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Category:Legal-Pre-Filed Exhibits
MONTHYEARML1025100362010-09-0808 September 2010
PIC000001-Christopher I. Grimes Resume
ML1025006142010-09-0707 September 2010
NRC000001-Boric Acid Corrosion Guidebook, Revision 1
ML1025006402010-09-0707 September 2010
NRC000030-Revised Policy Statement on the Conduct of Nuclear Power Plant Operations
ML1025006382010-09-0707 September 2010
NRC000025-Figures 2 Through Figure 6
ML1025006332010-09-0707 September 2010
NRC000027-The Nature of Safety Cultua Review of Theory and Research
ML1024607632010-09-0303 September 2010
NSP000061-SER Section 3.0.4
ML1024607282010-09-0303 September 2010
NSP000050-Revised Prefiled Testimony of Northard/Petersen/Peterson-2009 Corrective Action Program Self-Assessment
ML1024607252010-09-0303 September 2010
NSP000041-Revised Prefiled Testimony of Northard/Petersen/Peterson-Performance Recovery Plan
ML1024607682010-09-0303 September 2010
NSP000069-NRC December 21, 2007 PI&R Report
ML1024607712010-09-0303 September 2010
NSP000070-PINGP Pride Initiative Focus 2010
ML1024606332010-09-0303 September 2010
NSP000035B-Revised Testimony of Northard/Petersen/Peterson-Root Cause Evaluation Report 01157726
ML1024606292010-09-0303 September 2010
NSP000020-Revised Testimony of Northard/Petersen/Peterson-Petersen Resume
ML1024606272010-09-0303 September 2010
NSP000035C-Revised Testimony of Northard/Petersen/Peterson-Root Cause Evaluation Report 01157726
ML1024606242010-09-0303 September 2010
NSP000019-Revised Testimony of Northard/Petersen/Peterson-Northard Resume
ML1024605692010-09-0303 September 2010
Applicant Revised Exhibit NSP000001-Revised Testimony of Steven Skoyen Resume
ML1024605662010-09-0303 September 2010
NSP000017-Revised Testimony of Steven Skoyen-NUREG-1765 Excerpts
ML1024605152010-09-0303 September 2010
Northern States Power Co. October 2010 Evidentiary Hearing on Safety Culture Contention, Hearing Exhibits
ML1024605512010-09-0303 September 2010
NSP000015-Revised Testimony of Steven Skoyen-EFR 1160372-03
ML1024605522010-09-0303 September 2010
NSP000006-Revised Testimony of Steven Skoyen-2006 AES Letter
ML1024605532010-09-0303 September 2010
NSP000010-Revised Testimony of Steven Skoyen-EFR 1160372-04
ML1024605542010-09-0303 September 2010
NSP000008-Revised Testimony of Steven Skoyen-Dominion Evaluation R-4448-00-01
ML1024605552010-09-0303 September 2010
NSP000012-Revised Testimony of Steven Skoyen-EC 15651
ML1024605562010-09-0303 September 2010
2009/09/03-Applicant-Revised Exhibit NSP000003-PINGP License Renewal Application Excerpts
ML1024605572010-09-0303 September 2010
NSP000013-Revised Testimony of Steven Skoyen-ACRS Letter
ML1024605582010-09-0303 September 2010
NSP000007-Revised Testimony of Steven Skoyen-CAP 1160372
ML1024605592010-09-0303 September 2010
NSP000011-Revised Testimony of Steven Skoyen-EC 15044
ML1024605602010-09-0303 September 2010
NSP000016-Revised Testimony of Steven Skoyen-ACRS July 2009 Meeting Transcript Excerpts
ML1024605622010-09-0303 September 2010
Applicant Revised Exhibit NSP000004-Root Cause Evaluation Report 01160372-01
ML1024605632010-09-0303 September 2010
NSP000018-Revised Testimony of Steven Skoyen-CE 01140617-03
ML1024605642010-09-0303 September 2010
NSP000009-Revised Testimony of Steven Skoyen-CE 1233806-2
ML1024605652010-09-0303 September 2010
Applicant Revised Exhibit NSP000002-Schematic Representation of PINGP Containment
ML1022504662010-08-13013 August 2010
NRC Staff Exhibit 63 - Boric Acid Corrosion Guidebook, Revision 1: Managing Boric Acid Corrosion Issues at PWR Power Stations EPRI, Palo Alto, CA (2001) 1000975 Pages 4-25 & 4-26
ML1022504702010-08-13013 August 2010
2010/08/13-NRC Staff Exhibit 4A - NUREG-1800 Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants (SRP-LR) (September 2005), (Excerpt)
ML1022504772010-08-13013 August 2010
NRC Staff Exhibit List
ML1022504962010-08-13013 August 2010
Northard Exhibit 43, Section 3.0.4. of NRC SER
ML1022504982010-08-13013 August 2010
Northard Exhibit 48, WM-0491 - Prairie Island Corrective Backlog
ML1022505002010-08-13013 August 2010
Skoyen Exhibit 17, NUREG-1765 Excerpt
ML1022505082010-08-13013 August 2010
Northard Exhibit 52, Pride Initiative Focus/2010
ML1022505132010-08-13013 August 2010
Skoyen Exhibit 18, CE01140617-03 Potential IWE Non-Compliance
ML1022505142010-08-13013 August 2010
Northard Exhibit 42, Event Cross References in 2009
ML1022504952010-08-10010 August 2010
Skoyen Rebuttal Exhibits List
ML1022505012010-08-10010 August 2010
List of Northard Rebuttal Exhibits
ML1024607592010-08-0606 August 2010
NSP000066-PINGP Corrective Backlog 2010
ML1022504752010-08-0404 August 2010
NRC Staff Exhibit 62 - Summary of July 28, 2010 Public Meeting to Discuss Observations and Lessons Learned During the Pilot Application of the Nuclear Energy Institutes Nuclear Safety Culture Assessment Process (August 4, 2010)
ML1021607812010-08-0404 August 2010
Intervenor-Exhibit 21-Xcel Management Review Committee Meeting Summary No. 2010-01, Msrc Meeting Date March 17 and 18, 2010 (NSPM Prod 00000267)
