ML20199E050
| ML20199E050 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 01/23/1998 |
| From: | Grobe J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Kingsley O COMMONWEALTH EDISON CO. |
| References | |
| 50-254-97-27, 50-265-97-27, EA-97-591, NUDOCS 9802020082 | |
| Download: ML20199E050 (53) | |
See also: IR 05000254/1997027
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January 23, 1998
EA 97-5911
Mr. Oliver D. Kingsley
President, Nuclear Generation Group
Commonwealth Edison Company
ATTN: Regulatory Services
Executive Towers West 111
1400 Opus Place, Suite 500
Downers Grove,IL 60510
SUBJECT: NRC PREDECISIONAL ENFORCEMENT CONFERENCE SUMMARY
Dear Mr. Kingsley;
On January 9,1998, members of Commonwealth Edison staff met with NRC personnelin the
Region Ill office located in Lisle, Illinois, to discuss the apparent violatiors and unresolved item
Ider,tified in the NRC Inspection Report Nos. 50-254/97027(DRS); 50-265/97027(DRS), The
inspection report described several failures in your program to properly implement key aspects
pertaining to the American Society of Mechanical Engineers Code,Section XI, Class 1 and 2
pressure testing. The conference was held at the request of Region Ill.
Commonwealth Edison agreed with the facts detailed in the inspection report and did not i
contest the apparent violations. During the conference, Commonwealth Edison discussed a
number of causes for these violations which included: lack of senior management review, lack
of awareness of the applicable requirements, personneljudgement and performance errors, ,
and ineffective self assessment. Additionally, Commonwealth Edison addressed the
unresolved item and acknowledged that the VT-2 exaraination of the Class 1 system completed
on June 22,1997, was inadequate with respect to Code requirements. Commonwealth Edison
based this conclusion on a January 3,1998, re-enactment of the VT-2 examination, which
- demonstrated that leakage (if present) would not hcva been detected in several areas of the
. Class 1 system boundary. In response to these apparent violations and unresolved item,
Commonwealth Edison had initiated generally comprehensive corrective actions. NRC staff
review of these corrective actions is in process and will be addressed in subsequent
correspondence. Commonwealth Edison's analysis of these problems and the corrective
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actions, both planned and in process, was contained in a handout that was provided at the
conference.
A copy of that handout, a summary of the apparent violations, and an attendance list from the
conference are enclosed with this summary.
9902020082 990123 -
PDR ADOCK 05000254
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O. D. Kingsley 2 January 23. 1998
in accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of :: tis letter and its
enclosures wih he placed in the NRC Public Document Room (PDR).
Sincerely,
original signed by J. A. Grobe
John A. Grobe, Director
Division of Reactor Safety
Docket Nos. 50-254; 50-265
Enclosures: As stated
cc w/encls: M. Wallace, Senior Vice President,
Corporate Services
J. Perry, Vice President, BWR Operations
E. S. Kraft, Jr., Site Vice President ~
Liaison Officer, NOC-BOD
D. A. Sager, Vice Presloent,
Generation Support
D. Farrar, Nuclear Rogulatory
Services Manager
1. Johnson, Licensing Operations Manager
Document Control Desk - Licensing
Quad Cities Station Manager
C. C. Peterson, Regulatory Affairs Manager
Richard Hubbard
Nathan Schloss, Economist,
Office of the Attomey General
State Liaison Officer
Chairman, Illinois Commerce Commission
W. D. Leech, Manager of Nuclear,-
MidAmerican Energy Company
Distribution
Docket File w/encls Rlli PRR w/encls Rlil Enf. Coordinator w/encls
PUBLIC IE-01 w/encls SRI, Quad Cities w/encls TSS w/encls
LPM, NRR w/encls J. L. Caldwell, Rlli w/encls R. A. Capre, NRR w/encls
DRP w/encls A. B. Beach, Rill w/encls DOCDESK w/encls
DRS w/encls CAA1 w/encls J. Lieberman, OE
J. Goldberg, OGC R. Zimmerman, NRR
DOCUMENT NAME: G:DRS\qua01148.drs
t. . .. .m.,,, w c . co w.c n=.. m r.com.c u a v. %
OFFICE Rill (i Rill ,. C- Rlli l C- l Rlli , .,
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NAME MHolmberg:sd & JGavulaW MRing W JGrob(,[/W
DATE 01/H/98 01/ll/98 l{ V 01h5/98 01/759/
OFMCIAL RECORD COPY
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Appendix G .
