORC) is the onsite review committee and is responsible for the review of day-to-day operations. A quorum of at least four individuals drawn from the management staff at the site, including the Plant Manager or designated alternate as chairman, is required. The ORC shall meet at least once per calendar quarter. The SRC and the ORC review material as required by the QAPD.

6.3 CONTRACTOR USE The use of contractors at LACBWR will continue as required throughout the SAFSTOR and DECON periods. The use of contractors will complement areas where DPC expertise or staffing is inadequate to perform specific tasks. Contractor employment during the SAFSTOR and D-PLAN 6-3 November 2010 1

6. DECOMMISSIONING PROGRAM - (cont'd)

DECON periods will continue to be governed by the requirements of the QAPD. Contractors will be selected in each case on a basis of ability, price, past performance, and regulatory requirements.

DPC will retain full responsibility for the performance of contractor tasks and will provide the supervision necessary to ensure that the tasks performed by contractors are in full compliance with the QAPD, the purchase agreement, and other appropriate regulations.

6.4 TRAINING PROGRAM 6.4.1 Training Program Description 6.4.1.1 LACBWR has established General Employee Training (GET) requirements for all personnel who may be assigned to perform work at LACBWR.

6.4.1.2 In addition to GET, programs have been designed to initially qualify personnel, and maintain their proficiency, in the following areas:

1) Health Physics Technician (HPT)

2) Operator

3) Certified Fuel Handler (CFH)

These programs and requirements will change when all spent fuel is at the ISFSI. CFH training and proficiency will no longer be required. Training in ISFSI administration, security, monitoring, and maintenance will be fully implemented.

6.4.1.3 Special infrequently performed evolutions relating to decommissioning activities may be included for training as they approach. These evolutions may typically be:

1) Cask Handling

2) Systems Internals and Equipment Decontamination and Dismantling

3) Special Tests

4) Any other evolution determined by plant management to require special training.

6.4.2 General Employee Training (GET) 6.4.2.1 All personnel, either assigned to LACBWR or who may be assigned duties at LACBWR will receive GET commensurate with their assignment. This training will include, as appropriate:

1) Emergency Plan Training

2) Security Plan Training

3) Radiation Protection Training

4) Quality Assurance Training

5) Respiratory Protection Training

6) Industrial Safety, First Aid, and Fire Protection D-PLAN 6-4 November2010 I

6. DECOMMISSIONING PROGRAM - (cont'd) 6.4.3 Technical Training The following areas consist of a formal initial training program, followed by a recurring continuing training program.

6.4.3.1 The Health Physics Technician (HPT) Initial Training Program consists of the following topics:

1) Science Training a) Nuclear Theory b) Chemistry i) Non-radiological ii) Radiochemistry c) Radiological Protection and Control (including surveys)

2) Systems Training a) Effluent Systems Sampling and Control

3) Emergency Plan Training a) Onsite Survey Team Member b) Nearsite Survey Team Member c) Duty HP d) Re-entry Team Members e) PASS Sampling f) Medical Emergency

4) Environmental Program

5) Waste Disposal

6) Personnel Monitoring, including Internal Deposition Counting

7) Respiratory Protection Program

8) Radiation Monitoring and Instrumentation

9) Administrative Requirements

10) First Aid Training 6.4.3.2 The Health Physics Technician (HPT) Continuing Training Program consists of the following:

1) The program will be of 12-month duration, and will be repeated each 12 months.

2) Health and Safety management will review significant industry events and distribute, as required reading to all technicians, those events determined to be applicable to LACBWR HPT's.

D-PLAN 6-5 November 2010 1

6. DECOMMISSIONING PROGRAM - (cont'd)

3) Health and Safety management will review LACBWR events and distribute, as required reading to all technicians, those events determined to be relevant and significant.

4) Emergency Plan Training commensurate with duties.

5) Procedure changes will be reviewed by Health and Safety management and those determined to be relevant to the performance of a technician's duties will be distributed as required reading.

6) The Health and Safety/Maintenance Supervisor may initiate additional training for the technicians at any time. This training could be for, but not limited to, any of the following:

a) Equipment upgrade/replacement.

b) Infrequent and/or important tasks.

c) Significant procedure or department policy changes.

d) Significant performance problems.

7) The Health and Safety/Maintenance Supervisor will ensure all Journeyman Technicians successfully complete the HP continuing training. Records of satisfactory completion will be maintained by Health and Safety management. The continuing training will cover the following topics.

a) Intra-laboratory comparisons in radio-chemistry (crosscheck analysis).

b) Emergency Plan training.

8) A meeting will be conducted, at least semiannually, by the Health and Safety/Maintenance Supervisor for all technicians for the purpose of discussing any pertinent information on the following topics:

a) Significant Plant/Industry events.

b) Equipment Changes.

c) Management/Technician Concerns.

d) Performance Problems.

e) Minutes of these meetings will be taken.

6.4.3.3 Operator Training Program

1) Operators assigned to LACBWR will be qualified to perform the duties of Auxiliary Operator (AO) and Control Room Operator (CRO).

2) The Operator Initial Training Program consists of the following:

a) Part I - The initial GET and Indoctrination is presented to give the new employee background information concerning the LACBWR organization, radiation safety, payroll practices, and general plant description and administration.

b) Part II - The second part of the training program, "Initial Plant Qualification Program," provides a comprehensive outline of material considered necessary for D-PLAN 6-6 November 2010 1

6. DECOMMISSIONING PROGRAM - (cont'd) the training of individuals to qualify them for all operator duties. Periodic written and/or oral examinations, plus actual demonstrations of proficiency in practical factors, will be required of the trainees to determine their progress in the program.

c) Operator Sciences Training i) Nuclear Theory ii) Radiological Protection and Control iii) Electrical Theory, as applied to operators iv) Chemistry d) Operator Systems Training i) Plant Specific Systems ii) Design Bases iii) Flow Paths iv) Components v) Instrumentation and Control vi) Operational Aspects e) Control Room Training by familiarization and manipulation under the supervision of a qualified CRO, consisting of training and exercises which apply the operating philosophy, procedures, and attitudes needed as an operator at LACBWR. The Control Room Training will be documented in the operator Practical Factors Record. Topics include:

i) Normal operations ii) Malfunctions iii) Surveillances iv) Procedures v) Technical Specifications vi) Emergency response actions f) Emergency Training i) Emergency Plan and EPP's ii) Plant Emergency Procedures iii) Review of Incident Reports and LER's.

g) In addition, operator trainees will take part in the LACBWR Continuing Training Program when assigned to an operating crew. This program is intended as a review for personnel and as such is not intended to serve as the sole means of training for operator trainees. All quiz and examination scores attained by trainees in the requalification program will be used to aid the trainee and not to determine his status in the program. No lecture attendance or retraining requirements are to be based on test results.

h) The candidate will normally get the necessary signatures for the Auxiliary Operator Watch Card, then Control Room Operator Watch Card and, while standing these watches, work to complete each Progress Card. As the Progress Cards are completed, the training personnel shall prepare and administer a written exam. The trainee must receive a score of > 80% to pass exam.

D-PLAN 6-7 November2010 I

6. DECOMMISSIONING PROGRAM - (cont'd) 6.4.3.4 Certified Fuel Handler Training Program. A training and certification program has been implemented to maintain a staff properly trained and qualified to maintain the spent fuel, to perform any fuel movements that may be required, and to maintain LACBWR in accordance with the possession-only license while spent fuel is stored in the Fuel Element Storage Well.

This program provides the training, proficiency testing, and certification of fuel handling personnel. A detailed description of the Certified Fuel Handler Program is provided in Section 10.

6.4.4 Other Decommissioning Training It is anticipated that other technical topics will be presented to personnel on an as-needed basis.

Current administrative guidelines will be followed to establish new procedures and to ensure the training is completed.

6.4.5 Training Program Administration and Records The LACBWR Plant Manager is responsible for ensuring that the training requirements and programs are satisfactorily completed for site personnel. The Operations, Training/Relief Supervisor is responsible for the organization and coordination of training programs, for ensuring that records are maintained and kept up-to-date, and assisting in training material preparation and classroom instruction.

6.5 QUALITY ASSURANCE Decommissioning and SAFSTOR activities will be performed in accordance with the NRC-approved Quality Assurance Program Description (QAPD) for LACBWR. The QAPD has been developed to assure safe and reliable operation of LACBWR in a SAFSTOR condition and transition to an Independent Spent Fuel Storage Installation. The program is designed to meet the requirements of 10 CFR 50, Appendix B as applicable to the possession-only license condition and 10 CFR 72, Subpart G as applicable to the onsite dry cask storage of spent nuclear fuel.

The QAPD addresses all 18 criteria of 10 CFR 50, Appendix B, and 10 CFR 72, Subpart G, and applies to all activities affecting the functions of the structures, systems, and components that are associated with a possession-only license condition using a graded approach, and the Dry Cask Storage Project, respectively. These activities include design, installation, operations, maintenance, repair, fuel handling, testing, modifications, and radioactive waste shipments.

Design and fabrication of storage and shipping casks for radioactive material are addressed by the Dry Cask Storage System Certificate of Compliance holder's quality assurance program and will not be conducted under the QAPD. Safety Related as defined in 10 CFR 50.2 is no longer applicable in the possession-only license condition. A graded approach is used to implement the QAPD by establishing managerial and administrative controls commensurate with the complexity and regulatory requirements of the activities undertaken.

Scheduled activities during SAFSTOR shall be performed within schedule intervals. A schedule interval is a time frame within which each scheduled activity shall be performed, with a maximum allowable extension not to exceed 25 percent of the schedule interval.

D-PLAN 6-8 November 2010 1

6. DECOMMISSIONING PROGRAM - (cont'd)

For Important to Safety (ITS) activities, as defined in 10 CFR 72, Subpart G, a Quality Assurance Project Plan (QAPP) was established to define the quality assurance requirements to be implemented during the Dry Cask Storage Project at the LACBWR site. The QAPP does not supplant the QAPD which is used to assure safe and reliable operation of the LACBWR plant in a SAFSTOR condition. The QAPP utilizes the QAPD as the base document of the Dry Cask Storage Project's overall quality assurance program. The QAPP references and provides clarification to each applicable section of the QAPD that collectively meet the quality assurance requirements of 10 CFR 50 Appendix B, 10 CFR 71 Subpart H, and 10 CFR 72 Subpart G.

The QAPP applies to all activities associated with the design, fabrication, installation, and preparation for operation of an Independent Spent Fuel Storage Installation and any related plant modifications and other site activities as designated by the Plant Manager. Design and fabrication of the Dry Cask Storage System (DCSS) will not be performed under the QAPP; however, selection, qualification, and performance-based overview of the selected DCSS designer and DCSS fabrication will be conducted in accordance with the QAPP.

6.6 SCHEDULE The current schedule for decommissioning activities at LACBWR is depicted in Figure 6.2.

Following final reactor shutdown in April 1987, the transition from operating plant to possession-only facility required numerous administrative changes. Staff level was reduced, license required plans were revised, and operating procedures were curtailed or simplified as conditions and NRC approval allowed. The LACBWR Decommissioning Plan was approved in August 1991, and the facility entered the SAFSTOR mode. License renewal granted at the same time accommodated the proposed SAFSTOR period for a term to expire March 29, 2031. At the time of the original Decommissioning Plan in 1987, DPC anticipated the plant would be in SAFSTOR for a 30-50 year period.

To make better use of resources during the SAFSTOR period, some incremental decontamination and dismantlement activities were desirable. By Confirmatory Order from the NRC in 1994, changes in the facility meeting 10 CFR 50.59 requirements were permitted and limited gradual dismantlement progressed. As of November 2008, approximately 2 million pounds of material related to the removal of unused components or whole systems, completed in over 100 specific approved changes to the facility, have been reprocessed or disposed of as dry active waste. This total does not include reactor vessel and B/C waste disposal.

The 2-year Reactor Pressure Vessel Removal (RPV) Project was completed in June 2007 with disposal of the intact RPV at the Barnwell Waste Management Facility (BWMF). Disposal of the RPV was completed at this time prior to the planned closing of BWMF to out-of-compact waste in July 2008. RPV removal was not specifically addressed in the original decommissioning schedule. The removal of this large component, as defined in 10 CFR 50.2, was an activity requiring notice be made pursuant to 10 CFR 50.82, Termination of License, (a)(7). This notice was made by submittal to the NRC on August 18, 2005. Section 7.6 describes the RPV Removal Project in greater detail.

The original schedule indicated that during the SAFSTOR period, DPC expected to ship the activated fuel to a federal repository, interim storage facility, or licensed temporary monitored D-PLAN 6-9 November 2010 1

6. DECOMMISSIONING PROGRAM - (cont'd) retrievable storage facility. The timing of this action would be dependent on the availability of these facilities and their schedule for receiving activated fuel. In 2007, DPC began efforts to place an Independent Spent Fuel Storage Installation (ISFSI) on site by commencing the Dry Cask Storage Project. An onsite ISFSI is the available option that provides flexibility for license termination of the LACBWR facility. With respect to the federal repository option, a marker for transport of spent fuel offsite has also been added to Figure 6.2 as best available information can provide.

As to another option, DPC is a part of the consortium of utilities that formed the Private Fuel Storage (PFS) Limited Liability Company for the sole purpose of developing a temporary site for the storage of spent nuclear fuel for the industry. The Nuclear Regulatory Commission issued Materials License No. SNM-2513 pursuant to 10 CFR 72, dated February 21, 2006, for the PFS Facility.

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

This phase of decommissioning activity does not result in significant environmental impacts and has been reviewed as documented in the "Generic Environmental Impact Statement (GEIS) on Decommissioning of Nuclear Facilities," NUREG-0586, Supplement 1, November 2002. As stated in the GEIS, licensees can rely on information in this Supplement as a basis for meeting the requirements in 10 CFR 50.82(a)(6)(ii). The GEIS characterizes the environmental impacts resulting from this decommissioning activity as generic and small. Potential site-specific environmental impacts not determined in the GElS will be addressed in the License Termination Plan (LTP) for LACBWR.

DPC's review and analysis found that the Nuclear Decommissioning Trust (NDT) was sufficiently funded to allow dismantlement to begin in 2012, immediately after spent fuel removal is completed. Costs of the metal removal project will be funded from the NDT. DPC's approved strategy requires continuing evaluation of the costs of the decommissioning activity as it progresses. During this time the LTP will be formulated determining the disposition of concrete structures and site end use. The LTP will include an updated site-specific estimate of remaining decommissioning costs. DPC's decommissioning strategy for LACBWR with accelerated systems removal provides flexibility in that provisions are afforded to evaluate the costs and benefits of alternative methodologies for concrete removal, and delay LTP implementation if necessary to assure adequate NDT funds are available for the final decommissioning process. Figure 6.2 depicts the revised schedule.

