ML042260093
| ML042260093 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 08/12/2004 |
| From: | Reynolds S Division Reactor Projects III |
| To: | |
| References | |
| IR-04-003 | |
| Download: ML042260093 (124) | |
See also: IR 05000266/2004003
Text
August 12, 2004
Mr. Dennis L. Koehl
Site Vice-President
Point Beach Nuclear Plant
Nuclear Management Company, LLC
6590 Nuclear Road
Two Rivers, WI 54241-9516
SUBJECT: POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2
NRC INTEGRATED INSPECTION REPORT 05000266/2004003;
Dear Mr. Koehl:
On June 30, 2004, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated
inspection at your Point Beach Nuclear Plant, Units 1 and 2. The enclosed report documents
the inspection findings which were discussed on July 16, 2004, with you and members of your
staff.
The inspection examined activities conducted under your license as they relate to safety and to
compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
Based on the results of this inspection, one self-revealed finding and five NRC-identified
findings of very low safety significance were identified, five of which involved a violation of NRC
requirements. However, because the violations were of very low safety significance and
because the issues were entered into the licensees corrective action program, the NRC is
treating these violations as Non-Cited Violations (NCVs) consistent with Section VI.A of the
NRC Enforcement Policy. Additionally, a licensee-identified violation determined to be of very
low safety significance is listed in this report.
Two of the findings involved the installation of steam generator nozzle dams during the Unit 1
refueling outage while the reactor coolant system was at a reduced inventory level. The
performance issues associated with these findings could have resulted in nuclear and industrial
safety consequences of higher significance.
D. Koehl -2-
In addition to the routine NRC inspection and assessment activities, Point Beach performance
is being evaluated quarterly as described in the Annual Assessment Letter - Point Beach
Nuclear Plant, dated March 4, 2004. Consistent with Inspection Manual Chapter (IMC) 0305,
Operating Reactor Assessment Program, plants in the multiple/repetitive degraded
cornerstone column of the Action Matrix are given consideration at each quarterly performance
assessment review for (1) declaring plant performance to be unacceptable in accordance with
the guidance in IMC 0305; (2) transferring to the IMC 0350, Oversight of Operating Reactor
Facilities in a Shutdown Condition with Performance Problems, process; and (3) taking
additional regulatory actions, as appropriate. On April 30, May 24, and June 16, 2004, the NRC
reviewed Point Beach operational performance, inspection findings, and performance indicators
for the second quarter of 2004. Based on this review, we concluded that Point Beach is
operating safely. We determined that no additional regulatory actions, beyond the already
increased inspection activities and management oversight, are currently warranted.
If you contest the subject or severity of an NCV in this report, you should provide a response
within 30 days of the date of this inspection report, with the basis for your denial, to the
U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk,
Washington, DC 20555-0001, with a copy to the Regional Administrator, U.S. Nuclear
Regulatory Commission - Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532 - 4352;
the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; and the Resident Inspector Office at the Point Beach Nuclear Plant facility.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter,
its enclosure, and any response you provide will be available electronically for public inspection
in the NRC Public Document Room or from the Publicly Available Records System (PARS)
component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web
site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Steven A. Reynolds, Acting Director
Division of Reactor Projects
Docket Nos. 50-266; 50-301
Enclosure: Inspection Report 05000266/2004003; 05000301/2004003
w/Attachment: Supplemental Information
See Attached Distribution
D. Koehl -3-
cc w/encl: F. Kuester, President and Chief
Executive Officer, We Generation
J. Cowan, Executive Vice President
Chief Nuclear Officer
D. Cooper, Senior Vice President, Group Operations
J. McCarthy, Site Director of Operations
D. Weaver, Nuclear Asset Manager
Plant Manager
Regulatory Affairs Manager
Training Manager
Site Assessment Manager
Site Engineering Director
Emergency Planning Manager
J. Rogoff, Vice President, Counsel & Secretary
K. Duveneck, Town Chairman
Town of Two Creeks
Chairperson
Public Service Commission of Wisconsin
J. Kitsembel, Electric Division
Public Service Commission of Wisconsin
State Liaison Officer
To receive a copy of this document, indicate in the box:"C" = Copy without enclosure "E"= Copy with enclosure"N"= No copy
OFFICE RIII RIII RIII RIII
NAME MKunowski:ags SOrth for KRiemer PLouden SReynolds
DATE 8/12/04 8/12/04 8/12/04 8/12/04
OFFICIAL RECORD COPY
D. Koehl -4-
ADAMS Distribution:
WDR
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GEG
PGK1
CAA1
C. Pederson, DRS (hard copy - IRs only)
DRPIII
DRSIII
PLB1
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ROPreports@nrc.gov
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket Nos: 50-266; 50-301
Report No: 05000266/2004003; 05000301/2004003
Licensee: Nuclear Management Company, LLC
Facility: Point Beach Nuclear Plant, Units 1 and 2
Location: 6610 Nuclear Road
Two Rivers, WI 54241
Dates: April 1 through June 30, 2004
Inspectors: P. Krohn, Senior Resident Inspector
M. Morris, Resident Inspector
D. Karjala, Resident Inspector
P. Higgins, Resident Inspector
S. Ray, Senior Resident Inspector
M. Kunowski, Project Engineer
R. Alexander, Radiation Specialist
B. Jorgensen, NRC Consultant
M. Holmberg, Reactor Inspector
D. Schrum, Reactor Engineer
T. Ploski, Senior Emergency Preparedness Inspector
C. Zoia, Operator Licensing Examiner
Approved by: P. Louden, Chief
Branch 7
Division of Reactor Projects
Enclosure
TABLE OF CONTENTS
SUMMARY OF FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
REPORT DETAILS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
Summary of Plant Status . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
1. REACTOR SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
1R01 Adverse Weather Protection (71111.01) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
1R04 Equipment Alignment (71111.04) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
1R05 Fire Protection (71111.05) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
1R07 Heat Sink Performance (71111.07) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
1R08 Inservice Inspection Activities (IP 71111.08) . . . . . . . . . . . . . . . . . . . . . . . . . . 10
1R11 Licensed Operator Requalification (71111.11) . . . . . . . . . . . . . . . . . . . . . . . . . 14
1R12 Maintenance Effectiveness (71111.12) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
1R13 Maintenance Risk Assessment and Emergent Work Evaluation (71111.13) . . 15
1R14 Personnel Performance During Non-Routine Plant Evolutions and Events
(71111.14) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
1R15 Operability Evaluations (71111.15) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
1R16 Operator Workarounds (OWAs) (71111.16) . . . . . . . . . . . . . . . . . . . . . . . . . . 17
1R19 Post-Maintenance Testing (PMT) (71111.19) . . . . . . . . . . . . . . . . . . . . . . . . . 18
1R20 Refueling and Outage Activities (71111.20) . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
1R22 Surveillance Testing (71111.22) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20
1R23 Temporary Plant Modifications (71111.23) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
1EP4 Emergency Action Level and Emergency Plan Changes (71114.04) . . . . . . . . 22
1EP6 Drill Evaluation (71114.06) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22
2. RADIATION SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
2OS1 Access Control to Radiologically Significant Areas (71121.01) . . . . . . . . . . . . 23
2OS2 As-Low-As-Is-Reasonably-Achievable Planning And Controls (ALARA)
(71121.02) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28
4. OTHER ACTIVITIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30
4OA1 Performance Indicator (PI) Verification (71151) . . . . . . . . . . . . . . . . . . . . . . . . 30
4OA2 Identification and Resolution of Problems (71152) . . . . . . . . . . . . . . . . . . . . . . 31
4OA3 Event Follow-up (71153) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37
4OA4 Cross-Cutting Aspects of Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40
4OA5 Other Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41
4OA6 Meetings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 66
4OA7 Licensee-Identified Violation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 66
ATTACHMENT: SUPPLEMENTAL INFORMATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 67
KEY POINTS OF CONTACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
ITEMS OPENED, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
LIST OF DOCUMENTS REVIEWED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
LIST OF ACRONYMS USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38
Enclosure
SUMMARY OF FINDINGS
IR 05000266/2004003, 05000301/2004003; 04/01/2004 - 06/30/2004; Point Beach Nuclear
Plant, Units 1 & 2; Fire Protection, Inservice Inspection Activities, Access Control to
Radiologically Significant Areas, Identification and Resolution of Problems, Event Follow-up,
and Temporary Instruction (TI) 2515/153, Reactor Containment Sump Blockage.
This report covers a three-month period of baseline resident inspection and announced
inservice inspection (Inspection Procedure (IP) 71111.08), heat sink (IP 71111.07B), and
radiation protection (IP 71121) baseline inspections for Point Beach Nuclear Plant, Units 1
and 2. In addition, TI inspections were completed in the areas of reactor pressure vessel head
and vessel head penetration nozzles (TI 2515/150, Unit 1), reactor pressure vessel lower head
penetration nozzles (TI 2515/152, Unit 1), reactor containment sump blockage (TI 2515/153,
Units 1 and 2), spent fuel material control and accounting (TI 2515/154, Units 1 and 2), and
offsite power system operational readiness (TI 2515/156, Units 1 and 2). The inspections were
conducted by 12 inspectors: a radiation specialist inspector, an inservice inspection specialist
inspector, a heat sink specialist inspector, a senior emergency preparedness inspector, a
project engineer, an NRC consultant inspector, an operator licensing examiner, and five
resident inspectors. One Green finding that was not a violation of NRC requirements and five
Green findings that were Non-Cited Violations (NCVs) were identified. The significance of most
findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter
(IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not
apply may be Green or be assigned a severity level after NRC management review. The
NRCs program for overseeing the safe operation of commercial nuclear power reactors is
described in NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.
A. Inspector-Identified and Self-Revealed Findings
Cornerstone: Initiating Events
- Green. The inspectors identified an NCV of 10 CFR 50.48(a)(2)(i) having very low
safety significance when transient combustibles were stored in the Unit 1 containment
building and the turbine building without required administrative controls. The finding
also affected the cross-cutting area of human performance in that the licensee failed to
identify the transient combustible materials during tours required by the Fire Protection
Evaluation Report.
The inspectors concluded that the finding is more than minor because it affected the
Reactor Safety Initiating Events Cornerstone objective to limit the likelihood of events
that upset plant stability and challenge critical safety functions during shutdown,
specifically protection against external factors (fire). The inspectors determined that the
finding was of very low safety significance (Green), since the issue was assigned a low
degradation rating and the quantity of transient combustibles had been bounded by the
analysis contained in the Fire Hazards Analysis Report. The licensee has entered this
finding into its corrective action (CA) program. (Section 1R05.1)
1 Enclosure
- Green. The inspectors identified a finding associated with installing steam generator
nozzle dams and establishing a hot leg vent path during a portion of the Unit 1 cycle 28
refueling outage (U1R28). The primary cause of this finding was related to the cross-
cutting area of human performance, involving the decision by several licensed and
experienced personnel to allow nozzle dam installation to commence prior to
establishment of a vent path through the pressurizer manway.
The finding is considered more than minor because it affected: (1) the Reactor Safety
Initiating Events Cornerstone objective to limit the likelihood of those events that upset
plant stability and challenge critical safety functions during shutdown operations, and (2)
the human performance attribute of the Initiating Events Cornerstone. The finding was
considered to be of very low safety significance and did not require quantitative
assessment since: (1) conditions meeting a loss of control were not met in that no
inadvertent change in reactor coolant system temperature or change in reactor vessel
level actually occurred, and (2) the licensee had maintained adequate mitigation
capability for the existing plant conditions. No violation of regulatory requirements
occurred because: (1) the actual sequence of events showed that all four nozzle dams
had not been completely installed while the pressurizer manway was still in place, and
(2) an engineering analysis showed that an adequate hot leg vent path was available
while one of the A steam generator hot leg nozzle dam side pieces was not installed.
The licensee has entered this finding into its corrective action (CA) program. (Section
4OA2.5)
Cornerstone: Mitigating Systems
- Green. The inspectors identified an NCV of 10 CFR 50.55a(a)(3)(i) for the licensees
incorrect substitution of weld surface examinations into the risk-based portion of the
Inservice Inspection Program, which required volumetric weld examinations.
This finding is greater than minor because it affected the Mitigating Systems
Cornerstone objective of equipment reliability and, if left uncorrected, could allow
unacceptable piping system weld flaws to remain in-service and render safety-related
systems inoperable. The finding is of very low safety significance because the licensee
had sufficient time left in the Code interval to perform the required number of volumetric
examinations of piping welds in the affected risk-based category during future Unit 1
outages. The licensee has entered this finding into its corrective action (CA) program.
(Section 1R08)
- Green. An NCV of Technical Specification (TS) Surveillance Requirement (SR) 3.5.1.2
was self-revealed when the water volume in the Unit 2 safety injection (SI) accumulator,
2T-34A, exceeded the TS limit of 1136 cubic feet.
The finding is greater than minor because it affected the Reactor Safety Mitigating
Systems Cornerstone objective to ensure the availability, reliability, and capability of
systems that respond to initiating events. The finding was considered to be of very low
safety significance since: (1) the Nuclear Steam Supply System vendor performed an
analysis of the over-filled, as-found condition and determined that the 2T-34A
accumulator had been capable of performing the design basis function and would not
2 Enclosure
have challenged the 10 CFR 50.46 Loss-of-Coolant-Accident acceptance criteria, and
(2) the finding did not result in a design or qualification deficiency, an actual loss of
safety function, or involve internal or external initiating events. The licensee has entered
this finding into its corrective action (CA) program. (Section 4OA3.1)
- Green. The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion VI,
Document Control, having very low safety significance associated with Unit 1
emergency operating procedures when a software error deleted reference to two of five
indications intended to monitor primary containment sump performance during the
recirculation phase of a design basis accident. Specifically, the RHR Pump Operation -
NORMAL and SI Pump Operation - NORMAL substeps of Unit 1 emergency operating
procedure EOP-1, Loss of Reactor or Secondary Coolant, Step 29c, Revision 35, were
deleted by the software program and not detected by operations personnel for a period
of approximately 9 months. The primary cause of this finding was related to the
cross-cutting area of human performance in that despite previous knowledge of the
software problem and operations department management expectations to perform line-
by-line reviews prior to distribution, 16 errors occurred in safety-related emergency
operating, emergency contingency action, critical safety, and shutdown emergency
procedures for Units 1 and 2.
The inspectors determined that the finding is more than minor because it affected the
procedure quality attribute of the Mitigating Systems Cornerstone objective of ensuring
the availability, reliability, and capability of systems that respond to initiating events. The
finding was considered to be of very low safety significance because it did not result in a
design or qualification deficiency, an actual loss of safety function, or involve internal or
external initiating events. The licensee has entered this finding into its corrective action
(CA) program. (Section 4OA5.4.b.2)
Cornerstone: Occupational Radiation Safety (OS)
- Green. A finding of very low safety significance and an associated NCV were identified
through an NRC-identified event, when on April 9, 2004, while installing steam generator
nozzle dams, licensee staff increased supplied breathing air pressure in excess of
procedural requirements while attempting to mitigate lost or diminished air flow to
contract workers who were utilizing continuous flow, supplied-air respirator bubble
hoods. The inspectors determined that the licensee failed to meet the requirements of
10 CFR 20.1703, when the licensee increased the air line pressure in excess of the
procedural guidance, which resulted in the licensee utilizing a respiratory protection
device contrary to its National Institute for Occupational Safety and Health (NIOSH)
certification.
The inspectors determined that the finding is more than minor because use of a
respiratory protection device outside its specifications could impact internal dose, and if
left uncorrected, could become a more significant safety concern. The finding was
considered to be of very low safety significance because no internal exposure to
radioactive material resulted from the use of the bubble hoods with higher air line
pressure than allowed. The licensee has entered this finding into its corrective action
(CA) program. (Section 2OS1.2.b)
3 Enclosure
B. Licensee-Identified Violations
A violation of very low significance, which was identified by the licensee, has been
reviewed by the inspectors. Corrective actions taken or planned by the licensee have
been entered into the licensees corrective action program. This violation and corrective
action tracking number are listed in Section 4OA7 of this report.
4 Enclosure
REPORT DETAILS
Summary of Plant Status
Unit 1 began the inspection period in coastdown at 92 percent power prior to the Unit 1 cycle 28
refueling outage (U1R28), which began on April 2, 2004. Unit 1 achieved criticality on June 7,
after completion of the outage, and returned to full power on June 11. Power was reduced to
55 percent later the same day to address high vibrations associated with the 1P-28B main
feedwater pump. Following maintenance on the pump, Unit 1 returned to full power on June 14
and remained there through the end of the inspection period.
Unit 2 began the inspection period at full power and remained there until May 15, 2004, when
the Unit was manually tripped after communications were lost with a diver performing
inspections at the circulating water (CW) intake crib. Unit 2 returned to full power on May 20.
Unit 2 remained at or near full power for the remainder of the inspection period.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, and
1R01 Adverse Weather Protection (71111.01)
.1 High Wind/Tornado Preparations
a. Inspection Scope
During the week of April 20, 2004, the inspectors reviewed the facility design and
licensee procedures to evaluate the plants likely response to high winds and tornadoes.
Additionally, the inspectors walked down selected areas to evaluate plant buildings,
switchyard, and equipment susceptible to high winds and tornadoes. The inspectors
also reviewed Abnormal Operating Procedure (AOP), AOP-13C, "Severe Weather
Conditions," dated June 30, 2003, which prescribed station actions for severe weather
conditions and several corrective action program documents (CAPs) associated with
recent high wind conditions. This inspection activity constituted one inspection sample.
b. Findings
No findings of significance were identified.
.2 Hot Weather Preparations
a. Inspection Scope
During the week of June 14, 2004, the inspectors reviewed the facility design and
licensee procedures to evaluate preparations for summertime high temperatures.
Additionally, the inspectors walked down selected areas to evaluate plant equipment
susceptible to high temperatures. The inspectors evaluated changes to the
5 Enclosure
methodology used to perform hot weather preparations and compared the changes to
those made as a result of the lessons learned during cold weather preparations. This
inspection activity constituted one inspection sample.
b. Findings
No findings of significance were identified.
1R04 Equipment Alignment (71111.04)
.1 Partial System Walkdowns
c. Inspection Scope
The inspectors performed three partial walkdowns of accessible portions of
risk-significant systems to evaluate the operability of the selected systems. The
inspectors utilized valve and electrical breaker checklists (CLs), tank level books, plant
drawings, and selected operating procedures to determine if the components were
properly positioned and supported the systems as needed. The inspectors also
examined the material condition of the components and observed operating equipment
parameters to check for obvious deficiencies. The inspectors reviewed completed work
orders (WOs) and calibration records associated with the systems for issues that could
affect component or train function. The inspectors used the information in the
appropriate sections of the Final Safety Analysis Report (FSAR) to determine the
functional requirements of the system. These observations constituted three quarterly
inspection samples.
The inspectors verified the alignment of the following systems:
- Residual Heat Removal (RHR), Train A on April 8, 2004;
- RHR, Train B on May 12, 2004; and
- Spent Fuel Pool (SFP) Cooling, on May 13, 2004.
b. Findings
No findings of significance were identified.
.2 Complete System Walkdown
c. Inspection Scope
On May 18, 2004, the inspectors performed a complete system alignment inspection of
345-Kilo-Volt (KV) and 13.8-KV systems. These systems were selected because they
were considered both safety-significant and risk-significant in the licensees probabilistic
risk assessment. The inspection consisted of the following activities:
- a review of plant procedures (including selected abnormal and emergency
procedures), drawings, and the FSAR to identify proper system alignment;
6 Enclosure
- a review of outstanding or completed temporary and permanent modifications to
the systems;
and
- a walkdown of mechanical and electrical components in the systems to verify
proper alignment, component accessibility, availability, and current condition.
The inspectors also reviewed selected issues to determine if they had been properly
addressed in the licensees corrective action (CA) program. This inspection activity
constituted one inspection sample.
b. Findings
No findings of significance were identified.
1R05 Fire Protection (71111.05)
.1 Walkdown of Selected Fire Zones
a. Inspection Scope
The inspectors conducted walkdowns focused on availability, accessibility, and the
condition of fire fighting equipment, the control of transient combustibles and ignition
sources, and on the condition and operating status of installed fire barriers. The
inspectors selected nine fire areas for inspection based on the areas overall fire risk
contribution, as documented in the licensees Individual Plant Examination of External
Events, the areas potential to impact equipment which could initiate a plant transient, or
the areas impact on the plants ability to respond to a security event. The inspectors
used the documents listed in the attachment to this report to determine if fire hoses and
extinguishers were in their designated locations and available for immediate use, fire
detectors and sprinklers were unobstructed, transient material loading was within the
analyzed limits, and fire doors, dampers, and penetration seals were in satisfactory
condition. These observations constituted nine quarterly inspection samples.
The following areas were inspected by walkdowns:
- Fire Zone 505, Unit 1 Containment, 8 foot;
- Fire Zone 511, Unit 1 Containment, 21 foot;
- Fire Zone 516, Unit 1 Containment, 48 foot;
- Fire Zone 520, Unit 1 Containment, 66 foot;
- Fire Zone A23N, Auxiliary Feedwater (AFW) Pump Room (North);
- Fire Zone A23S, AFW Pump Room (South);
- Fire Zone 142, Component Cooling Water (CCW) Pump Room;
- Fire Zone 156, Motor Control Center Room - 1B32; and
- Fire Zone 166, Motor Control Center Room - 2B32
7 Enclosure
b. Findings
Introduction: The inspectors identified an NCV of 10 CFR 50.48(a)(2)(i) having very low
safety significance (Green) for transient combustibles stored in the Unit 1 containment
building and the turbine building without required administrative controls.
Description: Inspectors identified that significant quantities of transient combustible
materials were stored in the Unit 1 containment and the turbine buildings during U1R28
without required administrative controls. The materials in the containment building
included 11 drums of lubricating oil and storage shelves of radiation protection materials
(cloth and plastic contamination control clothing and supplies). The inspectors also
identified 8 drums of waste oil in the turbine building adjacent to AFW pump and
emergency diesel generator (EDG) rooms. Permits did not exist for the storage of these
materials as required by the Fire Protection Evaluation Report (FPER), Section 3.1.2.2,
Transient Combustible Control, and Section 7.3.1, Containment.
Analysis: The inspectors determined that failing to implement administrative controls for
transient combustible materials in areas containing, and adjacent to, safety-related
equipment was a performance deficiency warranting a significance evaluation. The
inspectors concluded that the finding was more than minor in accordance with
Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix
B, Issue Disposition Screening, issued on January 14, 2004. The finding affected the
Reactor Safety Initiating Events Cornerstone objective to limit the likelihood of events
that upset plant stability and challenge critical safety functions during shutdown,
specifically protection against external factors (fire). The finding also affected the cross-
cutting area of human performance in that personnel failure to identify transient
combustible materials during tours required by FPER, Section 3.1.2.2, Transient
Combustible Control, and Section 7.3.1, Containment.
The inspectors completed a significance determination of this issue using IMC 0609,
Significance Determination Process (SDP), dated March 21, 2003, and Appendix F,
Fire Protection SDP, issued May 28, 2004. The inspectors determined that the finding
was of very low safety significance (Green), since the issue was assigned a low
degradation rating and the quantity of transient combustibles had been bounded by the
analysis contained in the Fire Hazards Analysis Report (FHAR). This finding was
assigned to the Reactor Safety Initiating Events Cornerstone for Unit 1.
Enforcement: 10 CFR 50.48(a)(2)(i) requires that the fire protection plan include
administrative controls for fire prevention. The FPER, Section 3.1.2.2, Transient
Combustible Control, and Section 7.3.1, Containment, require that administrative
controls be maintained for use of transient combustible material. Contrary to these
requirements, significant quantities of transient combustible materials were found in
areas of the Unit 1 containment and turbine buildings without the required administrative
controls. Because this violation was of very low safety significance and it was entered
into the licensees CA program, this violation is being treated as an NCV, consistent with
Section VI.A of the NRC Enforcement Policy. (NCV 05000266/2004003-01).
8 Enclosure
1R07 Heat Sink Performance (71111.07)
.1 Resident Inspector Annual Review of Heat Sink Performance
a. Inspection Scope
The inspectors evaluated the performance of the containment air cooler heat
exchangers (HXs) by conducting a review of licensee programs and practices that
assess HX performance on an ongoing basis and maintain that performance at
acceptable levels. The inspectors also determined if potential HX vulnerabilities and
deficiencies were identified and were being addressed in the CA program. Finally, the
inspectors reviewed HX testing and performance data to evaluate the condition of the
containment air cooler HXs. This inspection activity constituted one inspection sample.
b. Findings
No findings of significance were identified.
.2 Biennial Review of Heat Sink Performance
a. Inspection Scope
The inspectors reviewed documents associated with inspection, cleaning, and
performance trending of HXs primarily focusing on the CCW HX, HX-12C, and the
coolant HXs for EDGs G01 and G02. These HXs were chosen based upon their
importance in supporting required safety functions, as well as relatively high risk
achievement worth in the plant specific risk assessment. The component cooling HX
was also selected to evaluate the licensee's thermal performance testing methods.
During the inspection, the inspectors reviewed calculations to determine if these
activities adequately ensured proper heat transfer. The inspectors reviewed the
documentation to confirm that the inspection methodology was consistent with accepted
industry and scientific practices, based on review of heat transfer texts and Electrical
Power Research Institute (EPRI) standards. Specifically, the inspectors reviewed the
licensees heat transfer related calculations and/or maintenance activities to confirm that
the minimum design heat transfer capability was maintained for these HXs, in
accordance with licensee commitments to Generic Letter (GL) 89-13, Service Water
System Problems Affecting Safety-Related Equipment, and limiting design performance
values identified in the FSAR.
The inspectors performed a field walkdown of the CCW HXs, the service water (SW)
chemical treatment system, and the copper ion generator system. In addition, the
inspectors observed the cleaning of the G01 coolant HX.
The inspectors' review of licensee activities and documents regarding the component
cooling HX and EDG coolant HXs constituted three samples (two required) for the
biennial review of heat sink performance in accordance with Section 71111.07-05 of
Inspection Procedure (IP) 71111.07, Heat Sink Performance.
9 Enclosure
The inspectors reviewed documents associated with licensee controls for the SW from
the ultimate heat sink (UHS) to prevent clogging due to macrofouling and biotic fouling.
These two attributes met the procedure requirements for verifying the performance of
the UHS.
The inspectors reviewed CAs concerning HX and UHS performance issues to determine
if the licensee had an appropriate threshold for identifying issues and entering them in
the CA program. The inspectors also evaluated the effectiveness of the CAs for
identified issues, including the engineering justification for operability.
The documents that were reviewed are included at the end of this report.
b. Findings
No findings of significance were identified.
1R08 Inservice Inspection Activities (IP 71111.08)
a. Inspection Scope
For Unit 1, the inspectors evaluated the implementation of the licensees inservice
inspection (ISI) program for monitoring degradation of the reactor coolant system
boundary and risk significant piping system boundaries, based on a review of
nondestructive examination (NDE) records.
From April 5 through 28, 2004, inside the Unit 1 containment building, the inspectors
observed ultrasonic (UT) examinations, which constituted one type (volumetric) of NDE
activity. Specifically, the inspectors observed UT examination of two pressurizer spray
line welds (RC-03-PS-1001-14 and 15), two auxiliary feedwater system welds (AF-03-
1002-76 and 77), and one feedwater system weld (FW-16-FW-1002-15). Additionally,
the inspectors observed a second and third type of NDE activity related to the under
head vent line dye penetrant (PT) examination of reactor vessel nozzle No. 26 J-groove
weld and a visual (VT-3) examination of a feedwater system hanger (EB-9-FW-1111).
The inspectors selected these components in order of risk priority as identified in
Section 71111.08-03 of IP 71111.08, Inservice Inspection Activities. The inspectors
evaluated these examinations for compliance with the American Society of Mechanical
Engineers (ASME) Boiler and Pressure Vessel Code Section XI and plant Technical
Specification (TS) requirements and to determine if indications and defects (if present)
were dispositioned in accordance with the ASME Code. This review counted as two
inspection samples as described in Section 71111.08-5 of IP 71111.08.
From April 5 through 28, 2004, in an office on the 8-foot level of the Technical Support
Building (TSB), the inspectors reviewed the licensees records related to three
examinations (summary report 004500 for control rod drive housings No. 1, reactor
pressure vessel head flange report 99U1-350P004, and reactor pressure vessel Stud
No. 44 report 99U-350P021) with recordable indication accepted for continued service.