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Pre-Decisional Enforcement
Conference
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January 9,1998 ,
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Agenda
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Introduction Edward S. Kraft, Jr.
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Onening Remarks Bill Pearce
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Background Chris Hebel l
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Operational Safety Focus Bob Svaleson !
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Findings Issues and Actions Jack Purkis
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Comed VT-2 Program John Hutchinson
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Closing Remarks Edward S. Kraft, Jr.
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Introduction ,
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Edward S. Kraft. a. .
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Quad Cities Station
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Site Vice President
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Opening Remarks
Bill Pearce
Quad Cities Station
General Station Manager
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Opening Remarks
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Comed and Quad Cities Station Enderstand the !
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Seriousness of the Issues
Recognize Increased Risk of L sing Nuclear Heat
and Saturated Conditions Before Completion of
Pressure and Leak Tests on the Reactor Pressure
Vessel
Complex Story Involving Several Poor Decisions
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We Accept the Violations
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Background
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Chris Hebel
Quad Cities Station
Test Director
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Background
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Explanation of Pressure Test
Pressure Control Methods
>> Temperature Control Methods
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Background
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June 1996 - Began Development ofNew
Pressure / Temperature Curves (P/T)
Needed Prior to 16 EFPY
Utilized New Methodology
>> Addressed Lower Head P/T Limits
P/T Limits Shifted 20 Higher
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Background
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July 10,1996 - Meeting to Discuss New P/T
Curves
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Concern Regarding Controlin Narrow t
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Temperature Band
Discussed Options
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September 20,1996 - Submitted New P/T
Curves to NRC
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P/T Curve
, 1020 _ #
$ 212 F Mode 3 Limit
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1010 Operating Margin For
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WITH THE NEW '
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( AT Bulk Water
TEMPERATURE
CURVE
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Metal Temp.
980
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180 190 200 210 220
Temperature ( a F )
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September 26,1996 - Implemented Upgraded
Technical Specifications
Defined Mode Change at 212 F
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October 1996 - Meeting to Consider Performing
Pressure Tests While Critical
Decided Not to Test Critical
Winter 1996 - Developed Plan to Establish
Primary Containment Prior to Pressure Test
Allow Reactor Water Temperature to Exceed 212 F
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Background
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March 25,1997 - New Pressure / Temperature
Curves Inserted Into Technical Specifications
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May 1997 - Pre-Test Review Recognized Mode 3
Requirements at 212 F
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Could Not Meet Mode 3 Requirements During
Test
Reconsidered Options
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Background
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Options Considered
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Control Temperature in Tight Band
Request Regulatory Relief
Test While Critical
Option Selected
Perform 850 psi Pre-Test and Perform Code
Pressure Vessel Test After Criticality
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Applied Routine Processes
Changed Procedures
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Performed Simulator Run
50.59 Screenmg
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Startua Checklist
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Background
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May 24,1997 - Performed Preliminary Pressure
Leakage Test at 850 psi
- June 8,1997 - E nit 2 Critical
- June 12,1997 - Enit 2 Critical
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June 22,1997 - Performed Code Required Vessel
Pressure Leakage Test !
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October 1,1997 - Recognized 10CFR50
Appendix G Requirement for Testing Before
Reactor Critical
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Operational Safety Focus
Bob Svaleson
Quad Cities Station
Operations Manager
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Operational Safety Focus
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Examples of Proper Actions
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What Should Have Happened?