D-PLAN 6-10 November 2010

6. DECOMMISSIONING PROGRAM - (cont'd) 6.7 SAFSTOR FUNDING AND DECOMMISSIONING COST FINANCING 6.7.1 SAFSTOR Funding Pursuant to 10 CFR 50.54(bb), Dairyland Power Cooperative (DPC) has promulgated the following SAFSTOR spent fuel management and funding plan for LACBWR. This plan applies also to ISFSI operations.

Independent of funding costs for SAFSTOR, DPC has established the Nuclear Decommissioning Trust (NDT) and reports annually to the Nuclear Regulatory Commission the status of NDT funds. DPC understands that none of the funds in the NDT may be used for spent fuel removal or for developing an Independent Spent Fuel Storage Facility (ISFSI). DPC has no plans to use any of the NDT funds for an ISFSI or for spent fuel removal purposes.

DPC continues to fund the expense of SAFSTOR activities, including wet or dry fuel storage costs, from the annual operating and maintenance budget. As part of generation expenses, SAFSTOR costs are recovered in rates that DPC charges distribution cooperative members under long-term, all requirements wholesale power contracts. DPC's rates to member cooperatives are annually submitted to the United States Rural Utilities Service (RUS) as part of RUS oversight of DPC operations. DPC is required by RUS lending covenants and RUS regulations to set rates at levels sufficient to recover costs and to meet certain financial performance covenants. DPC has always met those financial performance covenants and has satisfied the RUS regulations concerning submission and approval of its rates.

DPC's 25 member cooperatives set their own rates through participation in the DPC board of directors. The operations and maintenance budget approved by the DPC Board, and incorporated into rates submitted to and approved by the RUS, will be funded and available to pay SAFSTOR expenses as incurred.

DPC has found no need to separately fund SAFSTOR costs outside the regular operating and maintenance budget. SAFSTOR costs are relatively small compared to DPC's annual O&M costs for generation and transmission facilities, and DPC has continued the long-standing policy of recovering SAFSTOR costs as part of regular rates. DPC has seen no need to change the funding plan for SAFSTOR under those circumstances.

In August 2007, the DPC Board of Directors authorized to proceed with a dry cask storage project, including the planning, engineering and licensing, and construction and operation of an onsite ISFSI for the LACBWR spent fuel. Funds for ISFSI construction and operation will be generated through DPC operating and maintenance budgets. DPC does not intend to use any funds from the NDT for dry cask storage purposes.

DPC's annual budget for operating and maintenance activities at LACBWR accommodates SAFSTOR activities and includes funds for performing limited dismantlement at the LACBWR facility. Accomplishing limited dismantlement activities during SAFSTOR reduces the amount that will ultimately be necessary for decommissioning LACBWR after removal of the fuel. This approach takes advantage of the collective experience and familiarity of the LACBWR staff with the plant, and builds further conservatism into the funding plan for ultimate decommissioning of the facility.

D-PLAN 6-11 November 2010

6. DECOMMISSIONING PROGRAM - (cont'd) 6.7.2 Decommissioning Cost Financing In late 1983, the Dairyland Power Cooperative Board of Directors resolved to provide resources for the final dismantlement of LACBWR. DPC began making deposits to a decommissioning fund in 1984. The Nuclear Decommissioning Trust (NDT) was established in July 1990 as an external fund outside DPC's administrative control holding fixed income and equity investments.

The NDT, with requisite funding and accumulated earnings, was established to assure adequate funds would be available for the final decommissioning cost of LACBWR.

The cost of DECON was based on the selection of total radiological cleanup as the option to be pursued for LACBWR. At the time of preparation of this plan in 1987, decommissioning cost was based on studies by Nuclear Energy Services, Inc., available generic decommissioning cost guidance, and technology as it existed. In the Safety Evaluation Report dated August 7, 1991, related to the order authorizing decommissioning and approval of the Decommissioning Plan, the NRC found the estimate of $92 million in Year 2010 dollars reasonable for the final dismantling cost of LACBWR.

An improved site-specific decommissioning cost study was performed by Sargent & Lundy (S&L) in 1994 and provides basis for the current cost estimate and funding. The S&L study determined the cost to complete decommissioning to be $83.4 million in Year 1994 dollars with commencement of decommissioning assumed to occur in 2019. A cost study revision completed in July 1998 placed the cost to complete decommissioning at $98.7 million in Year 1998 dollars.

A cost study revision, prompted by significant changes in radioactive waste burial costs, as well as lessons learned on decontamination factors and methods, was prepared in November 2000 and placed the cost to complete decommissioning at $79.2 million in Year 2000 dollars. During 2003, the cost study was revisited again to include changes in escalation rates, progress in limited dismantlement, and a revised reactor vessel weight definition. This update placed the cost to complete decommissioning at $79.5 million in Year 2003 dollars.

In preparation for removal of the reactor pressure vessel (RPV), cost figures were brought current to $84.6 million in Year 2005 dollars. As of December 2006, NDT funds were approximately $83.4 million. NDT funds for B/C waste and RPV removals, approved by the Board of Directors, have been drawn in the amount of $18.2 million. Following B/C waste and RPV disposal a revision to the cost estimate was performed in September 2007 that placed the cost to complete decommissioning at $62.5 million in Year 2007 dollars.

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

D-PLAN 6-12 November 2010

6. DECOMMISSIONING PROGRAM - (cont'd)

Cooperative management believes that the balance in the nuclear decommissioning funds, together with future expected investment income on such funds, will be sufficient to meet all future decommissioning costs.

The DPC Board of Directors remains committed to assuring that adequate funding will be available for the final decommissioning of the LACBWR facility and ISFSI and is prepared to take such actions as it deems necessary or appropriate to provide such assurance, based upon its review of the most recent decommissioning cost estimate and other relevant developments in this area.

Every five years during the SAFSTOR period, a review of the decommissioning cost estimate will be performed in order to assure adequate funds are available at the time final decommissioning is performed.

6.8 SPECIAL NUCLEAR MATERIAL (SNM) ACCOUNTABILITY The LACBWR Accountability Representative is the person responsible for the custodial control of all SNM located at the LACBWR site and for the accounting of these materials. The representative is appointed in writing by the Dairyland Power Cooperative President & CEO.

The LACBWR spent fuel inventory is stored underwater in two-tiered storage racks within the Fuel Element Storage Well located in the Reactor Building or in dry storage casks located at the onsite ISFSI. 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 are 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 72, 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 while spent fuel is stored wet. With implementation of ISFSI operations fire protection planning for the LACBWR facility will adapt to the absence of fuel stored wet and will focus on ISFSI fire protection. As long as radiological hazards remain at the LACBWR facility, fire protection planning will be commensurate with the risks associated with the reduction in those radiological hazards.

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:

D-PLAN 6-13 November 2010

6. DECOMMISSIONING PROGRAM - (cont'd)

" 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 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 decontaminated 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 of the outside fire service.

The fire protection plan at the ISFSI is based on the fire hazards analysis performed in support of the Dry Cask Storage Project. The ISFSI fire hazards analysis demonstrated that the explosion and heat effects of credible fire and explosion hazards at the Genoa site will not significantly increase the risk of radioactivity release to the environment. Therefore storage of spent nuclear fuel at the LACBWR ISFSI is in accordance with 10 CFR 72.122(c) general design criteria. The D-PLAN 6-14 November2010 I

6. DECOMMISSIONING PROGRAM - (cont'd)

ISFSI fire protection features and administrative controls will ensure that the Genoa site fire and explosion hazards are acceptable and within the cask system design basis for fuel-loaded Vertical Concrete Casks (VCCs) located at or in route to the ISFSI.

6.9.2 Fire Protection Program The fire protection program for the ISFSI consists mainly of administrative controls to limit flammable liquids and combustible materials in the area of the fuel-loaded VCCs and is implemented by ISFSI procedures. There will be no organized fire brigade at the ISFSI.

Personnel monitoring dry cask storage conditions may extinguish incipient fires, but Genoa Fire Department will be summoned for fire emergencies at the ISFSI site.

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 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.

The ISFSI Administration Building is equipped with three 120-V AC, ceiling mounted, UL listed smoke detectors.

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.

D-PLAN 6-15 November 2010 1

6. DECOMMISSIONING PROGRAM - (cont'd) 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 Genoa Unit 3 jockey pump. Two HPSW diesel pumps provide fire suppression water when started manually or when started automatically by a decrease in HPSW pressure to <90 psig for HPSW Diesel Pump IA or

<80 psig for HPSW Diesel Pump lB. Fire suppression water can be supplied from Genoa Unit 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 lB 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 1 A 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 lB Diesel Generator Room is protected by a CO 2 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.

6.9.2.6 Portable Fire Extinguishers and Other Fire Protection Equipment. Dry chemical and C02 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.

The ISFSI is supplied with its own complement of portable fire extinguishers.

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 emergency locker. Spare sprinkler heads and other sprinkler equipment are located in the D-PLAN 6-16 November2010 I

6. DECOMMISSIONING PROGRAM - (cont'd)

Maintenance Shop emergency locker. Rechargeable 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 shall 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 Operations 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 Genoa Fire Department during an emergency at LACBWR or the ISFSI.

6.9.2.9 Reporting. Fire emergencies shall be documented under the following reporting guidelines:

1) Any fire requiring Fire Brigade response shall be reported by the Operations Shift Supervisor using a LACBWR Incident Report.

2) 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.

3) Any incident requiring outside fire service assistance within the ISFSI Controlled Area Boundary shall be reported by the ISFSL Security Shift Supervisor using an ISFSI Security Incident Report.

D-PLAN 6-17 November 2010 1

6. DECOMMISSIONING PROGRAM - (cont'd) 6.9.2.10 Training. Security badged personnel 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.

Operations and Security personnel have Fire Brigade responsibilities and are given basic practical fire fighting and specific fire protection program instruction annually. Fire Brigade members shall also participate in at least one drill annually.

ISFSI personnel will be termed designated employees, and as such will not be members of an organized fire brigade. These personnel will be properly trained to use portable fire extinguishers to fight incipient fires in the employee's immediate work area.

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 and ISFSI, security will be maintained at a level commensurate with the need to ensure 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, Security Force Training and Qualification Plan, Security Control Procedures, Fitness for Duty Program, Unescorted Access Authorization Program, and Behavior Observation Program. The Security Plan for Transportation of LACBWR Hazardous Materials is found in the Process Control Program. ISFSI security requirements are addressed and implemented as applicable.

6.11 TESTING AND MAINTENANCE OF SAFSTOR SYSTEMS Testing and maintenance continues for those systems designated as being required for SAFSTOR. Routine preventive maintenance is performed at specified intervals. Corrective maintenance is performed when identified as necessary. Instrument calibrations and other routine testing continue at specified intervals for equipment required to be operable during SAFSTOR.

The LACBWR Maintenance Rule Program implements requirements of 10 CFR 50.65. The program identifies structures, systems, and components (SSCs) to be monitored under the rule, establishes goals for those SSCs, and provides a process for corrective action implementation for failure of identified SSCs. When all spent fuel is at the ISFSI, Maintenance Rule Program requirements will no longer be applicable.

D-PLAN 6-18 November 2010 1

6. DECOMMISSIONING PROGRAM - (cont'd) 6.12 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 conditions are not deteriorating and environmental effects of the site are negligible.

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

" Specific area dose rates and contamination levels.

" Specified system piping and component contact dose rate.

" Radionuclide inventory in specified plant systems.

" 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.

6.12.2 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.

6.12.3 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.

6.12.3.1 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 concentration 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. With all spent fuel stored at the ISFSI, no concentration of Kr-85 will be available as a source of radioactivity in effluent releases.

D-PLAN 6-19 November 2010 1

6. DECOMMISSIONING PROGRAM - (cont'd) 6.12.3.2 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.

6.12.4 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.

6.13 RECORDS The Quality Assurance Program Description (QAPD) establishes measures for maintaining records which cover all documents and records associated with the decommissioning, operation, maintenance, repair, and modification of structures, systems, and components covered by the QAPD.

Any records which are generated for the safe and effective decommissioning of LACBWR and the ISFSI will be placed in a file explicitly designated as the decommissioning file. Examples of records which would be required to be placed in the decommissioning file are:

" Records of spills or spread of radioactive contamination, if residual contamination remains after cleanup.

" Records of contamination remaining in inaccessible areas.

" Plans for decontamination (including processing and disposal of wastes generated).

" Baseline surveys performed in and around the LACBWR facility and ISFSI.

" Analysis and evaluations of total radioactivity concentrations at the LACBWR facility and ISFSI.

" Records for ISFSI construction, dry cask storage system fabrication, and dry cask loading.

" Any other records or documents, which would be needed to facilitate decontamination and dismantlement of the LACBWR facility or ISFSI and are not controlled by other means.

D-PLAN 6-20 November 2010

0.

LA CROSSE BOILING WATER REACTOR SAFSTOR STAFF I PLANT MANAGER I ISFSI OPERATIONS & SECURITY *

+

ORC, SRC, QA I

-I-I I

OPERATIONS, TRAINING/RELIEF SUPERVISOR 4 Shift Supervisors 6 Operators HEALTH & SAFETY /

MAINTENANCE SUPERVISOR 3 Health Physics Technicians 2 Instrument Technicians 2 Electricians 3 Mechanics ADMINISTRATIVE ASSISTANT Additional Administrative Personnel as Needed LICENSING ENGINEER REACTOR / RADIATION PROTECTION ENGINEER **

Assumes Cooperative-wide Security & QA

• Duties to be performed by existing LACBWR staff and security force.

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

D-PLAN FIGURE 6.1 November2010

D-PLAN FIGURE 6.2 November 2010

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. With all spent fuel stored at the Independent Spent Fuel Storage Installation (ISFSI) in seal-welded canisters protected by concrete overpacks, there will be no concentration of the isotope Kr-85 available for release from the LACBWR plant.

7.3 SIGNIFICANT SAFSTOR LICENSING ACTIONS DPC's authority to operate LACBWR under Provisional Operating License DPR-45, pursuant to 10 CFR Part 50, was terminated by License Amendment No. 56, dated August 4, 1987, and a possess but not operate status was granted. The Decommissioning Plan was submitted December 1987 with a chosen decommissioning alternative of SAFSTOR. License Amendment No. 63, dated August 18, 1988, amended the operating license to Possession-Only License DPR-45 with a term to expire March 29, 2003.

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

D-PLAN 7-1 November 2010 1

7. DECOMMISSIONING ACTIVITIES - (cont'd)

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.

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

License Amendment No. 69, containing the SAFSTOR 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.

License Amendment No. 71 was submitted July 28, 2009, requesting changes to the LACBWR Appendix A, Technical Specifications in support of the LACBWR Dry Cask Storage Project.

The request seeks approval of a revised definition of FUEL HANDLING, approval of a reduction of the minimum water coverage over stored spent fuel from 16 feet to 11 feet, 6/2 inches, and a small number of editorial changes to clarify heavy load controls and reflect inclusion of the cask pool as part of an "extended" Fuel Element Storage Well. These changes were requested to accommodate efficient dry cask storage system loading operations and reduce overall occupational dose to personnel during these operations.

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.

7.4 AREA AND SYSTEM DECONTAMINATION The decontamination program during the SAFSTOR 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.