The inspectors evaluated these examinations for compliance with ASME Code Section
XI. This review counted as one inspection sample as described in Section 71111.08-5
of IP 71111.08.
10 Enclosure
From April 5 through 28, 2004, in an office on the 8-foot level of the TSB, the inspectors
reviewed the licensees records related to pressure boundary welding to replace pipe
and elbows on 2-inch lines to the T-34B safety injection system accumulator (Class 2
component). Specifically, the inspectors reviewed records for welds FW-1 and FW-2 to
determine if the welding acceptance and preservice examinations (e.g., pressure
testing, visual, dye penetrant, and weld procedure qualification tensile tests and bend
tests) were performed in accordance with ASME Code,Section III, Section V, Section
IX, and Section XI. This review counted as one inspection sample as described in
Section 71111.08-5 of IP 71111.08.
From April 5 through 28, 2004, in an office on the 8-foot level of the TSB, the inspectors
reviewed the licensees records associated with two ASME Code Section XI
replacement activities (replace pipe and elbows on 2-inch lines to the T-34B safety
injection system accumulator) for Code Class 2, to verify that the ASME Code
Section III,Section V, and Section XI requirements were met. This review counted as
one inspection sample as described in Section 71111.08-5 of Inspection Procedure
From April 5 through May 14, 2004, in Room 138 of the on-site training building, the
inspectors observed acquisition of steam generator (SG) tube eddy current (ET) data for
the Unit 1 SGs. The inspectors also reviewed the SG ET examination scope, expansion
criteria, analysis procedures, and examination reports for the Unit 1 A and B SGs to
confirm that:
C TS requirements were met;
C the inspection was consistent with the EPRI Guidelines;
C areas of potential degradation were inspected; and
C ET probes and equipment were qualified in accordance with the EPRI Guidelines
for the expected types of tube degradation.
The inspectors concluded that the review discussed above did not count as a completed
inspection sample as described in Section 71111.08-5 of IP 71111.08, but the sample
was completed to the extent possible. The specific activities that were not available for
review to complete this inspection sample are identified in the table below.
Inspection Procedure 7111108 Reason Activity was Reduction in Inspection
Section Number Unavailable For Procedure Samples
Inspection
Section 02.02.a 1 thru 4: The licensee did not The inspectors concluded
associated with review of identify any tubes that these unavailable
licensee in-situ pressure testing that required activities constituted a
of steam generator tubes. pressure testing. reduction by one from the
total number of procedure
samples required by
Section 71111.08-5 of
Inspection Procedure
11 Enclosure
Inspection Procedure 7111108 Reason Activity was Reduction in Inspection
Section Number Unavailable For Procedure Samples
Inspection
Section 02.02.f and g: confirm The licensee did not
that all repair processes used identify any tubes
were approved in the technical that required repair.
specifications for use at the site;
reviewed tube repair criteria.
Section 02.02.h: associated with The licensee
steam generator tube leakage reported that no
greater than 3 gallons per day. steam generator
tube leakage had
been observed.
Section 02.02.k: associated with The inspectors did
review of one to five samples of not identify any
eddy current data. serious questions
regarding the eddy
current data.
The specific list of documents reviewed by the inspectors in conducting this inspection
are listed in the attachment to this report.
b. Findings
b.1 Substitution of Weld Surface Examinations for Volumetric Examinations
Introduction: The inspectors identified a Green NCV of 10 CFR 50.55a(a)(3)(i) for the
licensees substitution of weld surface examinations into the risk-based portion of the ISI
program, which required volumetric weld examinations.
Description: On April 9, 2004, while performing the baseline ISI procedure
(IP 7111108), the inspectors identified that the licensee had inappropriately credited
surface examination of welds in the risk-based ISI program.
By letter dated July 3, 2002, the licensee requested approval to use a risk-informed ISI
program in accordance with EPRI TR-112657 as an alternative to the weld inspection
program required by the ASME Code for Class 1 and 2 piping welds. The NRC
approved this request under provisions allowed in 10 CFR 50.55a(3)(i) as an acceptable
alternative program which would provide for a comparable level of safety. Table 4-1 of
EPRI TR-112657 required volumetric examination of welds subject to all degradation
mechanisms except for microbiologically induced corrosion (MIC) and outside diameter
stress corrosion cracking (ODSCC). On January 17, 2003, the licensee submitted the
Owners Inservice Inspection Summary Report for Unit 1 to the NRC. In this report, the
licensee credited two Unit 1 safety injection (SI) system weld PT examinations,
completed in September 2002, as risk-based weld examinations (SIS-04-SI-1005-25
and SIS-04-SI-1005-25B). The licensee had not identified these welds as susceptible to
12 Enclosure
MIC or ODSCC or any other degradation mechanism (e.g., weld category R1.20 from
Code Case N-578-1). Therefore, by taking credit for these surface PT examinations,
the licensee reduced the number of volumetric examinations for this category of welds in
the risk-based ISI program. The inspectors concluded that the licensees use of surface
examinations changed the basis for the approved risk-based ISI program (EPRI
TR-11267), which required volumetric examinations to detect degradation that typically
originated from the inside surface of piping systems. The inspectors were concerned
that substitution of surface examinations for volumetric examinations could allow
unacceptable piping system weld flaws to remain in-service and render safety related
systems inoperable. The licensee has entered this issue in its corrective action
program.
Analysis: The licensees performance deficiency associated with this finding is the
failure to perform the required volumetric weld examinations by substitution of weld
surface examinations. The inspectors concluded that the finding was greater than minor
in accordance with IMC 0612, Power Reactor Inspections Reports, Appendix B, Issue
Disposition Screening, because, if left uncorrected, the substitution of surface
examinations in place of volumetric examinations could allow unacceptable piping
system weld flaws to remain in service. The finding was assigned to the Mitigating
System Cornerstone because the affected weld examinations identified were associated
with the SI system (mitigating system) and the finding affected the Mitigating System
Cornerstone objective of equipment reliability. The inspectors determined that the
finding could not be evaluated using the SDP in accordance with NRC IMC 0609,
Significance Determination Process, because the SDP for the Mitigating Systems
Cornerstone only applied to degraded systems/components, not to the program/process
failures that could result in failure to detect degraded systems/components. Therefore,
this finding was reviewed by the Regional Branch Chief in accordance with IMC 0612,
Section 05.04c, who agreed with the inspectors, that this finding was of very low safety
significance (Green). The inspectors determination of very low risk was based on the
fact that the licensee had sufficient time left in the Code interval to perform the required
number of volumetric examinations of piping welds in the affected risk based category
during future Unit 1 outages.
Enforcement: On April 9, 2004, while performing the baseline ISI, the inspectors
identified an NCV of 10 CFR 50.55a(a)(3)(i).
10 CFR 50.55a(a)(3)(i) states, in part, that alternatives to requirements of paragraph
10 CFR 50.55a(g) [ASME Section XI Code] may be used, when authorized by the NRC.
By letter dated July 2, 2003, in accordance with 10 CFR 50.55a(a)(3)(i), the NRC
approved the licensees use of a risk-based ISI program in accordance with EPRI TR-
112657, Revised Risk-Informed Inservice Inspection Evaluation Procedure, Revision
B-A. In EPRI TR-112657, Table 4-1, volumetric examinations of welds were identified
as the approved weld examination technique for all degradation mechanisms except
Contrary to these requirements, on January 17, 2003, the licensee took credit for
surface examinations of welds SIS-04-SI-1005-25 and SIS-04-SI-1005-25B, completed
in September of 2002 in their risk-based ISI program. These welds were not subject to
MIC or ODSCC and, therefore, the licensees use of weld surface examinations was
13 Enclosure
contrary to requirements of EPRI TR-112657 Table 4-1. However, because of the very
low safety significance of this finding and because the issue was entered into the
licensees corrective action program (CAP055529), it is being treated as an NCV,
consistent with Section VI.A.1 of the Enforcement Policy. (NCV 05000266/2004003-02)
1R11 Licensed Operator Requalification (71111.11)
a. Inspection Scope
On June 10, 2004, the inspectors observed the operating crew performance during
simulator training. The inspectors also reviewed some of the changes to the simulator
model against modifications made in the plant. This observation constituted one
quarterly inspection sample.
The inspectors evaluated crew performance in the areas of:
- clarity and formality of communications;
- understanding of the interactions and function of the operating crew during an
emergency;
- prioritization, interpretation, and verification of actions required for emergency
procedure use and interpretation;
- oversight and direction from supervisors; and
- group dynamics.
Crew performance in these areas was compared to licensee management expectations
and guidelines as presented in Procedure NP 2.1.1, Conduct of Operations,
Revision 1. The inspectors evaluated the licensees failure of the crew when two critical
objectives of the scenario were not successfully completed. The inspectors confirmed
that the crew received remedial training and passed a simulator exam prior to being
allowed to return to shift duties.
b. Findings
No findings of significance were identified.
1R12 Maintenance Effectiveness (71111.12)
a. Inspection Scope
The inspectors performed an issue/problem-oriented review of the system listed below,
completing one maintenance effectiveness inspection sample. The inspectors reviewed
repetitive maintenance activity to assess maintenance effectiveness, including
maintenance rule activities, work practices, and common cause issues. Inspection
activities included, but were not limited to, the licensee's categorization of specific
issues, including evaluation of performance criteria, appropriate work practices,
identification of common cause errors, extent of condition, and trending of key
parameters. Additionally, the inspectors reviewed implementation of the Maintenance
Rule (10 CFR 50.65) requirements, including a review of scoping, goal-setting,
14 Enclosure
performance monitoring, short-term and long-term CAs, functional failure determinations
associated with reviewed CAPs, and current equipment performance status.
For the system reviewed, the inspectors reviewed significant WOs and CAPs to verify
that failures were properly identified, classified, and corrected, and that unavailable time
had been properly calculated. The inspectors reviewed documents listed in the
attachment to this inspection report to determine if minor discrepancies in the licensees
maintenance rule reports were corrected. This observation constituted one quarterly
inspection sample.
Specific components and systems reviewed were:
- 125-volts direct current electrical system.
b. Findings
No findings of significance were identified.
1R13 Maintenance Risk Assessment and Emergent Work Evaluation (71111.13)
a. Inspection Scope
The inspectors reviewed risk assessments for the following maintenance activities,
completing risk assessment and emergent work control inspection samples. These
observations constituted five quarterly inspection samples.
during the week of April 26, 2004;
- unavailability of the electric firewater pump because of unplanned maintenance
during the U1R28, the week of April 18, 2004;
- unavailability of various equipment during the U1R28 refueling outage while
completing Unit 2 planned maintenance during the week of May 17, 2004;
- unavailability of the G01 EDG for planned maintenance during the week of
May 31, 2004; and
- unavailability of the D-106 battery for planned maintenance during the week of
June 13, 2004.
During these reviews, the inspectors compared the licensees risk management actions
to those actions specified in the licensees procedures for the assessment and
management of risk associated with maintenance activities. The inspectors determined
if evaluation, planning, control, and performance of the work were done in a manner to
reduce the risk and minimize the duration, where practical, and that contingency plans
were in place, where appropriate. The inspectors used the licensees daily configuration
risk assessment records, observations of shift turnover meetings, and observations of
daily plant status meetings to determine if the equipment configurations had been
properly listed, that protected equipment had been identified and was being controlled
where appropriate, and that significant aspects of plant risk were communicated to the
necessary personnel.
15 Enclosure
b. Findings
No findings of significance were identified.
1R14 Personnel Performance During Non-Routine Plant Evolutions and Events (71111.14)
.1 Unit 1 Shutdown for U1R28 Refueling Outage
a. Inspection Scope
On April 2, 2004, the inspectors observed the shutdown for the Unit 1 refueling outage
(U1R28). The inspectors observed operator procedure use and adherence,
communications, control of equipment, and response to minor equipment complications.
This observation constituted one inspection sample.
b. Findings
No findings of significance were identified.
.2 Unit 2 Start-up from a Reactor Trip
a. Inspection Scope
On May 18, 2004, the inspectors observed the start-up of Unit 2 following a manual
reactor trip. The Unit was tripped from 100 percent power after communications with a
diver at the circulating water intake structure were lost. The inspectors observed the
approach to criticality, power ascension to the point-of-adding-heat, blocking of trip
circuits, operator response to an unexpected turbine trip and feedwater recirculation
valve flange leak, and ascension to full power. This observation constituted one
inspection sample.
b. Findings
No findings of significance were identified.
.3 Unit 1 Nozzle Dam Removal
a. Inspection Scope
On May 23, 2004, the inspectors observed operator response to mid-loop operations
and steam generator nozzle dam removal. Following previous issues associated with
nozzle dam installation, the inspectors scrutinized control and conduct of the evolution
including communications, job planning, management of air lines to personnel bubble
hoods, procedure use and adherence, coordination with the Outage Control Center, and
management oversight. This observation constituted one inspection sample.
b. Findings
No findings of significance were identified.
16 Enclosure
1R15 Operability Evaluations (71111.15)
a. Inspection Scope
During this inspection period, the inspectors reviewed the following operability
evaluations:
- Unanalyzed Load on G-03 and G-04 EDGs;
- Trouble-Shooting after Inadvertent Letdown Isolation;
- Excore Source-Range Nuclear Instrument IN-31 Pre-amplifier Found Defective;
- Unit 1 and 2 Motor-Driven AFW Pump Flows; and
- Seismic Monitoring Operability in the North Warehouse.
These observations constituted five quarterly inspection samples.
The inspectors reviewed the technical adequacy of the operability evaluations against
TS, FSAR, and other design information; determined whether compensatory measures,
if needed, were taken; and determined whether the evaluations were consistent with
Procedure NP 5.3.7, Operability Determinations. The inspectors also reviewed CAPs
to determine if licensee personnel identified issues at an appropriate threshold and
entered them into the corrective action program in accordance with station procedures.
Documents reviewed during this inspection are listed in the attachment to this report.
b. Findings
No findings of significance were identified.
1R16 Operator Workarounds (OWAs) (71111.16)
a. Inspection Scope
The inspectors reviewed OWAs with particular focus on the method by which
instructions and contingency actions were communicated to and reviewed with on-shift
licensed operators. Documents reviewed during this inspection are listed in the
attachment to this report. This observation constituted one inspection sample.
The inspectors completed the sample by reviewing:
- CAP053487, relating to the Alarm Response Book for control room panel
1C20A, Window 2-5, Containment Hydrogen System Trouble, which contained
instructions directing operators to declare the wrong train out-of-service.
b. Findings
No findings of significance were identified.
17 Enclosure
1R19 Post-Maintenance Testing (PMT) (71111.19)
a. Inspection Scope
During this inspection period, the inspectors completed six quarterly inspection samples,
composed of the following PMT activities:
- Fuel Assembly Core Location Mapping following U1R28 on May 2, 2004,
- Safety injection pump 1P-15B following seal replacement on May 25, 2004,
- Unit 1 A Steam Generator Header Drain and Trap Isolation Valve (1MS-228)
Repack, on June 4, 2004,
- Unit 1 primary containment sump B valve (1SI-850B) testing on May 3, 2004,
- Local Leak Rate Test of Unit 1 Containment Purge Valves on June 2, 2004, and
May 4, 2004.
Documents reviewed during this inspection are listed in the attachment to this report.
During completion of the inspection samples, the inspectors observed in-plant activities
and reviewed procedures and associated records to determine if:
- testing activities satisfied the test procedure acceptance criteria,
- effects of the testing had been adequately addressed prior to the
commencement of the testing,
- measuring and test equipment calibration was current,
- test equipment was used within the required range and accuracy,
- applicable prerequisites described in the test procedures were satisfied,
- affected systems or components were removed from service in accordance with
approved procedures,
- testing activities were performed in accordance with the test procedures and
other applicable procedures,
- jumpers and lifted leads were controlled and restored, where used;
- test data/results were accurate, complete, and valid,
- test equipment was removed after testing,
- equipment was returned to a position or status required to support the operability
of the system in accordance with approved procedures,
- the impact of failed PMTs on primary containment isolation requirements were
evaluated, and
- all problems identified during the testing were appropriately documented in the
corrective action program.
b. Findings
No findings of significance were identified.
18 Enclosure
1R20 Refueling and Outage Activities (71111.20)
.1 Routine Refueling Outage Inspection Activities
a. Inspection Scope
The inspectors observed outage activities during U1R28, conducted between April 2 and
June 7, 2004. These inspection activities constituted one refueling outage inspection
sample.
This inspection consisted of an in-office review of the licensees outage schedule, safe
shutdown plan and administrative procedures governing the outage, periodic
observations of equipment alignment, and plant and control room outage activities.
Specifically, the inspectors determined the licensees ability to effectively manage
elements of shutdown risk pertaining to reactivity control, decay heat removal, inventory
control, electrical power control, and containment integrity.
The inspectors conducted in-plant observations of the following daily outage activities:
- attended outage management turnover meetings to determine if the current
shutdown risk status was accurate, well understood, and adequately
communicated;
- performed walkdowns of the main control room to observe the alignment of
systems important to shutdown risk;
- observed the operability of reactor coolant system (RCS) instrumentation and
compared channels and trains against one another;
- performed in-plant walkdowns to observe ongoing work activities; and
- conducted in-office reviews of selected issues that the licensee entered into its
corrective action program to determine if identified problems were being entered
into the program with the appropriate characterization and significance.
Additionally, the inspectors performed in-plant observations of the following specific
activities:
- observed the control room staff perform the Unit 1 shutdown and initial
cooldown;
- verified that RCS cooldown rates were within TS limits;
- observed control room staff operations during reduced inventory conditions;
- observed core unloading activities in the containment, SFP, and control room;
- observed core reload from the control room;
- observed operators align the RHR system for shutdown cooling;
- observed placement of the over-pressure protection system into operation;
- monitored a pre-job briefing for fuel handling evolutions;
- performed walkdowns of the auxiliary building to verify the placement of
clearance orders on Unit 1 electrical buses, RHR systems, and SW systems;
- observed lifting and transport of the reactor vessel head in preparation for core
offload;
19 Enclosure
- performed a walkdown of the control room and turbine building to verify
safety-related electrical alignments following battery charger and 4-KV electrical
bus routine maintenance;
- performed a closeout inspection of the Unit 1 containment including a review of
the results of the emergency core cooling sump inspection that had been
performed earlier by the licensee. As part of this inspection, the inspectors also
assessed whether all discrepancies noted during the walkdown were recorded
and corrected;
- walked down nozzle dam control panels to assess proper indications, installation,
removal, and alarm functions;
- observed steam generator nozzle dam installation and removal;
- reviewed shutdown margin calculations;
- reviewed reduced inventory level RCS transmitter configurations;
- reviewed the proper alignment and operation of the potential-dilution-in-progress
alarm;
- reviewed the evaluation of the fuel handling bridges in containment and the SFP
- reviewed Mode change checklists (CLs) to verify that selected requirements
were met while transitioning from the refueling Mode to full power operations;
- observed portions of low power physics testing and approach to criticality; and
- observed portions of the plant ascension to full power operations.
b. Findings
No findings of significance were identified.
1R22 Surveillance Testing (71111.22)
a. Inspection Scope
During this inspection period, the inspectors completed inspection samples, composed
of the following surveillance testing activities:
- Inservice Test (IT) 09A, Cold Start of Turbine-Driven Auxiliary Feed Pump and
Valve Test (Quarterly) Unit 2 on May 31, 2004, reviewed Mode change CLs to
determine if selected requirements were met while transitioning from the
refueling Mode to full power operations;
- Operations Refueling Test 3B, SI Actuation with Loss of Engineering Safeguards
AC (Train B) on April 4, 2004,
- IT-760, High Head SI Check Valve Full Flow Testing, on April 23, 2004,
- Operations Refueling Test 3A, SI Actuation with Loss of Engineering Safeguards
AC (Train A) on April 6, 2004,
- RHR Pump Vibration testing on April 20, 2004, and
- IT-03, Low Head SI Pump and Valve Test (Quarterly), on June 22, 2004.
Documents reviewed during this inspection are listed in the attachment to this report.
These observations constituted six quarterly inspection samples.
20 Enclosure
During completion of the inspection samples, the inspectors observed in-plant activities
and reviewed procedures and associated records to determine if:
- preconditioning occurred;
- effects of the testing had been adequately addressed by control room personnel
or engineers prior to the commencement of the testing;
- acceptance criteria were clearly stated, demonstrated operational readiness, and
were consistent with the system design basis;
- plant equipment calibration was correct, accurate, and properly documented; as-
left setpoints were within required ranges; and the calibration frequency was in
accordance with TSs, FSAR, procedures, and applicable commitments;
- measuring and test equipment calibration was current;
- test equipment was used within the required range and accuracy;
- applicable prerequisites described in the test procedures were satisfied;
- test frequency met TS requirements to demonstrate operability and reliability;
- the tests were performed in accordance with the test procedures and other
applicable procedures;
- jumpers and lifted leads were controlled and restored where used;
- test data/results were accurate, complete, within limits, and valid;
- test equipment was removed after testing;
- where applicable for in-service testing activities, testing was performed in
accordance with the applicable version of ASME Section XI, and reference
values were consistent with the system design basis;
- where applicable, test results not meeting acceptance criteria were addressed
with an adequate operability evaluation or the system or component declared
- where applicable for safety-related instrument control surveillance tests,
reference setting data have been accurately incorporated in the test procedure;
- prior procedure changes had not provided an opportunity to identify problems
encountered during the performance of the surveillance or calibration test;
- equipment was returned to a position or status required to support the
performance of its safety functions; and
- all problems identified during the testing were appropriately documented and
dispositioned in the corrective action program.
b. Findings
No findings of significance were identified.
1R23 Temporary Plant Modifications (71111.23)
a. Inspection Scope
The inspectors conducted in-plant observations of physical changes to the plant and
equipment, performed in-office reviews of documentation, and assessed, where
applicable, remote alarm capabilities to evaluate the TMs detailed below. The
inspectors reviewed design basis documents (DBDs) and safety evaluation screenings
to ensure that the modifications were consistent with applicable documents, drawings,
and procedures. The inspectors also reviewed the post-installation results to confirm
21 Enclosure
that any impacts of the TMs on permanent and interfacing systems were adequately
verified. These observations constituted two inspection samples.
The inspectors reviewed the following TMs:
- Installation of Blank Flange at VNSPE-3212/3244.
- Auxiliary Feed Tunnel Seismic Event Annunciation.
b. Findings
No findings of significance were identified.
1EP4 Emergency Action Level and Emergency Plan Changes (71114.04)
a. Inspection Scope
The inspectors reviewed Revision 40 of Section 2 and Revision 46 of Section 7 of the
Point Beach Emergency Plan to determine if the revisions reduced the Plans
effectiveness, pending on-site inspection of the implementation of these changes.
b. Findings
No findings of significance were identified.
1EP6 Drill Evaluation (71114.06)
.1 Emergency Plan Procedure Training Drills
a. Inspection Scope
During the weeks of June 14 and June 21, 2004, the inspectors observed the training
drills involving the revised EALs and Emergency Plan Implementing Procedures. The
inspectors observed classifications, notifications, facility activations, and facility critiques.
The observations were in the Control Room (simulator), Technical Support Center, and
Emergency Operations Facility. The inspectors also observed the training of new
Emergency Response Organization personnel. This observation constituted one
inspection sample.
b. Findings
No findings of significance were identified.
22 Enclosure
2. RADIATION SAFETY
Cornerstone: Occupational Radiation Safety (OS)
2OS1 Access Control to Radiologically Significant Areas (71121.01)
.1 Plant Walkdowns and Radiation Work Permit Reviews
a. Inspection Scope
The inspectors reviewed licensee controls and surveys for selected radiation areas, high
radiation areas and airborne radioactivity areas, as available, in the following
radiologically significant work areas within the plant and reviewed work packages which
included associated licensee controls and surveys for these areas to determine if
radiological controls (including postings and barricades) were acceptable:
- Primary Auxiliary Building, and
- Unit 1 Containment (all levels).
The inspectors reviewed the radiation work permits (RWP) and work packages used to
control work in these areas and other high radiation work areas to identify the work
control instructions and control barriers that had been specified. Electronic dosimeter
alarm setpoints for both integrated dose and dose rate were evaluated for conformity
with survey indications and plant policy. Workers were interviewed to assess their
knowledge of the actions required when their electronic dosimeters noticeably
malfunctioned or alarmed.
The inspectors walked down these areas to determine if the prescribed RWPs,
procedures, and engineering controls were in place, licensee surveys and postings were
complete and accurate, and air samplers (if necessary) were properly located.
The inspectors reviewed the RWPs and surveys for the steam generator nozzle dam
installation and ET activities which had the potential for creating an airborne radioactivity
area. The inspectors reviewed the RWPs to determine if barrier integrity and
engineering control contingency plans were in place and to determine if there was a
potential for individual worker internal exposures of greater than 50 millirem committed
effective dose equivalent. This and other work activities/areas having a history of, or the
potential for, airborne transuranic isotopes were evaluated to determine if the licensee
had considered the potential for transuranic isotopes and provided appropriate worker
protection.
The inspectors assessed the adequacy of the licensees internal dose assessment
process by reviewing personnel contamination event logs (and associated dose
assessments) for the refueling outage. As of April 21, 2004, no personnel
contamination events had resulted in dose assignments of greater than 10 millirem
committed effective dose equivalent.
These reviews represented four inspection samples.
23 Enclosure
b. Findings
No findings of significance were identified.
.2 Job-In-Progress Reviews
a. Inspection Scope
The inspectors observed the following four activities that were being performed in
radiation areas, airborne radioactivity areas, or high radiation areas to assess work
activities that presented the greatest radiological risk to workers:
- 1B Reactor Coolant Pump Motor Lift;
- Steam Generator ET Testing;
- Reactor Vessel Head Lift; and
- Cono-Seal Bullet Replacement.
The inspectors reviewed radiological job requirements for these four activities, including
RWP and work procedure requirements, and attended
as-low-as-is-reasonably-achievable (ALARA) pre-job briefings.
Job performance was observed with respect to these requirements to determine if
radiological conditions in the work areas were adequately communicated to workers
through pre-job briefings and postings. The inspectors also reviewed the adequacy of
radiological controls (including required radiation, contamination, and airborne surveys),
radiation protection job coverage (including audio/visual surveillance for remote job
coverage), and contamination controls. This included a review of the radiological
controls employed and resulting potential dose consequences related to the installation
of steam generator nozzle dams early in the refueling outage. The inspectors
completed their assessment of the nozzle dam installation activities by conducting an
in-office review of the licensees root cause evaluation (RCE) for the evolution during the
week of May 24, 2004.
Radiological work in high radiation work areas having significant dose rate gradients
was reviewed to evaluate the application of dosimetry to effectively monitor exposure to
personnel and to determine if licensee controls were adequate. In particular, the steam
generator ET activities and cono-seal bullet replacement involved evolutions where the
dose rate gradients were severe, which increased the necessity of providing multiple or
repositioned dosimetry and/or enhanced job controls.
These reviews represented three inspection samples.
b. Findings
Introduction: One NRC-identified Green finding and an associated NCV were identified
when, on April 9, 2004, licensee staff increased supplied breathing air pressure while
attempting to mitigate lost or diminished air flow to contract workers in continuous flow,
supplied air respirator bubble hoods who were installing steam generator nozzle dams.
24 Enclosure
The increased air pressure resulted in the use of respiratory equipment contrary to
Description: On April 9, 2004, during the Unit 1 refueling outage, installation of steam
generator nozzle dams commenced. The RWP for the job required the use of
continuous flow, supplied air respirator bubble hoods for whole body entries into the
steam generator bowl. At the start of the work, at least one contractor expressed a
concern to RP staff about low air pressure within the hood but he subsequently entered
the B steam generator. After working in the steam generator for 1:48 minutes, the
worker exited the generator because of continued low air flow.
The RP staff examined the quick disconnect Snap-Tite connection from the bubble
hood hose to the regulator air hose and identified no obvious problems. Also, no
problems were identified with the plant air system. Subsequently, RP supervision
approved an increase in the hood air line air pressure from 20 - 28 pounds per square
inch gauge (psig) to approximately 60 - 64 psig, for the 50-foot air line hose of the
bubble hood. The increase of the air pressure violated Health Physics Implementing
Procedure (HPIP) 4.58, Step 4.5.7, which stated, adjust air supply pressure so that air
flow is between 6 and 15 cubic feet per minute [cfm]. For an air line length of 50 feet, a
pressure range of 20 to 28 psig corresponds to a flow rate of 6 to 15 cfm.