. Should Have Known of Appendix G Change
>> Should Have Had PORC Review
On-Site Review Should Have Raised the Issue
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Findings -
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Issues and Actions
Jack Purkis
Quad Cities Station
System Engineering Supelvisor
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Violation of Appendix G
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NRC Inspection Finding
Failure to Perform Lnit 2 ASME Code
Section XI Class 1 Leakage Test Prior to
Criticality (EEI 01)
We Agree With This Finding
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Violation of Appendix G
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Lack of Awareness ofNew Appendix G Revision
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-Ineffective Tracking ofNew or Changed Rules and
Regulations
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-Insufficient Research to Identify Appendix G Change
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Violation of Appendix G
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Corrective Actions
SVP Will Discuss This Finding at Next Station "All
Hands" Meeting
An Assessment Has Been Completed to Determine
Which Procedures Should Receive an Additional
PORC Review Prior to Implementation.
Station Has Put in Place, in Interim, a Process to
Ensure Notifications of All Rule Changes
All Applicable Rule Changes Will Be Distributed and
Tracked by OPEX
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Violation of Appendix G
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Corrective Actions (Cont.)
The Rule Changes for the Last 3 Years Have Been
Reviewed for Applicability
>> Xew Corporate Wide System for Tracking and
Distributing New or Changed Rules and Regulations Is
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Station Startup Procedure Has Been Revised to Include
the Appendix G Requirement
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Operator Requalification Lessons Learned Which Will
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Violation of Appendix G
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Corrective Actions (Cont.)
Station Procedures and Processes Are Being Reviewed
to Verify All Requirements of Appendix G Are Being
Met. This Will Be Completed by February 28,1998
The 10CFR50 in the Technical Library Will Be
Handled As a Controlled Copy. This Will Be
Completed by January 31,1998
Training Will Be Completed for Persons Who Conduct
Cross Discipline Review to Ensure Applicable Codes
and Standards Are Reviewed.
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Violation of Appendix G
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Safety Significance
Safety Significance Was Mitigated .
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- Pre-Te.st at Approximately 840 psig Conducted in the Same i
Manner As Code Test
- Probability of Test Propagating a Crack Through Wall Is
Approximately the Same for a 850 or 1005 psig Test
- 55% Of Reactor Vessel Shell Welds Had Volumetric Exams
During Outage
- Unidentified Leakage at Startup Was Approximately 0.5
Gallons Per Minute
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Inadequate 50.59 Safety Evaluation
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NRC Inspection Finding
Failure to Perform Adequate 50.59 Safety
Evaluation for the Test Procedure (QCOS
0210-10} (EEI 06)
We Agree With This Finding
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! Inadequate 50.59 Safety Evaluation
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j Insufficient Research by the Preparer and
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Reviewer to Identify All Requirements
>> Xo Direct Tie From Applicable UFSAR Sections
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Inadequate 50.59 Safety Evaluation
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Corrective Actions
The Pressure Test Procedure, QCOS 0210-10, Has Been Retired
System Engineering Supervisor Discussed This Finding and
Expectations for Ad. equate Research With Qualified 50.59
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Training Will Be Conducted for 50.59 Preparers and Reviewers
on This Finding and Other Identified Weaknesses. This Will Be
Completed by March 31,1998
Engineering Assurance Group Will Perform Reviews on
Selected 50.59 Screenings Until Effectiveness Is Assured
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Inadequate 50.59 Safety Evaluation
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Corrective Actions (Cont.)