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 dpml/100 cm 2). 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.

D-PLAN 7-2 November 2010 1

7. DECOMMISSIONING ACTIVITIES - (cont'd)

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.

7.5 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 in accordance with 10 CFR 50.59 requirements.

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

7.6 REACTOR PRESSURE VESSEL REMOVAL DPC entered contract agreement July 2005 with Duratek, Inc. (later to become Energy Solutions, LLC) for removal and disposal of the intact Reactor Pressure Vessel (RPV) at the Barnwell Waste Management Facility (BWMF) in South Carolina. Major subcontractors included Bigge Power Constructors, ARES Corp., Bluegrass Concrete Cutting, Inc., Pacific Intemational Grout Co., and Patent Construction Systems. Engineering and project development progressed through 2005 and 2006. The RPV, forced circulation loop piping, and pumps were filled with low density cellular concrete (LDCC) in March 2006. The LDCC fixed in place components and contamination internal to the RPV. Site mobilization and major project activity commenced September 2006.

The opening in the Reactor Building (RB) required removal of one polar crane runway support column. Restoration of the polar crane to full capacity by installation of an alternate support plate and components was completed to allow use of the polar crane in support of project D-PLAN 7-3 November 2010 1

7. DECOMMISSIONING ACTIVITIES - (cont'd) activities. Three upper cavity shield plugs, 15' diameter, 15" thick, weighing 30 tons each, were staged, cut into 18 pieces, and removed to clear obstruction to the project. Portions of the concrete floor at elevation 701' were cut and removed to clear travel path obstruction and provide access to the octagonal biological shield (bio-shield) structure. Two floor and beam shoring supports were installed.

Core drilling was performed in areas of the RB wall and the bio-shield openings for rigging and cutting wire access. Cores were drilled to allow precise diamond wire saw cuts that created manageable sized blocks. An opening in the steel and concrete exterior RB wall was completed, then closed by installation of a weather-tight, insulated, roll-up, bi-parting door in November 2006. The RB opening (described in Section 4.2.1) and bi-parting door are depicted in Figures 4.6 and 4.7.

A 10'-6" opening was made in the 4' to 6' thick concrete bio-shield from elevation 701' down to elevation 667'. From access in the cavity, RPV nozzles and appurtenances were cut from the RPV to near bottom dead center and to a critical diameter of 119" by March 2007. Temporary lifting device erection and installation began as nozzle cutting was being completed.

The grout-filled RPV weighing 370,000 lbs. was disconnected from its support, lifted 20',

translated outside the RB, and placed upright into a staged steel cylindrical package. The package with RPV was filled with concrete, seal welded, down-ended, and heavy-hauled to an on-site rail siding. The RPV package weighing 624,500 lbs. was loaded onto a special transport rail car, and shipped with final burial at the BWMF completed June 6, 2007.

7.6.1 Temporary Lifting Device A Temporary Lifting Device/Gantry Rail System (TLD) was erected and installed inside and outside the RB. The TLD system consisted of a temporary runway structure and rolling trolley which incorporated hydraulic strand jacks for lifting the RPV. The runway structure consisted of 37-feet girders inside the RB and 74-feet girders outside the RB. The runway structure design inside the RB met NUREG-0612 criteria. The runway structure design outside the RB was not required to meet NUREG-0612 criteria, as NUREG-0612 pertains to lifts and equipment inside buildings where spent fuel is stored.

The trolley was a moveable platform with four two-wheeled bogie end trucks (8 total double flanged wheels) designed to run on the box girder rails. Two of the trucks had electric mechanical drives. Each drive consisted of a gearbox, motor, and brake. There were two driven/braked wheels in the 8 wheel set. The brake was automatically set when the momentary directional motion switch was released to the neutral position. The trolley had two travel speeds; 1.62 feet per minute (FPM) and 6.80 FPM. Both travel speeds were very slow. At the slower speed it would have taken over an hour to traverse the runway from south to north. Two hydraulic strand jack hoisting systems were mounted on top of the trolley platform. The strand jack systems were independent from each other and were specially fabricated to meet the specifications for the LACBWR RPV lift and transport. Hoisting speed was 0.5 FPM. The strand jacks were comprised of 36 strands per jack; failure of any given strand would not result in loss of control of the suspended load. Failure of over 75% of the strands would have had to occur before the remaining strands could not carry the load. Two separate electrical sources were used to power the two strand jack power packs and one trolley drive system through three D-PLAN 7-4 November2010 I

7. DECOMMISSIONING ACTIVITIES - (cont'd) dedicated load disconnect switches. The strand jack system was designed such that the load would remain secured at the height lifted upon loss of power or hydraulic pressure. The trolley assembly was designed to meet NUREG-0612 criteria.

The TLD was constructed of components within the Bigge equipment inventory along with new fabricated assemblies. Prior to TLD use for the RPV lift, a load test of 110% of the load lifted outside the RB (service load 639,000 lbs/test load 703,000 lbs) was conducted. Since a load test of 150% of the load lifted inside the RB (service load 380,000 lbs/test load 570,000 lbs) was less than the outside load test weight, the inside load test was not performed. The percent increases above static weight or service load were consistent with NQA-1 and ANSI N 14.6.

The custom built RPV attachment/handling fixture used inside the RB was load tested in accordance with ANSI N 14.6-1993, Section 7, "Special lifting devices for critical loads."

Section 7.3.1 (a) required the test load to be three times (3x) the weight the fixture would support.

The handling fixture load test was documented for record.

All TLD equipment was removed following RPV removal with the exception of two rocker bearing assemblies installed on the bio-shield at elevation 701' and two bearing assemblies mounted at the RB wall opening.

7.6.2 NUREG-0612 Compliance RPV removal project activities that occurred inside the RB were performed in compliance with NUREG-0612 due to the close proximity of the RPV lift to the stored spent fuel. Additional information relative to NUREG-0612, and TLD compliance with other applicable codes and standards, was provided in Bigge Document No. 2150-D 1. Compliance to NUREG-0612, Section 5.1, was met by implementing the following measures for each of the criteria:

5.1(1) The project implemented detailed operator training, handling system design, load handling instructions, and conducted equipment inspection to ensure reliable operation of the handling system.

5.1(2) The safe load travel path was included in procedures and operator training. The engineered load path was such that the RPV was not carried over or near irradiated fuel or important to safety (ITS) SSCs.

5.1(3) Mechanical stops were provided to prevent movement of heavy loads over irradiated fuel or in proximity to ITS equipment.

Guidance in NUREG-0612, Section 5.1 allows for certain deficiencies if alternate compensatory measures are credited; however, no deficiencies in defense-in-depth criteria were noted.

Additional measures were included that could be given credit for. The items providing defense-in-depth were: increased hoisting system reliability by providing an increased factor of safety, increased inspections, and significant spent fuel decay. The lift supervisor also held Certified Crane Operator (CCO), as designated by the National Commission for the Certification of Crane Operators.

D-PLAN 7-5 November 2010 1

7. DECOMMISSIONING ACTIVITIES - (cont'd)

Facility Change 30-06-08, "RPV Supports Disconnection, Lift and Transit," documented compliance with the seven criteria in NUREG-0612, Section 5.1.1, for overhead handling systems that handle heavy loads in the area of the spent fuel pool. The criteria fully addressed were: (1) safe load paths; (2) procedures; (3) crane operators training and qualification; (4) codes and standards for special lifting devices; (5) codes and standards for lifting devices not specifically designed; (6) codes and standards for crane inspection, testing and maintenance; and (7) codes and standards for crane design.

NUREG-0612, Section 5.1.4 discusses boiling water reactors, but does not address the RPV as a potential heavy load. Section 5.1.4 states that in addition to meeting the general guidelines of Section 5.1.1, either the lifting devices should meet the guidelines in Section 5.1.6, or the effects of a heavy load drop should be analyzed. TLD design met the requirements of Section 5.1.6.

The runway and trolley structural steel were designed to twice (2x) the design safety factor to meet the requirements of AISC and for SSE seismic conditions with the lifted load to the design criteria of Bigge Document No. 2150-D 1. Jacking components were designed to twice (2x) normal design safety factors of ANSI B30.1, and below the hook lifting devices were designed to twice (2x) normal design safety factors in accordance with ANSI N14.6-1993, Section 7. The in-place structure was designed for normal and seismic load cases with the lifted load to meet plant design basis acceptance limits.

TLD design met NUREG-0612, Section 5.1.6(3) requirements in that all interfacing lift points such as lifting lugs or trunnions had a design safety factor often times (1Ox) the maximum static plus dynamic load.

TLD design met the defense-in-depth approach. Although there were no deficiencies, certain additional measures could be given credit. The TLD was used solely for lifting and removal of the RPV. Compliance to NUREG-0612 was in the context that the TLD was not planned as new permanent plant equipment or replacing the existing polar crane, but was used as a Temporary Lifting Device to lift the RPV.

7.6.3 50.59 Evaluations RPV Removal Project work was performed under nine major Facility Changes (FCs) for which 50.59 Evaluations were conducted and are listed below:

" FC 30-06-05, LACBWR Reactor Pressure Vessel Grouting

" FC 37-06-31, Biological Shield Cutting and Removal

" FC 37-06-35, 50/5-Ton Polar Crane Runway Restoration

" FC 37-06-32, Reactor Building Modification

" FC 37-06-34, Reactor Building Restoration Activities FC 30-06-06, RPV Bottom Head Nozzle and Appurtenances Removal

" FC 30-06-07, Reactor Pressure Vessel Nozzle Removal D-PLAN 7-6 November2010 I

7. DECOMMISSIONING ACTIVITIES - (cont'd)

" FC 37-06-33, TLD/Gantry System Install, Test and Disassembly FC 30-06-08, RPV Supports Disconnection, Lift and Transit Calculations and analyses in support of the conclusions of these 50.59 Evaluations are listed in Section 7.6.4. RPV Removal Project work was performed under twenty-three additional FCs for which 50.59 Screens were adequate in determining that the proposed activities created no new failure modes or other adverse effects.

7.6.4 References 7.6.4.1 Bigge Document 2150-D1, "Engineering Design Basis, Engineering, Rigging and Onsite Transport Services, Phase 2 - Reactor Pressure Vessel Removal Project, La Crosse Boiling Water Reactor Nuclear Plant," April 26, 2007, Rev. 2.

7.6.4.2 Bigge Calculation No. 2150-C 10, "Strand Jack Trolley Adequacy," Rev. 1.

7.6.4.3 Bigge Calculation No. 2150-C30, "TLD Runway Structure," Rev. 2.

7.6.4.4 Bigge Calculation No. 2150-C50A, "RPV Lift Lug and Miscellaneous Hook Rigging,"

Rev. 0.

7.6.4.5 Bigge Calculation No. 2150-C50B, "RPV Head Adequacy," Rev. 0.

7.6.4.6 Bigge Calculation No. 2150-CTLD Seismic, "TLD Runway & Trolley Structure Seismic," Rev. 0.

7.6.4.7 ARES Calculation No. 0526301.1 1-S-001, "Regeneration of LACBWR 1982 Containment Building Model for Seismic and Structural Analysis," Rev. 0.

7.6.4.8 ARES Calculation No. 0526301.11-S-002, "Seismic Analysis of Modified LACBWR Containment Building with SAP2000," Rev. 1.

7.6.4.9 ARES Calculation No. 0526301.11-S-003, "Structural Analysis of LACBWR Modified Containment Building Outer Shield Wall to Support Crane Girder Loads During RPV Removal,"

Rev. 1.

7.6.4.10 ARES Calculation No. 0526301.11 -S-005, "Structural Reinforcing of the Containment Building Outer Shield Wall Opening to Maintain Polar Crane Capacity," Rev. 0.

7.6.4.11 ARES Calculation No. 0526301.11-S-006, "Structural Analysis for Shoring of Floor at El. 701' Due to Concrete Cutting Inside the Reactor Building," Rev. 0.

7.6.4.12 ARES Calculation No. 052630 1.11-S-007, "Structural Integrity Analysis of Spent Fuel Storage Well and Racks Inside the Reactor Building," Rev. 1.

D-PLAN 7-7 November 2010 1

7. DECOMMISSIONING ACTIVITIES - (cont'd) 7.6.4.13 ARES Calculation No. 0526301.11 -S-008, "Seismic Analysis of LACBWR's Temporary Lifting Device (TLD) Structure for Removing RPV from the Reactor Building," Rev.

0.

7.6.4.14 ARES Calculation No. 0526301.12-S-001, "Structural Analysis of Support Reinforcement for Bi-Parting Door at Containment Building Opening," Rev. 0.

7.6.4.15 ARES Report No. 0526301.11-002, "Phase 2, LACBWR Reactor Pressure Vessel Removal Structural Analysis and Design Criteria," Rev. 0.

7.7 B/C WASTE REMOVAL Included in the scope of work during the RPV project was removal of irradiated hardware and other B/C wastes. Processing of waste stored in the FESW began in April 2006. This waste consisted of 73 irradiated zircaloy fuel shrouds, 24 irradiated stainless steel fuel shrouds, 10 irradiated control rod blades, and 2 antimony-beryllium startup sources. A control rod extension shear, control rod blade crimper, and hydraulic crusher shear were used to process components into cylindrical waste liners. Startup sources were placed in liners without processing. The Duratek CNS 3-55 Shipping Cask was used in the transport of two liners of irradiated hardware waste and disposal was completed at the BWMF in July 2006. Other B/C waste included resins, filters, and waste barrel contents that were collected in three liners and shipped for disposal in June 2007.

7.8 DRY CASK STORAGE PROJECT The LACBWR Dry Cask Storage Project establishes an ISFSI under general license provisions of 10 CFR 72, Subpart K, on the Genoa site. The ISFSI is located 2,232 feet south-southwest of the Reactor Building center on land which was previously used for the access road between the two closed ash landfills of the Genoa site. The ISFSI will be used for interim storage of LACBWR spent fuel assemblies in the NAC International, Inc. (NAC) Multi-Purpose Canister (MPC) System. 10 CFR 72.212 requires a general licensee to conduct and document an array of reviews to confirm that the physical ISFSI site and the site organization are prepared to implement dry spent fuel storage and that the generically designed dry spent fuel storage cask chosen for use bounds applicable site-specific design criteria and conditions. This evaluation is documented in the LACBWR ISFSI 10 CFR 72.212 Report and meets the requirements set forth in 10 CFR 72.212(b)(2)(i), (b)(3), and (b)(4) that mandate such written evaluations prior to use of the cask system under a Part 72 general license.

Refer to the NAC-MPC Storage System Certificate of Compliance No. 1025 and Final Safety Analysis Report for details of the MPC-LACBWR design, operation, and safety analyses.

Decommissioning Plan Section 4.2.1 discusses modifications made to the Reactor Building for dry cask loading operations and Section 4.2.5 discusses the onsite ISFSI.