Ten minutes after the first contract worker exited the steam generator, a second
contractor entered the B steam generator bowl. According to the licensees RCE, while
inside the bowl (for 1:18 minutes), the contractor realized that he had lost air flow, but he
continued working until he believed he had 2 or 3 good breaths left at which point he
exited the steam generator. As with the first contractor, the second contractor was cut
out of the bubble hood upon exiting the steam generator bowl and incurred a minor
personal contamination event (PCE). The RP staff at the steam generator platform
determined that the Snap-Tite fitting on the hood hose had disconnected, resulting in
the loss of air flow. A third contractor then entered the bowl and completed the
installation of the nozzle dam.
During installation of the nozzle dam in the A steam generator, two additional loss of air
incidents occurred. In one instance, the contractor had partially entered the bowl when
the Snap-Tite fitting contacted the manway and disconnected. The contractor
immediately exited, the air line was reconnected, and the contractor re-entered the
steam generator to complete the installation. In the second instance, an air supply line
pinched against and was cut by equipment staged on the platform, resulting in
diminished air flow. The RP technician taped the cut air line and the worker completed
his work in the bowl. Around this time, an NRC inspector who was observing the nozzle
dam installation activities via video monitors raised concerns about the breathing air
problems to site management. Subsequently, a formal investigation was initiated by the
licensee which led to a formal root cause evaluation.
The licensees root cause evaluation (CAP055527/RCE 253, completed May 22, 2004),
detailed more than 20 inappropriate actions during the steam generator nozzle dam
installation activities. In addition to the performance issues detailed above, the
licensees RCE identified that three different RP department evaluations of calendar
year 2003 operating events/experience (OE), relative to the loss of supplied breathing
25 Enclosure
air due to separation of air line quick disconnect fittings, had failed to adequately assess
the stations susceptibility to similar occurrences. Specifically, OE031454, OE048685,
and OE010321 were evaluated by the licensee but were closed for reasons including
fittings are from a different manufacturer, fittings are taped, and procedures and
controls are adequate to minimize susceptibility to this event. However, the fittings
described in the OE are of similar design as those used for bubble hood air line
connections at Point Beach and no actions were taken to physically challenge the
Snap-Tite fittings during the evaluations.
The licensees corrective actions included: (1) a site-wide stand-down to discuss these
and related events with station/contractor staff; (2) an independent team assessment of
the stations procedures and processes relative to the use of supplied air respiratory
devices (and implementation of procedural and equipment changes, as necessary);
(3) development of a complete nozzle dam removal plan in accordance with the
applicable work planning procedures; (4) full mock-up training for the nozzle dam
removal, including bubble hood use and air pressure requirements; and (5) development
of a specific procedure for nozzle dam installation/removal activities, including lessons-
learned, supervisory oversight requirements, stop work authority, communications
protocol, and external operating experience.
During the in-office review of the RCE, the inspectors identified that when the RP staff
increased the air line pressure to approximately 60 psig, the resulting flow rate in the
bubble hood would have exceeded 15 cfm. The National Institute for Occupational
Safety and Health (NIOSH) certification requirements for continuous flow, supplied air
respirator bubble hoods are described in 42 CFR Part 84, Subpart J. Specifically, Table
8 of Subpart J, requires, in part, that for the bubble hoods used during the evolution
(Type C, loose fitting hood), the air supply hose with air regulating valve shall permit a
flow of not less than 6 cfm, and the maximum flow shall not exceed 15 cfm. Therefore,
in addition to violating station procedure HPIP 4.58, the inspectors determined that when
the RP staff increased the air pressure to approximately 60 psig, the licensee used a
respiratory protection device contrary to the NIOSH certification for the device, which is
a violation of 10 CFR 20.1703, Use of Individual Respiratory Protection Equipment.
Analysis: The inspectors determined that the licensee failed to meet the requirements
of 10 CFR 20.1703, when the licensee increased the air line pressure in excess of the
procedural guidance in HPIP 4.58, which resulted in the licensee utilizing a respiratory
protection device contrary to its NIOSH certification. This issue could reasonably be
viewed as a precursor to a significant event and, if left uncorrected, would become a
more significant safety concern. Also, the issue involved conditions contrary to licensee
procedures and NRC regulations which impact protective equipment related to
mitigating worker dose. Therefore, the issue was determined to be more than minor
and represents a finding which was evaluated using the significance determination
process (SDP) for the Occupational Radiation Safety Cornerstone.
The inspectors determined utilizing Manual Chapter 0609, Appendix C, Occupational
Radiation Safety SDP, that the finding did not involve ALARA/work controls. Further,
based on the inspectors review of: (1) dose rates and contamination levels in the
steam generator bowls; (2) licensee dose evaluations; and (3) worst case stay time
estimates within the bowls, there were no radiological exposures in excess of regulatory
26 Enclosure
limits, nor was there a substantial potential for an overexposure. Additionally, the
licensees ability to assess dose was not compromised. Consequently, the inspectors
concluded that the SDP assessment for this finding was of very low safety significance
(Green).
Finally, the inspectors identified that multiple cross-cutting areas were affected by these
events. Specifically, the root cause of inadequate supervisory oversight and the
contributing cause relative to procedure use and adherence were the result of poor
human performance. The contributing cause relative to the stations use and evaluation
of operating experience relates to Problem Identification and Resolution. Finally, an
inadequate level of questioning attitude was exhibited by station staff relative to: (1) the
cause of the apparent low air flow, and (2) the one contractors failure to immediately
leave the steam generator bowl upon loss of air.
Enforcement: 10 CFR 20.1703 requires, in part, that if a licensee permits the use of
respiratory protection equipment, the licensee: (1) use only equipment that is tested
and certified by the NIOSH, unless authorized by the NRC; and (2) implement and
maintain a respiratory protection program that includes written procedures regarding the
storage, issuance, maintenance, repair, testing, and quality assurance of respiratory
protection equipment. Contrary to the above, on April 9, 2004, the licensee increased
the air line pressure of the bubble hoods employed during steam generator nozzle dam
installation activities, beyond the procedural requirements of HPIP 4.58, Issuance of
Respiratory Equipment, which resulted in the licensee utilizing a respiratory protection
device contrary to the NIOSH certification and without NRC authorization. However,
because the licensee documented this issue in its corrective action program
(CAP055527/RCE 253), conducted a full evaluation into the causes of the events, and
took corrective actions to address staff knowledge of procedural adherence prior to
nozzle dam removal activities, and the violation is of very low safety significance, it is
being treated as an NCV (NCV 05000266/2004003-03).
.3 Radiation Worker Performance
a. Inspection Scope
During job performance observations, the inspectors evaluated radiation worker
performance with respect to stated radiation protection work requirements and
evaluated whether workers were aware of the significant radiological conditions in their
workplace, the RWP controls and limits in place, and that their performance accounted
for the level of radiological hazards present.
These reviews represented one inspection sample.
b. Findings
No findings of significance were identified.
27 Enclosure
.4 Radiation Protection Technician Proficiency
a. Inspection Scope
During job performance observations, the inspectors evaluated RP technician
performance with respect to radiation protection work requirements and evaluated
whether they were aware of the radiological conditions in their workplace, the RWP
controls and limits in place, and if their oversight of radiological activities was consistent
with their training and qualifications with respect to the radiological hazards and work
activities.
These reviews represented one inspection sample.
b. Findings
No findings of significance were identified.
2OS2 As-Low-As-Is-Reasonably-Achievable Planning And Controls (ALARA) (71121.02)
.1 Inspection Planning
a. Inspection Scope
The inspectors reviewed the U1R28 refueling outage work scheduled during the
inspection period and associated work activity exposure estimates for the following four
work activities which were likely to result in the highest personnel collective exposures:
- U1R28 RP Coverage [RWP No. 04-104];
- Bottom Mounted Instrumentation Inspection [RWP No. 04-133];
- Nozzle Dam Installation/Removal [RWP No. 04-141]; and
- Steam Generator ET Testing [RWP No. 04-142].
These reviews represented one inspection sample.
b. Findings
No findings of significance were identified.
.2 Radiological Work Planning
a. Inspection Scope
For those activities identified in Section 2OS2.1, the inspectors reviewed the ALARA
evaluations, exposure estimates, and exposure mitigation requirements to determine if
the licensee had established procedures, and engineering and work controls that were
based on sound radiation protection principles in order to achieve occupational
exposures that were ALARA.
28 Enclosure
The interfaces between radiation protection, operations, maintenance, planning,
scheduling, and engineering groups were evaluated by the inspectors to identify
interface problems or missing program elements. The inspectors evaluated if work
activity planning included consideration of the benefits of dose rate reduction activities,
such as shielding provided by water filled components/piping, job scheduling, and
shielding and scaffolding installation/removal activities. Finally, the inspectors evaluated
the integration of radiological job planning activities (pre-job ALARA reviews) into work
procedure and RWP documents.
These reviews represented three inspection samples.
b. Findings
No findings of significance were identified.
.3 Verification of Dose Estimates and Exposure Tracking Systems
a. Inspection Scope
The inspectors reviewed the licensees process for adjusting exposure estimates or
re-planning work, when unexpected changes in scope, emergent work, or higher than
anticipated radiation levels were encountered. This review included a determination if
adjustments to estimated exposures (intended dose) were based on sound radiation
protection and ALARA principles, rather than adjustments to account for failures to
adequately control the work. The frequency of these adjustments was reviewed to
evaluate the adequacy of the original ALARA planning process. In particular, the
inspectors reviewed and discussed with the RP staff the in-progress ALARA reviews
conducted for the bottom mounted instrumentation inspection and steam generator
nozzle dam installation/removal RWPs.
These reviews represented one inspection sample.
b. Findings
No findings of significance were identified.
.4 Job Site Inspections and ALARA Control
a. Inspection Scope
The inspectors observed the four activities identified in Section 2OS1.2 that were being
performed in radiation areas, airborne radioactivity areas, or high radiation areas for
observation of work activities that presented the greatest radiological risk to workers.
The licensees use of engineering controls to achieve dose reductions was evaluated to
determine if procedures and controls were consistent with the licensees ALARA
reviews, sufficient shielding of radiation sources was provided for, and the dose
expended to install/remove the shielding did not exceed the dose reduction benefits
afforded by the shielding.
29 Enclosure
These reviews represented one inspection sample.
b. Findings
No findings of significance were identified.
.5 Radiation Worker Performance
a. Inspection Scope
Radiation worker and RP technician performance was observed during work activities
performed in radiological areas that presented the greatest radiological risk to workers.
The inspectors evaluated whether workers demonstrated the ALARA philosophy in
practice by being familiar with the work activity scope and tools to be used, by utilizing
ALARA low dose waiting areas, and by complying with work activity controls. Also,
radiation worker performance was observed to determine whether individual training/skill
level was sufficient with respect to the radiological hazards and the work involved.
These reviews represented one inspection sample.
b. Findings
No findings of significance were identified.
4. OTHER ACTIVITIES
4OA1 Performance Indicator (PI) Verification (71151)
Cornerstone: Initiating Events
.1 Reactor Safety Strategic Area - Initiating Events Cornerstone
a. Inspection Scope
The inspectors reviewed the licensees recent PI submittal. The inspectors used PI
definitions and guidance contained in Revision 2 of Nuclear Energy Institute
Document 99-02, "Regulatory Assessment Performance Indicator Guideline," to
determine the accuracy of the PI data. The inspectors reviewed selected applicable
conditions and data from logs, licensee event reports, and CAPs from April 2002
through April 2004. The inspectors independently re-performed calculations where
applicable. The inspectors also compared that information to the information required
for each PI definition in the guideline, to ensure that the licensee reported the data
accurately.
These observations constituted four inspection samples. The following PIs were
reviewed:
30 Enclosure
Unit 1
- Unplanned Scrams per 7000 Critical Hours
- Unplanned Scrams with a Loss of Heat Removal
Unit 2
- Unplanned Scrams per 7000 Critical Hours
- Unplanned Scrams with a Loss of Heat Removal
b. Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems (71152)
.1 Routine Review of Identification and Resolution of Problems, Inservice Inspection
Activities
a. Inspection Scope
From April 5 through 28, 2004, in an office on the 8-foot level of the TSB, the inspectors
performed a review of a sample of ISI-related problems that were identified by the
licensee and entered into the corrective action program. The inspectors reviewed these
corrective action program documents to confirm that the licensee had appropriately
described the scope of the problems. Additionally, the inspectors review included
confirmation that the licensee had an appropriate threshold for identifying issues and
had implemented effective corrective actions. The inspectors evaluated the threshold
for identifying issues through interviews with licensee staff and review of licensee
actions to incorporate lessons learned from industry issues related to the ISI program.
The inspectors performed these reviews to ensure compliance with 10 CFR Part 50,
Appendix B, Criterion XVI, Corrective Action, requirements. The specific corrective
action documents that were reviewed by the inspectors are listed in the attachment to
this report.
b. Findings
No findings of significance were identified.
.2 Resident Inspector Routine Review of Identification and Resolution of Problems
a. Inspection Scope
As discussed in previous sections of this report, the inspectors routinely reviewed issues
during baseline inspection activities and plant status reviews to determine if they were
being entered into the licensees corrective action system at an appropriate threshold,
adequate attention was given to timely corrective actions, and adverse trends were
identified and addressed. Minor issues entered into the licensees corrective action
31 Enclosure
system as a result of inspector observations are included in the list of documents
reviewed which is attached to this report.
b. Findings
No findings of significance were identified.
.3 Resident Inspector Semi-Annual Trend Review
a. Inspection Scope
The inspectors performed a semi-annual review of licensee trending activities to
determine if emerging adverse trends that might indicate the existence of a more
significant safety issue were adequately identified, were entered into the licensees
corrective action system at an appropriate threshold, and timely corrective actions were
implemented. The effectiveness of the licensee trending activities was assessed by
comparing trends identified by the licensee with those issues identified by the NRC
during the conduct of routine plant status and baseline inspections.
The inspectors review nominally considered the six-month period of January 2004
through June 2004, although some examples expanded beyond those dates when the
scope of the trend warranted. The inspectors review was focused on operations and
engineering human performance errors, but also considered the results of daily
inspector corrective action program item screening discussed in Section 4OA2.2 above,
licensee trending efforts, and licensee human performance results. This inspection
effort completed one semi-annual trending inspection sample.
b. Findings and Observations
There were no findings of significance identified.
The inspectors evaluated the licensee trending methodology and observed that the
licensee had performed a detailed review. The inspectors compared the licensee
process results with the results of the inspectors daily screening and did not identify any
discrepancies or potential trends in the corrective action program data that the licensee
had failed to identify. The significant trends identified as a result of this review included
trends in the areas of Operations mis-positioning events, the number of human
performance events associated with inattention to detail, implementation of the work
management process, and declining performance associated with Operations internal
and external communications. Trends identified in the Engineering area included the
high number of mechanical and electrical design CAPs, inadequate documentation of
activities, configuration control process issues, and the number of corrective action
program actions being returned. The inspectors verified these trends were captured in
the corrective action system. Finally, the inspectors noted that although the licensee
had initiated an in-depth examination of the station scheduling activities and practices to
identify potential problem areas in the first quarter of 2004, the initiative was not
successful in preventing a large number of late schedule changes and adjustments
immediately prior to the beginning of the U1R28 refueling outage.
32 Enclosure
.4 Resident Inspector Review of an Unexpected Unit 2 Charging Pump Trip
a. Inspection Scope
During the week of April 15, 2004, the inspectors reviewed the circumstances
associated with an unexpected trip of a Unit 2 charging pump to evaluate potential
repetitive equipment failures or human performance issues which might warrant
additional follow-up. In addition, the inspectors reviewed the effectiveness of the CAs
taken on identified issues. This observation constituted one resident inspector
inspection sample
b. Findings and Observations
There were no findings of significance identified.
The inspectors reviewed the background, history, and licensee follow-up actions relating
to the unexpected trip of a Unit 2 charging pump (2P-2B) on January 12, 2004. The
cause of the trip was initially thought to be a failure of a drive belt, but was later traced to
actuation of a protective relay (the 86' over-current relay). Vibrations caused by
opening a sprung breaker cabinet door, while preparing for maintenance unrelated to
the charging pump, caused the actuation. Plant operators used applicable procedures
to assess and make notifications regarding the event, and to promptly restore charging
flow and balance it with letdown flow. The licensees evaluation of the event was
acceptable and the Fix it Now team was assigned to implement timely corrective
maintenance. Appropriate investigative and corrective actions were initiated and an
acceptable extent of condition review was performed. No previous events were
identified relating to 86' relays or to charging pumps, but in one previous event, a relay
vibrated when a cabinet door was opened and caused an inadvertent start of an EDG.
This event was properly classified as a Maintenance Rule Functional Failure as it related
to charging pump 2P-2B.
.5 Resident Inspector Review of Unit 1 Outage Performance
a. Inspection Scope
From April 2 through June 8, 2004, the inspectors reviewed selected Unit 1 refueling
outage events related to worker interface with operations, inattention-to-detail, adequate
self-checking, control of contractors, operators awareness of TS requirements, and
maintaining a critical safety focus in order to evaluate potential repetitive failures or
human performance issues which might warrant additional follow-up. In addition, the
inspectors reviewed the effectiveness of the CAs taken on identified issues. This
observation constituted one resident inspector inspection sample.
b. Findings and Observations
Introduction: The inspectors identified a Green finding concerning the decision by
several licensed and experienced personnel during U1R28 to authorize steam generator
nozzle dam installation prior to the establishment of a vent path through the pressurizer
manway.
33 Enclosure
Description: During the Unit 1 outage from April 2 through June 8, 2004, the inspectors
observed multiple human performance issues.
Potential Loss of Hot Leg Vent Path During Nozzle Dam Installation
At approximately 4:00 a.m. on April 9, senior outage, work control center, and control
personnel (including several senior reactor operators) were notified of difficulties in
removing one of the bolts on the pressurizer manway. In the outage schedule, this
activity preceded installation of the steam generator nozzle dams. According to the
licensee, the senior personnel reviewed the schedule and operating procedure (OP) 4F,
Reactor Coolant System Reduced Inventory Requirements. From the review, the
personnel concluded that the nozzle dam installation could precede while efforts to
remove the pressurizer continued. The personnel, however, did not review the outage
safety plan and the 10 CFR 50.59 evaluation for the nozzle dam installation procedure in
which was contained a discussion of the need for the pressurizer manway removal to
establish an appropriate-sized vent path prior to the complete installation of the nozzle
dams.
Installation of the cold leg nozzle dams was completed and installation of the hot leg
nozzle dams was in progress around 6:00 a.m., when oncoming dayshift personnel
questioned the status of the manway removal. Nozzle dam installation was then halted
and an investigation was begun. The licensees investigation identified that because of
a damaged bolt hole, the hot leg nozzle dam for the A SG was never completely
installed. A licensee engineering evaluation subsequently concluded that an adequate
vent path was provided by this configuration, even with the pressurizer manway in-place.
Other Outage Human Performance Issues Affecting Multiple Disciplines
- On May 15, divers entered the intake crib to begin planned repairs. Licensee
procedures required that dive boat operators be in constant telephone
communications with the control room. Despite this requirement, the dive boat
only communicated with the work control center on an as-needed basis. During
the repairs, a diver entered the north side of the intake that housed the CW
pipes for the operating Unit 2. The suction flow pulled the divers air line and
tether into the intake bell such that the diver could not free himself. According to
the licensee, the licensee contractor liaison assigned to the dive crew did not
exhibit positive control when informed by the contractors that they were also
going to inspect the Unit 2 crib while they were on the dive. When it was noticed
that the diver was being drawn into the Unit 2 CW pipes, the boat crew called the
work control center and requested that the CW pumps be secured. Work control
center personnel notified control room operators, who manually scrammed the
Unit 2 reactor and secured the CW pumps. This manual trip of a reactor
because of problems with a diver is similar to an occurrence in 2000 (Inspection
Report 50-266/00-14(DRP); 50-301/00-14(DRP) and Licensee Event Report
(LER) 266/2000-010). This most recent event demonstrated less than adequate
control of contractors.
- A temporary change was made to Procedure OP-4F on May 19, requiring shiftly
nozzle dam checks. During a procedure review three days later, the licensee
34 Enclosure
noticed that the checks were not being done, a minor violation of regulatory
requirements.
- On April 21, one of three incore thermocouple guides was inadvertently lifted
along with the Unit 1 reactor vessel head. This left the thermocouple
connections exposed and prevented refueling cavity flood up until the guide
could be replaced, lengthening the reduced time-to-boil periods. Visual
examinations were conducted, as required by the head lift procedure, when the
head was approximately 1 foot and 4 feet above the vessel flange, however, the
examinations did not identify the inadvertent lifting of the thermocouple guide.
This issue represented a lack of worker attention-to-detail.
- On April 23, electrical maintenance performed a breaker alignment in
preparation for an upcoming electrical bus outage. During the alignment, the
person performing the breaker alignment and the peer checker both verified that
the correct breaker was going to be manipulated. Both workers, however,
became distracted prior to the manipulation. When the person performing the
manipulation returned his attention to the task of opening the breaker, he opened
the wrong breaker. Approximately three hours later, an auxiliary operator
discovered that the SFP cooling flow was zero despite the A SFP cooling pump
(P-12A) having been previously operating. The auxiliary operator reported the
information to the control room and the operators entered AOP-8F, Loss of
Spent Fuel Pool Cooling, procedure. The operators identified that the breaker
for P-12A was in the OFF position. The breaker was closed and P-12A
restarted. This was an example of inadequate self-checking. No cooling
temperature limits were exceeded.
inspectors observed problems with the breathing air for steam generator bowl
workers. The inspectors discussed these problems with the attendant RP
personnel but the problems continued. The inspectors then provided this
information to licensee senior managers who stopped the nozzle dam installation
work pending a review and resolution of the problems. This is discussed in
section 2OS1.
- Several times during the refueling outage, contract workers entered the
radiological controlled area without properly logging into an RWP. One group of
contractors logged into an RWP for the primary auxiliary building and during the
course of work entered primary containment without returning to access control
to log into the appropriate containment RWP. During work in containment, their
dosimetry alarmed because they had exceeded the dose associated with their
primary auxiliary building RWP. This was an example of inadequate
self-checking.
- Technical Specification 3.9.1 requires boron concentration in the RCS, the
refueling canal, and the refueling cavity to be maintained within the limits
specified in the core operating limits report (COLR). This surveillance was
required every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, but was performed by operations every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. On
April 12, with core 28 still in the reactor vessel, operations noticed that boron
35 Enclosure
concentration was being verified using the COLR associated with core 29, the
core that was to be installed during the refueling outage. The COLR had been
replaced in the control room on April 8. The licensee immediately provided a
copy of COLR 28 to the control room operators and on April 14, COLR 28 was
re-issued. The boron concentration limit for COLR 28 was 2400 parts per million
(ppm), and the limit for COLR 29 was 2200 ppm. The inspectors determined
that the surveillance had been performed daily and boron concentration had
been greater than the required 2400 ppm between April 8 and 12. This issue
involved operator attention-to-detail and timely identification of Technical
Specification issues. The inspectors determined that the issue was minor
because the COLR 28 boron requirements were more conservative than the
COLR 29 requirements.
- 120-Volt Vital Instrument Panel 1Y-04 was de-energized for planned
maintenance at 10:21 p.m. on May 3, affecting the nuclear instrumentation
source range audible count rate circuit. Despite procedure 1-SOP-Y-Y04,
"1Y-04, Yellow 120V Vital Instrument Panel", providing guidance on affected
loads and TSs, senior reactor operators on multiple shifts were not aware of the
TSAC requirement to immediately isolate unborated water sources when the
source range audible count circuit was inoperable. TSAC 3.9.2.C until 4:20 a.m.
on May 4. This issue involved operator inattention-to-detail and timely
identification of Technical Specification requirements. The issue was entered
into the licensee Corrective Action Program (Activity Request CAP056363). The
inspectors determined that this was minor because a review of plant conditions
indicated that all unborated water sources were isolated during the event, even
though the operators had not taken action to perform this requirement.
Analysis: Issues associated with the nozzle dam and hot leg vent path indicated a lack
of safety focus on the part of licensed and experienced personnel such that if shutdown
cooling had been lost, significant consequences may have resulted. In addition, the lack
of safety focus on the part of licensed and experienced personnel was considered to
affect the cross-cutting area of human performance. Despite several barriers,
experienced, supervisory personnel incorrectly allowed nozzle dam installation to
commence prior to pressurizer manway removal.
The inspectors determined that allowing hot leg nozzle dams to be installed before the
intended pressurizer manway hot leg vent path was in place was a performance
deficiency warranting a significance evaluation in accordance with IMC 0612, Power
Reactor Inspection Reports, Appendix B, Issue Screening, issued on June 20, 2003.
The finding was considered more than minor because it affected: (1) the Reactor Safety
Initiating Events cornerstone objective to limit the likelihood of those events that upset
plant stability and challenge critical safety functions during shutdown operations, and (2)
the human performance attribute of the Initiating Events cornerstone.
The inspectors completed a significance determination of the nozzle dam and hot leg
vent path issue using IMC 0609, Significance Determination Process, dated
March 21, 2003, Appendix G, Shutdown Operations, dated May 25, 2004. The
inspectors determined that the finding was considered to be of very low safety
significance (Green) and did not require quantitative assessment since (1) conditions
36 Enclosure
meeting a loss of control were not met in that no inadvertent change in RCS
temperature or change in reactor vessel level actually occurred, and (2) the licensee had
maintained adequate mitigation capability for plant conditions associated with a PWR in
refueling operations with a time-to-boil of less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The inspectors noted,
however, that the mitigation capability was not achieved by intentional or controlled
actions on the part of the licensee, rather, the problems with the bolt hole in the A SG
hot leg precluded full installation of all four nozzle dams while the pressurizer manway
was still attached.
Enforcement: Because (1) the actual sequence of events showed that all four nozzle
dams had not been completely installed while the pressurizer manway was still in place;
and (2) an engineering analysis showed that an adequate hot leg vent path was
available while one of the A SG hot leg nozzle dam side pieces was not installed, no
violation of regulatory requirements occurred. This issue was considered a finding of
very low significance (FIN 05000266/2004003-04). The licensee entered the finding into
its corrective action program as CAP055538, Potential for No Hot Leg Vent Path during
Unit 1 SG Nozzle Dam Installation.
4OA3 Event Follow-up (71153)
.1 (Closed) Licensee Event Report (LER) 50-301/2004-001-00: SI System Accumulator
Operated With Fluid Level Out of Specification High.
Introduction: A Green NCV of TS Surveillance Requirement (SR) 3.5.1.2 was self-
revealed when the water volume in SI accumulator 2T-34A exceeded the Technical
Specification limit of 1136 cubic feet. The finding was considered self-revealing since it
was indicated through a change in the functionality of the accumulator level transmitters
during routine operations.
Description: SI accumulator 2T-34A was drained for maintenance on October 8, 2003,
during the 2003 Unit 2 refueling outage. The two associated level transmitters were
calibrated, filled, and vented while the accumulator was drained. The accumulator was
re-filled on October 29, 2003, and a discrepancy was noted between level indicators 2LI-
939 and 2LI-938. The level transmitters were vented and filled, and the transmitter
equalizing valves were opened and closed. However, opening the equalizing valves
improperly allowed water into the dry reference legs of the transmitters and introduced a
bias such that the indicated accumulator level was lower than the actual level.
Upon completion of the refueling outage, Unit 2 entered Mode 3 (on November 10,
2003), when the SI accumulators were required by TSs to be operable, and
subsequently resumed full power operations. On February 14, 2004, 2LI-939 was noted
to be drifting lower. Troubleshooting efforts commenced and the transmitter reference
leg was drained. Upon return to service, the 2LT-939 indication was high off scale.
Subsequently, the transmitter was found to be out-of-calibration. On February 15, the
transmitter was replaced with an on-site spare. When placed in service, the
replacement transmitter again indicated high off scale. A calibration check found the
output to be drifting and attempts to calibrate the transmitter were unsuccessful. On
February 22, a new transmitter from the manufacturer was installed. When placed in
service, the new transmitter indicated high off scale, for a third time. On March 19, 2LT-
37 Enclosure
939 was replaced with a different model level transmitter under a modification. When
placed in service, the new transmitter again indicated high off scale, for a fourth time.
On March 30, UT of the sensing and reference lines of both 2LT-938 and 2LT-939
identified that 2LT-939 was indicating correctly, water was present in the reference leg
of 2LT-938, and the actual accumulator level had exceeded the TS limit. Immediate
actions were taken to enter the applicable TS Action Condition, to restore the
accumulator to operable status. A root cause evaluation was performed and an LER
was submitted.
