ML111822085
| ML111822085 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 07/01/2011 |
| From: | Vincent Gaddy NRC/RGN-IV/DRP/RPB-C |
| To: | O'Grady B Nebraska Public Power District (NPPD) |
| References | |
| IR-11-008 | |
| Download: ML111822085 (40) | |
See also: IR 05000298/2011008
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGI ON I V
612 EAST LAMAR BLVD, SUITE 400
ARLINGTON, TEXAS 76011-4125
July 1, 2011
Brian J. OGrady, Vice President-Nuclear
and Chief Nuclear Officer
Nebraska Public Power District
72676 648A Avenue
Brownville, NE 68321
Subject: COOPER NUCLEAR STATION - NRC SPECIAL INSPECTION REPORT
Dear Mr. OGrady:
On May 3, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed a special
inspection at your Cooper Nuclear Station to evaluate the facts and circumstances surrounding
the exposure of three workers to higher than expected dose rates while removing an
intermediate range monitor shuttle tube from beneath the reactor pressure vessel. The
enclosed report documents the inspection findings that were discussed on June 9, 2011, with
Mr. A. Zaremba, Director, Nuclear Safety Assurance, and other members of your staff.
The inspection examined activities conducted under your license as they relate to safety and
compliance with the Commission's rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
plant personnel.
Based upon exceeding the deterministic criteria for radiation safety specified in NRC
Management Directive 8.3, "NRC Incident Investigation Program, the NRC initiated a special
inspection in accordance with Inspection Procedure 93812, "Special Inspection. The basis for
initiating the special inspection was the work activity led to unplanned changes in restricted area
dose rates in excess of 20 rem per hour in an area where personnel were present. The focus of
the inspection was the event that took place on April 3, 2011, when three workers removed an
intermediate range monitor shuttle tube from beneath the reactor pressure vessel and dose
rates in the area went from 120 millirem per hour to 39 rem per hour at 30 centimeters from the
tip of the shuttle tube, which was the source of the excess dose. The focus areas for review are
detailed in the Special Inspection Charter (Attachment 2). On April 5, 2011, the NRC
determined that the inspection would be conducted and the onsite inspection started on
April 11, 2011.
This report documents six NRC-identified findings of very low safety significance (Green). Five
of these findings were determined to involve violations of NRC requirements. However,
because of their very low safety significance and because they are entered into your corrective
action program, the NRC is treating these findings as noncited violations, consistent with
Section 2.3.2 of the NRC Enforcement Policy. If you contest the violations or the significance of
the noncited violations, 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,
Nebraska Public Power District -2-
ATTN: Document Control Desk, Washington, D.C. 20555-0001, with copies to the Regional
Administrator, U.S. Nuclear Regulatory Commission, Region IV, 612 E. Lamar Blvd, Suite 400,
Arlington, Texas, 76011-4125; the Director, Office of Enforcement, U.S. Nuclear Regulatory
Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the facility. In
addition, if you disagree with the cross-cutting aspect assigned to any finding in this report, you
should provide a response within 30 days of the date of this inspection report, with the basis for
your disagreement, to the Regional Administrator, Region IV, and the NRC Resident Inspector
at the facility. The information you provide will be considered in accordance with Inspection
Manual Chapter 0305.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its
enclosure, and your response, if you choose to provide one, will be made available
electronically for public inspection in the NRC Public Document Room or from the NRC's
document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-
rm/adams.html. To the extent possible, your response should not include any personal privacy
or proprietary, information so that it can be made available to the Public without redaction.