Applicable Sections of the UFSAR Are Being Revised to
- Reference Appendix G. This Will Be Completed by January 31,
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Inadequate 50.59 Safety Evaluation
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Safety Significance
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Lack of an Adequate 50.59 Is a Significant Issue
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Code Pressure Test
XRC Inspection Finding
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Missed or Inadequate Completion of ASME l
Code Section XI Class 1 and 2 Pressure Test -
Five Examples (EEI 03)
>> We Agree With This Finding
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Code Pressure Test
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Incomplete VT-2 Inspection of Unit 2-Reactor Head
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Failure to Perform Adequate VT-2 (Use of Leak
Detection System) of Unit 2 Reactor Vessel Flange
Failure to Perform Adequate VT-2 (Use of Leak
Detection System) of Unit 1 Reactr Vessel Flange
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Failure to Perform Class 2 Pressure Test of Unit 1
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Head Flange Leak Detection System
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Failure to Perform Class 2 Pressure Test of Unit 2
Head Flange Leak Detection System
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Code Pressure Test
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Incomplete Inspection
- Poor Judgment on Part ofIndivklual Who Inspected the ;
Reactor Head and Harsh Environmental Conditions l
- Belief That the Leak Detection System Did Not Ivieet the
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Functional Requirements ofIWA-5243
- Performance of the VT-2 Pressure Test of the Class 2 Portion
of tla Leak Detection System Was Not Captured in
Procedures or Predefines a
Ineffective Self Assessment
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Code Pressure Test
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Corrective Actions
Pressure Test of Unit 2 Will Be Performed Prior to
Taking the Reactor Critical
Unit 1 Leak Detection System Was Inspected at 700
Psig
Procedure Has Been Revised to Specifically Require
Walking Around Reactor Vessel Head and Opening the
Access Ports in the Insulation Wall to Perform the VT-2
Inspection Methods Now Require Individual
Acknowledgment of Each Inspection Location As They
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M-issed or Inadequate
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Code Pressure Test
Corrective Actions (Cont.)
VT-2 Examiners Have Been Retrained on Specific ;
VT-2 Requirements
Q&SA Will Overview Next Class 1 System Pressure
Test
>> Corporate ISI Assessment of Quad Cities Program
(January 5 - 16)
Corporate Pressure Test Assessment (November 18 -
January 16)
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Code Pressure Test
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Corrective Actions (Cont.)
Independent Industry Expert Assessment of Pressure
Test and ISI Programs (November 18 - January 16)
All Assessments at Quad Cities Will Be Completed
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and Issues Appropriately Dispositioned Before Taking
the Respective Reactor Critical
- All Identified Issues Will Be Shared With Other
Comed Sites
) Identified Issues at Quad Cities Will Be Resolved Prior
- to September 1,1998
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M.issed or Inadequate
Code Pressure Test
Documentation Issues
No Sign Off for Reactor Head Inspection
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- Procedural Omission
>> Programmatic and Procedural Deficiencies Caused the
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- - VT-2 of Class 2 Leak Detection System Not Incorporated
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Procedural Steps Were Completed After Exiting the
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Drywell From the Class 1 Test
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95E9d, Missed or Inadequate -
Code Pressure Test
Corrective Actions for Documentation Issues
Reactor Head Sign Off Has Been Included
All Pressure Testing Procedures Will Be Reviewed
for Omissions and Errors Prior to Use, or by
September 1998, Whichever Occurs First
Management Expectations for Procedural Adherence
to Assure Safety Will Be Reinforced
The Class 1 Test Procedure Now Requires Individual
Sign Offs As Inspections Are Performed
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Significance
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These Programmatic Deficiencies Are a Significant
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Opportunities to Identify
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In November 1997, Station Management
Discovered, That in July 1997, an Individual Felt l
That One Example of an Inadequate Inspection !
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This Has Been Discussed With the NRC Staff Earlier
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Adequacy of VT-2 Inspections
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Unresolved Item Regarding How All Accessible
Surfaces and Insulation Joints of Class 1 Piping
VT-2 Examinations Were Completed in the
Elapsed Times Noted
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The Station Has Determined That These
Inspections Were Not Adequate
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Pressure Test of Unit 1 and Unit 2 Will Be Performed
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- This Finding Will Be Included As Part of the
Comprehensive Assessments Currently Being
Performed. Additional Corrective Actions Will Be .