7.8.1 Cask Handling Crane American Crane and Equipment Corporation (ACECO) was contracted to supply an 85-ton capacity temporary cask handling system for the Dry Cask Storage Project. The temporary cask handling system consists of the cask handling crane being supplied by ACECO; and a temporary D-PLAN 7-8 November 2010 1

7. DECOMMISSIONING ACTIVITIES - (cont'd) runway system being supplied by Rigging International. The cask handling crane consists of a refurbished single failure proof trolley and hoist previously utilized at Maine Yankee designed, manufactured, and tested in accordance with ASME NOG-1-1998 and NUREG 0554. The temporary runway system is a new structure designed, manufactured, and tested in accordance with the requirements of ASME NOG-1-2004 for a Type 1 Crane (i.e. single failure proof crane).

Features are included in the crane design to assure that any credible failure of a single component will not result in the loss of capability to stop and hold the critical load.

Dry cask storage system component lifts will be made using the single failure proof cask handling crane. All other lifts of equipment in the Reactor Building are performed in accordance with the requirements of NUREG-0612, LACBWR heavy load control procedures, and activity-specific rigging plans.

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

The environmental impact of all completed or planned dismantlement activities is SMALL as determined by the "Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities (GEIS)," NUREG-0586, Supplement 1, November 2002. The environmental impact of ISFSI construction and operation at the LACBWR site is an activity not within the scope of the GEIS and is addressed in the licensing process for the activity.

Site-specific potential environmental impacts not determined in the GEIS are:

" Offsite land use activities

" Aquatic ecology as to activities beyond the operational area

" Terrestrial ecology as to activities beyond the operational area

" Threatened and endangered species

" Socioeconomic

" Environmental justice The License Termination Plan (LTP) for LACBWR will detail final dismantlement activities, processes for demolition of structures, site remediation, survey of residual contamination, and determination of site end-use. A final supplement to the Environment Report in support of the LTP will address all environmental impacts of the license termination stage.

D-PLAN 7-9 November 2010 I

8.

HEALTH PHYSICS During the SAFSTOR period of LACBWR, radiation protection and health physics programs will be provided to ensure the health and safety of LACBWR workers. The programs will also provide the necessary monitoring and control of radiological conditions to protect the health and safety of the general public and to ensure compliance with LACBWR license requirements. In addition, programs will be provided to maintain radiation exposures as low as reasonably achievable (ALARA) at LACBWR and the onsite ISFSI.

8.1 ORGANIZATION AND RESPONSIBILITIES The organization described below is the organization as it is expected to exist during the SAFSTOR activities. The organization may be changed slightly during the SAFSTOR period as staffing levels requirements change. Responsibilities assigned to a position which is deleted will be assigned to another individual in order to maintain continuity.

The LACBWR Plant Manager has the overall responsibility for all onsite activities including assurance that ALARA policies and the radiation protection program are carried out. He is the chairman of the ORC. He is also responsible for approving all plant procedures.

Health and Safety management shall provide the first-line supervision, training and technical assistance to the Health and Safety department. Management personnel will report directly to the Plant Manager. They shall assure that all ALARA policies and all aspects of the Radiation Protection Program are implemented. They shall also be members of the ORC. Health and Safety management will be responsible for all departmental budgeting and scheduling.

The Health Physics Technicians will perform chemical and radiological sampling, surveys and analysis as directed by Health and Safety management. In addition, they will also be responsible for conducting the personnel monitoring program, maintaining radiation protection records and monitoring work in progress within the radiologically restricted area.

8.2 ALARA PROGRAM 8.2.1 Basic Philosophy The radiation exposure criteria set forth shall be for the protection of personnel against radiation hazards arising from work associated with LACBWR. As good practice, no person under 18 years of age shall be employed by DPC to be occupationally exposed to ionizing radiation. A continuous effort should be made to reduce levels of radiation and radioactivity in order to maintain radiation doses at the lowest achievable value below the established limits of 10 CFR 20.1201.

A further goal of the Health and Safety Procedures in use at LACBWR shall be to reduce personnel exposures to radiation and radioactive material to As Low As Reasonably Achievable (ALARA).

D-PLAN 8-1 November 2010 1

8. HEALTH PHYSICS - (cont'd) 8.2.2 Application of ALARA 8.2.2.1 To obtain the goal of ALARA, the Total Effective Dose Equivalent (TEDE) to be received during a specific job and the total allowable for the year for the entire operation of the facility should be balanced.

8.2.2.2 The occupational dose received by an individual shall be considered with respect to his/her yearly internal and external accumulation. The individual's TEDE dose should be balanced with the TEDE dose received by the entire LACBWR work force including temporary DPC/contract employees to aid the overall ALARA program.

8.2.2.3 An ALARA review may be conducted if the following thresholds are expected to be exceeded:

1) Between 100 and 500 milliRem total collective deep dose equivalent (DDE) for performing a job.

2) Potential intake greater than 50 DAC-HRS for an individual and respiratory devices are not planned to be used. An ALARA review form used for this application should be governed by total Person Rem and DAC-HR estimates, based upon current surveys and job-time estimates, and total Person Rem for past similar jobs based upon an SWP dose accountability file.

8.2.2.4 An ALARA review shall be conducted if the following thresholds are expected to be exceeded:

1) Greater than 500 milliRem collective DDE for performing a job.

2) Potential intake greater than 100 DAC-HRS for an individual and respiratory devices are not planned to be used.

8.2.2.5 Documentation of ALARA engineering work and cost benefits shall be maintained in files.

8.2.2.6 Health and Safety management will conduct ALARA reviews of Actual versus Projected (goal) exposures. Person Rem exposures will be reviewed regularly with the Plant Manager.

Included should be a review of the effectiveness of specific steps that were taken to reduce radiation exposure (ALARA Engineering).

8.2.3 Radiation Exposure Limits Radiation exposure to individuals at LACBWR will be controlled and limited in accordance with the following:

8.2.3.1 Occupational Dose Limit Guideline. LACBWR will provide a guideline for the control of occupational dose to individual adults to the following annual limits. Any individual exceedance of these limits will require approval by the Plant Manager.

D-PLAN 8-2 November 2010 1

8.

HEALTH PHYSICS - (cont'd)

1) Total effective dose equivalent (TEDE) 2.5 Rem.

2) The sum of the deep-dose equivalent and the committed dose equivalent to any individual organ or tissue other than the lens of the eye being equal to 25 Rem.

3) An eye dose equivalent of 7.5 Rem.

4) A shallow-dose equivalent of 25 Rem to the skin or to each of the extremities.

5) The dose received by an embryo/fetus during the entire pregnancy due to the occupational exposure of a declared pregnant worker shall not exceed 0.25 Rem (250 mRem).

8.2.3.2 Occupational Dose Limit. LACBWR shall control the occupational dose to individual adults to the following annual limits. Any individual exceedance of these limits will require approval by the ORC.

1) Total effective dose equivalent (TEDE) less than 4 Rem.

2) The sum of the deep-dose equivalent and the committed dose equivalent to any individual organ or tissue other than the lens of the eye being equal to 40 Rem.

3) An eye dose equivalent of 12 Rem.

4) A shallow-dose equivalent of 40 Rem to the skin or to each of the extremities.

5) The dose received by an embryo/fetus during the entire pregnancy due to the occupational exposure of a declared pregnant worker shall not exceed 0.4 Rem (400 mRem).

8.2.3.3 The NRC establishes annual occupational dose limits to individual adults, except for Planned Special Exposures authorized under 10 CFR 20.1206, to the following:

1) Total effective dose equivalent (TEDE) 5 Rem.

2) The sum of the deep-dose equivalent and the committed dose equivalent to any individual organ or tissue other than the lens of the eye being equal to 50 Rem.

3) An eye dose equivalent of 15 Rem.

4) A shallow-dose equivalent of 50 Rem to the skin or to each of the extremities.

5) The dose received by an embryo/fetus during the entire pregnancy due to the occupational exposure of a declared pregnant worker should not exceed 0.50 Rem (500 mRem).

8.2.3.4 The Health and Safety Department shall be contacted as soon as possible when entry into an area of known or suspected high airborne radioactivity has or will occur. The Health and D-PLAN 8-3 November 2010 1

8. HEALTH PHYSICS - (cont'd)

Safety Department shall provide air sampling and recommend protective measures for entry to such areas. Engineering measures that can be taken to reduce or eliminate the airborne contamination areas are the recommended methods to reduce/eliminate internal deposition.

Respiratory protection will be used only after a review to ensure an excessive increase in the external dose received, due to the use of respirators, will not occur. Suspected exposure to airborne contamination for inhalation/ingestion or waterborne concentrations of radioactive materials shall be investigated by lung deposition counting and/or urinalysis.

8.2.3.5 An adult worker may be authorized to receive doses in addition to and accounted for separately from the doses received under the normal occupational limits. These are known as Planned Special Exposures and will be used only under unusual/emergency situations. Planned Special Exposures will be authorized only in accordance with 10 CFR 20.1206.

8.3 RADIATION PROTECTION PROGRAM The radiation protection program that will be utilized during the SAFSTOR period will be an extension of the program that was used during the period of reactor operations at LACBWR. This program is in compliance with the requirements of 10 CFR 20. Implementation of the radiation protection program will be done at LACBWR through Health and Safety procedures. The following section describes the radiation protection program.

8.3.1 Personnel Monitoring To ensure that the radiation exposure limits of 10 CFR 20 are not exceeded, a personnel radiation exposure monitoring system will be maintained. Two basic means shall be used to evaluate each individual's radiation exposure:

" Badges - to give integrated dose measurements over relatively long periods of time.

" Self-Reading Dosimeters - to give interim indication of accumulated doses.

Badges and self-reading dosimeters will be worn by all plant personnel entering the radiological controlled area. They will be worn at or above the waist and on the front of the body, unless the Health and Safety management specifies that the badges be worn differently. Extremity dosimetry will be worn by all personnel when conditions exist that could cause a significantly higher than whole body dose to be received by a worker's extremities.

Long-term visitors expecting to receive a radiation exposure of 50 mRem will be issued badges and dosimeters and will be monitored in the same manner as the regular plant personnel.

Casual and short-term visitors (those for whom exposures are expected to be insignificant) will be issued pocket dosimeters only.

Badge records received from the badge processor will be evaluated and maintained. Periodic quality testing of badges and pocket dosimeters will be conducted.

Bioassays will be performed in accordance with the requirements of 10 CFR 20.1204 and in conformance to the recommendations of Regulatory Guide 8.26, "Application of Bioassay for D-PLAN 8-4 November 2010

8. HEALTH PHYSICS - (cont'd)

Fission and Activation Products," and Regulatory Guide 8.32, "Criteria for Establishing a Tritium Bioassay Program."

The LACBWR internal lung deposition counter will be used to detect any internal lung contamination for:

" All new employees who will routinely work with radioactive material.

" Any individual suspected of having received any internal lung deposition.

" Upon termination of any employee who worked with radioactive material.

If it is determined that any employee has a significant internal lung deposition of any isotope, the individual may be required to submit a urine and/or fecal specimen.

All personnel leaving a restricted area will be required to conduct a personnel contamination survey using the contamination detection instrument provided at the exit.

8.3.2 Respiratory Protection Program A respiratory protection program will be maintained during the SAFSTOR period.

The Health and Safety/Maintenance Supervisor is responsible for the Respiratory Program at LACBWR. The Health and Safety/Maintenance Supervisor or designated alternate will evaluate the total job hazard, recommend engineering controls if appropriate, specify respiratory protection if control cannot be otherwise obtained, and forbid the use of respirators if conditions warrant.

The Health and Safety Department is responsible for the selection, care, and maintenance of all respiratory protection equipment that falls under the scope of the respiratory protection program.

The acceptable manner for limiting the internal exposure of personnel is to control radioactivity concentration in the air breathing zones. Whenever possible, this will be accomplished by the application of engineering control measures such as containment, decontamination, special ventilation equipment, and design. The use of personal respiratory protective equipment as a primary control is undesirable and is acceptable only on a non-routine basis or in an emergency situation.

Equipment such as hoods, blowers, and filtered exhaust systems will be used to provide controls for routine operations and, whenever possible, for non-routine operations. In some cases, such controls may be inadequate or impractical and the use of protective breathing apparatus will be approved on a short-term basis.

The periods of time for which respirators may be worn continuously, and the overall time of uses, should be kept to a minimum. The wearer shall leave the area for relief from respirator use in case of equipment malfunction, undue physical or psychological discomfort, or any other condition that, in the opinion of the user, his supervisor or the Health and Safety Department, might cause significant reduction in the protection afforded the user.

D-PLAN 8-5 November 2010 I

8. HEALTH PHYSICS - (cont'd)

Respiratory protection equipment will be issued to individuals only after documentation has been received that shows that the person has satisfactorily completed:

" a medical exam,

" respiratory protection training, and a respiratory fit test (does not apply to in-line supplied air hoods and Self-Contained Breathing Apparatus).

8.3.3 Protective Clothing Personnel working in contaminated areas of LACBWR are provided with protective clothing to minimize the potential for personnel contamination. Routine entry into a contaminated area will require a minimum protective clothing requirement of:

" coveralls

" head covering

" gloves

" shoe coverings Specific jobs may require additional protective clothing. These additional requirements will be determined by the Health and Safety Department and will be listed on the Special Work Permit for the job.

During the SAFSTOR period, the laundry facility will remain operational to ensure an adequate supply of clean protective clothing.

8.3.4 Access Control To limit radiation exposures, personnel access is controlled in areas where such exposure is possible. This control consists of a system of physical barriers, warning signs and signals.

A Special Work Permit (SWP) will be issued as authorization for personnel to perform work of a non-routine nature in a specific area which involves unusual hazards. SWPs will be used to inform personnel of these hazards and the safeguards/protective measures which need to be taken during the work to ensure their well being.

8.3.5 Postings Postings shall be in accordance with the requirements of 10 CFR 20, Subpart J, as applicable.

8.4 RADIATION MONITORING A program for routine surveys and monitoring will be continued during the SAFSTOR period at LACBWR. This program will continue to assure all personnel are aware of the possible hazards involved before entering a potential radiation area or a potentially contaminated area. This will be done to ensure that the potential hazards are adequately defined, that adequate controls are instituted so that radiation exposure to personnel working in radiation areas or working with D-PLAN 8-6 November 2010 1

8. HEALTH PHYSICS - (cont'd) radioactive materials is minimized, and that each person carries out his work in a radiologically safe manner.

Survey data records will be maintained to assist in the evaluation of the radiological conditions and trends at LACBWR during SAFSTOR activities.

The radiological monitoring program will include the following surveys:

" airborne activity surveys

" dose rate surveys

" contamination surveys

• liquid activity surveys

" environmental surveys 8.4.1 Airborne Radioactivity Surveys In addition to using the fixed location or mobile air monitors, particulate airborne activity shall also be determined as needed by drawing a sufficient quantity of air through a filter paper. The samples shall be counted for beta-gamma activity in gas-flow proportional detector and scaler equipment. Alpha activity of a sample shall be determined by means of a windowless gas-flow proportional detector and a scaler when alpha radioactivity is suspected of being present.