Analysis: The inspectors determined that operating with the Unit 2 A safety injection
accumulator with levels above TS limits was a performance deficiency warranting a
significance evaluation. The inspectors concluded that the finding was greater than
minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B,
Issue Screening, issued on June 20, 2003, because it affected the Reactor Safety
Mitigating Systems cornerstone objective to ensure the availability, reliability, and
capability of systems that respond to prevent initiating events.
The inspectors completed a significance determination of this issue using IMC 0609,
Significance Determination Process, dated March 21, 2003, Appendix A, Significance
Determination of Reactor Inspection Findings for At-Power Situations, dated
March 18, 2002. The inspectors determined that the finding was considered to be of
very low safety significance (Green) since (1) the Nuclear Steam Supply System vendor
performed an analysis of the over-filled, as-found condition and determined that the
2T-34A accumulator had been capable of performing the design basis function and
would not have challenged the 10 CFR 50.46 Loss-of-Coolant-Accident acceptance
criteria, and (2) the finding did not result in a design or qualification deficiency, an actual
loss of safety function, or involve internal or external initiating events. This finding was
assigned to the Reactor Safety Mitigating Systems Cornerstone for Unit 2.
Enforcement: Technical Specification 3.5.1 requires that two SI accumulators be
operable in Modes 1 and 2, and in Mode 3 with RCS pressure greater than 1000 psig.
Technical Specification SR 3.5.1.2 requires that borated water volume in each
accumulator be verified to be greater than or equal to 1100 cubic feet, and less than or
equal to 1136 cubic feet every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Contrary to these requirements, the water
volume in safety injection accumulator 2T-34A was greater than 1136 cubic feet
between November 10, 2003, and March 30, 2004.
This violation was entered into the licensee's corrective action system as CAP055204
Troubleshooting Reveals 2T34A SI Accumulator Level Out of Specification High.
Because this violation was of low safety significance and it was entered into the
licensees corrective action program, this violation is being treated as an NCV,
consistent with Section VI.A of the NRC Enforcement Policy. (NCV 05000301/2004003-
05). This LER is closed.
38 Enclosure
.2 (Closed) LER 50-266/2004-001-00: Reactor Pressure Vessel Head Penetration 26 Flaw
Indications.
During NDE Unit 1 reactor pressure vessel head, possible flow indications were
observed using UT of the J groove weld area for control rod drive mechanism head
penetration 26. To further characterize these indications, multiple PTs of the
penetration 26 J groove weld area were performed, including PT following minor
excavation of the weld surface area. The results of these examinations confirmed the
existence of flaws in the J groove weld. On May 6, 2004, based upon preliminary
analyses, the licensee determined that these indications would probably not be found
acceptable under ASME Code standards. Therefore, the condition was reported under
10 CFR 50.72(b)(3)(ii)(A) as a significant degradation of a principal safety barrier.
Research by the licensee on the observed UT indications concluded that they were likely
the result of weld repairs during fabrication and were not related to primary water stress
corrosion cracking.
The licensee conducted an extent-of-condition evaluation. The evaluation concluded
that the tubing, counterbore region, and other areas of the reactor vessel head were
free of defects, wastage and boric acid deposition and, therefore, the structural integrity
and leak integrity of the reactor vessel head was assured. The licensee based this
conclusion on:
- Review and comparison of the ultrasonic signatures obtained during the U1R27
and U1R28 underhead inspections;
- Review of available fabrication records;
- Inspection of the Alloy 600 control rod drive mechanism tubing, counterbore
region and reactor vessel head in accordance with NRC Order EA-03-009; and
- Verification of the leakage integrity of the RCS boundary by visual examination of
the top of the reactor pressure vessel head with no evidence of defects, wastage
or boric acid crystal deposition.
A weld repair of the penetration 26 control rod drive mechanism nozzle was completed
on May 22, in accordance with an approved plant modification. On June 4, the NRC
approved a relaxation of the first revised order EA-03-009 regarding the upcoming
Unit 1 operating cycle. This allowed removal of a Unit 1, Mode 2 restraint and
authorized full power operations of Unit 1 for one operating cycle. Unit 1 achieved
criticality on June 7 and returned to full power operations on June 11, 2004.
Further information concerning NRC review of the Unit 1 reactor vessel head inspection
is provided in Section 4OA5.1 of this report. This LER was reviewed by the inspectors
and no findings of significance were identified. This LER is closed.
39 Enclosure
.3 Unit 2 Manual Reactor Trip
a. Inspection Scope
On May 15, 2004, the inspectors observed operator response to a manual reactor trip
when a diver entered the north side of the CW intake crib and could not free himself.
The inspectors reviewed operator actions to trip both CW pumps immediately after the
reactor was manually tripped and maintain RCS temperature using the steam generator
atmospheric steam dump valves. The inspectors reviewed the event for equipment
problems and ensured Unit 2 was stabilized at normal operating temperature and
pressure. The inspectors observed portions of the Unit 2 restart efforts, including
operator response to turbine electro-hydraulic control system malfunctions.
b. Findings
No findings of significance were identified.
4OA4 Cross-Cutting Aspects of Findings
.1 A finding described in Section 1R05.1 of this report had, as a contributing cause, a
human performance deficiency, in that the licensee failed to identify transient
combustible materials during tours required by the FPER and NP procedures.
.2 A finding described in Section 2OS1.2 of this report, had, as primary causes, human
performance deficiencies, in that: (1) there was inadequate supervisory oversight of
bubble hood use during steam generator nozzle dam installation; and (2) RP staff failed
to appropriately apply the procedure for bubble hood issuance and the maximum air flow
criteria.
.3 A finding described in Section 2OS1.2 of this report, had, as a contributing cause,
inadequate evaluation of operating experience (an element of problem identification and
resolution), in that certain quick disconnect air line fittings had recently been identified
as being susceptible to inadvertently separating. However, the licensee, in its
evaluations of these operating experiences, failed to physically challenge the bubble
hood air fittings used at the station.
.4 A finding described in Section 4OA2.5 of this report had, as its primary cause, a human
performance deficiency. Several licensed and experienced personnel incorrectly
authorized the installation of steam generator nozzle dams prior to the establishment of
a vent path through the pressurizer manway.
.5 A finding described in Section 4OA5.4 of this report had, as its primary cause, a human
performance deficiency, in that despite problems with a software program being
previously known and operations department management expectations to perform line-
by-line reviews prior to distribution having been established, 16 undetected errors in
emergency operating, emergency contingency action, critical safety, and shutdown
emergency procedures for Units 1 and 2 existed between October 3, 2003, and
July 9, 2004.
40 Enclosure
4OA5 Other Activities
.1 Reactor Pressure Vessel (RPV) Head and Vessel Head Penetration Nozzles
a. Inspection Scope
On February 11, 2003, the NRC issued Order EA-03-009 (ADAMS Accession Number
ML030410402). This order required examination of the reactor pressure vessel head
and associated vessel head penetration (VHP) nozzles to detect primary water stress
corrosion cracking (PWSCC) of VHP nozzles and corrosion of the vessel head. The
purpose of TI 2515/150, Reactor Pressure Vessel Head and Vessel Head Penetration
Nozzles, Revision 2, was to implement an NRC review of the licensee's head and VHP
nozzle inspection activities required by NRC Order EA-03-009. The inspectors
performed a review in accordance with TI 2515/150 of the licensees procedures,
equipment, and personnel used for examinations of the Unit 1 RPV and VHP to confirm
that the licensee met requirements of NRC Order EA-03-009 (as revised by NRC letter
dated February 20, 2004). The results of the inspectors review included documentation
of observations and conclusions in response to the questions identified in TI 2515/150.
From April 5 through May 26, 2004, in an office on the 8-foot level of the TSB building,
(unless otherwise stated), the inspectors performed a review of the licensees Unit 1
head inspection related activities in response to NRC Order EA-03-009. To evaluate the
licensees efforts in conducting the required examinations, the inspectors:
C performed direct visual examination of the head-to-nozzle interface for portions
of 30 VHP nozzles inside the Unit 1 containment from access doors in the
service structure surrounding the head;
C observed, inside the Unit 1 containment building, licensee personnel conducting
a remote visual examination of the RPV head for portions of 12 VHP nozzles;
C conducted interviews with the licensees nondestructive examination personnel
performing nondestructive examinations of the vessel head in the head
inspection trailer within the site protected area;
C reviewed the head inspection procedures;
C reviewed the certification records for the nondestructive examination personnel
performing examinations of the vessel head;
C reviewed the procedures used for identification and resolution of boric acid
leakage from systems and components above the vessel head;
C reviewed the licensees procedures and corrective actions implemented for boric
acid leakage;
C reviewed in an on-site trailer the videotaped PT examinations conducted on the
VHP nozzle No. 26 J-weld;
41 Enclosure
C reviewed in an on-site trailer the videotaped cleaning and visual examination of
portions of six head-to-nozzle interface areas;
C reviewed in an on-site trailer automated UT data for rotating and blade probes
collected during the Unit 1 vessel head at 20 VHP nozzle locations;
C reviewed in an on-site trailer automated UT data collected for VHP nozzles No.
32 and No. 33 during the previous Unit 1 outage; and
C observed from a remote camera monitor in an on-site trailer manual UT
examination of the lower portions of VHP nozzles No. 32 and No. 33.
The inspectors conducted these reviews to confirm that the licensee performed the
vessel head examinations in accordance with requirements of NRC Order EA-03-009 (or
Order relaxation requests), using procedures, equipment, and personnel qualified for
the detection of PWSCC in VHP nozzles and detection of vessel head wastage.
From May 11 through 26, 2004, in an office on the 8-foot level of the TSB building,
(unless otherwise stated), the inspectors performed a review of the licensees repair
activities for VHP nozzle No. 26. The inspectors reviewed the licensees weld
procedures, certified mill test reports for the weld materials, process traveler steps, and
weld control records, and observed portions of the repair welding in the Unit 1
containment to confirm that ASME Code Section III and Section IX requirements were
met (as amended by a licensee's Code relief request).
From April 5 through 28, 2004, in an office on the 8-foot level of the TSB, the inspectors
reviewed the licensees VHP nozzle susceptibility ranking calculation C11470, Reactor
Vessel Head Effective Degradation Year (EDY), to:
C verify that appropriate plant-specific information was used as input;
C confirm the basis for the head temperature used by licensee; and
C determine if previous VHP cracks had been identified, and if so, documented in
the susceptibility ranking calculation.
b. Observations
Summary
The licensee performed a remote visual examination of the top surface of the Unit 1
vessel head using a robotic crawler with a high-resolution camera supplemented with
direct visual examinations to complete inspection of the 49 Unit 1 VHP nozzles and the
head vent line penetration. Based upon this inspection, the licensee did not identify any
leaking VHP nozzles or evidence of vessel head wastage. The licensee also conducted
UT examinations for each of the 49 VHP nozzles and for the head vent line penetration
nozzle. Due to limitations in UT examination coverage at the bottom end of 17 VHP
nozzle locations, the licensee requested relaxation from Order EA-03-009 requirements.
42 Enclosure
The licensee also performed PT examinations of the head vent line and VHP nozzle No.
26 J-groove weld locations. During the PT examination of the VHP nozzle No. 26 J-
groove weld, the licensee identified linear indications (cracks) which required repair.
The licensee subsequently removed the cracked nozzle No. 26 J-groove weld and
completed a temper bead weld repair.
Evaluation of Inspection Requirements
In accordance with requirements of TI 2515/150, the inspectors evaluated and
answered the following questions:
1. For each of the examination methods used during the outage, was the
examination performed by qualified and knowledgeable personnel? (Briefly
describe the personnel training/qualification process used by the licensee for this
activity.)
Above Head Visual Examinations
Yes. The licensee conducted a remote and direct visual examination of the top
surface of the vessel head with knowledgeable staff members certified to Level II
or Level III as VT-2 examiners in accordance with procedure NDE-3, Written
Practice For Qualification And Certification For NDE Personnel. This
qualification and certification procedure met the industry standard ANSI/ANST
CP-189, Standard for Qualification and Certification of Nondestructive Testing
Personnel. Additionally, the licensees VT-2 personnel had access to
photographs of each penetration location taken during the last Unit 1 visual head
inspection, completed in 2002.
Under Head Automated UT Examinations
Yes. The licensees vendor personnel that performed the automated UT were
certified to Level II or III in UT examination in accordance with vendor
(Framatome) procedure 54-ISI-30-01, Written Practice for the Qualification and
Certification of NDE Personnel. This procedure met the industry standard
ANSI/ANST CP-189, Standard for Qualification and Certification of
Nondestructive Testing Personnel. Additionally, the licensees vendor UT
acquisition and analysis personnel had a minimum of 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> training on the
automated UT examination techniques used.
Under Head Manual Ultrasonic Examinations
Yes. The licensee conducted a manual UT examination of the lower portions of
VHP nozzles No. 32 and No. 33 below the J-groove weld with a knowledgeable
staff member certified to Level III as for UT examination in accordance with
procedure NDE-3, Written Practice For Qualification And Certification For NDE
Personnel. This procedure met the industry standard ANSI/ANST CP-189,
Standard for Qualification and Certification of Nondestructive Testing
Personnel.
43 Enclosure
Under Head PT Examinations
Yes. The licensee conducted a solvent removable PT examination of the head
vent and penetration VHP nozzle No. 26 J-groove weld locations with a
knowledgeable staff member certified to Level III in PT examination in
accordance with procedure NDE-3, Written Practice For Qualification And
Certification For NDE Personnel. This procedure met the industry standard
ANSI/ANST CP-189, Standard for Qualification and Certification of
Nondestructive Testing Personnel.
2. For each of the examination methods used during the outage, was the
examination performed in accordance with demonstrated procedures?
Above Head Visual Examinations
Yes. The licensee performed a bare metal inspection of the vessel head in
accordance with procedure NDE-757, Visual Examination For Leakage of
Reactor Pressure Vessel Penetrations. The licensee considered this procedure
to be demonstrated because examination personnel could resolve lower case
alpha numeric characters 0.158 inches in height at a maximum of 6 feet under
existing lighting, which met the Code visual VT-2 examination criterion.
However, the inspectors identified parameters that could impact the
quality/effectiveness of the inspection which were not controlled by the
procedure. Specifically, the procedure did not provide:
C guidance for when and how to collect samples of deposits if any had
been identified near the interface of lower head penetrations. Further, no
guidance existed to identify what analysis would be performed to
determine the source of deposits identified. Instead, the licensee staff
stated that they would follow a Bottom Mounted Instrument Inspection
Decision Tree Diagram to make decisions on sampling of deposits on the
upper head;
C guidance or threshold for identification and documentation of corrosion or
wastage (e.g., 1 percent or 10 percent wastage etc.). Note that the
licensee and NRC inspectors did not identify any significant corrosion or
wastage in the visual examinations of the RPV head;
C demonstration of the near distance resolution capability for the remote
camera system; and
C demonstration of color resolution capability for the remote camera
system.
For the items discussed above, the licensee provided verbal direction or
controlled the parameters, such that the inspectors did not consider the quality of
the visual examination to be compromised.
44 Enclosure
The inspectors observed licensee personnel performing the remote visual
examination of the upper surface of the reactor head under the insulation using a
camera mounted to a robotic crawler in accordance with procedure NDE-757 for
portions of 12 vessel head VHP nozzle locations. The licensee was able to
position the inspection camera within a few inches of the vessel head penetration
VHP nozzle interface with sufficient lighting such that a sharp/clear visual image
was obtained. The inspectors judged the resolution capability of the remote
visual camera system to be very good, based upon the ability to resolve very
small debris particles at the penetration nozzle-to-head interfaces.
The inspectors reviewed the licensees demonstration of visual resolution and
noted that it was consistent with the procedure requirements. The inspectors
also performed a direct visual inspection for portions of 30 VHP nozzles viewable
at 5 of the 6 inspection ports in the service structure. Based on this examination,
the inspectors noted that the remote picture quality appeared to provide for a
superior inspection to that achievable by a direct visual examination from the
service structure access doors.
Under Head UT and PT Examinations
Yes. The licensees vendor performed automated UT examinations in
accordance with Framatome ANP Nondestructive Examination Procedure 54-ISI-
100-11, Remote Ultrasonic Examination of Reactor Head Penetrations. The
licensees vendor demonstrated an earlier version of this procedure on mockup
VHP nozzles which contained cracks or simulated cracks as documented in
EPRI MRP-89, Materials Reliability Program Demonstrations of Vendor
Equipment and Procedures for the Inspection of Control Rod Drive Mechanism
Head Penetrations. The inspectors reviewed the revisions to procedure 54-ISI-
100-11 implemented since the licensees vendor had demonstrated this
procedure in EPRI MRP-89, to ensure that any equipment configuration changes
did not affect flaw detection capability. Additionally, the licensees vendor had
demonstrated the capability to detect a leakage path in the interference zone
using this procedure on a mockup with a simulated leak path and at other
nuclear power plants with observed leakage paths such as the Oconee Units.
However, the inspectors noted that this UT procedure/method was not designed
to detect PWSCC contained entirely within the J-groove welds of VHP nozzles.
The inspectors identified a potential weakness in the licensees implementation
of procedure 54-ISI-100-11, Remote Ultrasonic Examination of Reactor Head
Penetrations. The inspectors noted that the licensees vendor typically ran the
blade UT probe to failure which precluded a final calibration check of the failed
UT probe. If the vendor had elected to incorporate the ASME Code Section XI
rules into this procedure, the examination data would have been considered
invalid back to the last known UT equipment calibration check. The licensees
vendor UT analyst typically accepted the UT data up to point of probe failure.
This practice was allowed by the licensees procedure; however, the inspectors
concluded that it placed greater reliance on the licensees vendor UT data
analyst which could increase the probability of missing cracks due to human
errors.
45 Enclosure
Unknown. The licensee conducted under head automated UT examinations of
the vessel head vent line nozzle penetration in accordance with procedure
54-ISI-137-03, Remote Ultrasonic Examination of Reactor Vessel Head Vent
Line Penetrations. The licensees vendor considered this procedure
demonstrated based upon the ability to see electric discharge machined (EDM)
notches in the UT calibration standard (reference 54-PQ-137-01, Remote
Ultrasonic Examination of Reactor Vessel Head Vent Line Penetrations). The
inspectors noted that this type of demonstration would not assure the capability
of this equipment to detect PWSCC. Therefore, the inspectors could not
independently confirm the ability of this equipment to detect PWSCC in the head
vent line nozzle base material.
Yes. The licensee conducted manual UT examinations of the lower portions of
VHP nozzles No. 32 and No. 33 below the J-groove weld in accordance with
procedure NDE-141, Manual Ultrasonic Examination of Reactor Head
Penetrations. The licensee demonstrated this procedure in a blind test on a
control rod drive penetration tube mockup with EPRI. EPRI considered this
procedure qualified for detection only and not for sizing of flaws. This manual
UT examination did not include the J-groove weld region of VHP nozzles No. 32
and No. 33.
Yes. To detect PWSCC in the J-groove weld area of the head vent line and VHP
nozzle No. 26, the licensee performed a PT examination in accordance with
procedure NDE-451 Visible Dye Penetrant Examination Temperature
Applications 45 degrees Fahrenheit to 125 degrees Fahrenheit. The licensee
considered the use of an ASME Code qualified solvent removable visible PT
procedure to detect surface breaking PWSCC flaws in the J-groove welds as
demonstrated. This procedure allowed the licensee to use a greater
temperature range over the standard band specified in Article 6, of Section V of
the ASME Code. The ASME Code allows expanded temperature ranges if the
procedure is demonstrated at the limits of the expanded temperature band. The
inspectors confirmed that the licensee had appropriately demonstrated the
procedure on a quench cracked aluminum comparator block in accordance with
the ASME Code Section V, Article 6 requirements.
3. For each of the examination methods used during the outage, was the
examination able to identify, disposition, and resolve deficiencies and capable of
identifying the PWSCC and/or head corrosion phenomena described in Order
Above Head Visual Examinations
Yes. The inspectors determined through direct observation of the bare metal
head, interviews with inspection personnel, reviews of procedures and inspection
reports, and reviews of videotape documentation that the licensee was capable
of detecting and characterizing leakage from cracking in VHP nozzles.
The upper head had been cleaned during the previous outage and was relatively
free of debris or deposits which would mask evidence of leakage. The
46 Enclosure
inspectors performed a direct visual examination through five of six viewing ports
in the service structure and observed the licensee performing the remote video
inspection of the bare metal head conducted under the insulation with a camera
mounted to a magnetic crawler. The licensee also supplemented the remote
visual with direct visual examinations and performed frequent checks of the VT-2
visual examination quality indicator card during these examinations. Overall, the
inspectors concluded that the remote visual examination resolution and picture
quality equal or superior to a direct visual examination. The licensee was able to
obtain a visual examination at each of the 49 VHP nozzles and the head vent line
nozzle penetration, with no obstructions or interferences. Therefore, the
inspectors concluded that the inspection performed was capable of detecting
evidence of leakage at the VHP nozzle penetrations caused by PWSCC or
corrosion of the vessel head caused by boric acid.
Under Head VHP Automated UT Examinations
Yes. For the VHP nozzle base metal material, the UT equipment, techniques
and procedures had been demonstrated as effective in detection of PWSCC.
The licensee used automated UT equipment with two different configurations. A
blade-type UT probe was used to acquire data for sleeved VHP nozzles and
relied on a single transducer pair optimized for detection of circumferentially
oriented flaws using a time of flight diffraction (TOFD) UT technique. A rotating
head type UT probe was used to acquire data from VHP nozzles without thermal
sleeves. The rotating probe contained multiple TOFD transducer configurations
and shear wave transducers which were designed to optimize detection of both
circumferential and axial oriented flaws. Both the blade and rotating head UT
probes were configured to detect evidence of leakage/corrosion in the
interference zone behind the VHP nozzle based on the pattern in the UT
backwall response. During the Unit 1 VHP examinations, the licensees vendor
identified that a rotating probe shear wave transducer failed to detect the
reflectors in the calibration block during the post-examination calibration check
because it was too noisy. The licensee determined that loss of data from this
one transducer had no effect on the rotation probe's ability to detect PWSCC
due to the multiple transducers on the rotating probe which still functioned
properly. The inspectors agreed with the licensees evaluation that failure of this
UT transducer would not affect the ability of the rotating probe to detect PWSCC.
No. The licensees UT examination methods implemented on the VHP nozzles
were not designed to detect J-groove weld cracking and, therefore, had not been
demonstrated for detection of PWSCC or other flaws contained entirely within
the J-groove welds. Therefore, for PWSCC contained entirely within the
J-groove weld, the inspectors concluded that the licensees UT examination
method would not be effective for detection of PWSCC.
Under Head Vent Line Penetration Automated UT Examinations
Unknown. A rotating probe with pulse-echo type shear and longitudinal wave
transducers was used to acquire data from the head vent line penetration. The
licensees vendor considered the UT method used on the head vent nozzle as
47 Enclosure
demonstrated based upon the ability to see simulated cracks (EDM notches) in
the UT calibration standard (reference 54-PQ-137-01, Remote Ultrasonic
Examination of Reactor Vessel Head Vent Line Penetrations). The EDM
process results in a uniform notch with a relatively wide air filled gap
perpendicular to one surface that is readily detected by UT examination. In
contrast, PWSCC gaps are very small (e.g., tight), are not uniform in nature and
may not be perpendicular to the surface, which represents a more significant
challenge for detection by UT examination. Therefore, the inspectors concluded
that demonstration of this UT technique on EDM notches in the calibration
standard was not sufficient to confirm the ability of this UT probe to detect
Under Head Manual UT Examinations Of VHP Nozzle No. 32 And No. 33
Yes. The licensee performed manual UT examinations of the lower portions of
VHP nozzles No. 32 and No. 33 below the J-groove weld in accordance with
procedure NDE-141, Manual Ultrasonic Examination of Reactor Head
Penetrations. The licensee demonstrated this procedure during a blind test on
a VHP nozzle mockup containing EDM notches at an EPRI facility. The
licensees inspector also examined samples of VHP nozzles with PWSCC
removed from the Oconee plant. The EPRI staff confirmed that the licensees
inspector was able to detect the PWSCC flaws in the Oconee samples.
Therefore, the inspector concluded that the licensee procedure was qualified for
detection of PWSCC flaws in the VHP nozzle base material.
Under Head Penetration PT Examinations
Yes. The licensee conducted a PT examination of the head vent line and VHP
nozzle No. 26 J-groove weld in accordance with procedure NDE-451. The
inspectors observed the videotaped PT examination conducted on the head vent
line penetration J-groove weld and confirmed that the licensee met Code
penetrant dwell time and developer times and observed that no recordable
indications were identified. For the VHP nozzle No. 26 J-groove weld, the
licensee performed a series of PT examinations (with intermediate
buffing/grinding steps) and confirmed two patches of multiple linear indications in
the J-groove weld. The inspectors observed the videotaped PT examinations
conducted on the VHP nozzle No. 26 J-groove weld that identified the two areas
of small linear indications. Therefore, the inspectors concluded that the Code
qualified PT examination of these J-groove welds was capable of detecting
PWSCC based on identification of flaw-like indications in VHP nozzle No. 26 and
based upon a review of vendor data that clearly showed the ability of Code PT
examinations to detect PWSCC at other reactor sites.
48 Enclosure
4. What was the physical condition of the reactor head (debris, insulation, dirt,
boron from other sources, physical layout, viewing obstructions)?
Above Head Visual Examinations
The Unit 1 vessel head insulation consisted of reflective metal insulation panels
installed on a support structure over the top of the reactor head with access for
visual examinations through six viewing ports in the metal service structure
surrounding the top of the head. The inspectors viewed the bare metal head
condition through five of these six viewing ports and considered the head
condition relatively clean. The outer surface of the penetration tubes above the
head generally contained a sprayed-on white mastic coating which had been
applied as a sealer in the original head insulation design. The bare metal head
was covered with a light gray colored coating applied by the head fabricator,
which provided an adequate surface for visual resolution of boric acid deposits.
The inspectors also observed portions of the licensees visual examination and
portions of videotapes of examinations completed on other shifts. The remote
camera visual inspection was conducted under the insulation support structure
and the as-found head condition was generally clean (free of debris, insulation,
dirt). For some penetration locations, the annulus gap contained loose debris
(presumed to be mastic which was scraped off the upper penetration tube
housings during installation of new insulation during the last outage), which did
not hinder the licensees evaluation of the penetrations, because the licensee
vacuumed, blew air, or used a soft brush to remove this loose debris. The
licensee supplemented the remote camera inspection with direct visual
examinations at some VHP nozzles. The licensee did not identify any
obstructions which limited their visual inspection and licensee inspection
personnel were able to fully examine the 49 VHP nozzles and the head vent line
penetration.
The inspectors identified that the licensee had not determined if the visual
examination scope would meet NRC Order EA 03-009 requirements. NRC order
EA-03-009, dated February 20, 2004, required the licensee to complete a 95
percent surface area examination of the upper head including areas upslope and
downslope of the service structure. The service structure and vertical insulation
panels represented areas where the vessel head surface was not examined.
The inspectors questions as to the adequacy of the visual examination coverage
prompted the licensee to document, in CAP056522, the need to develop a
calculation to estimate the area of visual examination coverage. The licensee
subsequently decided to document coverage in an internal memorandum dated
May 17, 2004. In this memorandum, the licensee determined through review of
drawings related to the head, head service structure, and insulation package,
that the total head area not available for visual examination was 1.5 percent.
The inspectors questions as to how this number was calculated prompted the
licensee to issue a new memorandum dated May 24, 2004, which documented
the square inches of surface area obstructed. In this memorandum, the licensee
changed the total obstructed area to 5 percent and concluded that the visual
examination scope would be able to achieve the 95 percent coverage required
by the Order.
49 Enclosure
5. Could small boron deposits, as described in Bulletin 2001-01, be identified and
characterized?
Above Head Visual Examinations
Yes. Based upon the quality and scope of the licensees visual examination, and
independent direct observations, the inspectors concluded that any boron
deposits characteristic of coolant leakage would have been identified (if any had
been present). The inspectors noted that no boric acid deposits were found on
the 49 VHP nozzles and head vent line penetration nozzle. The inspectors
independently observed the remote visual examination for portions of 12 VHP
nozzles and direct examinations of portions of 30 VHP nozzles and did not
observe white deposits (boric acid) with characteristics (adherent popcorn-like)
indicative of reactor coolant system leakage. The licensee performed a
systematic inspection and documented the visual examination results for every
nozzle-to-vessel interface location. No indications of head leakage were
recorded.
6. What material deficiencies (i.e., cracks, corrosion, etc) were identified that
require repair?