Sincerely,
/RA/
Vincent G. Gaddy, Chief
Project Branch C
Division of Reactor Projects
Docket: 50-298
License: DPR-46
Enclosure: NRC Inspection Report 05000298/2011008
w/Attachments:
Attachment 1: Supplemental Information
Attachment 2: Special Inspection Charter
Attachment 3: Pictures and Diagrams
cc w/Enclosures:
Distribution via Listserv
Nebraska Public Power District -3-
Electronic distribution by RIV:
Regional Administrator (Elmo.Collins@nrc.gov)
Deputy Regional Administrator (Art.Howell@nrc.gov)
DRP Director (Kriss.Kennedy@nrc.gov)
Acting DRP Deputy Director (Jeff.Clark@nrc.gov)
(Troy.Pruett@nrc.gov)
DRS Director (Anton.Vegel@nrc.gov)
DRS Deputy Director (Tom.Blount@nrc.gov)
Senior Resident Inspector (Jeffrey.Josey@nrc.gov)
Resident Inspector (Michael.Chambers@nrc.gov)
Branch Chief, DRP/C (Vincent.Gaddy@nrc.gov)
Senior Project Engineer, DRP/C (Bob.Hagar@nrc.gov)
Project Engineer, DRP/C (Rayomand.Kumana@nrc.gov)
CNS Administrative Assistant (Amy.Elam@nrc.gov)
Public Affairs Officer (Victor.Dricks@nrc.gov)
Public Affairs Officer (Lara.Uselding@nrc.gov)
Project Manager (Lynnea.Wilkins@nrc.gov)
Acting Branch Chief, DRS/TSB (Dale.Powers@nrc.gov)
RITS Coordinator (Marisa.Herrera@nrc.gov)
Regional Counsel (Karla.Fuller@nrc.gov)
Congressional Affairs Officer (Jenny.Weil@nrc.gov)
OEMail Resource
ROPreports
OEDO RIV Coordinator (Stephaine.Bush-Goddard@nrc.gov)
R:\_REACTORS\_CNS\2011\CNS 2011008-RP-DHO.docx
ADAMS: No Yes SUNSI Review Complete Reviewer Initials:
Publicly Available Non-Sensitive
Non-Publicly Available Sensitive
RI:PBA RI/PBE SPE/PBC DRP/PBC
BKTharakan DOverland BHagar VGaddy
E-mail/Vgaddy for E-mail/Vgaddy for /RA/ /RA/
6/28/11 6/28/11 6/14/11 6/29/11
C:DRS/PS2 C:DRS/PS2 DD:DRP
GWerner LRicketson KMKennedy
E-mail/Vgaddy for /RA/
6/29/11 6/15/11
OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket: 05000298
License: DPR-46
Report: 05000298/2011008
Licensee: Nebraska Public Power District
Facility: Cooper Nuclear Station
Location: 72676 648A Avenue
Brownville, NE 68321
Dates: April 11 through May 3, 2011
Inspectors: D. Overland, Resident Inspector, Waterford 3 Steam Electric Station
B. Tharakan, CHP, Resident Inspector, South Texas Project
Approved By: Vince Gaddy, Chief
Project Branch C
Division of Reactor Projects
-1- Enclosure
TABLE OF CONTENTS
SUMMARY OF FINDINGS ..........................................................................................................3
REPORT DETAILS .....................................................................................................................7
4OA3 Event Follow-Up................................................................................................................7
1.0 Special Inspection Scope .....................................................................................7
2.0 Event Description and Chronology
2.1 Event Summary ........................................................................................7
2.2 Sequence of Events .................................................................................8
2.3 Immediate Actions Taken ........................................................................10
3.0 Work Planning and Execution
3.1 Work Order Planning ..............................................................................10
3.2 Job Preparation ......................................................................................15
3.3 Work Execution ......................................................................................17
4.0 Radiation Protection Performance
4.1 ALARA Planning .....................................................................................19
4.2 Radiation Work Permit Adequacy ...........................................................19
4.3 ALARA Briefing.......................................................................................20
4.4 Job Coverage .........................................................................................22
4.5 Dose and Dose Rate Assessment ..........................................................27
5.0 Review of Previous Activity Performance ...........................................................28
6.0 Review of Causal Determination and Corrective Actions .................................... 28
4OA6 Meetings........................................................................................................................29
ATTACHMENTS
1 Supplemental Information .............................................................................. A1-1
Key Points of Contact .................................................................................... A1-1
List of Items Opened, Closed, and Discussed ................................................ A1-1
Documents Reviewed .................................................................................... A1-2
2 Special Inspection Charter ............................................................................. A2-1
3 Pictures and Diagrams ................................................................................... A3-1
-2- Enclosure
SUMMARY OF FINDINGS
IR 05000298/2011008; 04/11/11 - 05/03/11; Cooper Nuclear Station; Special inspection to
evaluate unexpected doses to workers performing under-vessel maintenance activities.