Samples are analyzed for specific isotopic concentrations, by the use of a gamma analyzer.

Particulate samples of the stack releases will be obtained and analyzed weekly to determine release rates.

Non-routine air samples to establish protection requirements for maintenance activities or to verify airborne radioactivity conditions during work activities are obtained and analyzed when routine samples are not sufficient for monitoring plant conditions.

8.4.2 Radiation Surveys Radiation surveys are conducted for the following purposes:

" Measure and document radiation and contamination levels in areas of interest.

" Identify trends in radiation and contamination levels, particularly during work in progress.

" Determine appropriate protective measures for personnel working in restricted areas.

• Provide information so that workers can maintain their doses ALARA.

• Identify locations and situations where special dosimetry is required.

In addition to the measurements made by the fixed-location area radiation monitors, the measurement of external dose-rates shall be accomplished by portable survey instruments. The operation of the survey instruments shall be in accordance with the operating instructions outlined in each particular instrument manual or by procedure. Instruments covering high, intermediate, and low ranges shall be available on site.

D-PLAN 8-7 November 2010 1

8. HEALTH PHYSICS - (cont'd)

Surveys will be conducted by the Health and Safety Department to determine general area dose rates. They will also monitor areas to locate any radiological hot spots. Surveys will be performed on a routine basis established by procedures.

Special radiation surveys of particular items or areas are performed on an "as needed" basis.

Examples of special radiation surveys are the removal of equipment or materials from a restricted area, leak testing of sealed radioactive sources, or the shipment or receipt of radioactive material packages.

8.4.3 Contamination Surveys Contamination surveys will be conducted routinely by the Health and Safety Department as established by procedure to determine area contamination levels.

Special contamination surveys of particular items or areas are performed on an "as needed" basis.

Examples of special contamination surveys are the removal of equipment or materials from a restricted area, leak testing of sealed radioactive sources, or the shipment or receipt of radioactive material packages.

A dry filter paper or cloth disc will be wiped over approximately one square foot (12"x12" square or 12'-long S-shaped) of the surface being monitored. Swipes will be counted for beta-gamma activity in a gas-flow proportional detector or with a 27r GM probe or equivalent in fixed geometry sample holder as necessary. Alpha activity of a swipe will be determined by means of a windowless gas-flow proportional detector and a scaler or equivalent, when alpha radioactivity is suspected of being present.

8.4.4 Liquid Activity Surveys Samples of water containing radioactivity are collected and analyzed on a routine basis. Spent fuel pool water is analyzed to detect indications of degradation of the fuel stored in the pool.

Samples of liquid radioactive wastes and processed wastes are analyzed to ensure levels of radioactivity are below the levels permitted for release. Samples are analyzed by Health and Safety Department personnel in accordance with established procedures.

8.4.5 Environmental Surveys Environmental samples will be taken within the surrounding areas of the plant. These samples will be analyzed to determine any effects plant effluent releases may have on the environment.

This program will be conducted as per the ODCM.

8.4.6 ISFSI Radiation Monitoring Spent fuel will be placed in dry storage at the ISFSI in the MPC-LACBWR system which will provide an inert environment; passive shielding, cooling and criticality control; and a confinement boundary closed by welding. The structural integrity of the system precludes the release of contents in any of the design basis normal conditions and off-normal or accident events, thereby assuring public health and safety during use of the system.

D-PLAN 8-8 November 2010 1

8. HEALTH PHYSICS - (cont'd)

The MPC-LACBWR 5-cask array is evaluated to determine the minimum distance necessary to achieve a controlled area boundary dose of 25 mrem/year as required by 10 CFR 72.104(a). In the NAC-MPC FSAR, Section 1 0.A, annual exposures, based on an 8760-hour residence year, were determined from the center of a single cask and a 5-cask array. The NAC-MPC FSAR includes a plot of the 25 mrem/year footprint and the boundary required. A rectangular boundary a minimum of 300 feet from the pad center around the ISFSI ensures compliance with the requirements of 10 CFR 72.104(a) that dose rate will not exceed 25 mrem/year at the Controlled Area Boundary.

Prior to ISFSI construction, baseline radiation sampling and surveys were performed at the ISFSI site. With implementation of ISFSI operations, the fuel-loaded Vertical Concrete Cask (VCC) dose rates will be verified to be compliant with limits specified in Technical Specifications for the NAC-MPC System to maintain dose rates ALARA at locations on the VCCs where surveillance is performed and to reduce offsite exposures. Radiological conditions at the ISFSI will be monitored routinely to evaluate the continued effectiveness of the dry storage cask confinement boundary.

8.5 RADIATION PROTECTION EQUIPMENT AND INSTRUMENTATION A variety of equipment and instruments are used as part of the radiation protection program.

Equipment and instrumentation are selected to perform a particular function. Sensitivity, ease of operation and maintenance, and reliability are factors that are considered in the selection of a particular instrument. As the technology of radiation detection instrumentation improves, new instruments are obtained to more accurately measure radioactivity and ensure an effective radiation protection program.

This equipment can be broken down into several specific groups each with its own dedicated functions. These groups are:

" Portable Instruments

" Installed Instrumentation

" Personnel Monitoring Instrumentation

" Counting Room Instrumentation This equipment will be used, checked and calibrated by trained personnel according to in-plant procedures.

8.5.1 Portable Instruments There will be sufficient types and quantities of portable instruments to provide adequate beta, gamma, and alpha surveys at LACBWR. This equipment will have the ability to detect these types of radiation over the potential ranges that will be present during SAFSTOR. Portable dose rate instruments will be source checked prior to use, and they will be calibrated semiannually.

8.5.2 Installed Instrumentation There will be sufficient types and quantities of installed instrumentation to provide continuous in-plant and effluent release monitoring. This will assure the safe reliable monitoring of both area D-PLAN 8-9 November 2010

8. HEALTH PHYSICS - (cont'd) dose rates and airborne activity concentration throughout the area. These instruments will be response tested monthly and calibrated once every 18 months.

8.5.3 Personnel Monitoring Instrumentation Friskers and personnel instrumentation monitors will be provided throughout the plant to provide personnel contamination monitoring. These monitors will be of the type and sensitivities necessary to minimize the spread of in-plant contamination and prevent the introduction of contamination to outside areas. This equipment will be checked daily during normal workdays and calibrated semiannually.

8.5.4 Counting Room Instrumentation Laboratory equipment will be available to perform gross alpha and beta analyses and gamma isotopic analyses of samples collected in the plant. There will also be equipment available in a low background area to provide adequate analysis of environmental samples. A quality control program will be in effect for this equipment to ensure the accurate and proper operation of the equipment. Gross alpha/beta counters wilt be calibrated annually. The HPGe detectors will be calibrated every two years.'

8.6 RADIOACTIVE WASTE HANDLING AND DISPOSAL Radioactive waste generated at LACBWR during the SAFSTOR period will primarily consist of the following:

" Resin

" Dry active waste (DAW)

" Dismantlement (Metallic)

Radioactive waste generation will be maintained as low as possible to minimize the volume of material requiring reprocessing and disposal.

8.6.1 Resin Spent resin will be transferred to the spent resin receiving tank where it will be held until there is a sufficient quantity available for shipment to an approved processing facility. The resin will be transferred to an approved shipping container where it will be dewatered and made ready for shipment.

8.6.2 Dry Active Waste (DAW)

Any material used within the restricted area will be considered radioactive and will be disposed of as DAW, unless it can be demonstrated to be within established releasable limits. The generation of this material will be maintained as low as possible to reduce the total waste volume generated S

onsite. The material generated will be placed into approved shipping containers.

D-PLAN 8-10 November 2010 1

8.

HEALTH PHYSICS - (cont'd) 8.6.3 Dismantlement (Metallic)

During the SAFSTOR period, LACBWR employees will pursue limited dismantlement of the facility. This project will generate metallic wastes from system removal. This metallic waste will be placed in approved shipping containers and sent to an approved reprocessor.

Disposal of all radioactive waste will be in accordance with all pertaining guidelines.

8.7 RECORDS Records generated in the performance of the radiation protection program will be maintained as required to provide the necessary documentation of the program and in accordance with the QAPD. These records will be maintained in a designated storage area.

8.8 INDUSTRIAL HEALTH AND SAFETY LACBWR will continue to participate in Dairyland Power Cooperative's industrial safety program as prescribed by the DPC Safety Department. These programs will include:

" Accident prevention

" Hazardous waste management and control

" Asbestos control

" Hearing conservation D-PLAN 8-11 November 2010 1

9.

SAFSTOR ACCIDENT ANALYSIS

9.1 INTRODUCTION

The probability of an accident occurring during the SAFSTOR period is considerably less than during plant operation. The focus of the potential accidents has also changed. During operation, the focus was on minimizing the plant transient and cooling the reactor core. While spent fuel is in wet storage during SAFSTOR, the only major concern is protecting the fuel in the Fuel Element Storage Well (FESW). Once all spent fuel is placed in dry storage, accidents associated with the onsite ISFSI are discussed in the NAC-MPC Final Safety Analysis Report and the LACBWR ISFSI 10 CFR 72.212 Report. Accidents associated with dry cask storage at the ISFSI will not be addressed in this Decommissioning Plan.

The spent fuel stored in the FESW, while not benign, is not as much a hazard as the fuel in the operating reactor was. Since April 30, 1987, the fission product inventory has decreased and the decay heat generation is significantly less. These factors reduce the consequences of any accident affecting the spent fuel. As time passes, the consequences will continue to decrease.

The reactor's design basis accidents were reviewed to determine which could still occur during SAFSTOR. Some other accident scenarios which were not previously considered design basis accidents were also evaluated. A list of 8 postulated accidents was identified. These events are:

" Spent Fuel Handling Accident

" Shipping Cask or Heavy Load Drop into FESW

" Loss of FESW Cooling

" FESW System Pipe Break

" Uncontrolled Liquid Waste Discharge

" Loss of Offsite Power

" Earthquakes

" Wind and Tornado Each of these postulated events was evaluated based on the revised plant status to identify their potential consequences during the SAFSTOR period. The following sections discuss these accidents.

One additional event involving fire was examined. Fire protection is covered in Section 6.9.

The potential safety consequences of any fire scenario fall within the scope of other evaluated events.

9.2 SPENT FUEL HANDLING ACCIDENT This accident postulates a fuel assembly falling from the hoist into the FESW. The probability of this accident is extremely small, since minimal fuel handling will be performed during the SAFSTOR period until the fuel assemblies are removed from the FESW. Periodic inspections may be conducted during the years the fuel remains onsite. In the almost 20 years of operation and associated fuel handling at LACBWR, no fuel assemblies were ever dropped.

In this event, it is assumed that the cladding of all the pins in two fuel assemblies ruptures. The fuel handling crew evacuates when the local area radiation monitor alarms. Reactor Building D-PLAN 9-1 November 2010

9. SAFSTOR ACCIDENT ANALYSIS - (cont'd) ventilation dampers would be closed manually from the Control Room on high activity, but for this analysis, no containment integrity is assumed.

The assumptions used in evaluating this event during SAFSTOR were similar to those used in the FESW reracking analyses (References 9.10.1 and 9.10.2). 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 9.10.1, 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 2010 only 22.6% of the Kr-85 activity remains-97.7 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 two (2) hours 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 boundary (Reference 9.10.3) for a ground level release is also calculated. Adverse meteorology is assumed for all cases.

O 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/m 3 Kr-85 average concentration at 500m E 6.00 E-2 Ci/sec x 2.3 E-4 sec/m 3 = 1.38 E-5 Ci/m 3 Immersion Dose Conversion at 500m E Kr-85 Gamma Whole Body Dose Factor (Regulatory Guide 1.109) 1.61 E+I mRem/yr x 106 gCi x 1.142 E-4 y = 1,839 mRem/hr

ýtCi/m3 Ci hr Ci/m 3 Whole Body Dose at 500m E 1839 mRem/hr x 1.38 E-5 Ci/m3 x 2 hr = 0.05 mRem (as of 10/10 - 0.01 mRem)

Ci/m 3 D-PLAN 9-2 November 2010 1

9. SAFSTOR ACCIDENT ANALYSIS - (cont'd)

Kr-85 Beta/Gamma Skin Dose Factor (Regulatory Guide 1.109) mRem/yr 106 iCi yr mRem/hr 1.34E+/-3 x

xl.142E-4-=l1.53 E5 VtCi/m 3 Ci hr Ci/m 3 Skin Dose at 500m E mRem/hr 1.53 E5 x 1.38E -5 Ci/m 3 x 2hr = 4.2 mRem (as of 10/10 = 1.0 mRem)

Ci/m 3

Ground Level Release at EAB Worst Case X for 2 hrs at 338m NE or 33 8m SSE using Regulatory Guide 1.25 Q

2.2 E-3 sec m3 Whole Body Dose at 338m Skin Dose at 339m 10/87 = 40.4 mRem 10/10 = 9.2 mRem 10/87 = 0.49 mRem 10/10 = 0.11 mRem Ground Level Release at Emergency Planninz Zone Boundary Worst Case X for 2 hrs at 1 00m E Q

1.02 E-2 sec m 3 Whole Body Dose at 100m E Skin Dose at 100m E 10/87 = 187 mRem 10/10 = 42.3 mRem 10/87 = 2.25 mRem 10/10 = 0.51 mRem 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.

9.3 SHIPPING CASK OR HEAVY LOAD DROP INTO FESW This accident postulates a shipping cask or other heavy load falling into the FESW. Reference 9.10.1 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 D-PLAN 9-3 November 2010 1

9. SAFSTOR ACCIDENT ANALYSIS - (cont'd) released within the 2-hour period. The doses calculated are as follows. (Due to decay, as of Oct. 2010 only 22.6% of the Kr-85 activity remains - 8,096 Curies.)

Elevated Release Whole Body Dose at 500m E Skin Dose at 500m E 10/87 = 350 mRem 10/10 = 79.2 mRem 10/87 = 4.2 mRem 10/10 = 1.0mRem Ground Level Release at EAB Whole Body Dose at 338m Skin Dose at 338m 10/87 = 3.34 Rem 10/10 = 0.8 Rem 10/87 = 40.2 mRem 10/10 = 9.1 mRem Ground Level Release at Emergency Plannine Zone Boundary Whole Body Dose at 100m E Skin Dose at 100m E 10/87 = 15.6 Rem 10/10 = 3.5 Rem 10/87 = 186 mRem 10/10 = 42 mRem 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.

9.4 LOSS OF FESW COOLING This accident postulates a loss of FESW cooling. The most likely causes of a loss of cooling are:

" Both FESW pumps fail or FESW piping has to be isolated for maintenance;

" The Component Cooling Water (CCW) System is out of service due to failure of both pumps or other reason. The CCW System removes heat from the FESW cooler.

" The Low Pressure Service Water (LPSW) System is out of service due to failure of both pumps or other reason. The LPSW System removes heat from the CCW coolers.

If the third possibility is the cause, cooling to the CCW coolers can be restored by cross-connecting the High Pressure Service Water System to the coolers, in lieu of LPSW.