At VHP nozzle No. 26, the licensees UT examination identified a
circumferentially oriented indication (60 - 70 degree extent) located in the
J-groove weld and which extended for 20 to 25 percent through-wall into the
penetration tube. The licensee determined that this indication was likely due to
original construction J-groove weld repair activities and was not considered a
flaw. To confirm this conclusion, the licensee performed four PT examinations of
the VHP nozzle No. 26 J-weld with intermediate buffing/grinding steps to attempt
to remove axial indications. In the final PT examination the licensee identified
two patches of flaw-like axial indications at the surface of the J-groove weld.
One area of linear indications measured approximately 1.5 inch by 0.6 inch and
the other area measured 2.5 inch by 0.6 inch. The licensee did not record the
actual size, number, or spacing of these indications. The licensee documented
their basis for not to perform additional PT examinations of other J-groove welds
in an internal memorandum dated May 13, 2004, and letter to the NRC dated
May 23, 2004.
The licensee decided to repair VHP nozzle No. 26, based upon the PT
examination results which identified linear indications in the J-groove weld. The
licensees repair technique involved removal of the lower portion of the VHP
nozzle up through the existing J-groove weld and installation of a new temper
bead weld that overlapped a portion of the existing J-groove weld. The licensee
performed this new temper bead weld repair in accordance with vendor traveler,
Ambient ID Temper Bead Repair for CRDM [Control Rod Drive Mechanism]
Nozzles, and the welding occurred in accordance with weld procedure
specification (WPS) 5S-WP3/43/ F43TBSCA301. The inspectors reviewed the
certified mill test reports for the weld filler materials, process traveler steps, weld
control records and observed portions of the machine operator repair welding to
confirm ASME Code Section III and Section IX requirements (as amended by the
50 Enclosure
licensees Code relief request) were met. Additionally, the inspectors performed
independent calculations of weld heat input for weld passes No. 1 through No. 3,
to confirm that weld heat input remained within 10 percent of that qualified in
accordance with Code Case N-638 requirements. The inspectors also reviewed
final weld UT examination records to confirm that no flaws were identified in the
VHP nozzle No. 26 repair weld.
The licensees vendor used non-structural attachment (tack) welds on the
existing J-groove weld at VHP nozzle No. 26 to mount tooling used in machining
and welding. The inspectors identified that the repair process traveler steps did
not include a PT examination following removal of this tack weld as required by
the ASME Code Section III, paragraph NB-4435. Initially, the licensee staff
considered that the existing J-groove weld was no longer part of the pressure
boundary and, therefore, did not consider the ASME Code Section III
requirements to apply. However, based upon followup discussions with the
inspectors and NRR staff, the licensee staff submitted a supplement to the relief
requests for VHP nozzle No. 26 (MR 02-018-1 and MR 02-018-2) on
May 21, 2004, to request relief and to justify this deviation from Code
requirements. By phone conference held on May 26, 2004, NRR staff granted
the licensee verbal approval to use this relief request. The inspectors
considered this violation of the ASME Code to be of minor significance, because
it involved an issue of regulatory compliance, which did not have any potential
safety significance.
7. What, if any, impediments to effective examinations, for each of the applied
methods, were identified (e.g., centering rings, insulation, thermal sleeves,
instrumentation, nozzle distortion)?
Above Head Visual Examinations
None.
Under Head PT Examination of Head Vent Line and VHP Nozzle No. 26
None.
Under Head Ultrasonic Examinations
NRC Order EA-03-009 dated February 20, 2004, required licensees to scan to
at least 1 inch below the lowest point at the toe of the J-groove weld for each
penetration and all areas with greater than 20 ksi (1,000 pounds per square inch)
tension residual and normal operating stress. For 17 VHP nozzle locations, the
licensee was not able to obtain at least a full 1 inch below the J-groove weld.
For these nozzles, the maximum extent volumetrically scanned at the tube
outside diameter below the downhill side of the weld was less than the 1 inch
due to the short length of nozzle existing below the J-groove weld and the UT
transducer configuration. Specifically, the axially aligned transducer pair used on
the blade probe resulted in a small volume of uninspected tube material at the
inside corner of these sleeved VHP nozzle locations. On conference calls with
51 Enclosure
NRR and Region-based staff held on May 6, 2004, and May 11, 2004, the
licensee discussed their intent to justify this limitation in a relaxation request to
the NRC Order EA-03-009 using a deterministic fracture mechanics approach
which assumed the uninspected area contained flaws. On May 14, 2004, the
licensee issued a letter requesting relaxation to Order EA-03-009, which
identified the 17 VHP nozzles to which this condition applied.
For VHP nozzles No. 32 and No. 33, the licensee was not able to get full
360 degree UT examination coverage with the blade UT probe due to nozzle
distortion which created an insufficient clearance gap between the thermal
sleeves and VHP nozzles. The licensee had similar inspection problems with
these locations during the last Unit 1 outage and had to replace these thermal
sleeves to allow access during the previous outage. The licensee determined
that this previous replacement work would complicate another thermal sleeve
removal and reinstallation activity which would be necessary to support additional
UT examination coverage. The extent of uninspected area below the J-groove
welds for VHP nozzles No. 32 and No. 33 was 42 degrees and 306 degrees
respectively. The licensee also identified an additional 60 degrees of
uninspected area in and above the J-groove weld for VHP nozzle No. 33. On
conference calls with NRR and Region-based staff held on May 6, 2004, and
May 11, 2004, the licensee discussed their intent to further justify this limitation in
a supplemental relaxation request to the NRC Order EA-03-009. On May 14,
2004, the licensee completed additional manual UT examinations on the lower
end of VHP nozzles No. 32 and No. 33 such that the examination coverage
required by the Order was met for VHP nozzle No. 32. On May 14, 2004, the
licensee issued a letter requesting relaxation to Order EA-03-009, for the limited
UT coverage on VHP nozzle No. 33 which included a deterministic fracture
mechanics analysis approach to support continued operation. On May 19, 2004,
the licensee elected to remove the thermal sleeve from VHP nozzle No. 33 to
permit access for the rotating UT probe to complete the examination coverage
for VHP nozzle No. 33 rather than pursue the request for Order relaxation. On
May 20, 2004, the licensee completed the rotating UT probe examination for
VHP nozzle No. 33, such that this VHP nozzle no longer required relaxation from
Order EA-03-009 requirements.
8. What was the basis for the temperatures used in the susceptibility ranking
calculation, were they plant-specific measurements, generic calculations,
(e.g., thermal hydraulic modeling, instrument uncertainties), etc.?
NRC Order EA-03-009 required licensees to calculate the susceptibility category
of each reactor head to PWSCC-related degradation. The susceptibility
category in EDY established the basis for the licensee to perform appropriate
head inspections during each refueling outage. The licensee documented the
Unit 1 RPV head EDY in calculation C11470, Reactor Vessel Head Effective
Degradation Year (EDY). In this calculation, the licensee used the formula
required by NRC Order EA-03-009 and determined the EDY for each operating
Unit. As of April 1, 2004, Unit 1 was at 15.5 EDY which placed this Unit in the
high susceptibility category. The inspectors also reviewed the examination
52 Enclosure
records from the previous Unit 1 head examinations and confirmed that no
PWSCC of VHPs had been previously identified.
NRC Order EA-03-009 also required the licensee to have used best estimate
values in determining the susceptibility category for the vessel head. The
inspectors reviewed Table 2-1 of EPRI MRP-48, PWR Materials Reliability
Program Response to NRC Bulletin 2001-01, which documented operating head
temperatures of 559 through 592 degrees Fahrenheit over the operating life of
Unit 1. The current operating head temperature was identified as 592 degrees
Fahrenheit in MRP-48 and this value had been used in the licensees
susceptibility ranking calculation. The inspectors questioned the licensee staff
as to the source of the head temperature used in MRP-48, which prompted the
licensee to document additional information obtained from their vendor. In a
memorandum to file dated April 22, 2004, the licensee documented that an
upper head bulk mean fluid temperature of 591.6 degrees Fahrenheit had been
calculated by the licensees vendor using a proprietary THRIVE computer model.
This model was used to produce a range of head temperatures based on vessel
core inlet operating temperatures. The temperature for the Point Beach Unit 1
head was determined by graphical interpolation from the THRIVE computer runs.
Therefore, the inspectors concluded that the licensee had used a combination of
plant specific information and a generic analytical model to determine operating
head temperatures for Point Beach Unit 1.
9. During non-visual examinations, was the disposition of indications consistent
with the guidance provided in Appendix D of this TI? If not, was a more
restrictive flaw evaluation guidance used?.
The inspectors determined that this question was not applicable, because the
licensee did not identify any flaws that required evaluation and return to service.
10. Did procedures exist to identify potential boric acid leaks from pressure-retaining
components above the vessel head?
Yes. The licensee performed inspections of components within containment to
identify leakage which included the area above the vessel head. This inspection
was conducted by Operations and Maintenance Department personnel during
the conduct of the reactor coolant system leakage test in accordance with
procedure 1-PT-RCS-1 Reactor Coolant System (RCS) Pressure Test-
Inside/Outside Containment Unit 1." The licensee stated that this procedure was
implemented four to five weeks prior to the outage with the plant at power to
complete an as-found leakage inspection, but the scope at this point did not
include areas above the reactor head. The licensee implemented this procedure
a second time just after plant shutdown and once again just prior to plant startup
from the refueling outage. During the two inspections with the plant shutdown,
the licensees inspection scope included areas above the reactor head. The
licensee's staff were required to document indications of boric acid or active
leakage (none were identified) on evaluation sheets of Appendix C of the Boric
Acid Leakage and Corrosion Monitoring Program. The overall division of
responsibilities and integrated actions to address boric acid leakage was
53 Enclosure
identified in NP 7.4.14, Boric Acid Leakage and Corrosion Monitoring, and the
Boric Acid Leakage and Corrosion Monitoring Program.
11. Did the licensee perform appropriate follow-on examinations for boric acid leaks
from pressure retaining components above the vessel head?
Not applicable. The licensee did not identify any instances of active boric acid
leakage from components above the Unit 1 head. The inspectors independently
reviewed data records of leakage identified during the last Unit 1 RCS leakage
tests to confirm that no indications of boric acid leakage were recorded for areas
near the reactor vessel head. Additionally, the NRC had confirmed that no
evidence of boric acid leakage had contacted the Unit 1 head during the prior
outage bare metal head examination (reference NRC inspection report 50-
266/02-13; 50-301/02-13).
c. Findings
No findings of significance were identified.
.2 Reactor Pressure Vessel Lower Head Penetration Nozzles (TI 2515/152)
a. Inspection Scope
On August 21, 2003, the NRC issued Bulletin 2003-02, Leakage from Reactor Pressure
Vessel Lower Head Penetrations and Reactor Coolant Pressure Boundary Integrity.
The purpose of this Bulletin was to: (1) Advise pressurized water reactor (PWR)
licensees that current methods of inspecting the vessel lower heads may need to be
supplemented with additional measures (e.g., bare-metal visual inspections) to detect
reactor coolant pressure boundary leakage; (2) request PWR addressees to provide the
NRC with information related to inspections that have been or will be performed to verify
the integrity of the reactor vessel lower head penetrations, and; (3) require PWR
addressees to provide a written response to the NRC in accordance with 10 CFR
50.54(f).
The objective of TI 2515/152, Reactor Pressure Vessel Lower Head Penetration
Nozzles, was to support the NRC review of licensees vessel lower head penetration
inspection activities that were implemented in response to Bulletin 2003-02. The Point
Beach licensee had committed to perform a bare metal inspection of the lower vessel
head for Unit 1 in response to the NRC Bulletin 2003-02. The inspectors performed a
review in accordance with TI 2515/152, Revision 0, of the licensees procedures,
equipment, and personnel used for reactor vessel lower head penetration examinations
to confirm that the licensee met commitments associated with Bulletin 2003-02. The
results of the inspectors review included documenting observations and conclusions in
response to the questions identified in TI 2515/152.
From April 5 through 23, 2004, in an office on the 8-foot level of the TSB (unless
otherwise stated), the inspectors reviewed activities associated with licensee inspection
of the Unit 1 lower vessel head. Specifically, the inspectors:
54 Enclosure
C performed a direct visual examination inside the Unit 1 containment from a
staging platform under the reactor vessel of the nozzle-to-head interface for
portions of each of the 36 bottom head penetrations;
C interviewed nondestructive examination personnel in the head inspection trailer
within the site protected area;
C reviewed the lower head visual inspection procedure NDE-757, Visual
Examination For Leakage of Reactor Pressure Vessel Penetrations;
C reviewed the certification records for the nondestructive examination personnel;
- reviewed the licensees procedure for certification of visual examination
personnel; and
- reviewed visual examination and evaluation of indication records.
b. Observations
Summary
Based upon a bare metal direct visual examination of the lower head, the licensee did
not identify evidence of reactor coolant system leakage near the instrument nozzle
penetrations. One quadrant of the vessel at the 270 to 360 degrees azimuth had
evidence of corrosion stains that were caused by rundown from liquid sources above the
bottom of the vessel. The licensee believed that these stains were caused by
condensed moisture corrosion of the vessel support steel. A few penetrations in this
quadrant were contacted by this rust stain, but did not result in debris/deposits in the
nozzle-to-head interface.
Evaluation of Inspection Requirements
In accordance with requirements of TI 2515/152, the inspectors evaluated and
answered the following questions:
a. For each of the examinations methods used during the outage, was the
examination:
1. Performed by qualified and knowledgeable personnel? (Briefly describe
the personnel training/qualification process used by the licensee for this
activity.)
Yes. The licensee conducted a direct visual examination of the Unit 1
lower vessel head penetration interface and lower vessel head surface
for leakage or boric acid deposits with knowledgeable staff members
certified to Level III as VT-2 examiners. One examiner was a licensee
staff member certified to licensee procedure NDE-3, Written Practice For
Qualification And Certification For NDE Personnel, and the other was a
licensee contractor certified to the contractors procedure 2-NDES-001,
55 Enclosure
Nondestructive Examination Personnel Qualification and Certification.
These qualification and certification procedures met the industry standard
ANSI/ANST CP-189, Standard for Qualification and Certification of
Nondestructive Testing Personnel. Additionally, the VT-2 examination
personnel had reviewed photographs of the boric acid deposits indicative
of penetration leakage found at the South Texas Nuclear Power Plant.
2. Performed in accordance with demonstrated procedures?
Yes. The licensee performed a bare metal inspection of the lower head
in accordance with procedure NDE-757, Visual Examination For
Leakage of Reactor Pressure Vessel Penetrations. The licensee
considered this procedure to be demonstrated because their examination
personnel could resolve the lower case alpha numeric characters 0.158
inches in height at a maximum of 6 feet under existing lighting to meet
the Code VT-2 inspection criterion.
The inspectors identified lack of procedure guidance which could
potentially impact the quality/effectiveness of the inspection. Specifically,
the procedure did not provide:
C guidance for when and how to collect samples of deposits if any
had been identified near the interface of lower head penetrations.
Further, no procedure guidance existed to identify what analysis
would be performed to determine the source of deposits identified.
Instead, the licensee staff stated that they would follow a Bottom
Mounted Instrument Inspection Decision Tree Diagram to make
decisions on sampling of deposits on the lower head;
C guidance or threshold for identification and documentation of
corrosion or wastage (e.g,. 1 percent or 10 percent wastage etc.).
Note that the licensee and NRC inspectors did not identify any
significant corrosion or wastage in the visual examinations of the
vessel head; or
C useful orientation and penetration numbering figure/schematic for
the bottom mounted instrument (BMI) penetrations. Specifically,
the procedure used a top down schematic vice a bottom up
picture (actual view that the licensees visual examiners were
presented with) and the BMI numbers marked by examination
personnel did not match the designated numbers on vendor
drawings. The licensee had physically marked each penetration
with numbers (1 through 36) to assist in the lower head
examination.
The inspectors performed an independent direct bare metal visual
examinations for most of the 36 lower head penetration nozzles from the
platform under the vessel head used by licensee's inspection staff. The
inspectors determined that each penetration was readily accessible such
56 Enclosure
that the licensee's inspection staff were able to conduct the visual
examination from within a few inches of each penetration location.
Additionally, the inspectors reviewed a sample of licensee photographs
taken at each penetration nozzle. Based upon this inspection and
interviews with the licensee's inspection staff, the inspectors did not
identify any concerns associated with implementation of the visual
inspection procedure for the lower head.
3. Able to identify, disposition, and resolve deficiencies?
Yes. The lower vessel head at the 270 to 360 degree (south) quadrant
contained corrosion stains in a pattern that suggested a flow of liquid had
run down from a source above. This flow pattern impacted several lower
head penetrations. In most cases this flow pattern did not reach the BMI
head-to-nozzle interface because of a raised metal pad that extended for
several inches around the surface of the lower vessel head at each
penetration. Based upon the visual examination, the licensee did not
identify any penetration nozzles with deposits at the nozzle-to-head
interface, indicative of boric acid leakage.
4. Capable of identifying pressure boundary leakage as described in the
bulletin and/or vessel lower head corrosion?
Yes. The inspectors performed a direct visual inspection of portions of
the 36 lower BMI penetration nozzles. Based on this examination, and
interviews with licensee examiners, the inspectors concluded that the
visual examination was capable of detecting deposits indicative of
pressure boundary leakage and head corrosion as described in the
bulletin.
b. Could small boric acid deposits representing reactor coolant system leakage as
described in the Bulletin 2003-02, be identified and characterized, if present by
the visual examination method used?
Yes. If small boric acid deposits characteristic/indicative of leakage had existed,
the inspectors concluded that the licensees examination would have identified
these. However, the licensee did not identify any boric acid deposits indicative of
leakage.
c. How was the visual inspection conducted (e.g., with video camera or direct
visual by examination personnel).
Licensee personnel conducted a direct visual examination of each of the lower
head penetration nozzles. This examination included a bare metal visual
examination of the lower head up to the transition to the vertical vessel shell wall.
The licensee's inspection staff also reported looking for evidence of boric acid
deposits or head corrosion during this inspection.
57 Enclosure
d. How complete was the coverage (e.g., 360 degrees around the circumference of
all the nozzles)?
The licensees visual examination coverage included a 360 degree unobstructed
view of each of the 36 lower head penetration nozzles at the interface of the
vessel head. Because the lower insulation was removed, the entire lower head
was accessible to the licensee staff for the visual examination.
e. What was the physical condition of the vessel lower head (e.g., debris,
insulation, dirt, deposits from any source, physical layout, viewing obstructions)?
Did it appear that there are any boric acid deposits at the interface between the
vessel and the penetrations?
The Point Beach Unit 1 lower head was surrounded by mirror-type insulation.
The original insulation configuration conformed with the contour of the lower
vessel dome with a 3-inch gap between the vessel and insulation. Each BMI
penetration had a slight gap that varied in size and was normally covered by
metal flashing. For the Unit 1 visual examination, this insulation had been
removed to provide unobstructed access to the BMI penetrations. The licensee
intended to install a revised lower head insulation structure with a tub-type
configuration (e.g., horizontal insulation floor with vertical walls). This revised
insulation design provided for access doors in the vertical and horizontal walls to
allow access for future bare metal head inspections.
On the lower head, the inspectors observed scattered patches of what the
licensee staff believed was a corrosion-resistant coating applied to the vessel
head by the original fabrication vendor prior to installation. The remnants of this
coating did not interfere with the inspection. The lower vessel at the 270 to 360
degree quadrant contained corrosion and stains in a pattern that suggested a
flow of liquid had run down from a source above the lower head.
f. What material deficiencies (i.e., crack, corrosion, etc.) were identified that
required repair?
None. The licensee did not identify any boric acid deposits indicative of leakage
and, therefore, no repairs were required.
g. What, if any, impediments to effective examinations, for each of the applied
nondestructive examination method, were identified (e.g., insulation,
instrumentation, nozzle distortion)?
None. The direct visual examination required access to the vessel lower head
and BMI nozzle penetrations by climbing down a ladder, into the keyway (a sump
area under the vessel). This area was a confined space, a high radiation area,
and was congested by the instrument tubes and their supports. Scaffold had
been installed to support removal of the lower insulation and to allow access for
direct inspection of the BMI penetrations. With the insulation removed, each
penetration was accessible from this platform for direct visual inspection.
58 Enclosure
h. Did the licensee perform appropriate follow-on examinations for indications of
boric acid leaks from pressure-retaining components above the vessel lower
head?
The licensee did not identify indications of boric acid leakage from pressure-
retaining components above the lower head.
i. Did the licensee take any chemical samples of the deposits? What type of
chemical analysis was performed (e.g., Fourier Transform Infrared), what
constituents were looked for (e.g., boron, lithium, specific isotopes), and what
were the licensees criteria for determining any boric acid deposits were not from
RCS leakage (e.g., Li-7, ratio of specific isotopes, etc.)?
Not applicable. The licensee did not identify any boric acid deposits on the lower
head and, therefore, did not perform any chemical samples.
j. Is the licensee planning to do any cleaning of the head?
Yes. The licensee staff stated that the lower head would be cleaned with
deionized water, rags, and scotch-bright pads prior to reinstalling the lower head
insulation.
k. What are the licensees conclusions regarding the origin of any deposits present
and what is the licensees rationale for the conclusions?
The licensee did not identify any deposits on the Unit 1 lower head. The
inspectors questioned the licensee staff as to the source of the corrosion stains
at the 270 to 360 degree quadrant on the head in a pattern that suggested a flow
of liquid had run down from a source above the lower head. The licensee staff
stated they believed that this flow pattern was the result of condensed moisture
which had run down the side of the vessel from corrosion occurring on the vessel
support steel. The licensee had not been able to visually confirm the source of
these rust contrails due to the narrow gap between the vessel wall and mirror
insulation.
In July of 2003, the licensee identified boric acid deposits at the lower head
insulation seams and where the BMI tubes penetrated the insulation (reference
CAP034123). The licensee concluded that the leak source for these deposits
was the sand box covers or top hat covers in the refueling cavity (e.g., refueling
water seal leakage) and that this leakage would not likely contact the vessel.
The licensee had chemically tested the boric acid found on the lower head
insulation seams and based on the absence of lithium confirmed that source of
boric acid deposits was not reactor coolant leakage.
.3 (Closed) Unresolved Item URI 50-266/03-09-01: On September 16, 2003, the
licensees vendor identified that, during the Unit 1 vessel head UT inspection completed
in September 2002, the rotating UT probe head stalled due to coupling slippage which
resulted in partial data acquisition in 10 of the 19 VHP nozzles (reference Framatome
NCR 6028873-Lack of UT Coverage During U1 Refueling Outage No. 27 Head
59 Enclosure
Inspection). The licensee documented this issue in the corrective action system as
corrective action CA053202 and condition evaluation CE012362. The licensees vendor
implemented corrective actions, which included a redesigned coupling on the rotating
UT probe and use of backup analysts to prevent recurrence prior to using this tool
during the Unit 2 VHP examinations. Additionally, the licensee performed an analysis of
the coverage limitations and determined that there was sufficient Unit 1 data for the
examination results to remain valid. The licensee subsequently performed UT of the
affected VHP nozzles during U1R28 and no flaws were identified by UT. The inspectors
did not identify any violations of NRC requirements for this issue and this URI is
considered closed.
c. Findings
No findings of significance were identified.
.4 Reactor Containment Sump Blockage (NRC Bulletin 2003-01, Point Beach Units 1 & 2)
a. Inspection Scope
The inspectors performed a preliminary review of licensee activities in response to NRC
Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump
Recirculation at Pressurized Water Reactors (PWRs)," in accordance with NRC TI
2515/153, "Reactor Containment Sump Blockage (NRC Bulletin 2003-01)," dated
October 3, 2003. The inspectors reviewed the licensees completed and proposed
compensatory measures submitted in accordance with Bulletin 2003-01, Option 2, to
verify they have been implemented or were planned and scheduled for implementation
consistent with the licensees response.
Visual inspections of the containment sumps, sump screens and flow paths were
performed by the inspectors during the refueling outage. The inspectors also walked
down the Units 1and 2 containments to verify that the condition of the containment
coatings, piping insulation, post Loss-of-Coolant-Accident (LOCA) drainage paths, and
Emergency Core Cooling System (ECCS) recirculation sumps were consistent with
condition reported in station documents. The inspectors interviewed operating and
engineering personnel and reviewed training records, procedures for foreign material
control and containment inspection, and the results of containment coating and debris
generation inspections.
b. Findings
b.1 TI Reporting Requirements
No findings of significance were identified relative to the potential impact of debris
blockage on emergency sump recirculation. The following information is provided as
called for in the Reporting Requirements section of TI 2515/153.
60 Enclosure
During this inspection period Point Beach Unit 2 had completed an October 2003
refueling outage (Unit 2 refueling outage number 26) and subsequently returned to
power. During that refueling outage a containment walkdown to quantify potential debris
sources was conducted by the licensee. The walkdown conducted on Unit 2 checked
for gaps in the sump screen and for major obstructions in the containment upstream of
the sump.
During this inspection period Point Beach Unit 1 was in a refueling outage (Unit 1
refueling outage number 28). A containment walkdown to quantify potential debris
sources was conducted by the licensee during the refueling outage. The walkdown
conducted on Unit 1 checked for gaps in the sump screen and for major obstructions in
the containment upstream of the sump.
The licensee is making advance preparations for the installation of new sump strainers,
should it be found necessary. The preparations consist of evaluating the conceptual
design of new sump strainers, including calculation of ECCS net positive suction head
following debris accumulation, and an evaluation of downstream effects from suction
ingestion of debris.
b.2 Failure to Control Unit 1 Emergency Operating Procedure (EOP) Sub-Steps
Introduction: The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion
VI, "Document Control," associated with Unit 1 emergency operating procedures when a
software error, that deleted two of five indications intended to monitor primary
containment sump performance during the recirculation phase of a design basis
accident, went undetected for a period of approximately nine months.
Description: While reviewing the licensees 60-day response to Bulletin 2003-01, the
inspectors noted that the licensee had revised EOPs to include indications to be
monitored for containment sump performance and the required actions to be taken if
sump blockage developed. Specifically, the licensee revised EOPs to monitor:
(1) containment sump level, (2) RHR Pump Operation - NORMAL, (3) SI Pump
Operation - NORMAL, (4) Low Head Injection Flow - STABLE, and (5) High Head
Injection Flow - STABLE. The licensee included these indications in the Unit 1 and 2
EOPs by adding Step 29c to EOP 1, "Loss of Reactor or Secondary Coolant,"
Revision 34.
While reviewing EOP-1, Revision 35, however, the inspectors discovered that only three
of the five parameters were described in EOP-1, Step 29c, for Unit 1 while the
corresponding Unit 2 EOP continued to list all five indications. Specifically, the
inspectors identified that the RHR Pump Operation - NORMAL and SI Pump Operation -
NORMAL sub-steps were missing from Unit 1, EOP-1, Step 29c, Revision 35.
The Point Beach main control room boards contain RHR discharge pressure, SI pump
discharge pressure, and SI pump motor amperage loading meters, parameters that
would provide indications of loss of containment sump recirculation capabilities and be
used to evaluate normal RHR and SI pump operations. The difference between the
Unit 1 and 2 EOP procedures was determined to have been caused by the applicability
function within the software program for the affected substeps having been turned off.
61 Enclosure
Since this was a known problem and had been previously documented in the corrective
action program during September 2003, the Operations department expectation was to
perform a line-by-line review of EOP revisions to ensure all applicable steps and
substeps were included prior to delivering the procedure to document control personnel
for distribution. In the case of EOP-1, Revision 35, Step 29c, for Unit 1, however, the
operations review failed to identify that the substeps associated with RHR Pump
Operation - NORMAL and SI Pump Operation - NORMAL had been deleted. In
response to the issue, the licensee issued temporary procedure change 2004-0643 to
replace the missing substeps in Unit 1, EOP-1, step 29c.
As a result of the inspectors finding, the licensee performed an extent-of-condition
review of all other EOPs, emergency contingency action (ECA), critical safety procedure
(CSP), and shutdown emergency procedures (SEP) for Units 1 and 2. Fifteen other
cases where the software program had deleted safety-related procedure steps without
operations review having identified the error prior to distribution were identified. While
14 of the errors were level-of-detail discrepancies involving, for example, missing valve
numbers, one procedure had a missing response-not-obtained (RNO) column step.