Included As Part of an Overall Improvement Plan
i Developed From These Assessments
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Corporate Assessment and VT
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John Hutchinson
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VT Inspection Program
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Preliminary Assessment Of SPPVT-2-1
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Implementation
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A Review of the Recent L nidentified Leakage on
Startup From Our Lnits
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These Numbers Indicate That VT-2 Program in a
Gross Sense Is Identifying Leakage
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Based on a Review, the Requirements of SPPVT-
2-1 and SPP2-1-0 Are Being Met at the Other
Sites
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Comed Program Assessment
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VT Program Administered by Common
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Procedures SPPVT-2-1 and SPP2-1-0
Inspectors Are Qualified and Certified to
Procedure SPP2-1-0
SPPVT-2-1 Contains Requirements Provided in
IWX 5000
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Validation of the VT Pro ~ gram Implementation Is
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Included in a Broader Assessment of Section XI
Implementation Resulting From the Quad Cities
Inspection
Six Site Assessment of the In-Service Inspection
Program Relating to Piping and Supports, Pressure
Testing, and the Repair and Replacement
Programs
Assessment Based on Inspection Procedures
73051,73052,73753,73755 and NUPEG 0800 4s
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Comed Program Assessment
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Team Composed of Team Lead, Corporate ISI
Engineer, Q&SA Representative, Plus Industry
Consultants
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Duration of 3 Months Beginning at Quad Cities on .
January 5,1998 and Completing at Zion on March
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Focus on Consistency Between Comed Sites and
Comed and the Industry. Full Compliance With
Applicable Regulatory and ASME Code
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A Unicom Company
Closing Remarks
Edward S. Kraft, Jr.
Quad Cities Station
Site Vice President
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. , , .., Attachment 2
The apparent violations discussed at the predecisional enforcement conferehce are -
subject to further review and subject to change prior to any resulting enforcement action
1. 10 CFR 50.60, " Acceptance criteria for fracture prevention measures to lightwater
nuclear power reactors for normal operation,' requires, in part, that all light-water reactor
plants must meet the fracture toughness requirements for the reactor coolant pressure
boundacy as set forth in Appendix G, ' Fracture Toughness Requirements."
10 CFR 50, Appendix G, IV.2(d), " Pressure-Temperature Limits and Minimum
Temperature Requirements," requires, in part, that pressure tests ar,d leak tests of the
reactor vessel that are required by Section XI, " Rules for inspection of Nuclear Power
Plant Components," of the American Society for Mechanical Engineers (ASME) Code
must be completed before the core is critical.
ASME Code Section XI (1989 Edition, no Addenda), Table IWB-2500-1, " Examination
Category C P, All Pressure Retaining Components,' at item B15.10 required a system
leakage test (lWB-5221) and visual VT-2 examination of the pressure retaining
boundary of the reactor vessel each refueling outage.
Contrary to the above, on or about June 22,1997 the licensee had failed to perform an
ASME Code Section XI leakage test of the reactor vessel prior to core criticality from the
Unit 2 refueling outage.
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2. 10 CFR 50.59(a)(1), " Changes, tests, and experiments,' states, in part, that a licensee
may make changes in the facility as described in the safety analysis report (currently
referred to as the Updated Final Safety Analysis Report (UFSAR)) and may conduct
tests not described in the safety analysis report, without prior Commission approval,
unless the proposed change or test involves an unreviewed safety question.
l 10 CFR 50.59(a)(1) states, in part, that a proposed change shall be deemed to involve
an unreviewed safety question if the malfunction of equipment important to safety
previously evaluated in the safety analysis report may be increased.
10 CFR 50.59(b)(1) requires, in part, that the licensee shall maintain records of changes
in the facility as described in the safety analysis report and tests pursuant to paragraph
(a) of this section. These records must include a written safety evsluation which
provides the bases for the determination that the chai.ge or test does not involve an
unreviewed safety question.