After the final discharge of fuel to the FESW, a conservative calculation of the FESW heatup rate was performed using the estimated decay heat source in the spent fuel on January 1, 1988.

This calculation indicated that coolant boiling could occur approximately 5 days after the loss of cooling.

In July 1993, a test was conducted to determine the actual heat-up rate of the FESW with all cooling and coolant circulation to the pool isolated. This test, as documented in LACBWR D-PLAN 9-4 November 2010 1

9. SAFSTOR ACCIDENT ANALYSIS - (cont'd)

Technical Report, LAC-TR-137, showed that the pool temperature increased from 80'F to only 1

14°F in 15.5 days. The test was terminated at 1 14°F to limit increasing radioactivity in the pool water, but extrapolation of the data indicates the temperature would stabilize at approximately 150 0F.

Substantial time is therefore available for restoration of FESW cooling. No immediate action is necessary during this postulated accident.

9.5 FESW PIPE BREAK This accident postulates a break in the FESW system piping, other than in the pump discharge piping between the redundant check valves and the pool liner. A load analysis was performed on this approximately 20 feet of piping. It was concluded that all stresses are within ASME Code allowable. (Reference 9.10.1 calls this line the spent fuel pool drain line.) The series check valves were added during the 1980 FESW reracking. In November 1999, the FESW return line was rerouted to enter the top of the storage well and extends down to discharge underwater in the FESW. The bottom inlet line now ends at the biological shield wall and is sealed with a welded plug.

If the postulated break occurs, the lowest the FESW could drain is approximately elevation 679 feet. At this level all spent fuel will remain covered. 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. 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 would 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. No release of contamination is associated with this event. Active FESW cooling would be lost during this accident, but a test conducted during July 1993 with normal water level in the FESW indicates that considerable time is available to take action. This test, as documented in LACBWR Technical Report, LAC-TR-137, showed that with all cooling and coolant circulation to the pool isolated, FESW water temperature increased from 80'F to only 1 14'F in 15.5 days. This test was terminated at 1 14TF to limit increasing radioactivity in the pool water. Extrapolation of the data indicated the temperature would have stabilized at approximately 150TF. Due to the smaller water volume to act as the heat sink in the FESW pipe break accident the initial heat up rate of the FESW water would be approximately twice as great as that during the 1993 test. The heat removal rate from the FESW at a given temperature will be reduced somewhat since the wetted wall area is reduced by approximately 42%. The heat removal at a given temperature, by evaporation of the FESW coolant and condensation on the FESW cover and walls, will be essentially unchanged. Since heat removal rate increases rapidly as the temperature in the FESW increases, engineering judgment indicates the temperature in the FESW will stabilize at a temperature somewhat above 150TF, but boiling is not expected to occur. The total heat source in the FESW is only about 12.2 kW.

As with the loss of FESW cooling event, if water is added to the FESW, any consequences of water heat up can be delayed or prevented. Water can be added from the Demineralized Water System or the Overhead Storage Tank.

D-PLAN 9-5 November 2010

9.

SAFSTOR ACCIDENT ANALYSIS - (cont'd) 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 from the Control Room. All that is needed is for an operator to cross-connect HPSW to the Component Cooling Water (CCW) coolers.

If an HPSW Diesel and 1B Emergency Diesel Generator start, FESW cooling can be provided.

If 1 A Emergency Diesel Generator (EDG) starts, but lB does not, adequate cooling can be provided only if the essential buses are tied together.

If one or more EDG's 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, FESW and CCW pumps cannot 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 9.10.4-9.10.7). The major concern of the previous evaluation was to safely shut D-PLAN 9-6 November2010 I

9. SAFSTOR ACCIDENT ANALYSIS - (cont'd) down the plant and maintain adequate core cooling to prevent fuel damage. The focus now is to prevent damage to the spent fuel stored in the FESW.

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 9.10.1 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 Offsite Power Event (Section 9.7) and FESW Pipe 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).

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 9.10.4-9.10.9). 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 spent fuel stored in the FESW.

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.

An opening in the steel and concrete exterior Reactor Building wall was created, then closed by installation of a weather-tight, insulated, roll-up, bi-parting door in November 2006. The Reactor Building opening (described in Section 4.2.1) and bi-parting door are depicted in Figures 4.6 and 4.7. The 50.59 Evaluation of this modification to the Reactor Building structure concluded that since the governing load case does not include wind loading, but does include seismic loading, the seismic event governs. If the Reactor Building bi-parting door is open, or the closed bi-parting door is breached during a wind or tornado event, there are no structures, systems, or components important to safety that could be impacted in a direct linear path by wind-driven material entering the Reactor Building opening. The FESW is on the opposite side of the Reactor Building from the opening and at a low oblique angle. If an impact with equipment were to occur from wind-driven material, the event is bounded by the FESW Pipe Break analysis (Section 9.5). If wind-driven material were to enter the FESW, the event is bounded by the analysis of the Shipping Cask or Heavy Load Drop into FESW event (Section 9.3). The aluminum FESW cover is typically installed during the short periods of time the Reactor Building is open providing a defense-in-depth feature against wind-driven material entering the FESW. The bi-parting door panels are designed to withstand a 25-psf Exposure "B" wind load. The original Reactor Building wall design was 20 psf external wind load. The D-PLAN 9-7 November 2010

9. SAFSTOR ACCIDENT ANALYSIS - (cont'd) modification to the Reactor Building does not result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated.

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

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

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

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

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

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

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

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

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

9.10.10 FC 37-06-34, Reactor Building Restoration Activities.

D-PLAN 9-8 November 2010 1

10.

SAFSTOR OPERATOR TRAINING AND CERTIFICATION PROGRAM

10.1 INTRODUCTION

This program describes the training and certification for Operations Department personnel who as supervisors and operators are associated with the operation, surveillance, and maintenance of LACBWR while spent fuel is in wet storage in the FESW.

The Operator Training and Certification Programs ensure that people trained and qualified to operate LACBWR will be available during the SAFSTOR period. Licensee certification of personnel makes it unnecessary for the NRC to periodically conduct license examinations for persons involved in infrequent activities and prevents delays due to obtaining NRC Fuel Handler Licenses for any evolutions that may require fuel movements. When all spent fuel is in dry storage at the ISFSI, CFH training and proficiency will no longer be required.

10.2 APPLICABILITY LACBWR Technical Specifications require that all fuel handling shall be directly supervised by a Certified Fuel Handler (CFH). The following members of the plant staff attain certification, maintain qualification, and demonstrate proficiency in accordance with the CFH Training Program:

Shift Supervisors Control Room Operators

" Staff members appointed by the Plant Manager Additional personnel assigned Operations Department responsibilities are required to maintain CFH qualification.

10.3 INITIAL CERTIFICATION Candidates for CFH shall participate in a training program covering the following topic areas:

10.3.1 Reactor Theory (as applicable to the storage and handling of spent fuel) training will include characteristics of the stored spent fuel, subcritical multiplication, factors affecting reactivity and criticality, and the basis for fuel handling restrictions and procedures.

10.3.2 Design and Operating Characteristics (spent fuel handling and storage equipment) will include training in the functions and use of fuel handling tools, cranes, FESW, and FESW support systems and equipment. Prior to dry cask storage operations this training will include dry storage cask handling and loading, dry storage cask preparation, cask handling equipment, and procedures.

10.3.3 Monitoring and Control Systems will include training on the FESW monitoring systems and area radiation monitors.

D-PLAN 10-1 November2010

10. SAFSTOR OPERATOR TRAINING AND CERTIFICATION PROGRAM - (cont'd) 10.3.4 Radiation Protection Radiation training will include theory of radioactive emissions, control of radiation exposure, use of radiation detection and monitoring equipment, protective clothing and respiratory protection, and contamination control procedures. Training will emphasize the principles and practices associated with maintaining exposures as low as reasonably achievable (ALARA).

10.3.5 Normal and Emergency Procedures training will include the Emergency Plan and any operations and emergency procedures associated with the operation of LACBWR systems and equipment during SAFSTOR. This area shall also include training in the handling and processing of radioactive wastes.

10.3.6 Administrative Controls (applicable during the SAFSTOR period) training will include LACBWR Technical Specifications, Security Plan, Quality Assurance Program Description and plant administrative procedures associated with the operation, surveillance, and maintenance of LACBWR.

Training will be provided through a combination of classroom instruction, audio-visual instruction, self-study, and on-the-job training.

Satisfactory completion of the training shall be based on passing of a comprehensive written examination including each of the above areas and an oral examination. Minimum passing grade for the written examination shall be 70% in each area and 80% overall. The oral examination shall be administered by a member of the plant management staff. Results of the oral examination shall be on a pass/fail basis. Weaknesses noted as a result of the written or oral examination shall be documented and remedial training provided.

10.4 PROFICIENCY TRAINING AND TESTING Proficiency training shall be used to maintain the qualification level of certified personnel.

Proficiency training will include periodic training through the use of classroom training, audio/visual instruction, self-study assignments, and/or on-the-job training. Frequency and topics to be included in the proficiency training will depend on actual activities planned or in progress and identified weaknesses. As a minimum, training in the six areas included in the initial certification program shall be covered at least once every 2 years.

A biennial written examination, combined with an annual oral examination administered in those years when no written exam is given, shall be used to demonstrate the proficiency of certified personnel. Examinations will be similar to, but not as comprehensive as, the initial certification examinations. Minimum passing grade for proficiency examinations shall be 70% in each section and 80% overall. Oral examinations shall be on a pass/fail basis.

10.5 CERTIFICATION Upon successful completion of the initial certification training program, the Plant Manager or his delegate shall certify the individual as a Certified Fuel Handler. Normally an employee will complete the initial certification within one year after entering the program. After initial certification, personnel will be recertified every two years based on the successful completion of the Proficiency Training and Testing Program D-PLAN 10-2 November 2010 1

10. SAFSTOR OPERATOR TRAINING AND CERTIFICATION PROGRAM - (cont'd) 10.6 PHYSICAL REQUIREMENTS As a prerequisite to acceptance into the training program and for recertification, a candidate must successfully pass a medical examination designed to ensure that the candidate is in generally good health and is otherwise physically qualified to safely perform the assigned work. Minor correctable health deficiencies, such as eyesight or hearing, will not per se prevent certification.

The medical examination will meet or exceed the requirements of ANSI Standard N546-1976, "American National Standard - Medical Certification and Monitoring of Personnel Requiring Operator Licenses for Nuclear Power Plants."

10.7 DOCUMENTATION Initial Certification and Proficiency Training shall be documented and maintained for certified personnel while employed at LACBWR. The records shall include the dates of training, results of all quizzes and examinations, copies of written examinations, oral examination records, and information on results of physical examinations.

D-PLAN 10-3 November2010 I

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

Larry Nelson Health and Safety/Maintenance Supervisor October 1995 Revised: October 2010 Dairyland Power Cooperative 3200 East Avenue South La Crosse, WI 54601

LAC-TR-138 PAGE 1 LACBWR INITIAL SITE CHARACTERIZATION SURVEY FOR SAFSTOR

1.0 INTRODUCTION

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

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 the Dairyland Power Cooperative. The Allis-Chalmers Company was the original licensee; the AEC later sold the plant to the Dairyland Power Cooperative (DPC) and provided them with a provisional operating license.

La Crosse Boiling Water Reactor 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.

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

The La Crosse Boiling Water Reactor was permanently shut down on April 30, 1987.

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.

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%.

LAC-TR-138 PAGE 2 2.0 OPERATING EVENTS WHICH COULD AFFECT PLANT CLEANUP (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 (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 1 gph. FESW leakage has stabilized over the years to an average of approximately 21 gallons per day. This leakage is contained within the steel shell of the Reactor Building.

(3)

Release of Contaminated Water to the Controlled Area, July 2, 1982 at 0630 The failure to close the resin inlet valve in the resin addition line to the number one full flow demineralizer following the addition of resins subsequently caused the release of water to the Turbine Building Floor through the gasket on a bulls-eye sight glass in that line. Approximately 75 gallons of contaminated water could not be accounted for in the waste water storage tanks. Approximately 20 gallons of this water is estimated to have entered the ground in the radiological controlled area outside the west turbine hall door and the turbine hall truck bay door. Contaminated ground was removed over a 3 sq. ft area by the west turbine hall door and 2 sq. yard area by the truck bay door. It was placed in waste storage barrels and later sent to burial at Barnwell, S.C.

LAC-TR-138 PAGE 3 July 2, 1982 Spill Area

LAC-TR-138 PAGE 4 3.0 CLASSIFICATION OF THE GENOA SITE AREA BY CONTAMINATION POTENTIAL As per NUREG 5849 "Manual For Conducting Radiological Surveys in Support of License Termination" the LACBWR site will be placed into two separate classifications in accordance with section 4.2 of the NUREG. These classifications are defined as follows:

Affected Areas: Areas that have potential radioactive contamination (based on plant operating history) or known radioactive contamination. This would normally include areas where radioactive materials were used and stored, and where records indicate spills or other unusual occurrences that could have resulted in spread of contamination. Areas immediately surrounding or adjacent to locations where radioactive materials were used, stored, or spilled are included in this classification because of the potential for inadvertent spread of contamination.

At LACBWR the affected area will be that area located within the LSE boundary.

Movement of radioactive material occurred throughout this area, therefore a potential for the inadvertent spread of contamination in this area could have occurred. (See map).

Unaffected Areas: All areas not classified as affected. These areas are not expected to contain residual radioactivity, based on knowledge of site history and previous survey information.

The area outside the LACBWR LSE fence will be classified as the unaffected area.

LAC-TR-138 PAGE 5 LACBWR AFFECTED AREA MAP 4,

LAC-TR-138 PAGE 6 4.0 CHARACTERIZATION SURVEYS In order to provide an initial site characterization survey for LACBWR the following radionuclide determination tests were performed.

(1)

Spent fuel radionuclide inventory (2)

Core internal /RX radionuclide inventory (3)

Plant loose surface radionuclide inventory (4)

Plant systems internal radionuclide inventory (5)

Reactor biological shield activation survey (6)

Circulating cooling water outfall area radionuclide inventory (7)

Site soil contamination determination survey The results of these individual surveys provide the initial LACBWR site characterization data that will be needed during the decommissioning period. This data will be used to assist in the performance of the final Site Termination Survey which will be performed at the end of the decommissioning period.

4.1 Spent Fuel Radionuclide Inventory During June 1987 all fuel assemblies were removed from the reactor vessel.

Currently there are 333 spent fuel assemblies stored in the spent fuel pool.

The 72 fuel assemblies removed from the reactor in June 1987 have assembly average exposures ranging from 4,678 to 19,259 megawatt-days per metric ton of uranium. The exposures of the 261 fuel assemblies discharged during previous refuelings range from 7,575 to 21,532 MWD/MTU. The oldest fuel stored was discharged from the reactor in August 1972. Forty-nine of the A-C fuel assemblies discharged prior to May 1982 contain one or more fuel rods with visible cladding defects and 54 additional A-C fuel assemblies discharged prior to December 1980 contain one or more leaking fuel rods as indicated by higher than normal fission product activity observed during dry sipping tests.