The report covered one week of onsite inspection and in-office review through May 3, 2011.
Two resident inspectors performed the inspection. Five Green noncited violations and one
Green finding were identified. The significance of most findings is indicated by their color
(Green, White, Yellow, or Red) using Inspection Manual Chapter 0609, "Significance
Determination Process." The cross-cutting aspect is determined using Inspection Manual
Chapter 0310, Components within the Cross Cutting Areas. Findings for which the
significance determination process does not apply may be Green or be assigned a severity level
after NRC management review. The NRC's program for overseeing the safe operation of
commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process,"
Revision 4, dated December 2006.
A. NRC-Identified and Self Revealing Findings
Cornerstone: Occupational Radiation Safety
- Green. The inspectors identified a noncited violation of Technical
Specification 5.4.1, for a failure to implement procedures described in Regulatory
Guide 1.33, Appendix A. Specifically, the licensee failed to implement
procedures that provide guidance on creating clear, accurate work instructions.
As a result, the work instructions were not able to be completed as written and
needed parts were not available. This directly contributed to three
instrumentation and control technicians receiving an unexpected radiation dose.
A site stand-down was held to discuss the lessons learned and the event was
entered into the licensees corrective action program as Condition
Report CR-CNS-2011-4431.
This deficiency was reasonable for the licensee to foresee and prevent
occurrence. The finding was more than minor because it is associated with the
human performance attribute of the Occupational Radiation Safety Cornerstone
and affected the cornerstone objective to ensure the adequate protection of the
worker health and safety from exposure to radiation from radioactive material
during routine civilian nuclear reactor operation. The inspectors evaluated this
finding using Inspection Manual Chapter 0609, Appendix C, Occupational
Radiation Safety Significance Determination Process. The inspectors
determined that the finding is of very low safety significance (Green) because it
was not associated with ALARA planning or work controls, there was no
overexposure, there was no substantial potential for an overexposure, and the
licensees ability to assess dose was not compromised. The finding has a cross-
cutting aspect in the work practices component of the human performance area
because the licensee did not effectively communicate expectations regarding
procedural compliance and that personnel follow procedures. Specifically, the
licensee displayed a cultural behavior that unacceptable behaviors, such as
failing to follow procedures, are acceptable as long as the outcome is desirable
[H.4.(b)](Section 3.1).
-3- Enclosure
- Green. The inspectors identified a noncited violation of Technical
Specification 5.4.1, for failure to implement procedures described in Regulatory
Guide 1.33, Appendix A. Specifically, the licensee failed to implement
procedures that provide guidance on recognizing risk associated with a
maintenance activity and properly accounting for that risk. This directly
contributed to three instrumentation and control technicians receiving an
unexpected radiation dose. A site stand-down was held to discuss the lessons
learned and the event was entered into the licensees corrective action program
as Condition Report CR-CNS-2011-4435.
This deficiency was reasonable for the licensee to foresee and prevent
occurrence. The finding was more than minor because it is associated with the
human performance attribute of the Occupational Radiation Safety Cornerstone
and affected the cornerstone objective to ensure adequate protection of the
worker health and safety from exposure to radiation from radioactive material
during routine civilian nuclear reactor operation. The inspectors determined that
the finding is of very low safety significance (Green) because it was not
associated with ALARA planning or work controls, there was no overexposure,
there was no substantial potential for an overexposure, and the licensees ability
to assess dose was not compromised. The finding has a cross-cutting aspect in
the work control component of the human performance area because the
licensee did not plan work activities by incorporating risk insights. Specifically,
the licensee developed a work package that failed to recognize the risk
associated with the activity H.3(a)(Section 3.1).