UFSAR Section 5.2.4.7, " System Leakage and Hydrostatic Pressure Tests" stated, in
part, that system leakage and hydrostatic tests are conducted in accordance with
IWB-5000, " System Pressure Tests."
IWB-5210, " Test," stated, in part, that pressure retaining components shall be visually
examined by the method specified in Table IWB-2500-1, " Examination Category B-P."
The apparent violations discussed at the predecisional enforcement conference are
subject to further review and subject to change prior to any resulting enforcement action
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The apparent violations discussed at the predecisional enforcement conference are
subject to further review and subject to change prior to any resulting enforcement action
Table IWB-2500-1, Note 5, required, in part, that the system leakage test (lWB-5221)
shall be conducted prior to plant startup following each refueling outage."
Contrary to the above, the safety evaluation screenings, which provided the bases for
the determination that the change to procedure QCCS 0201-10, ' Reactor Vessel and
Class One Piping Leak Test at Power Operation," (Revision 0, authorized June 3,1997
and Revision 1, authorized on June 6,1997) did not involve an unreviewed safety
question, were inadequate. The evaluation screenings did not address UFSAR Section
5.2.4.7 which stated (by reference to IWB-5000) leakage tests shall be conducted prior
to plant startup following each refueling outage. Specifically the evaluation screenings
stated that the SAR does not define when the visual examination is performed and the
SAR does not describe leak testing of the reactor vessel with the reactor at power. As a
result of these inadequate safety evaluation screenings, operation of Unit 2 was
permitted prior to completing this leakage test which constituted an unreviewed safety
question, since the probability for a loss of coolaat accident had been increased.
3. Quad Cities Unit 1 and 2 Technical Specification 4.0.E " Surveillance Requirements'"
required implementation of the ASME Code Section XI inservice inspection and testing
requirements for Code Class 1 and 2 components.
ASME Code Section XI (1989 Edition, no Addenda), IWB-5210(b), ' System Test
Requirements - Test," stated, in part, that system pressure tests cnd visual
examinations shall be conducted in accordance with IWA-5000 " System Pressure
Tests."
i
a. iWA-5242(a), " Insulated Components," stated that visual examination VT-2 may
be conducted without the removal of insulation by examining the accessible and
exposed surfaces and joints of the insulation. Essentially vertical surfaces of
insulation need only be examined at the lowest elevation where leakage may be
detectable.
IWA-5242(b) stated that when examining insulated components, the examination
of surrounding area (including floor areas or equipment surfaces located
undemeath the components) for evidence of leakage, or other areas to which
such leakage may be channeled, shall be required.
Contrary to the above, on June 22,1997, the licensee failed to perform an
adequate VT-2 inspection of the reactor vessel head area during the ASME
Code Section XI pressure test of Class 1 systems for Unit 2. Specifically, the
VT-2 examination performed failed to include the lower edge and floor areas of
the refueling cavity at radial locations along the vertical insulation wall
surrounding the reactor vessel head. Further, the licensee failed to utilize an
inspection port in this vertical head insulation to perform direct VT-2 inspections
The apparent violations discussed at the predecisional enforcement conference are
subject to further review and subject to change prior to any resulting eniorcemens action
a - )
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3
e'
.
The apparent violations discussed at the predecisional enforcement confere"nce are -
subject to further review and subject to change prior to any resulting enforcement action
of the vessel head.
b. IWA-5243, " Components With Leakage Collection Systems," stated that where
leakagos from components are normally expected and collected (such as valve
stems, pump sea's, or vessel flange gaskets) the visual examination VT-2 shall
be conducted bj verifying that the leakage collection system is operative.'
l. Contrrty to the above, on June 22,1997, the licensee failed to perform
an ade.quate VT-2 examination of the reactor vessel head flange joint
during the Unit 2 ASME Code Section XI Class 1 system leakage test.