Specifically, Unit 2 safety-related ECA procedure 3.2, "SGTR [Steam Generator Tube
Rupture] With Loss of Reactor Coolant - Saturated Recovery Desired," Revision 28,
Step 17b, RNO was missing the operator actions for RCS subcooling being less than
desired with RCS temperature greater than 350 degrees Fahrenheit. The inspectors
determined that the result of the missing RNO step would be a delay in recovery actions
and the potential unnecessary cycling of an SI pump. The licensee issued temporary
procedure change 2004-0648 to replace the missing action on the same afternoon the
Unit 2 ECA 3.2 error was identified. The inspectors determined that none of the errors
would have impacted the ability to perform an intended safety function.
Analysis: The undetected software error allowed safety-related EOP, ECA, CSP, and
SEP procedures to be issued with unauthorized changes in that the deleted substeps
were not reviewed by the Plant Operating Review Committee. The inspectors
determined that issuing the procedures for operator use while not being aware of
deleted steps during the document revision process was a performance deficiency
warranting a significance evaluation in accordance with IMC 0612, "Power Reactor
Inspection Reports," Appendix B, "Issue Disposition Screening," issued on June 20,
2003. The inspectors determined that the finding was more than minor because it
affected the procedure quality attribute of the Mitigating Systems cornerstone objective
of ensuring the availability, reliability, and capability of systems that respond to initiating
events. The inspectors determined that the issue also affected the cross-cutting area of
human performance. Despite the problem with the software being previously known and
operations expectations to perform line-by-line reviews to ensure all applicable steps
were in the procedures prior to distribution, 16 EOP, ECA, CSP, and SEP procedure
errors occurred.
The inspectors completed a significance determination of this issue using IMC 0609,
"Significance Determination Process," dated March 21, 2003, Appendix A, "Significance
Determination of Reactor Inspection Findings for At-Power Situations," dated
March 18, 2002. The inspectors determined that the finding did not result in a design or
qualification deficiency, an actual loss of safety function, or involve internal or external
62 Enclosure
initiating events. Therefore, the finding was considered to be of very low safety
significance (Green).
Enforcement: Appendix B, Criterion VI, of 10 CFR Part 50, "Document Control,"
requires, in part, that measures shall assure that documents, including changes, are
reviewed for adequacy, approved for release, and are distributed to the location where
the prescribed activity is performed. Contrary to this requirement, between October 3,
2003, and July 9, 2004, two sub-steps associated with monitoring primary containment
sump performance during the recirculation phase of a design basis accident were
deleted by a software error. Specifically, the inspectors identified that, although
approved by the plant operating review committee and intended for distribution, the RHR
Pump Operation - NORMAL and SI Pump Operation - NORMAL portions of Unit 1
emergency operating procedure EOP-1, "Loss of Reactor or Secondary Coolant," Step
29c, Revision 35, were deleted by the software program and not detected by operations
personnel.
This violation was entered into the licensee's corrective action system as CAP05785,
"Vendor Program Applicability Problems Deleted Monitored Parameters for Sump
Blockage." Because this violation was of very low safety significance and it was entered
into the licensee s CAP, this violation is being treated as an NCV consistent with
Section VI.A. of the NRC Enforcement Policy. (NCV 05000266/2004003-06) This issue
and the results of the licensees extent-of-condition review that identified 15 additional
errors did not represent an immediate safety concern.
.5 Spent Fuel Material Control and Accounting at Nuclear Power Plants (TI 2515/154)
a. Scope
The inspectors completed Phase I and Phase II of the subject TI and provided the
appropriate documentation to NRC management as required by the TI.
b. Findings
No findings of significance were identified.
.6 Offsite Power System Operational Readiness (TI 2515/156)
a. Scope
The inspectors reviewed licensee event reports, CAPs, procedures, and other
documents; interviewed engineering, operations, and other personnel; and walked down
pertinent equipment in the main control room and in the switchyard to collect data
necessary to complete TI 2515/156, Offsite Power System Operational Readiness,
dated April 29, 2004. This review was conducted to confirm the operational readiness of
the offsite power systems in accordance with NRC requirements, such as Appendix A to
10 CFR Part 50, General Design Criterion 17, or similar requirements; Criterion XVI of
Appendix B to 10 CFR Part 50, Point Beach TSs for offsite power systems; 10 CFR
50.63, Loss of All Alternating Current; 10 CFR 50.65(a)(4), Requirements for
Monitoring the Effectiveness of Maintenance at Nuclear Power Plants; and licensee
63 Enclosure
procedures. Specifically, the inspectors reviewed the licensee's procedures and
processes for ensuring that grid reliability conditions were appropriately assessed during
periods of maintenance in accordance with 10 CFR 50.65(a)(4). The inspectors also
assessed the reliability and grid performance through a review of historical and current
data to verify compliance with the station blackout rule 10 CFR 50.63, TSs, and General
Design Criterion 17 (Point Beach was a pre-Appendix A plant, but had a similar
requirement). Finally, the inspectors assessed the licensee's implementation of
operating experience that was applicable to the site, as well as CAPs, to ensure issues
were being identified at an appropriate threshold, assessed for significance, and then
appropriately dispositioned. Documents reviewed for this TI are listed in the attachment
to this inspection report.
b. Findings
No findings of significance were identified. Based on the inspection, no immediate
operability issues were identified and Point Beach was operationally ready for the
summer of 2004 regarding the offsite grid. In accordance with TI 2515/156 reporting
requirements, the inspectors provided the required data in the TI work sheets to the
NRC headquarters staff for further analysis. A summary of the responses of Point
Beach to the questions of the TI is provided below.
1. What is the nature of the agreement between Point Beach and the regional
transmission organization?
The agreement is a formal contract, Generation - Transmission Interconnection
Agreement between American Transmission Company, LLC [ATC] as
Transmission Provider and Wisconsin Electric Power Company for the Point
Beach Plant Interconnection Facilities, dated November 1, 2000.
2. What voltage information has been transmitted by Point Beach to ATC?
The agreement incorporated several Point Beach procedures that specified
offsite power 345-KV bus limits and timeliness of notification by ATC to Point
Beach. The agreement included the operating voltage, shutdown voltage, and
post-trip load. The operating and shutdown voltage range specified in the
procedure was normal: 351 - 358 KVs; preferred: 352 - 354 KVs; and absolute,
for operability of 345-KV: 348.5 - 362 KVs. The range was such that initiation of
the degraded voltage relay timer was avoided. American Transmission
Company automatically calculated post-trip voltages every 5 minutes,
automatically following the trip of any transmission breaker (equal to or higher
than 115 KVs), as requested by Point Beach personnel, and on ATCs own
initiative.
3. What type of communication exists between Point Beach and ATC?
In the event of a problem on the grid, ATC would call the Point Beach control
room via a direct telephone line within 15 minutes, per a procedure incorporated
into the contract, but according to an ATC representative, in practice, the call
would be immediately.
64 Enclosure
4. How are current grid conditions factored into the licensees maintenance rule
activities?
Prior to testing or extensive maintenance on the EDGs, the licensee assesses
current grid conditions but does not address potential post-contingency grid
conditions. Included in this assessment was a review of local weather
conditions: there was no seasonal restriction on EDG testing but it was not
conducted during severe weather. The licensee also notified ATC of testing and
extensive maintenance of the EDGs. The licensee accounted for a loss of offsite
power (LOOP) for the 10 CFR 50.65(a)(4) pre-maintenance review but did not
always assume it would occur. The licensees risk modeling did not credit any
recovery of offsite power.
Risk management actions were not required but were available for use by the
operating crew. Many of these actions were listed in Attachment D of Procedure
NP 10.3.7, On-Line Safety Assessment, Revision 8, which included an action to
minimize work in the switchyard. A similar consideration was provided in
procedure NP 10.3.6, Outage Safety Review and Safety Assessment, Revision
11. Access to the switchyard was jointly controlled by security and the control
room. The stations work control and system engineering groups coordinated
switchyard work activities. Station procedure NP 2.1.5, Electrical
Communications, Switchyard Access and Work Planning, Revision 3,
specifically governed switchyard activities.
The licensees offsite power system (the 345-KV system) was scoped in the
Maintenance Rule as risk significant. The boundaries included the Unit 1 and
Unit 2 main (output) step-up transformers (1X01 and 2X01) on the side of the
switchyard closest to the plant, the five breakers associated with the four offsite
power lines coming in on the side of the switchyard farthest from the plant, and
various bus sections, breakers, motorized disconnect switches, and circuit
switches in-between. As part of the licensees review of this TI, it was identified
that the basis for the five breakers on the four offsite lines being in or
out-of-scope had not been documented. This issue was entered into the
corrective action program as CAP056406, Boundary of 345 KV System for
Maintenance Rule Not Documented.
5. How does the LOOP information in the TI compare to previous station blackout
(10 CFR 50.63) information?
The information listed in the TI for the six LOOP events at Point Beach in the last
20 years was accurate except for some minor corrections and clarifications. No
conflict with the licensees station blackout information was identified.
6. Has the licensee entered the August 14, 2003, eastern United States and
Canada grid disturbance into its corrective action system?
Yes. The licensee evaluated the event as industry operating experience after it
was entered into the corrective action program on December 8, 2003. The event
was tracked as CAP052189, SEN 242 Loss-of-Grid Event, August 14, 2003.
65 Enclosure
The licensee evaluated the event from the perspective of its impact on the
emergency response facilities and organization, such as availability of flashlight
batteries and backup power for computers, ventilation, and other equipment
associated with the emergency operations facility. Corrective actions as
appropriate were taken.
4OA6 Meetings
.1 Exit Meeting
On July 16, 2004, the resident inspectors presented the inspection results to Mr. Dennis
L. Koehl and members of his staff, who acknowledged the findings. The licensee did
not identify any information, provided to or reviewed by the inspectors, as proprietary.
.2 Interim Exit Meetings
Interim exits were conducted for:
- an Emergency Preparedness inspection with Ms. M. Ray on April 13, 2004.
- a Heat Sink inspection with Mr. J. Connolly on May 16, 2004.
- an Occupational Radiation Safety ALARA and access control programs
inspection with Mr. G. Van Middlesworth on April 23, 2004. A re-exit to discuss
the finding relative to bubble hood use was conducted telephonically on
July 9, 2004, with Mr. J. Shaw.
- TI 2515/150, TI 2515/152, and the ISI procedure (IP 71111.08) inspections with
Mr. J. Shaw on April 23, April 28, and May 26, 2004. The licensee confirmed
that none of the potential report input discussed was considered proprietary.
4OA7 Licensee-Identified Violation
The following violation of very low significance (Green) was identified by the licensee
and is a violation of NRC requirements which met the criteria of Section VI of the NRC
Enforcement Manual, NUREG-1600, for being dispositioned as an NCV.
10 CFR 26.20(e) requires, in part, that procedures ensure that persons called in to
perform an unscheduled working tour are fit to perform the task assigned. In addition,
the procedure must (1) require a statement to be made by a called-in person as to
whether he or she has consumed alcohol within the length of time stated in the pre-duty
abstinence policy, and (2) address other factors that could affect fitness-for-duty. On
April 7, 2004, the licensee identified that, contrary to this requirement, the automatic
call-in system did not ensure that the appropriate statements were made regarding
fitness-for-duty for emergency responders. The licensee entered the condition in the
corrective action program as CAP055425, Callout Process Does Not Meet
Requirements of 10CFR26.20(e)(1).
66 Enclosure
ATTACHMENT: SUPPLEMENTAL INFORMATION
67 Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
J. Brander, Maintenance Manager
B. Carberry, Radiation Protection - ALARA
G. Casadonte, Fire Protection Coordinator
J. Connolly, Regulatory Affairs Manager
B. Dungan, Operations Manager
C. Hill, Assistant Operations Manager
M. Holzmann, Nuclear Oversight Manager
P. Holzman, Heat Exchanger Engineer
R. Hopkins, Internal Assessment Supervisor
B. Jensen, Level III
C. Jilek, Maintenance Rule Coordinator
B. Kemp, Reactor Vessel Head Engineer
T. Kendall, Program Engineering
D. Koehl, Site Vice-President (June 2004 to end of inspection period)
B. Kopetsky, Security Coordinator
C. Krause, Senior Regulatory Compliance Engineer
R. Ladd, Fire Protection Engineer
K. Locke, Regulatory Specialist
J. McCarthy, Site Director of Operations
R. Milner, Business Planning Manager
T. Petrowsky, Design Engineer Manager
M. Ray, Emergency Preparedness Manager
C. Richardson, Design Engineer
P. Russell, Site Assessment Manager
D. Schoon, Production Planning Manager
J. Schroeder, Service Water System Engineer
M. Schug, Assistant Operations Manager
J. Schweitzer, Site Engineering Director
J. Shaw, Plant Manager
G. Sherwood, Engineering Programs Manager
C. Sizemore, Training Manager
P. Smith, Operations Training Supervisor
J. Strharsky, Planning and Scheduling Manager
S. Thomas, Radiation Protection Manager
R. Turner, Inservice Inspection Coordinator
G. Van Middlesworth, Site Vice-President (until June 2004)
K. Zastrow, Root Cause Assessment Coordinator, Kewaunee Nuclear Plant
Nuclear Regulatory Commission
H. Chernoff, Point Beach Project Manager, NRR
P. Louden, Chief, Reactor Projects, Branch 7
1 Attachment
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
05000266/2004003-01 NCV Loss of Transient Combustible Control in the
Containment and Turbine Buildings During a Unit 1
Refueling Outage (Section 1R05)05000266/2004003-02 NCV Substitution of Weld Surface Examinations for
Volumetric Examinations (Section 1R08)05000266/2004003-03 NCV Failure to Follow Procedures in the Issuance and Use
of Bubble Hood-type Respiratory Protective Devices
(Section 2OS1.2)05000266/2004003-04 FIN Potential Loss of Hot Leg Vent Path During Nozzle
Dam Installation (Section 4OA2.5)05000301/2004003-05 NCV Safety Injection System Accumulator Operated With
Fluid Level Above Technical Specification
Surveillance Requirement Limits (Section 4OA3.1)05000266/2004003-06 NCV Failure to Control Unit 1 Emergency Operating
Procedure Sub-Steps Committed to as Compensatory
Measures in Accordance With NRC Bulletin 2003-01,
Option 2 (Section 4OA5.4)05000301/2004001-00 LER SI System Accumulator Operated With Fluid Level
Out Of Specification High (Section 4OA3.1)05000266/2004001-00 LER Reactor Pressure Vessel Head Penetration 26 Flaw
Indications (Section 4OA3.2)
Closed
05000266/2004003-01 NCV Loss of Transient Combustible Control in the
Containment and Turbine Buildings During a Unit 1
Refueling Outage (Section 1R05)05000266/2004003-02 NCV Substitution of Weld Surface Examinations for
Volumetric Examinations (Section 1R08)05000266/2004003-03 NCV Failure to Follow Procedures in the Issuance and Use
of Bubble Hood-type Respiratory Protective Devices
(Section 2OS1.2)05000266/2004003-04 FIN Potential Loss of Hot Leg Vent Path During Nozzle
Dam Installation (Section 4OA2.5)
2 Attachment
05000301/2004003-05 NCV Safety Injection System Accumulator Operated With
Fluid Level Above Technical Specification
Surveillance Requirement Limits (Section 4OA3.1)05000266/2004003-06 NCV Unit 1 Emergency Operating Procedure Sub-Steps for
Containment Sump (Section 4OA5.4)05000301/2004001-00 LER SI System Accumulator Operated With Fluid Level
Out Of Specification High (Section 4OA3.1)05000266/2004001-00 LER Reactor Pressure Vessel Head Penetration 26 Flaw
Indications (Section 4OA3.2)05000266/2003009-01 URI Partial Data Acquisition Due To Coupling Slippage
(Section 4OA5.3)
Discussed
None.
3 Attachment
LIST OF DOCUMENTS REVIEWED
1R01 Adverse Weather
AOP-13C; Severe Weather Conditions; Unit 0, Revision 14
CAP032792; Entered AOP-13C Due to High Winds; dated May 11, 2003
CAP033403; Emergency Plan Implementing Procedure 1.1 Course of Actions Entered
Due to AOP 13C Severe Weather; dated June 8, 2003
CAP057358; Training Needed to Support Station Readiness for Severe Weather
Conditions; dated June 11, 2004
Periodic Check (PC) 49 Part 6; Securing From Cold Weather; Unit 0, Revision 13
1R04 Equipment Alignment
OP 71; Placing RHR System in Operation; Revision 42
CAP055691; Reactor Vessel Level Requirements per GL 88-17; dated April 14, 2004
OP 4F; RCS Reduced Inventory Requirements; Revision 19
OP 4D; Draining the RCS; Revision 62
Tag-Out Sheet; 1 OP-4D Part 1 RVLIS Ops; Revision 1-1
Westinghouse Drawing 541F091 Sheet 2, RCS P&ID [Pipe and Instrument Diagram]
1R05 Fire Protection
Fire Hazard Analysis Report (FHAR), Fire Zone 505; Containment - Unit 1 - 8 ft.;
January 2003
FHAR, Fire Zone 511; Containment - Unit 1 - 21 ft.; January 2003
FHAR, Fire Zone 516; Containment - Unit 1 - 46 ft.; January 2003
FHAR, Fire Zone 520; Containment - Unit 1 - 66 ft.; January 2003
FHAR, Fire Zone 263; South Service Building - First Floor; January 2003
Fire Protection Evaluation Report; Revision 2
FHAR; Revision 1
Point Beach Nuclear Plant - Fire Area Analysis Summary Report, January 2003
4 Attachment
NP 1.9.6; Plant Cleanliness and Storage; Revision 11
NP 1.9.9, Transient Combustible Control; Revision 6
CAP056427; NRC Questions on Combustible Material Control in Containment; dated
May 5, 2004
CAP057055; Concerns Over Management of Transient Combustible Loading During
U1R28; dated May 27, 2004
FHAR, Fire Zone 304, Fire Area A23, AFW Pump Room; January 2003
Calculation M-09334-357-HE.1, Appendix K, Effects of Fire Wall Addition in AFW Pump
Rooms on HELB [High Energy Line Break] Results; Revision 2
CAP054382; OM [Operations Manual] 3.27 Fire Rounds Not Initiated @ 30 days as
Required by Procedure; dated March 2, 2004
CAP051647; Non-Conservatism in Design Methodology For Loss Of RCP [Reactor
Coolant Pump] Seal Cooling; dated November 6, 2003
CA053746; Non-Conservatism in Design Methodology For Loss Of RCP Seal Cooling;
dated November 10, 2003
OM 3.27; Control of Fire Protection & Appendix R Safe Shutdown Equipment;
Revision 24
1R07 Heat Sink Performance
HX-01; Condition Assessment Program; Revision 1
GL 89-13 Program Document; Revision 3
Design Basis Document (DBD) 30; Containment Heating and Ventilation; Revision 2
TS Test 33; Containment Accident Recirculation Fan-Cooler Units (Monthly) Unit 1;
Revision 25
TS Test 34; Containment Accident Recirculation Fan-Cooler Units (Monthly) Unit 2;
Revision 30
Operating Instruction (OI) 130; Performance Test of 1HX-15D1-D8 Containment Fan
Cooler Unit 1; Revision 6
OI 131; Performance Test of 2HX-15D1-D8 Containment Fan Cooler Unit 2; Revision 7
CAP055143; HX Program Document Update Issues; dated March 26, 2004
5 Attachment
Calculations
Calculation 98-0051; SW System HX-55 A/B Flow Requirements; Revision 2
Calculation N-94-059; CCW, HX-012A-D, SW Flow Versus Temperature Requirements;
Revision 1
Calculation 2003-007; CCW Tube Plugging and Stabilization Criteria; Revision 0
Calculation 2003-008; CCW HX Plugging Limit; Revision 1
Calculation 2003-0037; Diesel Cooler Lakegrass Fouling Acceptance Criteria; dated
September 5, 2003
Calculation PGT-2000-1382; Point Beach Nuclear Plant CCW HXs HX-012C and
HX-012D Thermal Performance Test Data Evaluation and Uncertainty Analysis;
Revision 1
Calculation PGT-2002-1270; Point Beach Nuclear Plant CCW HXs HX-012C and
HX-012D Thermal Performance Test Data Evaluation and Uncertainty Analysis;
Revision 0
Drawings
Drawing Number D-9643; CCW HX; Atlas Industrial Manufacturing Co.; Revision 4
Drawing Number D-322730; HX Equip. #1 and 2, HX-55A1 and B1; Young Radiator Co.;
dated May 3, 1990
Condition Reports Reviewed During the Inspection
CAP011404; Significant Amount of Silt in Seal and Baseplate Leakage - SW Pumps;
dated January 25, 2000
CAP001861; Diesel Cooler Fouling; dated January 14, 2002
CAP028437; G-01 Diesel Cooler Zebra Mussel and Lake Weed Fouling; dated
June 11, 2002
CAP029092; G-02 Diesel Cooler Fouling; dated August 20, 2002
CAP030353; Continuing G0-2 Diesel Cooler Fouling; dated December 9, 2002
CAP030499; Major G0-1 Diesel Cooler Fouling; dated December 19, 2002
CAP033365; G0-2 Diesel Cooler Fouling; dated June 6, 2003
CAP033890; G0-2 Diesel Cooler Fouling; dated July 2, 2003
6 Attachment
CAP034365; G0-2 Diesel Cooler Fouling; dated July 28, 2003
CAP0500040; G0-1 Diesel Cooler Fouling and G0-2 Concerns; dated
September 10, 2003
CAP050119; G0-2 Diesel Cooler Fouling - Post Operability Determination Required;
dated September 11, 2003
CAP051874; Significant G0-2 Diesel Cooler Fouling Past Operability Determination
Required; dated November 17, 2003
CAP051944; G0-1 Diesel Cooler Fouling; dated November 20, 2003
CAP052753; G02 EDG H/X-055B-1 & HX-055B-2 Inspection Results; dated
January 12, 2004
CAP053209; Jan 04 G0-1 Diesel Cooler Fouling; dated January 26, 2004
CAP053569; G0-1 Diesel Cooler Fouling; dated February 5, 2004
CAP053900; Feb 13 G0-1 Diesel Cooler Fouling; dated February 16, 2004
CAP054615; March G0-2 Diesel Cooler Fouling; dated March 9, 2004
CAP054789; Mar 14 G0-1 Diesel Cooler Fouling; dated March 15, 2004
CAP055100; Mar 25 G0-1 Diesel Cooler Fouling; dated March 25, 2004
CAP055182; Mar 29 G0-2 Diesel Cooler Fouling; dated March 29, 2004
CAP055905; April 19 G0-1 Diesel Cooler Fouling; dated April 20, 2004
CAP056354; May 3 G0-2 Diesel Cooler Fouling; dated May 3, 2004
CAP056853; May 18 G0-1 Diesel Cooler Fouling; dated May 20, 2004
CAP057186; June 3 G0-2 Diesel Cooler Fouling; dated June 3, 2004
Condition Reports Written as a Result of the Inspection
CAP057330; GL 89-13 2002 Annual Report Never Written; dated June 10, 2004
CAP057403; Revise GL 89-13 2002 Program Document; dated June 15, 2004
CAP057406; OI-151 Procedure Compliance; dated June 15, 2004
CAP057409; FSAR Section 9.6 Contains Outdated Information on Zebra Mussel
Control; dated June 15, 2004
7 Attachment
CAP057420; ARB Actions Inconsistent With OI-70 Att. B for Low SW Flow to G01/G02
H/Xs; dated June 16, 2004
Other (OTH) 013682; Calculation 98-0051 Typographical Error; dated June 15, 2004
Condition Report CR-00-0267; Revision 1; dated March 17, 2000
MRE000147; Significant G0-2 Diesel Cooler Fouling Past Operability Determination
Required; dated December 12, 2003
Apparent Cause Evaluations (ACEs)
ACE001157; Apparent Cause Evaluation of CAP030619 and CAP030640; dated
February 7, 2003
ACE001657; K-3A Service Air Compressor SW Strainer Found Plugged With Grass;
dated April 21, 2004
Procedures
OI 70; SW System Operation; Revision 49
Alarm Response Book (ARB) C02 D 3-6; G-01 Emerg Diesel Cooler Low Flow;
Revision 5
ARB C02 F 3-1; G-02 Emerg Diesel Cooler Flow Low; Revision 9
System Operating Procedure 1-SOP-CC-001; Component Cooling System; Revision 11
Chemistry Analytical Methods and Procedures 917; Copper Ion Generator; Revision 6
HX-01; Condition Assessment Program; Revision 2; dated May 18, 2004
HX-01; Condition Assessment Program, Appendix C, Unit 1 Outage Cycle Inspection
Schedule; Revision 1; dated February 25, 2004
HX-01; Condition Assessment Program, Appendix D, Unit 2 Outage Cycle Inspection
Schedule; Revision 1; dated February 25, 2004
HX-01; Condition Assessment Program, Appendix E; Annual Cycle Inspection Schedule;
Revision 1; dated February 25, 2004
AM 3 -19; Biofouling Control Program; Revision 1
NP 7.7.15; Biofouling Control Methods; Revision 1
Diesel Generator CH01; Zebra Mussel Tracking and Evaluation; Revision 0
8 Attachment
CD 5.25; GL 89-13 Standard; Revision 0
Completed Inspections/Surveillance Procedures
Completed Bio/Silt Fouling Inspection Forms for Diesel Generator HXs
HX-055A-1 and HX-055A-2; dated various between March 18, 2003, through
June 14, 2004
Completed Bio/Silt Fouling Inspection Forms for Diesel Generator HXs
HX-055B-1 and HX-055B-2; dated various between August 20, 2002, through
June 3, 2004
Completed Miscellaneous HXs Cleaning and Inspection Program, Appendix A; HX
