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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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SKOYEN EXHIBIT 17 NUREG-1765 Basis Document for Large Early Release Frequency (LERF)
Significance Determination Process (SDP)
Inspection Findings that May Affect LERF U.S. Nuclear Regulatory Commission Office of Nuclear RegulatoryResearch Washington, DC 20555-0001
If the breach in the drywell pressure boundary results in a leakage2 to the environment greater than 200 x La it can constitute a large early release. Drywell sprays, if available, reduce the amount of the release. Data generated in the IPE program and reported in published PRAs suggests that for BWRs with Mark I containments, on average, about one-third of the core damage frequency consists of early containment failure sequences and about a third of the early containment failure sequences are large releases. Hence on average, about 0.1 of the core damage frequency in BWRs with Mark I containments constitutes LERF. Thus if a finding implies the existence of a breach in the drywell pressure boundary that would result in a drywell leakage rate > 200 x La, the large release probability of 0.1 increases essentially to 1.0. The conversion factor for Type B findings is, therefore, approximately (1.0-0.1) = 0.9 for findings of this type. This assumption neglects the effect of pool scrubbing for those sequences in which the in-vessel release passes through the suppression pool.
The risk significance can be determined by using the relationship given in Section 2.3 and assuming a total CDF of 10"S/ry for BWRs:
ALERF = 0.9 x I1V x (multiplier based on duration of degraded condition)
Using the multipliers given in Section 2.3 for each of the three (degraded condition) durations the following three ALERFs and the corresponding risk significance categories are obtained:
Duration ALERF Significance Category
> 30 days 9 x10"6 yellow 30-3 days 9 x10-7 white
< 3 days 9 x10s green If a finding identifies a degraded condition that involves a breach of the drywell pressure boundary that can potentially result in a leakage rate in excess of 200 x La and the duration of the degraded condition is also determined, one of the significance categories given above can be assigned to the finding.
2 Several studies, including NUREG/CR-4330, "Review of Light Water Reactor Regulatory Requirements," NUREG 1493, "Performance-Based Containment Leak-Test Program," and NUREGICR-6418, "Risk Importance of Containment and Related ESF System Performance Requirements," have been performed to determine the risk significance of various levels of containment leakage. While the results vary by plant and containment type, a containment leak rate of about 100 volume percent per day appears to constitute an approximate threshold beyond which the release may become significant to LERF. Design basis leakage from containment is determined by regulatory requirements to assure the containment leakage will be below the maximum allowable leak rate (denoted as Lj)set by Title 10 of the Code of FederalRegulations Part 100 dose limits that is incorporated in the plant technical specifications. Typical values of L, are 0. 1 containment volume percent per day for PWRs and 0.5 volume percent per day for Mark I and Mark II BWRs, and 0.2 volume percent per day for Mark III BWRs. Thus a LERF significant leakage rate from containment would be a rate greater than or equal to about 1000 1, for PWRs, 200 L, for Mark I and II BWRs, and 500 L. for Mark III BWRs. The 100 volume percent per day leakage rate is approximately equivalent to a hole size in containment of 2.5 - 3 inches in diameter for PWRs with large dry containments, 2 inches for PWRs with ice condenser containments, I inch for BWRs with Mark I and II containments, and 2.5 inches for BWRs with Mark III containments (Palla, 2001).
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