Specifically, the VT-2 examination performed failed to verify the absence
of leakage from the head flange, as monitored and collected by the
reactor pressure vessel flange seat leakage detection system. Further,
the VT-2 examination failed to verify that this system was operative.
ii. Contrary to the above, on May 3,1996, the licensee failed to perform an
adequate VT-2 examination of the reactor vessel head flange joint during
the Unit 1 ASME Code Section XI Class 1 system leakage test.
Specifically, the VT-2 examination performed failed to verify the absence
of leakage from the head flange, as monitored and collected by the
reactor pressure vessel flange sealleakage detection system. Further,
the VT-2 examination failed to verify that this system was operative.
c. Table IWC-2500, " Examination Category C-H, All Pressure Retaining
Components," required a pressure test (IWC-5221) and VT-2 examination of
pressure retaining boundaries of Code Class 2 systems during each inspection
period.
i. Contrary to the above, as of February 18,1996 (the end of first inspection
period of the third code interval for Unit 1), the licensee failed to perform
l a Code Class 2 system leakage test (IWC-5221 or Code relief PR-02) of
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the Unit 1 reactor pressure vessel head flange seal leak detection system
within the required codo inspection period. ,
ii. Contrary to the above, as of March 10,1996 (the end of first inspection
yriod of the third code interval for Unit 2), the licensee failed t Nrform
a Code Class 2 system leakage test (IWC-5221 or Code relie 4-02) of
the Unit 2 reactor pressure vessel head flange seal leak dek.,on system
within the required code inspection period.
The apparent violations discussed at the predecisional enforcement conference are
subject to further review and subject to change prior to any resulting enforcement action
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} ),. Attachment 3 !
ATTENDANCE LIST
(Partial)
DATE: JANUARY 9,1998
COMED
S. Perry, BWR Vice President
E. Kraft, Jr., Quad Cities Site Vice President
J. Williams, Project Manager, Nuclear Generation Group
L Pearce, Station General Manager, Quad Cities
D. Cook, Station Manager, Quad Cities
R. Freeman, Site Engineering Manager, Dresden
B. Rybak, Senior Nuclear Licensing Administrator
R. Ruffin, Regulatory Assistant Coordinator, Dresden
G. Contrady, Site Engineer Programs Lead, Byron .
k
G. Perkins, inspection and Testing Group Lead (Acting), Quad Cities
G. Knapp, Inservice Testing Coordinator, Quad Cities
T. Wojcik, Senior Engineer, Quad Cities
M. Plumadore, Design Engineer, Quad Cities
G. Wald, Corporate Communication
'
R. Wychocki. ISI Engineer
D. Evans, Liaison
J, Arnold, Programs Engineer, Quad Cities .
H. Do, ISI Engineer
B. Helfrich, Sr. Counsel-Nuclear
K. Bethard, Regulatory Assurance, Quad Cities
J. Blomgren, S/G & RPV Projects Manager
T. Fuchs, Nuclear Licensing Administrator .
C. Peterson, Regulatory Affairs, Quad Cities
J. Lewarid, Licensing Opemtionc
R. Svaleson, Operations Manager, Quad Cities
J. Purkis, System Engint. ' ing Supervisor, Quad Cities
C. Hebel, Test Director, Quad Cities
J. Hutchinson, System Material Manager
L. Waldinger, Nuclear Oversight Manager
' F. Famulari, Q&SA Manager, Quad Cities
Nf1C
A. Beach, Regional Adininistrator, Rill
G. Grant, Director, Division of Reactor Projects, Rill
J. Jacobson, Deputy Director, Division of Reactor Safety, Rill
. J. Heller, Enforcement Coordinator, Rlll
M. Holmberg, inspector, Rlli
M. Ring, Chief, Projects Branch 1, Rlli
J.' Gavula, Ch!ef Engineering Specialist Branch 1, Rill
K. Walton, NRC Resident inspector, Quad Cities
lilinois DeLartment of Nuclear Safety
R. Ganser, Quad Cities Resident inspector
The Disoatch and R.I. Argus Newscacers
R. Pearson, Reporter
StructuralIntegritv. Associates
P. Riccardella, Consultant
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