Internal Visual Inspection CL for HX-12C; dated August 19, 2002
Completed OI 151; HX-012C and D Component Cooling System HX Data Collection;
dated October 15, 2000
Completed OI 151; HX-012C and D Component Cooling System HX Data Collection;
dated April 14, 2002
Miscellaneous
Heat Exchanger Specification Sheet; Component Cooling HX; dated February 24, 1992
System Health Report SW; dated January 30, 2004
System Health Report SW; dated April 13, 2004
System Health Report; CCW System; dated January 8, 2004
System Health Report; Diesel Generator System; dated April 23, 2004
System Health Rating Status - CW; dated May 2004
Performance Criteria Assessments for CC since 6/1/2001; dated June 3, 2004
Performance Criteria Assessments for Diesel Generator since 6/1/2001; dated
June 3, 2004
Performance Criteria Assessments for SW since 6/1/2001; dated June 3, 2004
DBD-02; Component Cooling System Design Basis Document; Revision 4
DBD-10; RHR System Design Basis Document; Revision
DBD-12; SW System Design Basis Document; Revision 6
9 Attachment
DBD-16; EDG System Design Basis Document; Revision 4
FSAR Section 9.1; CCW; dated June 2003
FSAR Section 9.6; SW System; dated June 2002
FSAR Appendix A; Shared System Analysis; dated June 1998
FSAR Change Request FCR 04-007; dated February 3, 2004
TS 3.7.7; CCW System and Associated Bases B 3.7.7; Unit 1 - Amendment No. 201;
Unit 2 - Amendment No. 206
TS 3.7.8; SW System and Associated Bases B 3.7.8; Unit 1 - Amendment No. 201;
Unit 2 - Amendment No. 206
SW System Operational Performance Inspection; dated October 6, 1993
TIN No. 97-1177, Revision 1; Test Protocol Wisconsin Electric Power Company Point
Beach Nuclear Plant CCW HX; dated January 17, 2001
PBSA-ENG-03-02; Component Cooling Water System Self-Assessment; dated
September 8, 2003
GL 89-13 Annual Report for 2001; dated March 22, 2002
Point Beach GL 89-13 Program - 2003 Annual Report; dated March 22, 2004
Point Beach GL 89-13 Program Self-Assessment # PBSA-ENG-03-15; dated
December 23, 2003
Program Health Status; SW/Microbiologically Induced Corrosion; dated
May 21, 2004
Program Health Status; HX Safety Related NRC GL 89-13; dated
May 28, 2004
GL-89-13 Program Document; Revision 3
2003 EVAC Treatment Effectiveness Report; dated September 29, 2003
Zebra Mussel Program Effectiveness Report - Annual; dated November 13, 2003
4th Quarter 2003 SW System Maintenance Rule Summary; dated
January 15, 2004
GL 89-13 Program; 2003 SW System Engineer Report; dated February 13, 2004
10 Attachment
Response to GL 89-13, Safety Related SW Problem Point Beach Nuclear Plant; dated
January 12, 1990
Work Orders
WO9912379; Component Cooling HX-012C; dated January 5, 2001
WO0206227; Component Cooling HX-012C; dated August 28, 2002
WO9925508; G-01 EDG Coolant HX (East); dated May 29, 2001
WO9934401; G-01 EDG Coolant HX (East); dated August 1, 2001
WO9940300; G-01 EDG Coolant HX (East); dated May 25, 2001
WO9935151; G-01 EDG Coolant HX (East); dated June 20, 2002
WO0205750; G-01 EDG Coolant HX (East); dated July 3, 2002
WO0207039; G-01 EDG Coolant HX (East); dated December 26, 2002
WO0300008; G-01 EDG Coolant HX (East); dated March 24, 2003
WO0300015; G-01 EDG Coolant HX (East); dated May 28, 2003
WO0300017; G-01 EDG Coolant HX (East); dated November 11, 2003
WO0300018; G-01 EDG Coolant HX (East); dated December 6, 2003
WO0301022; G-01 EDG Coolant HX (East); dated January 14, 2003
WO0303667; G-01 EDG Coolant HX (East); dated June 18, 2003
WO0303668; G-01 EDG Coolant HX (East); dated July 21, 2003
WO0303674; G-01 EDG Coolant HX (East); dated September 11, 2003
WO0310370; G-01 EDG Coolant HX (East); dated January 11, 2004
WO0310371; G-01 EDG Coolant HX (East); dated February 21, 2004
WO9925507; G-01 EDG Coolant HX (West); dated May 29, 2001
WO9934400; G-01 EDG Coolant HX (West); dated August 1, 2001
WO9935150; G-01 EDG Coolant HX (West); dated June 20, 2002
WO9940299; G-01 EDG Coolant HX (West); dated May 25, 2001
11 Attachment
WO0205749; G-01 EDG Coolant HX (West); dated July 3, 2002
WO0207037; G-01 EDG Coolant HX (West); dated December 26, 2002
WO0300007; G-01 EDG Coolant HX (West); dated March 24, 2003
WO0300012; G-01 EDG Coolant HX (West); dated September 2, 2003
WO0300013; G-01 EDG Coolant HX (West); dated November 11, 2003
WO0300014; G-01 EDG Coolant HX (West); dated December 6, 2003
WO0301023; G-01 EDG Coolant HX (West); dated January 14, 2003
WO0303662; G-01 EDG Coolant HX (West); dated June 18, 2003
WO0310360; G-01 EDG Coolant HX (West); dated January 11, 2004
WO0310361; G-01 EDG Coolant HX (West); dated February 21, 2004
WO9926689; G-02 EDG Coolant HX (East); dated August 14, 2000
WO9926693; G-02 EDG Coolant HX (East); dated October 3, 2000
WO9927125; G-02 EDG Coolant HX (East); dated July 18, 2000
WO9927420; G-02 EDG Coolant HX (East); dated February 19, 2002
WO9934403; G-02 EDG Coolant HX (East); dated August 1, 2001
WO9936321; G-02 EDG Coolant HX (East); dated February 4, 2002
WO9938065; G-02 EDG Coolant HX (East); dated February 25, 2002
WO0205752; G-02 EDG Coolant HX (East); dated September 6, 2002
WO0207043; G-02 EDG Coolant HX (East); dated December 11, 2002
WO0216411; G-02 EDG Coolant HX (East); dated March 17, 2003
WO0216412; G-02 EDG Coolant HX (East); dated June 12, 2003
WO0216413; G-02 EDG Coolant HX (East); dated October 21, 2003
WO0216414; G-02 EDG Coolant HX (East); dated November 9, 2003
WO0301037; G-02 EDG Coolant HX (East); dated January 14, 2003
WO0303672; G-02 EDG Coolant HX (East); dated July 8, 2003
12 Attachment
WO0303673; G-02 EDG Coolant HX (East); dated July 29, 2003
WO0310390; G-02 EDG Coolant HX (East); dated February 21, 2004
WO9934402; G-02 EDG Coolant HX (West); dated August 1, 2001
WO9936367; G-02 EDG Coolant HX (West); dated February 4, 2002
WO9938063; G-02 EDG Coolant HX (West); dated February 25, 2002
WO0205751; G-02 EDG Coolant HX (West); dated January 16, 2002
WO0207041; G-02 EDG Coolant HX (West); dated December 11, 2002
WO0216406; G-02 EDG Coolant HX (West); dated March 17, 2003
WO0216407; G-02 EDG Coolant HX (West); dated June 13, 2003
WO0216408; G-02 EDG Coolant HX (West); dated October 21, 2003
WO0216409; G-02 EDG Coolant HX (West); dated November 19, 2003
WO0301038; G-02 EDG Coolant HX (West); dated January 15, 2003
WO0303669; G-02 EDG Coolant HX (West); dated July 8, 2003
WO0303670; G-02 EDG Coolant HX (West); dated July 29, 2003
WO0303852; G-02 EDG Coolant HX (West); dated April 1, 2003
WO0310380; G-02 EDG Coolant HX (West); dated February 21, 2004
WO0310381; G-02 EDG Coolant HX (West); dated May 18, 2004
WO0310384; G-02 EDG Coolant HX (West); dated May 18, 2004
1R08 Inservice Inspection Activities
Documents Associated with Two Types of Nondestructive Testing
RC-03-PS-1001-14; Primary ISI Isometric PBNP [Point beach Nuclear Plant] Unit 1
Pressurizer Spray From Loop A; Revision 2
Point Beach Nuclear Plant Ultrasonic Calibration Record; RC-03-PS-1001-14; dated
April 6, 2004
Point Beach Nuclear Plant Ultrasonic Piping Examination Record; RC-03-PS-1001-14;
dated April 6, 2004
13 Attachment
Point Beach Nuclear Plant Ultrasonic Calibration Record; RC-03-PS-1001-15; dated
April 7, 2004
Point Beach Nuclear Plant Ultrasonic Piping Examination Record; RC-03-PS-1001-15;
dated April 7, 2004
AF-03-AFW-1002; ISI Isometric Auxiliary Feedwater to Steam Generator B; Revision 1
Point Beach Nuclear Plant Ultrasonic Calibration Record; AF-03-AFW-1002-76; dated
April 27, 2004
Point Beach Nuclear Plant Ultrasonic Piping Examination Record; AF-03-AFW-1002-76;
dated April 27, 2004
Point Beach Nuclear Plant Ultrasonic Calibration Record; AF-03-AFW-1002-77; dated
April 27, 2004
Point Beach Nuclear Plant Ultrasonic Piping Examination Record; AF-03-AFW-1002-77;
dated April 27, 2004
FW-16-FW-1002; Primary ISI Isometric PBNP Unit 1 Loop B Feedwater Inside
Containment; Revision 4
Point Beach Nuclear Plant Ultrasonic Calibration Record; FW-16-FW-1002-15; dated
April 27, 2004
Point Beach Nuclear Plant Ultrasonic Piping Examination Record; FW-16-FW-1002-15;
dated April 27, 2004
EB-9-FW-H10; Pipe Hanger Support Detail; Revision 0
Point Beach Nuclear Plant Visual Examination Record; EB-9-FW-H10; dated
April 23, 2004
NDE 109; Manual Ultrasonic Examination Using Longitudinal Wave Straight Beam
Techniques; Revision 6
NDE 163; Manual Ultrasonic Examination of Ferritic Pressure Vessel Welds Greater
Than 2 Inches In Thickness; Revision 10
NDE-172; PDI Generic Procedure For The Ultrasonic Examination Of Ferritic Piping
Welds; Revision 7
NDE-173; PDI Generic Procedure For The Ultrasonic Examination Of Austenitic Piping
Welds; Revision 6
NDE-350; Magnetic Particle Examination Alternating Current AC Yoke; Revision 24
14 Attachment
NDE-451; Visible Dye Penetrant Examination Temperature Applications 45o F to 125o F;
Revision 21
NDE-753; Visual Examination (VT-2) Leakage Detection of Nuclear Power Plant
Components; Revision 10
Memorandum to G. Sherwood, DE Oakley, R. Turner from W.A. Jenson; ASME Section
XI IWA-2240 Demonstration of the Performance Demonstration Initiative Generic
Procedure As a Replacement for NDE-163 and NDE-170; dated March 19, 2003
NDE Procedure Qualification NDE-451; Visible Dye Penetrant Examination Temperature
Applications 45o F to 125o F; dated March 12, 2002
Documents Associated With Relevant Indications
Indication Disposition Report; Magnetic Particle Examination and Technique Record;
Component ID: RPV-HFLANGE-C; Component Description: Head to Flange (AZ 240-
360); dated February 15, 2000
Indication Disposition Report; Liquid Penetrant Examination Record; Component ID:
RPV; Component Description: CRDM Nozzle #1; dated October 01, 2002
Indication Disposition Report; Magnetic Particle Examination and Technique Record;
Component ID: RPV-STUD-44; Component Description: Closure Stud; dated
February 15, 2000
Documents Related to Code Pressure Boundary Welding
WO0212615; Cut Weld and Remove Pipe from SI Accumulator Nozzle at 1SI-833C for
PT Exam of Nozzle Inner Diameter; dated October 28, 2002
WO0212682; Cut Weld and Remove Pipe from B SI Accumulator Nozzle at 1SI-833B
for PT Exam of Nozzle Inner Diameter; dated January 14, 2003
2.P8-GT-SM; Welding Procedure for Austenitic Stainless Steels ASME Group P-8
GTAW- SMAW; Revision 0.
Fillet to Socket Weld Data Sheet; Component: FW-1 and FW-2 T-034B Nozzle; dated
October 3, 2002
Documents Related to Code Repairs or Replacements
WO0212615; Cut Weld and Remove Pipe from SI Accumulator Nozzle at 1SI-833C for
PT Exam of Nozzle Inner Diameter; dated October 28, 2002
WO0212682; Cut Weld and Remove Pipe from B SI Accumulator Nozzle at 1SI-833B
for PT Exam of Nozzle Inner Diameter; dated January 14, 2003
Repair Replacement Form 2002-0095; T-34B; dated September 30, 2002
15 Attachment
ASME Section XI Code Reconciliation Checklist; SI System Weld Filler Material; dated
September 30, 2002
Visual Weld Examination Record; FW-2; dated October 1, 2002
Visual Weld Examination Record; FW-1; dated October 2, 2002
Liquid Penetrant Examination Record; FW-2; dated October 1, 2002
Liquid Penetrant Examination Record; FW-1; dated October 2, 2002
ASME Section XI R/R/M Pressure Test Data Sheet; FW-1, FW-2; dated
October 13, 2002
WPS 2.P8-GT-SM; Welding Procedure For Austenitic Stainless Steels ASME Group P-8
GTAW-SMAW; Revision 0
PQR WP-2; Revision 4
Other Documents
PBNP Indication Disposition Report; IDR No. 02U1-E008; Component No. 1CH-10;
Component Description: Core Drilled Hole; dated April 28, 2001
PBNP Indication Disposition Report; IDR No. 01U1-L004; Component No. U1C;
Component Description: Unit 1 Containment; dated October 2, 2002
PBNP Fillet/Socket Weld Data Sheet; Equipment No. ISI 00853D; WO No. 0212465;
dated October 4, 2002
Drawing No. PBC-309; ISI Classification Drawing: Keyway Sump A/Tunnel; dated
August 13, 1998
Drawing No. PBC-312; ISI Classification Drawing: Electrical Penetrations; dated
September 1, 1998
SEM 7.11.2; ISI Data Sheet Review and Indication Evaluation Guideline; dated
March 19, 2004
Documents Related to Steam Generator Tube Inspection Activities
NMC-400-002; Multifrequency ET Testing of Non-Ferromagnetic Steam Generator
Tubing; Revision 2
NMC-400-004; Analysis of Rotating Eddy Current Data; Revision 3
NMC-400-003; Analysis of Bobbin Coil Eddy Current Data; Revision 3
NMC-400-007; Eddy Current Site Specific Performance Demonstration; Revision 0
16 Attachment
Point Beach Unit 1 Steam Generator Eddy Current Examination Report; dated
May 4, 2004
Memorandum from G. Sherwood (NMC) from P. Nelson (WE); dated April 30, 2004
CAP056028; Possible Loose Parts in SG; dated April 24, 2004
Steam Generator Degradation Assessment for Point Beach Unit 1 U1R28; dated
April 2004
MRS-TRC-1468; Use of Appendix H Qualified Techniques at Point Beach Unit 1 for the
Spring 2004 Steam Generator Inspection; dated April 13, 2004
Westinghouse Electric P-BOB-001; Steam Generator Eddy Current Inspection
Examination Technique Specification Sheets; dated April 9, 2004
NP 7.7.17; Requirements for Steam Generator Primary Side Activities; Revision 2
1R11 Licensed Operator Qualifications
Licensed Operator Requalification LOR 04-03, SES 114; Licensed Operator
Requalification Training Simulator Scenario 114; dated June 8, 2004, Revision 0
CAP057356; Crew Failure of a LOR Cycle 04-03 Simulator As-Found; dated
June 11, 2004
1R12 Maintenance Effectiveness
System Health Report; 125-Volt DC System; April 8, 2004
1R13 Maintenance Risk Assessment and Emergent Work Evaluation
E-1 Report; Work Week Schedule for Week of April 18, 2004
E-1 Report; Work Week Schedule for Week of April 26, 2004
E-1 Report; Work Week Schedule for Week of May 17, 2004
E-1 Report; Work Week Schedule for Week of May 31, 2004
E-1 Report; Work Week Schedule for Week of June 13, 2004
NP 10.3.6; Outage Safety Review and Safety Assessment; Revision 11
NP 10.3.7; On-Line Safety Assessment; Revision 8
CAP055974; P-38A AFW Pump Not Identified on Weekly Safety Monitor Risk Profile;
dated April 23, 2004
17 Attachment
1R14 Non-Routine Evolutions
CAP056804; Unit 2 Electro-Hydraulic Control Failure During Unit 2 Start-up; dated
May 19, 2004
1R15 Operability Evaluations
NP 5.3.7: Operability Determination (OD); Revision 10
NP 5.3.10; Operability Recommendation (OPR); Revision 1
OPR000107 (CAP054534) - Unanalyzed load on Emergency Diesel Generator
PBO G-03 and PBO G-04; Revision 0
AOP-22; Unit 1 EDG Load Management; Revision 2
CALC-WE0005-18, Diesel Fuel Oil Consumption
OI 92A; Fuel Oil Ordering, Receipt, Sampling and Offloading; Revision 11
CAP053555; Unit 2 Inadvertent Letdown Isolation; dated February 4, 2004
CAP053560, Application of Temporary Modification Procedure; dated February 5, 2004
CAP053565, Lack of Contingencies for Letdown Problems on U-2; dated
February 5, 2004
Instrument and Control Procedure (ICP) 5.21; Pressurizer Level Control; Revision 17
Westinghouse Drawing 883D195, Sheet 18, Logic Diagram - Pressurizer Pressure and
Level Control
Westinghouse Drawing 883D195, Sheet 19, Logic Diagram-Pressurizer Heater Control
Foxboro Drawing 10665 CD-13, Sheet 2, Wiring Diagram-Interconnect Reactor Control
System Rack 2-PLPP(2C110) Middle
Foxboro Drawing 10665 CD-13, Sheet 3, Wiring Diagram-Interconnect Reactor Control
System Rack 2-PLPP(2C110) Bottom
Foxboro Drawing 10665 BD-14, Block Diagram-Instrument-Reactor Control
System-Pressurizer Level Control
FP-E-TS-01; NMC Engineering Fleet Procedure - Troubleshooting Process; Revision 0
Routine Maintenance Procedure (RMP) 9201; Control and Documentation for
Troubleshooting and Repairs; Revision 0
AOP-1D; Chemical and Volume Control System Malfunction; Revision 2
18 Attachment
CAP055332; IN-31 Source Range Pre-Amp Found Defective During the Performance of
HCP 04.024; dated April 4, 2004
1ICP 04.024; Nuclear Instrumentation Source Range Channels; Revision 4
Station Logs: April 2, 2004 through April 4, 2004
CAP056170; IT-10 Acceptance Criteria Does Not Ensure Adequate AFW Flow Without
Operator Action; dated April 28, 2004
Byron-Jackson Test No. T-30944, Pump Curve for P-38A AFW Pump
Byron-Jackson Test No. T-30945, Pump Curve for P-38B AFW Pump
Point Beach Form PBF-1608, (Draft) Calculation #96-0244, Minimum Allowable IST
Acceptance Criteria for Turbine and Motor-Driven AFW Pump Performance
PBF-1621, Vendor Calculation/Evaluation Review Form, Calculation #97-114 Rev. A,
Development of Point Beach Auxiliary Feed Water System PROTO-FLO Hydraulic
Model
Proto-Power Corporation - Serial #PFL-1007 Test Data, Point Beach Nuclear Plant AFW
System, Flow with Degraded Curve
OPR000109; Motor-Driven AFW Pumps, P38A and P38B, IT-10 Acceptance Criteria
Does Not Ensure Adequate AFW Flow Without Operator Action; dated April 30, 2004
1R16 Operator Workarounds
CAP053487; ARB References Declaring Wrong Train OOS [Out-Of-Service];
February 3, 2004
ARB 1C20 A 1-5; Unit 1 and 2 Containment Hydrogen System Trouble (White)
ARB 1C20 A 2-5; Unit 1 and 2, Containment Hydrogen System Trouble (Yellow)
Point Beach Operations Procedure Change Report OPS-2001-20416; ARB 1C20 A 2-5,
Unit 1 and 2 Containment Hydrogen System Trouble (Yellow); dated
September 30, 2001
Point Beach Operations Procedure Change Report OPS-2001-00168, ARB 1C20 A 2-5,
Unit 1 and 2 Containment Hydrogen System Trouble (Yellow); dated February 3, 2004
1ICP 13.002; Containment Hydrogen Monitor Quarterly Gas Calibration; Revision 4
1R19 Post-Maintenance Testing
NF-NMC-03-190; Point Beach Unit 1, Cycle 29, Region and Fuel Assembly Locations;
November 24, 2003
19 Attachment
IT 01A; High Head SI Pumps and Valves (Cold Shutdown) Unit 1; Revision 12 with
Temporary Change Notice 2004-0418
WO0407597; MS-00228 HX-1A SG Header Drain and Trap Isolation Valve Needs
Repack; April 4, 2004
Regulatory Guide 1.141; Containment Isolation Provisions for Fluid Systems; April 1978
CAP057196; Body to Bonnet Steam Leak on 1MS-228 HX-1A SG Header Drain and
Trap Isolation; dated June 4, 2004
FSAR Section 5.2; Containment Isolation System; June 2003
FSAR Section 14.2; Steam Generator Tube Rupture; June 2002
Point Beach Nuclear Plant Containment Leakage Rate Testing Program; Basis
Document; Revision 6
TS Test 35; Local Leak Test of Containment Purge Valves Unit 1; Revision 24
IT 03; Low Head SI Pumps and Valves (Quarterly) Unit 1, Revision 45 with Temporary
Change
PBF-0026e; Temporary Change Review and Approval for IT 03; dated April 29, 2004
PBF-026m; Partial Procedure Performance; IT-03/Low Head SI
CAP056290; 1SI-897B Appears To Be Bound; dated May 1, 2004
Analysis of Return To Service Testing For 1P-10B RHR Pump; dated May 1, 2004
IT 531; Leakage Reduction and Preventive Maintenance Program Test of Containment
Sump B Suction Line Mode 5, 6, or Defueled, Unit 1; Revision 14
PBF-0026e; Temporary Change Review and Approval for IT 531; May 3, 2004
1R20 Refueling and Outage Activities
WO0407919, Remove and Repair Stuck Bullet
PBF-2114, Return to Service Testing Reviews (WO0407919)
PBF-9812, Categorization and Mitigation of Risk (WO0407919)
PBF-1958, Approval for Hard Hat Exemption (WO0407919)
PBF-9205, High Risk Work Pre-Job Briefing CL (WO0407919)
20 Attachment
CAP056617; Evaluate Need to Do Inspection Inside Vessel Prior to Setting Upper
Internals; dated May 12, 2004
CL-7A; SI System CL Unit 1; Revision 19
CAP056650; Safety Representative Was Not Present During the Internals Lift; dated
May 12, 2004
CAP056649; Missed Management Observation of High Risk Activity; dated
May 12, 2004
CAP056651; Management Expectation for Overview of High Risk Work Not Met; dated
May 12, 2004
CAP056477; Mode 5 HOLD for Reactor Head Repair; dated May 7, 2004
CAP056490; Z-14 Turbine Hall Crane Coil Failure; dated May 7, 2004
CAP056499; Turbine Building Crane Trolley Brake Transformer Smoking; dated
May 7, 2004
CAP056487; Unit 1 SFP Upender Stopped Unexpectedly; dated May 7, 2004
CAP056551; Inconsistencies in Nozzle Dam Protected Equipment Expectations; dated
May 9, 2004
CAP056336; Communication Issues Between Engineering, Projects Group, and RP;
dated May 2, 2004
CAP056296; Worn Thrust Bearings Being Re-used on 1P-1B-M; dated May 1, 2004
CAP056293; Lift Oil Flow Orifices Not Removed During Previous Mod on RCP Motor;
dated May 1, 2004
CAP055691; Reactor Vessel Level Requirements Per GL 88-17; dated April 14, 2004
CAP056387; Clarification on Management Expectations for Performing Observations
Required; dated May 4, 2004
CAP055811; Potential Adverse Trend in Operations Inattention to Detail; dated
April 17, 2004
CAP055764; Inadequate Staffing of Work Control Center for the U1 Outage; dated
April 16, 2004
CAP055738; 1RMP 9007, Precautions and Limitations Not Maintained; dated
April 15, 2004
21 Attachment
CAP055570; Un-clear Expectation on Management Briefing Before Starting Work; dated
April 11, 2004
CAP055365; Potential Unauthorized Modification on U1 Reactor Cavity Manipulator
1Z-16; dated April 5, 2004
CAP055842; Inadequate Response to Shutdown Risk Qualification Issue; dated
April 18, 2004
CAP055888; Unit 2 T-34B SI Accumulator Levels Calibrated with Wrong Tolerance
M&TE; dated April 20, 2004
CAP055957; Refueling Questions Needing Resolution; dated April 22, 2004
CAP055951; Incore Thermocouple Guide (Bullet Nose) Inadvertently Lifted with RX
Head; dated April 22, 2004
CAP055979; Extended Period of Time with No Direct RCS Level Indication in Control
Room; dated April 23, 2004
CAP055995; Refueling Questions Need Resolution; dated April 23, 2004
CAP056010; P-12A Spent Fuel Pump Breaker Found OFF - Results in Loss of Spent
Fuel Cooling; dated April 23, 2004
CAP056015; Spent Fuel Pool Cooling Safety Assessment; dated April 23, 2004
CAP055975; Refueling Cavity Level Raised Without Monitoring RCP Leakoff Collection;
dated April 23, 2004
CAP055567; Screening Was Not Obtained Until After Procedure Was Worked; dated
April 11, 2004
CAP055876; Z-013 Polar Crane Aux Hoist Fuses Not Per Drawing; dated April 19, 2004
CAP055880; Mis-communication on Return to Service of the Auxiliary Hoist on the Z-13
Crane; dated April 19, 2004
CAP055897; 1Z-013 Auxiliary Hoist Fuse Discrepancy; dated April 20, 2004
CAP055882; Unit 1 Z-13 Polar Crane Phase Firing Thyristors Parts Discrepancy; dated
April 20, 2004
CAP055915; U1 Refueling 28 Outage Schedule Conflicts; dated April 21, 2004
CAP056183; SK-37 Lead-in Guide Pulled Up; dated April 28, 2004
CAP056269; Potential Human Performance Error Trap Using RCC [Rod Cluster Control]
Change Tool Indication; dated April 30, 2004
22 Attachment
CAP056239; 1HX-001A Inspection Port Cover Bolts Torqued in Excess of 1RMP 9032
Requirement; dated April 29, 2004
CAP056236; Control Rod Contacted Top of Fuel Assembly During RCCA [rod Cluster
Control Assembly] Changing; dated April 29, 2004
CAP056159; 1B-42, Bolted Short Mod Back in Outage Schedule; dated April 28, 2004
CAP055691; Reactor Vessel Level Requirements Per GL 88-17; dated April 14, 2004
1R22 Surveillance Testing
IT-9A; Cold Start of Turbine-Driven Auxiliary Feed Pump and Valve Test, Unit 2,
Revision 34
WO0408743; MS-02019-O, HX-1B SG Header P-29 AFP [Auxiliary Feedwater Pump]
Steam Supply MOV [Motor-Operated Valve] Operator; dated June 1, 2004
RMP 9376-2; Limitorque MOV Static/DP [Differential Pressure] Testing for Gate and
Globe Valves; Revision 10
WO0408741; MS-02020-O, HX-1A SG Header P-29 AFP Steam Supply MOV Operator;
dated May 31, 2004
RMP 9376-7; MOV Troubleshooting Guide; Revision 2
CAP057117; Steam Supply Valves to 2P-29 AFW Pump Failures During IT-09A; dated
May 31, 2004
ORT 3A; SI Actuation With Loss of Engineered Safeguards AC (Train A) Unit 1;
Revision 37
CAP030028; Quarterly Vibration Readings on the RHR Pumps; dated April 30, 2003
IT 03; Low Head SI Pumps and Valves (Quarterly) Unit 1; dated February 5, 2004
IT 04; Low Head SI Pumps and Valves (Quarterly) Unit 2; dated September 8, 2003
Trend Data IT-3 Series (RHR); 1P-10A RHR Pump; dated March 30, 2004
Trend Data IT-3 Series (RHR); 1P-10B RHR Pump; dated March 30, 2004
Trend Data IT-3 Series (RHR); 2P-10A RHR Pump; dated March 30, 2004
Trend Data IT-3 Series (RHR); 2P-10B RHR Pump; dated March 30, 2004
IT 760; Flow Test of High Head SI Check Valves (Refueling) Unit 1; dated
September 12, 2002
23 Attachment
PBF-0026e; Temporary Change 2004-0321; dated April 21, 2004
PBF-0026; Temporary Change Affected Manual Location (TC#2004-0321)
PBF-1515b; 10 CFR 50.59/72.48 Pre-Screening Review, Temp Procedure Change to
IT-760 (TC#2004-0321); dated April 21, 2004
PBF-0026e; Temporary Change 2004-0322; dated April 21, 2004
PBF-0026h; Temporary Change Affected Manual Location (TC#2004-0322)
PBF-1515b, 10 CFR 50.59/72.48 Pre-Screening Review; Temp Procedure Change to
IT-760 (TC#2004-0322); dated April 21, 2004
PBF-1515c, 10CFR 50.59/72.48 Screening (New Rule) SCR 2003-0329; Steam
Generator Nozzle Dams; dated September 22, 2003
RP 1A; Preparation for Refueling; dated April 22, 2004
IT 03; Low Head SI Pumps and Valves (Quarterly) Unit 1; Revision 46
1R23 Temporary Plant Modifications
TM 04-001; Temporary Replacement Of Unit 1 Purge Supply/Return Valves; Revision 0
OI 58; Leak Testing of Containment Isolation Valves - Unit 1 and 2 General Instructions
and Information; Revision 22
TS Bases 3.9.3 Change; TS Basis B 3.9.3 - Containment Penetrations Change; dated
June 2, 2004
TM 04-005; Auxiliary Feed Tunnel Seismic Event Annunciation; Revision 0
1EP4 Emergency Action Level and Emergency Plan Changes
Point Beach Nuclear Plant Emergency Plan; Section 2; Revision 40
Point Beach Nuclear Plant Emergency Plan; Section 7; Revision 46
1EP6 Drill Evaluation
LOR 04-03 Scenario SES 114; Licensed Operator Requalification Training Scenario;