ML1021607642010-08-0404 August 2010
Intervenor-Exhibit 1-Resume for Christopher I. Grimes
ML1025006752010-08-0404 August 2010
NRC000058-Summary of July 28, 2010 Public Meeting to Discuss Observations and Lessons Learned During the Pilot Application of the Nuclear Energy Institute'S Nuclear Safety Culture Assessment Process (August 4, 2010)
ML1021503902010-07-30030 July 2010
Applicant-Northard Exhibit 4-INPO V NRC Nuclear Safety Culture Components
ML1021503922010-07-30030 July 2010
Applicant-Northard Exhibit 5 - Nuclear 2010 Business Plan Overview
ML1021503932010-07-30030 July 2010
Applicant-Northard Exhibit 10-FP-PA-HU-03, Rev. 6, Human Performance Observation Program.
2010-09-08
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Prairie Island Nuclear Generating Plant Application for Renewed Operating Licenses Technical and Administrative Information 2.4.7 Reactor Containment Vessels Units 1 and 2 Description The containment for each Unit was designed by Pioneer Service and Engineering Company and consists of two systems:
- A primary containment consisting of a free-standing, low-leakage steel vessel, including its penetrations, isolation systems, and heat removal systems designed to withstand the internal pressure accompanying a loss-of coolant accident, and
The primary containment system, also referred to as the Reactor Containment Vessel, consists of steel cylinder walls, a hemispherical dome, and an ellipsoidal bottom. A five foot wide annular space exists between the Reactor Containment Vessel walls and the Shield Building walls, and a seven foot clearance exists between the top of the vessel and Shield Building roof dome permitting in-service inspection and collection of containment out-leakage. The steel pressure vessel and all penetration assemblies form the primary containment boundary. The Reactor Containment Vessel cylinder wall outside diameter is 105'-3", dome radius is 52'-6", overall height is 206'-7 7/8", vessel wall thickness varies from 3/4" to 1 1/2", and vessel material is ASME SA516 Grade 70 steel.
With the exception of the unreinforced concrete placed underneath and near the ellipsoidal knuckle sides of the vessel, there are no structural ties between the Reactor Containment Vessel and the Shield Building above the foundation. The unreinforced concrete supporting its ellipsoidal bottom is tightly bonded to the outside of the vessel and the underlying reinforced concrete mat foundation, and is approximately 2'-10 1/2" thick (minimum). The mat foundation, which is common to the Shield Building, is 4'-0" thick and is placed on a 4" mud mat and weatherproofing membrane resting on controlled recompacted soils. The mat foundation also has structural continuity with the Auxiliary Building and Turbine Building foundations.
The Reactor Containment Vessel internal structure is for the most part conventionally reinforced concrete. The concrete forms floor slabs and compartments that support and protect the reactor pressure vessel (RPV) and components associated with engineered safeguards systems, and it provides the primary biological shield for the RPV. At various levels, concrete slabs are supported by structural steel framing which is supported off the central concrete core and peripheral steel columns. The internal structure is supported by reinforced concrete placed in the bottom and knuckle region of the Reactor Containment Page 2.4-36
Prairie Island Nuclear Generating Plant Application for Renewed Operating Licenses Technical and Administrative Information Vessel. Except for the contact at the base, the internal structure is completely isolated from the inside face of the Reactor Containment Vessel.
Reactor Containment Vessel major internal structural components include:
- Reactor/refueling cavity/biological shield wall
- Refueling floor - El. 755-0
- Operating floor - El. 733-9
- Mezzanine floor - El. 711-6
- Basement floor - El. 697-6 System Function Listing A comprehensive listing of functions associated with the Reactor Containment Vessels Units 1 and 2, or specific components contained in the structure, is provided in the summary below.
Code RCV-01 Cri 1 Cri 2 Cri 3 Reactor Containment Vessels and their internal FP EQ PTS AT SB structures provide structural support to safety X related components.
Comment: Reactor Containment Vessels and their internal structures are designed to provide structural support to safety related components relied upon to remain functional during and following design-basis events to ensure satisfactory accomplishment of any of the functions identified in 10 CFR 54.4(a)(1).
Code RCV-02 Cri 1 Cri 2 Cri 3 Reactor Containment Vessels and their internal FP EQ PTS AT SB structures provide flood protection from internal X flooding events.
Comment: Reactor Containment Vessels and their internal structures provide flood protection from internal flooding events.
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