The established radioactivity inventory in the 333 spent fuel assemblies was performed by using the computer program Fact 1 and hand calculations performed by Dr. S. Raffety (Nuclear Engineer) during July of 1987. Activity in the fuel assembly hardware is based on neutron activation on this hardware. All activity values have been decay corrected to January 1988.

LAC-TR-138 PAGE 7 Spent Fuel Radioactivity Inventory January 1988 Radionuclide Ce-144 Cs-137 Ru-106 Zr(Nb)-95 Cs-134 Kr-85 Ag-il0mn Sr-89 Te-127m Co-60 Ru-103 Pm-147 Ni-63 Ce-141 Cm-242 Am-241 Pu-238 Pu-239 Pu-240 Eu-154 Cm-244 Cr-51 Te-129m H-3 Fe-59 Eu-152 Am-242m Half Life (Years) 7.801 E-1 3.014 E+I1 1.008 E+0 1.754E-1 (9.58E-2) 2.070 E+0 1.072 E+I 6.990 E-I 1.385 E-1 2.990 E-1 5.270 E+0 1.075 E-1 2.620 E+0 1.000 E+2 8.890 E-2 4.459 E-I 4.329 E+2 8.774 E+1 2.410 E+4 6.550 E+3 8.750 E+0 1.812 E+1 7.590 E-2 9.340 E-2 1.226 E+I 1.220 E-i 1.360 E+1 1.505 E+2 Activity (Curies) 2.636 E+6 1.666 E+6 1.524 E+6 3.555 E+5 3.291 E+5 1.160 E+5 1.018 E+5 1.009 E+5 8.238 E+4 6.395 E+4 6.334 E+4 4.129 E+4 3.540 E+4 2.638 E+4 1.858 E+4 1.474 E+4 1.262 E+4 8.837 E+3 7.165 E+3 4.020 E+3 3.603 E+3 3.002 E+3 1.170 E+3 5.5 10 E+2 5.120 E+2 5.110 E+2 4.900 E+2 Radionuclide Sr-90 Pu-241 Fe-55 Zr-95 Ni-59 Tc-99 Sb-125 Eu-155 U-234 Am-243 Cd-I 13m Nb-94 Cs-135 U-238 Eu-156 Pu-242 U-236 Sn-121m Np-237 U-235 Sm-151 Sn-126 Se-79 1-129 Zr-93 1-131 Half Life (Years) 2.770 E+1 1.440 E+I 2.700 E+0 1.750 E-i 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+I 2.000 E+4 3.000 E+6 4.470 E+9 4.160 E-2 3.760 E+5 2.340 E+7 7.600 E+1 2.140 E+6 7.040 E+8 9.316 E+I 1.000 E+5 6.500 E+4 1.570 E+7 1.500 E+6 2.200 E-2 Activity (Curies) 1.147 E+6 1.138 E+6 5.254 E+5 3.52 E+2 2.87 E+2 2.76 E+2 2.73 E+2 1.68 E+2 6.37 E+I 6.31 E+I 1.78 E+I 1.59 E+I 1.40 E+I 1.22 E+I 8.63 8.58 6.32 4.44 2.19 1.89 1.51 7.01 E-1 5.52 E-i 3.90 E-i 1.11 E-i 2.00 E-3 Total Activity = 1.00 E7 curies NOTE: Attachment 1 is an inventory of the Spent Fuel Radioactivity decay corrected.

LAC-TR-138 PAGE 8 4.2 Core Intemals/RX Components Radionuclide Inventory Reactor components inand near the reactor core during power operation become radioactive due to nuclear interaction with the large neutron flux present in this region. Most of the residual radioactivity is produced by n,y reactions with the atomic nuclei of the target material although n,P reactions, for example the production of C-14 from N-14, are also significant.

The residual radioactivity in the materials of the various LACBWR reactor components has been estimated using activation analysis theory and, where available, actual data from laboratory analyses of irradiated metal samples. Best estimates of neutron fluxes in and irradiation histories of specific components were used in these calculations. Original material chemical compositions were obtained from actual material certification records when readily available but standard compositions for specified materials were used in some cases.

Radioactive decay of the activation product nuclides has been taken into account to obtain the best estimate values for the residual radioactivity as of January 1, 1988.

As of January 1, 1988, the largest contribution (210,000 Ci) to the radioactivity inventory in LACBWR activated metal components is the isotope Fe-55 (HL = 2.7y). The next largest contributor (101,000 Ci) is Co-60 (HL = 5.27y).

The activity of these two relatively short-lived isotopes will decrease very significantly during the proposed SAFSTOR period. The major long-lived contributor to the radioactivity inventory (10,700 Ci) is Ni-63 (HL =100y). The activity of other activation product nuclides have been lumped together in two categories, those with half lives less than 5 years and those with half lives greater than 5 years. The group with HL <5y consists mostly of Zr-95 (64d) in Zircaloy components and Cr-51 (27.7d) and Fe-59 (44.6d) in stainless steel components along with small quantities of Sn-113 (115d), Sn-119 (293d), Sn-123 (129d), Hf-175 (70d), Hf-181 (42.4d), W-181 (121d) and W-185 (75.1d). The group with HL > 5y consists mostly of Ni-59 (8x10 4y) with small quantities of C-14 (5730y),

Zr-93 (1.5x10 6y), Sn-121 (50y), Cd-113 (14.6y), Nb-94 (2x10 4y) and Tc-99 (2.13x10 5y).

0 LAC-TR-138 PAGE 9 Core Intemal/RX Component Radionuclide Inventory - January 1, 1988 Estimated Curie Content Other Nuclides Components Co-60 Fe-55 Ni-63 T112 < 5y TI/2 > 5y Total In Reactor Fuel Shrouds (72 Zr, 8 SS) 22,109 63,221 1,352 2,810 15 89,507 Control Rods (29) 4,886 4,826 817 24 15 10,568 Core Vertical Posts (52) 1,270 594 63 2,396 4

4,327 Core Lateral Support Structure 9,108 21,477 770 105 8

31,468 Steam Separators (16) 33,439 78,851 2,826 386 30 115,532 Thermal Shield 1,443 3,402 123 17 1

4,968 Pressure Vessel 347 1,029 10 2

-0 1,388 Core Support Structure 6,458 15,230 546 75 6

22,315 Horizontal Grid Bars (7) 173 408 15 2

-0 598 Incore Monitor Guide Tubes 307 188 611 7

5 1,118 Total 79,540 189,226 7,133 5,824 84 281,807 In FESW Fuel Shrouds (24 SS) 13,667 14,988 2,384

-0 26 31,065 Fuel Shrouds (73 Zr) 918 1,007 95 27 3

2,050 Control Rods (10) 3,456 2,386 910

-0 17 6,769 Start-up Sources (2) 3,177 2285 156 5

3 5,626 Total 21,218 20,666 3,545 32 49 45,510 NOTE: Core Internals/RX Components were removed and disposed of in 2007.

LAC-TR-138 PAGE 10 4.3 Plant Loose Surface Radionuclide Inventory A plant smear survey was performed of all accessible interior building surfaces in an attempt to determine the amount of loose surface contamination in the plant.

The specific isotopic identification of the contamination was also determined.

Each smear was gamma scanned to determine not only a correlation factor in ptCi per DPM/l 00 cm2 but also the percentile of each radioisotope present in the mixture. From previous, part 61, analysis of plant smears, it has been determined that Fe-55 is the major beta emitter in the plant and is in approximately the same percentage as Co-60. Fe-55 will be the only beta emitting isotope listed as a contaminant. Alpha activity on the surfaces has been checked by the use of an Internal Proportional Counter and has been found to be negligible and so will not be considered. The survey did indicate that the major isotopes present in the plant's loose surface contamination are the following isotopes:

Isotope 1/2 Life Co-60 5.27 years Cs-137 30.1 years Mn-54 312.2 days Fe-55 2.7 years It must be realized when reviewing the results of this survey that a 100%

smearing of plant surfaces was not performed and therefore the following data is subject to significant error.

The majority of the loose surface contamination throughout the plant is found in the plant contaminated areas. The plant areas that were classified as contaminated areas at plant shutdown in 1987 are listed below.

a)

Waste Treatment Building (1) decon area (2) basement (3) resin liner room (4) high level storage pit b)

Turbine Building (1) stop valve area (2) full flow room (3) condensate bay (4) area under main condenser (5) feed pump bed plates (6) tunnel - includes 4500 WT cubicle and crawlway to stack c)

Reactor Building (1) 701 south by FESW (2) mezz around core spray pumps (3) west NI platform (4) purification platform (5) FESW IX cubicle (6) purification pump area (7) basement - includes UCRD platform, purification IX cubicle, retention tank cubicle, subbasement, and FCP cubicles.

LAC-TR-130 PAGE 11 PLANT LOOSE SURFACE CONTAMINATION - JANUARY 1988 Isotopes Present, in tCi Total Area ktCi Location Co-60 Cs-137 Mn-54 Ce-144 Co-57 Cs-134 Fe-55 Content Turbine Building, (TB) a) Main Floor 0.83 0.07 0.83 1.73 b) Mezzanine - including stop valve area 0.49 0.14 0.04 0.49 1.16 c) Grade Floor - includes feedwater heater area 0.42 0.06 0.02 0.42 0.92 d) Tunnel 0.81 0.18 0.06 0.81 1.86 Reactor Building (RB) a) Above grade 3.16 0.20 0.39 3.16 6.91 b) Below grade 31.44 7.40 2.36 0.04 0.04 0.08 31.44 72.80 Waste Treatment Building 7.57 0.48 0.66 7.57 16.28 Totals 44.72 8.53 3.53 0.04 0.04 0.08 44.72 101.66

LAC-TR-138 PAGE 12 4.4 Plant System Internal Radionuclide Inventory The internal surfaces of many inplant systems were exposed to radionuclide contaminants during plant operation. To obtain the most accurate analysis of these systems, piping/component destruction would be necessary. This was determined to be an undesirable method of analysis at this time.

A method of nondestructive sampling was developed. Each system that would be sampled was looked at and a nondestructive entry point was found. Once the system was open, the piping was dried and a 1 cm2 area was scraped to bare metal. As the scraping was being done, the corrosion layer that was being removed was vacuumed into a glass fiber filter 47 mm in diameter. This filter was then gamma scanned for radionuclide identification using the computer-based gamma spectroscopy unit connected to a GeLi detector.

A sample of the crud layer from the bottom of the FESW was obtained using a vacuum system. This crud was mixed with the water and an aliquot was taken and gamma scanned. This was used to determine the activity in the FESW crud layer. Because of the inaccessibility of the bottom of the reactor vessel, the gamma spectrum obtained from the FESW sample was used as the gamma spectrum for any crud layer in the vessel bottom.

Conversation with representatives from other facilities who have looked at their reactor vessel bottom indicates a very small crud layer in the vessel.

Each sample was also alpha counted to determine the total alpha content. The individual alpha isotopic mixture was not determined. Alpha analysis was performed with a Canberra Internal Proportional counter.

In systems where the piping/component radiation levels varied significantly, the radiation level of the sample area was found. The remainder of the system was surveyed and the piping/component area was classified by radiation level. This technique allowed the sampled area uCi/cm2 value to correspond with a radiation level. As the system radiation level varied, a uCi/cm2 radiation level value was used to proportion the system areas to that of the sample area, thus allowing a total system uCi content to be determined. After sampling, each system was returned to an as-found condition.

Fe-55, will be the only beta emitting isotope to be considered in the plant isotopic content. This will be in the same percentage as that of the Co-60 found in the systems.

The isotopic composition occurring in the piping systems consists of the major nuclides listed as follows:

Isotope 1/2 life Co-60 5.27 years Cs-137 30.1 years Mn-54 312.2 days Fe-55 2.7 years This testing of piping systems is not an absolute value. Significant error can be associated with this sampling technique due to the inability to actually analyze a system component or pipe.

The values listed are subject to error due to the required non-destructive sampling techniques used and the inability to sample all components and areas.

LAC-TR-138 PAGE 13 PLANT SYSTEMS INTERNAL RADIONUCLIDE INVENTORY - JANUARY 1988 Nuclide Activity, in [tCi ystem Total Plant System Fe-55 Alpha Co-60 Cs-137 Mn-54 Cs-134 Nb-95 Co-57 Zn-65

[tCi Content RB Ventilation Offgas -

upstream of filters Offgas -

downstream of filters TB drains RB drains TB Waste Water RB Waste Water Main Steam Turbine Primary Purification Emergency Core Spray Overhead Storage Tank Seal Inject 1.6 E3 6.0 E2 6.5 E2 1.7 E4 3.8 E4 3.6 E3 2.1 E5 2.6 E5 9.3 E2 8.9 E4 1.8 E3 1.3 E4 1.6 E3 1.6 E3 1.7 E2 1.3 E2 1.40 6.0 E2 4.4 E4 1.0 E2 4.0 El 3.2 6.8 7.9 El 2.9 E2 1.8 1.2 El 5.0 3.4 El 3.8 6.5 E2 1.7 E4 3.8 E4 3.6 E3 2.1 E5 2.6 E5 9.3 E2 8.9 E4 1.8 E3 1.3 E4 1.6 E3 8.3 E2 5.0 E3 7.8 E2 2.4 E3 2.6 E3 1.2 E2 1.5 E2 2.3 E3 1.7 E4 2.0 E4 2.0 E2 1.8 E2 8.8 E3 1.1 E2 1.4 E2 7.8 E2 9.5 E2 5.5 El 1.5 E2 5.6 El 1.0 E3 3.5 E3 4.5 E4 2.1 E3 4.0 E4 8.1 E4 7.5 E3 4.4 E5 5.4 E5 2.2 E3 1.9 E5 3.9 E3 2.8 E4 3.4 E3 1.3E2 1.4E3 2.1 E3 7.5

LAC-TR-138 PAGE 14 PLANT SYSTEMS INTERNAL RADIONUCLIDE INVENTORY - JANUARY 1988 - (cont'd) k......