Revision 0, dated May 25, 2004
LOR TRPR 33.0; Operations Crew Failure PRC Minutes, dated June 10, 2004
TI 8.0, Attachment 2; Conduct of Simulator Training and Simulator Evaluation;
Revision 6
24 Attachment
2OS1 Access Control to Radiologically Significant Areas
CAP 055366; Worker Received Electronic Dosimeter Dose Alarm; dated April 5, 2004
CAP 055587; S/G Nozzle Dam Installation Dose Exceeded Estimate; dated
April 11, 2004
CAP 055595; Air Supply to Bubble Hoods Not Within Procedure Limits; dated
April 12, 2004
CAP 055951; Incore Thermocouple Guide (Bullet Nose) Inadvertently Lifted with
Reactor Head; dated April 22, 2004
CAP 055986; Evaluate Use of RP Greeter at Containment Hatches During Outage
Periods; dated April 23, 2004 [NRC-Identified Issue]
HP 3.2; Radiological Labeling, Posting, and Barricading Requirements; Revision 39
HPIP 4.58; Issuance of Respiratory Equipment; Revision 13
HPIP 3.52; Airborne Radioactivity Surveys; Revision 30
PCE No. 04-02-018; Personnel Contamination Event (PCE) Report; dated April 8, 2004
PCE No. 04-02-019; Personnel Contamination Event (PCE) Report; dated April 9, 2004
PCE No. 04-02-020; Personnel Contamination Event (PCE) Report; dated April 8, 2004
RCE 253/CAP055527; Industrial Safety Issues and Poor Work Practices During Nozzle
Dam Installation; dated May 22, 2004
RWP No.04-104; RP Coverage; Revision 0
RWP No.04-113; Reactor Head Lift; Revision 0
RWP No.04-122; Reactor Coolant Pump Maintenance; Revision 0
RWP No.04-133; BMI Inspection; Revision 0
RWP No.04-141; Nozzle Dam Install/Remove; Revision 0
RWP No.04-142; Steam Generator Eddy Current Testing; Revision 1
RWP No.04-171; NRC Walkdowns for U1R28; Revision 0
RWP No.04-182; Replace Cono-Seal Bullet; Revision 0
25 Attachment
2OS2 ALARA Planning And Controls
ALARA Review No. 2004-0012; Level 3 Pre-Job ALARA Review for BMI Inspection
(RWP No.04-133); dated April 6, 2004
ALARA Review No. 2004-0017; Level 3 Pre-Job and In-Progress ALARA Reviews for
Nozzle Dam Install/Remove (RWP No.04-141); dated April 2 and 11, 2004
ALARA Review No. 2004-0018; Level 3 Pre-Job ALARA Reviews (Revisions 0 and 1)
for Steam Generator Eddy Current Testing (RWP No.04-142); dated March 16 and
April 12, 2004
ALARA Review No. 2004-0027; Level 3 Pre-Job ALARA Review for Replace Cono-Seal
Bullet (RWP No.04-182); dated April 22, 2004
JIT Briefing Activity MM-8480D3; Just In Time Briefing for Reactor Vessel Head Lift;
dated April 21, 2004
HPIP 4.40; TEDE ALARA Evaluation; Revision 0 (February 6, 2002)
NP 4.2.1; ALARA Program; Revision 11 (November 19, 2003)
Point Beach Nuclear Plant U1R28 Estimated RWP Dose Spreadsheet; dated
April 6, 2004
TEDE ALARA Evaluation for RWP 04-141; dated April 3, 2004
TEDE ALARA Evaluation for RWP 04-182; dated April 22, 2004
4OA1 Performance Indicator Verification
NP 5.2.16; NRC Performance Indicators Attachment C for Unplanned Scrams, Unit 1 PI
Data Calculation, Review and Approval; Revision 7, April 23, 2004
NP 5.2.16; NRC Performance Indicators Attachment C for Unplanned Scrams, Unit 2 PI
Data Calculation, Review and Approval; Revision 7, April 23, 2004
NP 5.2.16; NRC Performance Indicators Attachment C for Scrams With Loss of Heat
Removal, Unit 1 PI Data Calculation, Review and Approval; Revision 7, April 23, 2004
NP 5.2.16; NRC Performance Indicators Attachment C for Scrams With Loss of Heat
Removal, Unit 2 PI Data Calculation, Review and Approval; Revision 7, April 23, 2004
4OA2 Identification and Resolution of Problems - Inservice Inspection
Corrective Action Documents
CAP047990; OE 14934 Problems with Ultrasonic Testing Caused Unnecessary Pipe
Replacement; dated August 21, 2003
26 Attachment
CAP054136; Unit 2 S/G Tube Leakage Exceeded 5 gpd; dated February 23, 2004
CAP053177; Increased Fluoride Contamination in the Unit 1 S/Gs; dated
January 25, 2004
CAP033575; OE 16308 Incorrect Diameter Probe Used During Eddy Current Inspection;
dated June 16, 2003
CAP029936; Service Water Intrusion into A and B S/Gs for Unit 1; dated
October 26, 2002
CAP003372; NSAL-02-13 Fatigue Life of CE [combustion Engineering] Steam
Generator Primary Manway Studs; dated August 20, 2003
CAP032045; New AFW Restricting Orifices May Not Meet Section XI R/R
[Repair/Replacement] Requirements; dated April 6, 2003
CAP051046; SW Pipe Wall Thinning Noted During Execution of U2R26 WO 9905610;
dated October 14, 2003
CAP032290; Inservice Inspection Limited Examinations; dated April 17, 2003
CAP051206; Small Wires Found in the Secondary Side on the A Steam Generator;
dated October 18, 2003
CAP051407; Small Wires Found in the Secondary Side on the B Steam Generator;
dated October 24, 2003
CAP029413; Accumulator Nozzle Have Unidentified Indications on the Inside Surface;
dated September 19, 2002
CAP010698; Accumulator Nozzles Have Unidentified Indications on the Inside Surface;
dated September 21, 2002
OTH026613; Accumulator Nozzles Have Unidentified Indications on the Inside Surface;
dated October 07, 2002
OTH026615; Accumulator Nozzles Have Unidentified Indications on the Inside Surface;
dated October 07, 2002
OTH026616; Accumulator Nozzles Have Unidentified Indications on the Inside Surface;
dated October 07, 2002
CE012362; Framatome NRC 6028873 - Lack of UT Coverage During U1R27 RPV
Inspection; dated September 18, 2003
CA053202; Framatome NRC 6028873 - Lack of UT Coverage During U1R27 RPV
Inspection; dated October 15, 2003
27 Attachment
CAP022754; Liner Plate Degradation; dated April 25, 2002
CAP012575; Liner Plate Degradation-U1R26 Restart Issue; dated April 13, 2001
CAP012576; Liner Plate Degradation; dated April 13, 2001
Corrective Action Reports Initiated as a Result of NRC Inspection
CAP055529; NIS-1 Report Contains Information That Could Be Misunderstood; dated
April 09, 2004
CAP055517; Repair/Replacement Documentation May Have Incomplete Information;
dated April 09, 2004
CAP055652; Wrong Size of Weld Filler Metal Used; dated April 13, 2004
CAP055664; Procedure NDE-750 Does Not Require Recording Boric Acid on Stainless
Steel Bolts; dated April 13, 2004
CAP055678; Feedback Regarding NDE From NRC Exit on April 9, 2004; dated
April 14, 2004
CAP056011; Tracking Mechanism for ISI Relief Requests Not Clear; dated
April 23, 2004
OTH012761; Calculate New RPV Head Temperatures - Post RPV Head Replacement;
dated April 26, 2004
4OA2 Identification and Resolution of Problems - Resident Inspector Samples
CAP052757; 2P-2B Charging Pump Tripped; dated January 12, 2004
CAP052764; 2P-2B tripped while opening door on 2B-03; dated January 12, 2004
CAP052838; Seismic Question of B03 and B04 Buses; dated January 15, 2004
Maintenance Rule Evaluation MRE000152: Maintenance Rule Evaluation, 2P-2B Trip;
dated January 12, 2004
CE012929; 2P-2B Trip; dated January 12, 2004
WO0400401; Door for Relay Cubicle is Sprung (Ref CAP052764)
WO0400534; Document Maintenance Rule Functional Failure of 2P-2B Charging Pump
NP 7.7.2; Seismic Qualification of Equipment; Revision 2
NP 13.1.1; Self-Assessment Program; Revision 10
28 Attachment
CAP056602; Ineffective CA on 1RH-713A/B, Evidence During IT-530C; dated
May 11, 2004
WEST110E018 Sheet 1; Auxiliary Coolant System Unit 1; Revision 57
DBD-10; RHR System, Section 3.7; Revision 1
CAP056580; Unexpected Cavity Level Decrease During IT-530C; dated May 11, 2004
NMC Incident Response Team For Issues Encountered During Unit 2 Trip Due to Diver
Trapped in Intake Structure at PBNP; dated May 17, 2004
CAP056785; CA Designated as CATPR [Corrective Action To Prevent Recurrence] in
RCE 00-093 Not Implemented; dated May 18, 2004
CAP056774; Wrong Cubicle Opened, WO 0407727; dated May 18, 2004
CAP056776; 1X-04 Annunciator Alarm Activated by D52A Selector Switch Operation,
dated May 17, 2004
Point Beach Nuclear Plant Corrective Action Program Report, 4th Quarter 2003
Point Beach Nuclear Plant Corrective Action Program Report, 1st Quarter 2004
Nuclear Plant Memorandum NPM 2004-00344; First Quarter Human Performance
Analysis; dated April 30, 2004
CAP057010; 1st Quarter Human Performance Data Analysis; dated May 26, 2004
CAP057705; 2004 CAP Self-Assessment - Minor Issues; dated July 1, 2004
CA054949; Mispositioning Adverse Trend Identified - Wrong Unit, Wrong Train, Wrong
Device; dated January 13, 2004
4OA3 Event Follow-up
LER 301/2004-001-00; SI System Accumulator Operated With Fluid Level Out Of
Specification High; May 21, 2004
CAP055204; Troubleshooting Reveals 2T34A SI Accumulator Level Out of
Specification High; dated March 30, 2004
RCE000251; Root Cause Evaluation; 2T-34A SI Accumulator Level Instruments
Returned to Service Without Proper Post Maintenance Testing; Revision 0
CAP055415; Organization Response to Unit 2 SI Accumulator Level Transmitter Issues;
dated April 7, 2004
CAP056175; Nuclear Safety Culture Assessment Required; April 28, 2004
29 Attachment
CAP056363; Missed TSAC Entry Following Denergizing of 1Y-04; dated May 4, 2004
4OA5.1 Reactor Pressure Vessel Head and Vessel Head Penetration Nozzles (TI 2515/150)
Nondestructive Examination Reports
Point Beach Unit 1 (U1R28) - Extent of UT Coverage in RVHP Nozzle Material; dated
May 6, 2004
Point Beach Nuclear Power Plant Liquid Penetrant Examination Record; Nozzle 26;
dated April 29, 2004
Point Beach Nuclear Power Plant Liquid Penetrant Examination Record; Nozzle 26;
dated May 2, 2004
Point Beach Nuclear Power Plant Liquid Penetrant Examination Record; Nozzle 26;
dated May 5, 2004
Videotaped of Dye Penetrant Examinations of Nozzle 26; Performed April 29, 2004,
April 30, 2004, May 2, 2004, and May 5, 2004
Point Beach Nuclear Power Plant Visual Examination Record; RPV Closure Head; dated
April 26, 2004
Point Beach Nuclear Power Plant Visual Examination Record; RPV Closure Head; dated
April 27, 2004
Point Beach Nuclear Power Plant Visual Examination Record; RPV Closure Head; dated
May 1, 2004
Point Beach Nuclear Power Plant Remote Visual Examination Record; RPV Closure
Head; dated May 23, 2004
Point Beach Nuclear Power Plant Visual Examination Record; RPV Closure Head; dated
May 6, 2004
Point Beach Nuclear Power Plant Ultrasonic Calibration Record; Penetrations 32 and 33;
dated May 14, 2004
Videotaped Upper Head Examination and Cleaning from April 26, 2004 through
May 6, 2004
Ultrasonic Calibration Data Sheets; Penetration No. 26 J-Groove Weld After Machining 0,
45 Degree, and OD Creeping Wave Scans; dated May 12, 2004
51-5045099-00; Point Beach Unit 1 (U1R28) RVH Nozzle UT Inspection Final Report;
Draft; dated May 26, 2004
30 Attachment
Other Documents
Westinghouse Letter Report LTR-RCDA-0377; Revision 2
C11470; Reactor Vessel Head Effective Degradation Year (EDY); dated May 29, 2003
EPRI MRP-89; Materials Reliability Program Demonstrations of Vendor Equipment and
Procedures for the Inspection of Control Rod Drive Mechanism Head Penetrations; dated
September 2003
WCAP-15950; Structural Integrity Evaluation of Reactor Vessel Upper Head Penetration
to Support Continued Operation of Point Beach Units 1 and 2; dated September 2002
PWR Materials Reliability Program Response to NRC Bulletin 2001-01 (MPR-48); EPRI
1006284; dated August 2001
Calculation Cover Sheet and Review Report; Reactor Vessel Head Effective
Degradation; Calc # C11470; dated May 29, 2003
Letter from B. Rassler (EPRI) to B. Jenson (Nuclear Management Company); Blind
Demonstration Testing of UT Procedure; dated May 3, 2004
54-ISI-30-01; Written Practice for the Qualification and Certification of NDE Personnel;
dated August 18, 2004
Framatome ANP Certificate of Personnel Qualification for:
C Jonathan D. Buttram, UT Level III; dated February 5, 2004;
C Jason D. Breza, UT Level II; dated January 29, 2004;
C Michael W. Key, UT Level III; dated January 29, 2004;
C Kent Gebetsberger, UT; dated September 14, 2002;
C Chuck Martin, UT Level II; dated September 14, 2002;
C John Touhalisky, UT Level II; dated September 14, 2002; and
C Robert Kellerhall, UT Level II; dated September 14, 2002.
NMC Record of Certificate of NDE Personnel as UT Level III for William Jensen; dated
August 19, 2003
1-PT-RCS-1; Reactor Coolant System Pressure Test - Inside/Outside Containment
Unit 1; dated October 15, 2002
1-PT-RCS-1; Reactor Coolant System Pressure Test - Inside/Outside Containment
Unit 1 Appendix B; dated October 13, 2002
SEM 7.11.5; RCS Leak Test for Unit 1; dated April 13, 2001
WO9923859; Visual Examination Leak Test Record Data Sheets (13 pages); dated
May 9, 2001
31 Attachment
Boric Acid Walkdown Data Sheets Refueling Outage: U1R27; dated September 15, 2002
Organizational Assessment Audit Plan and Checklist: First Quarter 2001 Engineering
Audit; Scope: Repair and Replacement Modification Activities Relating to ASME
Section XI, Inservice Testing per ASME Section XI; Document #: A-P-01-03; dated
January 15, 2001
Record of Certification of NDE Personnel; William Jensen; Visual Level III; dated
August 13, 2003
Record of Certification of NDE Personnel; Patric Turner; Visual Level II; dated
August 12, 2003
Weld Control Records; Layers 1 through 14; dated May 12 and 13, 2004
Drawing 5019702; Point Beach Unit 1 CRDM [Control Rod Drive Mechanism] Nozzle ID
[Inner Diameter] Temper Bead Weld Repair; Revision 3
Quality Assurance Data Package No. 23-5044625-00; Welding Filler Material For NMC,
Point Beach Unit 1 Reactor Vessel Head Repair; dated May 7, 2004
Process Traveler; Ambient ID Temper Bead Repair For CRDM Nozzles; dated
May 7, 2004
Repair/Replacement Form No. 2004-03; Repair Nozzle 26; dated May 11, 2004
Weld Procedure Specification 55-WP3/43/F43TBSCA301; Revision 1
Procedure Qualification Record 55- PQ7164-03; dated May 23, 2003
Procedure Qualification Record 55- PQ7183-03; dated May 8, 2004
WCAP 14929; Probabilistic Evaluation of Reactor Vessel Closure Head Penetration
Integrity for Point Beach Units 1 and 2; Revision 0
Point Beach U1R27 Reactor Vessel Head CRDM Nozzle Ultrasonic Examination Report;
dated October 5, 2002
Memorandum to File; Point Beach Nuclear Plant Vessel Closure Head Temperature;
dated April 22, 2004
MRP-89; Materials Reliability Program Demonstrations of Vendor Equipment and
Procedures for the Inspection of Control Rod Drive Mechanism Head Penetrations;
dated September 2003
54-5016639-00; Framatome ANP Reactor Vessel Head Penetration Leak Path
Qualification Report; dated February 6, 2002
54-5040736-00; Framatome ANP Demonstration of CRDM Leak Path Detection
Technique; dated February 26, 2004
32 Attachment
Letter from A. Johnson (WE) to USNRC; GL 97-01 120 Day Response Point Beach
Nuclear Plant, Units 1 and 2; dated July 30, 1997
Letter from A. J Cayia (NMC) to USNRC; Supplemental Response to NRC Bulletins
2001-01, 2002-01, and 2002-02 for Reactor Vessel Head and Head Penetration Nozzle
Inspection Findings; dated November 24, 2003
CA053202; Framatome NCR [Non-Conformance Report] 6028873-Lack of UT Coverage
During U1R27 RPV Inspection; dated October 15, 2003
CE012362; Framatome NCR 6028873-Lack of UT Coverage During U1R27 RPV
Inspection; dated September 18, 2003
Memorandum; Obstructed Area of Unit 1 Reactor Vessel Dome; dated May 24, 2004
Procedures
NDE-757; Visual Examination For Leakage of Reactor Pressure Vessel Penetrations;
Revision 3
NDE-451; Visible Dye Penetrant Examination Temperature Applications 45o F to 125o F;
Revision 21
54-ISI-100-11; Remote Ultrasonic Examination of Reactor Head Penetrations;
Revisions 9 through 11
54-ISI-137-03; Remote Ultrasonic Examination of Reactor Vessel Head Vent Line
Penetrations; Revision 3
54-PQ-137-01; Remote Ultrasonic Examination of Reactor Vessel Head Vent Line
Penetrations; dated February 22, 2002
54-PQ-137-01; Remote Ultrasonic Examination of Reactor Vessel Head Vent Line
Penetrations; dated September 20, 2002
54-PQ-137-01; Remote Ultrasonic Examination of Reactor Vessel Head Vent Line
Penetrations; dated November 21, 2002
1-PT-RCS-1; Reactor Coolant System (RCS) Pressure Test- Inside/Outside
Containment Unit 1; Revision 1
NP 7.4.14; Boric Acid Leakage and Corrosion Monitoring; Revision 0
Boric Acid Leakage and Corrosion Monitoring Program; Revision 0
NDE-141; Manual Ultrasonic Examination of Reactor Head Penetrations; Revision 0
33 Attachment
4OA5.2 Reactor Pressure Vessel Lower Head Penetration Nozzles (TI 2515/152)
Drawings
TP-3609-4; Section Thru Bottom of Reactor Vessel; Revision 0
RT-49006-RI; RVCH Insulation System General Arrangement Drawing; Revision 0
West 685J441, Sheets A, B, C, and D; NIS Bottom Mounted Instrumentation Point
Beach NP; Revision 9
Nondestructive Examination Reports
Point Beach Visual Examination Record; Reactor Pressure Vessel BMI Tubes; dated
April 6, 2004
Procedures
NDE-757; Visual Examination For Leakage of Reactor Pressure Vessel Penetrations;
Revision 3
NDE-3; Written Practice For Qualification And Certification For NDE Personnel;
Revision 28
2-NDES-001; Nondestructive Examination Personnel Qualification and Certification;
Revision 2
Other Documents
Point Beach Nuclear Plant Visual Examination Record; Reactor Pressure Vessel BMI
Tubes; dated April 6, 2004
Record of Certification NDE Personnel; William Jensen; dated August 19, 1983
IHI Southwest Technologies, INC. Statement of NDE Certification; Victor Morton; dated
January 5, 2004
4OA5.4 Reactor Containment Sump Blockage Point Beach Units 1 & 2 (TI 2515/153)
CAP05785; Vendor Program Applicability Problems Deleted Monitored Parameters for
Sump Blockage; dated July 9, 2004
CAP050529; Problems With Vendor Program Delayed Preparations for EOP Issuance;
dated July 9, 2004
EOP -1 Unit 1; Loss of Reactor or Secondary Coolant; Revision 35
EOP -1 Unit 2; Loss of Reactor or Secondary Coolant; Revision 35
34 Attachment
Document Review and Approval Form for EOP-1, Unit 1; Loss of Reactor or Secondary
Coolant, Revision 34; dated June 18, 2003
Document Review and Approval Form for EOP-1, Unit 1; Loss of Reactor or Secondary
Coolant, Revision 35; dated October 1, 2003
NMC Letter NRC 2003-0068; Nuclear Regulatory Commission Bulletin 2003-01:
Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized
Water Reactors - 60-Day Response; dated August 8, 2003
NMC Letter NRC 2004-0050; Supplement to 60-Day Response to Bulletin 2003-01:
Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized
Water Reactors; dated May 14, 2004
NP 7.2.28; Containment Debris Control Program; Revision 0
NP 8.4.15; Protective Coating Program; Revision 2
CL 20; Post Outage Containment Closeout Inspection Unit 2, Revision 11; Completed
November 2003
PC 24; Containment Inspection Checklist Unit 1, Revision 58
PC 24; Containment Inspection Checklist Unit 2; Revision 57
IT 536; Leakage Reduction and Preventative Maintenance Program Test of
Containment Sump B Suction Line Mode 5,6, or Defueled Unit 2; Revision 17
IT 531; Leakage Reduction and Preventative Maintenance Program Test of
Containment Sump B Suction Line Mode 5,6, or Defueled Unit 1; Revision 15
FSAR Section 14.3; Radiological Consequences of Loss of Coolant Accident; June 2002
FSAR Section 6.2; Safety Injection System; June 2003
EPRI Boric Acid Corrosion Guidebook; Managing Boric Acid Corrosion Issues at PWR
Power Stations; Revision 1
Unit 2 Refueling Outage 26 Containment Insulation and Debris Inventory for GSI
[Generic Safety Issue]-191; February 11, 2004
Final Letter Report Point Beach Unit 2 Parametric Debris Transport Calculation;
Framatome ANP, Inc; dated September 24, 2003
Point Beach Unit 2 - Parametric Debris Generation; Framatome ANP Document
32-5031657-00; dated August 18, 2003
Report on Containment Coating Assessment Point Beach Nuclear Plant - Unit 2; dated
April 29, 2002
35 Attachment
Report on Containment Coating Assessment Point Beach Nuclear Plant - Unit 1, U1R28
(Draft); dated July 2004
Wisconsin Electric Power Company Project Number 5613; Gibbs and Hill Inc., Coatings
Failure Analysis, Point Beach Nuclear Plant Final Report Submittal; dated
August 1, 1990
Wisconsin Electric Calculation DIT-PB-EXT-0494-00; Containment Sump Blockage Due
to Failure of Unqualified/Undocumented Coatings; dated January 21, 1999
Point Beach Calculation N92-086; ECCS Pump NPSH Calculation; Revision 3
U2R26 Containment Coating Assessment Point Beach Nuclear Plant; dated
April 15, 2004
Letter from Framatome ANP; Subject: Proposal for Phase 2 Evaluation for Point Beach
Nuclear Plant Unit 2; dated January 16, 2004
Letter from Framatome ANP; Subject: Engineering Scope to Address GSI-191
Concerns at Point Beach Unit 1; dated January 23, 2004
Temporary Procedure Change 2004-0643; EOP-1, Loss of Reactor or Secondary
Coolant - Unit 1; dated July 9, 2004
Temporary Procedure Change 2004-0648; ECA-3.2, Steam Generator Tube Rupture
With Loss of Reactor Coolant-Saturated Recovery Desired - Unit 2; dated July 13, 2004
CAP051547; PB2 Sump B Strainer Concerns; dated October 30, 2003
Bechtel Drawing M-276; Containment Safety injection Sump Requirements for Screens;
Revision 2
Wisconsin Electric Drawing PBC-309; ISI Classification Drawing Keyway Sump A
Tunnel; Revision D
4OA5.5 Spent Fuel Material Control and Accounting at Nuclear Power Plants (TI 2515/154)
Nuclear Plant Memorandum NPM 2004-0332; Response to NRC TI 2515/154, Point
Beach Nuclear Plants, Units 1 and 2; dated May 20, 2004
4OA5.6 Offsite Power System Operational Readiness (TI 2515/156)
Generation - Transmission Interconnection Agreement between American Transmission
Company, LLC [ATC] as Transmission Provider and Wisconsin Electric Power Company
[Point Beach owner] for the Point Beach Plant Interconnection Facilities, dated as of
November 1, 2000.
DBD-20; 345 KVAC [Kilovolt Alternating Current] System; Revision 2
36 Attachment
NP 2.1.5; Electrical Communications, Switchyard Access and Work Planning; Revision 3
NP 2.1.8; Protected Equipment; Revision 0
NP 10.3.6; Outage Safety Review and Safety Assessment; Revision 11
NP 10.3.7; On-Line Safety Assessment; Revision 8
TS Test 81; EDG G-01 Monthly; Revision 67
CAP050179; Industry OE - Callaway - Inoperability of Both Offsite Power Sources;
dated September 15, 2003
CAP050343; Evaluate Combinations of Offsite Power Conditions That Could Challenge
Plant Ops; dated September 23, 2003
CAP050414; Potential to Separate From Grid When Both X02 Are OOS [Out-Of-
Service] and a X03 Failure Occurs; dated September 25, 2003
CAP052189; SEN 242 Loss-of-Grid Event, August 14, 2003; dated December 8, 2003
CAP056406; Boundary of 345 KV System for Maintenance Rule Not Documented; dated
May 5, 2004
OE054270; Request an OE Evaluation of SEN 242 Loss-of-Grid Event, August 14,
2003; dated December 8, 2003
OTH055556; From OE Evaluation of SEN 242 - Evaluate the Need for a Backup Power
Supply at EOF [Emergency Operations Facility]; dated February 6, 2004
OP 2A, Normal Power Operation, Revision 50
37 Attachment
LIST OF ACRONYMS USED
ACE Apparent Cause Evaluation
ADAMS Agency Wide Access Management System
ALARA As-Low-As-Is-Reasonably-Achievable
AOP Abnormal Operating Procedure
ARB Alarm Response Book
ASME American Society of Mechanical Engineers
ATC American Transmission Company
BMI Bottom Mounted Instrument
CA Corrective Action
CAP Corrective Action Program Document
CCW Component Cooling Water
CE Condition Evaluation
CFR Code of Federal Regulations
CL Checklist
COLR Core Operating Limits Report
CRDM Control Rod Drive Mechanism
CSP Critical Safety Procedure
CW Circulating Water
DBD Design Basis Document
ECA Emergency Contingency Action
ECCS Emergency Core Cooling System
EDG Emergency Diesel Generator
EDM Electric Discharge Machined
EDY Effective Degradation Years
EOP Emergency Operating Procedure
EPRI Electric Power Research Institute
FHAR Fire Hazard Analysis Report
FPER Fire Protection Evaluation Report
FSAR Final Safety Analysis Report
GL Generic Letter
HPIP Health Physics Implementing Procedure
HX Heat Exchanger
ICP Instrument and Control Procedure
IMC Inspection Manual Chapter
IP Inspection Procedure
ISI Inservice Inspection
IT Inservice Test
KV Kilo-Volt
LER Licensee Event Report
LOCA Loss of Coolant Accident
LOOP Loss of Offsite Power
LOR Licensed Operator Requalification
MIC Microbiologically Induced Corrosion
MOV Motor-Operated Valve
38 Attachment
NCV Non-Cited Violation
NDE Nondestructive Examination
NIOSH National Institute for Occupational Safety and Health
NMC Nuclear Management Company
NP Nuclear Plant Procedures Manual
NRC Nuclear Regulatory Commission
No. Number
OCC Outage Control Center
ODSCC Outside Diameter Stress Corrosion Cracking
OE Operating Events/Experience
OI Operating Instruction
OM Operations Manual
OP Operating Procedure
OPR Operability Recommendation
OS Occupational Radiation Safety
OTH Other (Corrective Action Program Document)
OWA Operator Workaround
PBF Point Beach Form
PBNP Point Beach Nuclear Plant
PC Periodic Check
PCE Personnel Contamination Event
PI Performance Indicator
PMT Post-Maintenance Testing
ppm Parts Per Million
psig pounds per square inch gauge
PT Dye Penetrant Test
PWSCC Primary Water Stress Crack Corrosion
PWR Pressurized Water Reactor
RCE Root Cause Evaluation
RCP Reactor Coolant Pump
RMP Routine Maintenance Procedure
RNO Response Not Obtained
RP Radiation Protection
RWP Radiation Work Permit
SDP Significance Determination Process
SEP Shutdown Emergency Procedure
SFP Spent Fuel Pool
SI Safety Injection
SR Surveillance Requirement
TEDE Total Effective Dose Equivalent
TI Temporary Instruction
TOFD Time Of Flight Diffraction
TSB Technical Support Building
39 Attachment
TS Technical Specification
TSAC Technical Specification Action Condition
U1R28 Unit 1 Refueling Outage 28
URI Unresolved item
UT Ultrasonic Test
VHP Vessel Head Penetration
WO Work Order
WPS Weld Procedure Specification
40 Attachment