]

Nuclide Activity, in 1iCi System Total Plant System Fe-55 Alpha Co-60 Mn-54 Co-57 Co-58 Zn-65 Other 1iCi Content 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 1.0 E5 1.4 E5 9.9 E3 2.0 E4 2.3 E5 1.5 E5 1.5 E6 4.9 E2 6.6 E2 4.6 El 9.4 El 1.1 E3 7.2 E2 7.0 E3 1.0 ES 1.4 E5 9.9 E3 2.0 E4 2.3 E5 1.5 E5 1.5 E6 3.1 E4 4.2 E4 2.9 E3 5.9 E3 6.9 E4 4.6 E4 4.4 E5 1.6 E2 2.1 E2 1.5 El 3.0 El 3.5 E2 2.3 E2 2.3 E3 3.2 E3 4.3 E3 3.0 E2 6.1 E2 7.1 E3 4.7 E3 4.5 E4 3.5 E3 4.7 E3 3.3 E2 6.7 E2 7.8 E3 5.1 E3 5.0 E4 2.4 E5 3.3 E5 2.3 E4 4.7 E4 5.5 E5 3.6 E5 3.5 E6 5.9 E6 4.6 E5 9.3 E4 2.5 E6 1.2 E4 2.5 E6 7.6 E5 3.9 E3 7.8 E4 8.6 E4 2.1 E5 3.9 E4 2.8 E2 3.1 El 2.1 E5 3.9 E4 3.2 E4 1.3 E4 1.6 E3 1.5E1 6.3E2 3.1 E3 6.7 E2 Fe-59 = 5.2 E2 Nb -95 = 1.1 E2 Ru-103 = 4.9 El Ce-144 = 1.2 E2 I ________________

0 LAC-TR138 PAGE 15 PLANT SYSTEMS INTERNAL RADIONUCLIDE INVENTORY - JANUARY 1988 - (cont'd)

Nuclide Activity, in [tCi System Total Plant System Fe-55 Alpha Co-60 Mn-54 Cs-137 Ce-144 Zn-65 Other tCi Content Fuel Element Storage Well System 8.5 E5 3.9 E2 8.5 E5 1.4 E4 1.0 E4 1.7 E6 Fuel Element Storage Well - all but floor 1.3 E3 4.9 1.3 E3 6.0 E2 4.6 E3 4.5 E2 8.3 E3 Fuel Element Storage Well floor 2.6 E7 7.6 E3 2.6 E7 5.0 E5 4.1 E4 1.1 E5 Cs-134

= 1.3 E2 5.3 E7 Co-58

= 1.3 E2 Resin Lines 1.3 E5 1.0 E2 1.3 E5 4.2 E4 4.0 E2 2.2 E3 Fe-59

= 1.7 E3 3.1 E5 Co-57

= 4.8 El Co-58

= 2.1 E3 Nb-95

= 3.5 E2 Ru-103 = 1.6 E2 Main Condenser 1.1 E7 8.5 E3 1.1 E7 3.6 E6 3.4 E4 1.9 E5 Fe-59

= 1.4 E5 2.6 E7 Co-57 = 4.1 E3 Co-58 = 1.7 E5 Nb-95 = 3.0 E4 Ru-103 = 1.4 E4 NOTE: Attachment 3 is an inventory of the plant system - Internal Radionuclide Inventory decay corrected.

LAC-TR-138 PAGE 16 4.5 Reactor Biological Shield Activation Survey During 1993 LACBWR revised the decommissioning cost study. It was felt necessary at this time that a determination be made as to the extent of activation of the biological shield. The amount of biological shield activated will directly effect the cost of removal and disposal.

On October 20, 1993, an outside contractor completed boring of the biological shield. The initial boring site had to be abandoned due to the presence of shield cooling system piping. This hampered the drilling and at one point caused the bit to become stuck. The boring site was moved approximately two feet below the original site.

The boring through the biological shield was surveyed and analyzed on October 21, 1993. The borings were removed in several sections and are designated as below. These are listed as from the outside of the shield wall to the inside. These boring sections are being kept on the West TB Grade for further analysis.

1A-2A - 9.5 inches long 2A-3A -

14inches long 3A-4A -

27 inches long 4A-5A -

15 inches long 5A-6A -

2 inches long 6A-7A -

39 inches long Section 6A-7A, being closest to the reactor, was surveyed using an RO-3 portable radiation detector to determine activities present. The following dose rates were obtained. These are listed as distances from the inside shield wall.

0 inches - 130 mrem/hour 6 inches - 130 mrem/hour 12 inches -

55 mremihour 18 inches -

16 mrem/hour 24 inches -

6 mrem/hour 30 inches -

2.5 mrem/hour 36 inches -

1.2 mrem/hour 39 inches -

<1 mrem/hour A smear of this core section was taken with the following results:

Co-60 8.42 E-3 VtCi/smear Eu-152 1.27 E-2 ptCi/smear No other surveys of this core were taken due to the fact that the dose reading indicated complete activation of this section.

LAC-TR-138 PAGE 17 One inch sections of the remaining cores were cut and gamma-scanned to determine the depth of the concrete activation.

Distance from inside shield wall 39 inches 42 inches 48 inches 49 inches 52 inches 56 inches Co-60 (h!Ci/sample) 1.79 E-3 4.49 E-4 1.77 E-4 8.13 E-5 2.01 E-5 NP Eu-152 (LLCi/sample) 9.57 E-4 3.82 E-4 3.67 E-4 2.36 E-4 9.08 E-5 NP Surveys of these sections with a frisker indicated <100 CPM increase over background. Smears indicated <MDA loose surface contamination. Analysis indicates that the biological shield is activated at a distance of 56" from the inside of the shield wall.

LAC-TR-13 8 PAGE 18 BIOLOGICAL SHIELD CORE RADIATION SURVEY 160 150 140 130 120 110 100 90 80 70 60 50 40 30 20 10 0

7 I II I~

I - I II_ I I I I I i I

I I

I

__0__

__ I iii 2 H I-I I

1 II_____

0 so cm 5

10 15 20 25 30 35 40 45 DISTANCE-INCHES

LAC-TR-138 PAGE 19 4.6 Circulating Cooling Water Outfall Area Radionuclide Inventory As part of LACBWR's dismantlement cost study it was determined that the extent of contamination in the circulating cooling water outfall area be determined.

During February 1994 EMS (Environmental Marine Services) was contracted to send divers to LACBWR to perform a sampling survey in the outfall area. This sampling survey would be used to estimate the radionuclide inventory that exists in the plant outfall. The divers report that the area from the Mississippi River's edge to approximately 60' from the edge is covered with large rip rap and no sample could be obtained. From the 60' area a silt sample was taken approximately every 10' out and 110' from the rivers edge. The divers then turned downstream with the river current. They continued to obtain silt samples every 10' out to 150' from the outfall. The following table lists the activities found in these samples.

1Ci/kg as of February 1994 Sample Distances From Outfall (ft)

Co-60 Cs-137 1

60 1,070 1,980 2

70 1,260 2,200 3

80 1,330 2,230 4

90 813 1,870 5

100 216 362 6

110 82.3 127 7

120 8.96 21.4 8

130 NF 7.22 9

140 NF 20.11 10 150 9.33 22.6 (NF = None Found)

An attempt to obtain samples at various depths was tried. The divers didn't have the necessary equipment to perform this sampling so cross contamination occurred. Because of this no determination of the depth of the contamination can be made. The data we presently have would indicate a clean up area 20' wide by 95' long.

4.7 Initial Site Soil Contamination Determination Survey In accordance with NUREG/CR-5849 "Manual for Conducting Radiological Surveys in Support of License Termination" an initial site soil survey was performed during August - September 1995. The LACBWR affected area (area within the LSE fence) was marked out into 10 m2 areas for sample location identification. No samples were taken under surface coverings such as blacktop.

After gridding of the area was complete one surface soil sample was collected from the top 15 cm of the soil in each grid location. Each sample collected was approximately 1 - 2 kg.

LAC-TR-138 PAGE 20 Grass, rocks, sticks and foreign objects were removed from each sample to the degree practical. Each sample was then dried and placed in a disposable marinelli container for gamma analysis. All samples were allowed to sit for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to reduce the affect of any short lived natural decay chains. All samples were then counted for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> on the ENV HPGE. The following is a listing of activity found in these samples. No natural occurring isotope is listed. The actual sample gamma scans are being kept for further review if needed. All sample locations can be found using their corresponding number using the enclosed site grid map.

NOTE: All activity values are in pCi/kg Sample Sample Grid #

Co-60 Cs-137 Grid#

Co-60 Cs-137 Eu-155 1

10.5 106 27 13.9 58.4 2

34.8 81.6 28

<MDA 78.7 3

2.9 126 29

<MDA 74.1 4

16.7 78.5 30

<MDA 68.3 5

4.8 92.1 31

<MDA 64.2 6

8.8 80.2 32

<MDA 27.7 7

<MDA 51 33 31.2 129 8

<MDA 12.2 34

<MDA 171 9

<MDA 23.2 35

<MDA 91.7 53.5 10

<MDA 38.9 36

<MDA 61 20.3 11

<MDA 71 37 40.9 124 40.4 12

<MDA 93.3 38

<MDA 44.6 13 767 983 39

<MDA 130 14 100 312 40

<MDA 49 15 5.8 58.9 41

<MDA 74.9 16

<MDA 22.5 42

<MDA 55.6 17

<MDA 15 43 20.7 211 18

<MDA 14.4 44 9.1 145 19

<MDA 43.5 45 36.5 310 20

<MDA 44.2 46 19.5 197 30.8 21

<MDA 72.6 47 85.1 177 27.7 22

<MDA 55.6 48 12 158 23

<MDA 45.5 49 56.4 329 24

<MDA 67.9 50 23.8 99 25 7.4 31 51 19.3 192 33.4 26 10.7 28.7 52 17.4 353 11.9

LAC-TR-138 PAGE 21 Sample Grid #

Co-60 Cs-137 Eu-155 53 6.2 291 23 54 8.6 150 55 19.1 307 11.4 56 11.3 213 40 57 28.9 367 56.9 58 27 109 59 12.8 84.7 32.8 60

<MDA 30.7 61 48.5 137 62 76.9 206 63 28.3 226 64

<MDA 200 65

<MDA 122 66 26.7 66.8 67 19.2 185 27.3 68 276 1,300 52.2 69 44.8 252 26

LAC-TR-138 PAGE 22 SITE SOIL SAMPLING GRID MAP StACo RUIACTOR BOX WINC TURBINE

BUIIDII, ADKIN BUILDING

\\

4

.~5 ~33

,,30

\\

'Ile SWITCHYARD.

7

.q7

LAC-TR-138 PAGE 23 The unaffected area around LACBWR but within the owner controlled area was also checked. These samples were taken and analyzed as the samples taken from the affected area. The following results were obtained.

Location Co-60 Cs-137 Area W of #2 warehouse

<MDA 77.3 Area S of parking lot

<MDA 25.5 Area N of LACBWR admin building

<MDA 149 Area by G-3 ash silo

<MDA 38.2 Area outside G-3 offices

<MDA 76.3 Area by G-3 outfall

<MDA 17.2 All results are in pCi/kg To determine potential background Cs-i 37 levels for the area surrounding LACBWR several soil samples were taken at various locations outside the owner control area. These samples were collected and analyzed as per NUREG/CR-5849 as all the other samples. The following results were obtained.

Location 3.3 miles S @ boat landing along the Bad Axe River Pedretti Farm Substation E of plant Radio Tower NE of plant Junction of Cty Hwy 0 and K E of Stoddard @ Junction of Cty Hwy O and 162 NOTE: All a Co-60 22.7

<MDA

<MDA

<MDA Cs-137 381 227 35.00 188.00 66.5

<MDA ctivities are in pCi/kg

LAC-TR-138 PAGE 24 ATTACHMENT 1 SPENT FUEL RADIOACTIVITY INVENTORY Decay-Corrected to October 2010 Half Life Activity Half Life Radionuclide (Years)

(Curies)

Radionuclide (Years)

(Curies)

Ce-144 Cs-137 Ru-106 Cs-134 Kr-85 Co-60 Pm-147 Ni-63 Am-241 Pu-23 8 Pu-239 Pu-240 Eu-154 Cm-244 H-3 Eu-152 Am-242m 7.801 E-1 3.014 E+1 1.008 E+0 2.070 E+0 1.072 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+I 1.360 E+I 1.505 E+2 4.40 E-3 9.87 E+5 0.25 162 2.67 E+4 3.21 E+3 101 3.02 E+4 1.42 E+4 1.05 E+4 8.83 E+3 7.15 E+3 663 1.51 E+3 152 160 441 Sr-90 Pu-241 Fe-55 Ni-59 Tc-99 Sb-125 Eu-155 U-234 Am-243 Cd-i13m 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 2.770 E + 1 1.429 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+I 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+I 1.000 E+5 6.500 E+4 1.570 E+7 1.500 E+6 6.49 E+5 3.81 E+5 1.53 E+3 287 276 0.90 7.00 63.7 63.0 5.58 15.9 14.0 12.2 8.58 6.32 3.61 2.19 1.89 1.27 0.70 0.55 0.39 0.11 Total Activity = 2.12 E+6 Curies

0 LAC-TR-1*

PAGE2f ATTACHMENT 2 CORE 1NTERNAL/RX COMPONENT RADIONUCLIDE INVENTORY Estimated Curie Content Components Co-60 Fe-55 Ni-63 Other Nuclides]

T1/2 > 5y Total 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)

REACTOR VESSEL WAS PROCESSED, PACKAGED AND DISPOSED OF IN 2007 "IN FESW" COMPONENTS LISTED WERE PROCESSED, PACKAGED AND DISPOSED OF IN 2006 f

LAC-TR-1 PAGE 2"W ATTACHMENT 3 PLANT SYSTEMS INTERNAL RADIONUCLIDE INVENTORY - OCTOBER 2010 Nuclide Activity, in LiCi System Total ptCi Content Plant System

-I-RB Ventilation Offgas -

upstream of filter Offgas -

downstream of filters TB drains RB drains TB Waste Water RB Waste Water Main Steam Turbine Primary Purification Emergency Core Spray Overhead Storage Tank Seal Inject Fe-55 5

SYSTEM SYSTEM 49 111 10 611 757 3

259 SYSTEM 38 5

Alpha Co-60

-~1 I

I-t 80 101 Cs-137 REMOVED REMOVED 40 3

7 79 290 2

12 REMOVED 34 4

854 1,908 181 10,543 13,054 47 4,468 653 80 2,963 1,422 71 1,363 119 462 33 186 3,906 3,444 269 12,596 14,101 171 4,739 1,187 122

LAC-TR-13 8*&

PAGE 27TW ATTACHMENT 3 PLANT SYSTEMS INTERNAL RADIONUCLIDE INVENTORY - OCTOBER 2010

- (cont'd)

Nuclide Activity, in jiCi System Total

__Ci Content Plant System Fe-55 Alpha Co-60 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 291 SYSTEM SYSTEM 58 SYSTEM 437 4,367 SYSTEM SYSTEM SYSTEM 490 REMOVED REMOVED 94 REMOVED 720 7,000 REMOVED REMOVED REMOVED 5,021 1,004 7,531 75,310 5,802 1,156 8,688 86,667 i

LAC-TR-138 PAGE 283 ATTACHMENT 3 PLANT SYSTEMS INTERNAL RADIONUCLIDE INVENTORY - OCTOBER 2010

- (cont'd)

Nuclide Activity. in jiCi Plant System Fe-55 Alpha Co-60 System Total pCi Content Cs-137 Fuel Element Storage Well System Fuel Element Storage Well

- all but floor, Fuel Element Storage Well floor Resin lines 2,475 4

75,693 378 32,024 390 5

7,600 100 8,500 42,676 65 1,305,379 6,527 552,276 2,726 24,300 45,541 2,800 1,412,972 7,005 592,800 Main Condenser