- Green. The inspectors identified a finding for a failure to implement human
performance procedures. Specifically, the licensee failed to implement
procedures that provided guidance on conducting pre-job briefs, preparing work
in the field, and informing technicians on what to do when the workers
encountered a problem. This contributed to three instrumentation and control
technicians receiving an unexpected radiation dose. A site stand-down was held
to discuss the lessons learned from the event. This was entered into the
licensees corrective action program as Condition Report CR-CNS-2011-4258.
The finding was more than minor because it is associated with the human
performance attribute of the Occupational Radiation Safety Cornerstone and
affected the cornerstone objective to ensure the adequate protection of the
worker health and safety from exposure to radiation from radioactive material
during routine civilian nuclear reactor operation. The inspectors evaluated this
finding using Inspection Manual Chapter 0609, Appendix C, Occupational
Radiation Safety Significance Determination Process. The inspectors
determined that the finding is of very low safety significance (Green) because it
was not associated with ALARA planning or work controls, there was no
overexposure, there was no substantial potential for an overexposure, and the
licensees ability to assess dose was not compromised. The inspectors
determined that the apparent cause of this finding was the licensees failure to
promote the use of human performance tools to ensure job tasks were properly
completed. Therefore, this finding has a cross-cutting aspect in the work
practices component of the human performance area because the licensee did
not adequately communicate human error prevention techniques such that work
activities are completed safely H.4(a)(Section 3.2).
-4- Enclosure
- Green. The inspectors identified a noncited violation of Technical
Specification 5.4.1, for a failure to comply with procedures described in
Regulatory Guide 1.33, Appendix A. Specifically, the licensee failed to
implement procedures and a work order instruction that required the work order
to be returned to work planners and revised if the original work scope is changed
or a problem is encountered. This directly contributed to three instrumentation
and control technicians receiving an unexpected radiation dose. A site stand-
down was held to discuss the lessons learned from the event. This was entered
into the licensees corrective action program as Condition
Report CR-CNS-2011-4428.
The finding was more than minor because it is associated with the human
performance attribute of the Occupational Radiation Safety Cornerstone and
affected the cornerstone objective to ensure the adequate protection of the
worker health and safety from exposure to radiation from radioactive material
during routine civilian nuclear reactor operation. The inspectors evaluated this
finding using Inspection Manual Chapter 0609, Appendix C, Occupational
Radiation Safety Significance Determination Process. The inspectors
determined that the finding is of very low safety significance (Green) because it
was not associated with ALARA planning or work controls, there was no
overexposure, there was no substantial potential for an overexposure, and the
licensees ability to assess dose was not compromised. The finding has a cross-
cutting aspect in the decision making component of the human performance area
because the licensee did not use conservative assumptions in decision-making.
Specifically, the licensee did not validate the assumptions made when
considering the change in work scope H.1(b)(Section 3.3).
- Green. The inspectors identified a noncited violation of Technical
Specification 5.7.2, for the failure to adequately brief radiation workers entering a
locked high radiation area. Specifically, the radiation protection pre-job briefing
failed to make workers knowledgeable of the radiation dose rates that may be
encountered when pulling the intermediate range monitor shuttle tube from under
the reactor pressure vessel and failed to identify any change in work scope or
breach of the nuclear instrument system. This resulted in the workers being
exposed to higher than expected dose rates. The workers immediately
evacuated the area and contacted radiation protection. The licensee held a site
stand-down to discuss lessons learned and this finding was entered into the
licensees corrective action as Condition Report CR-CNS-2011-04441.
The finding was more than minor because it is associated with the human
performance attribute of the Occupational Radiation Safety Cornerstone and
affected the cornerstone objective to ensure the adequate protection of the
worker health and safety from exposure to radiation from radioactive material
during routine civilian nuclear reactor operation because workers were exposed
to higher dose rates. The inspectors evaluated the significance of the finding
using NRC Inspection Manual 0609, Appendix C, Occupational Radiation Safety
Significance Determination Process, dated August 19, 2008. The inspectors
determined that the finding is of very low safety significance because it was not
associated with ALARA planning or work controls, there was no overexposure,
there was no substantial potential for an overexposure, and the licensees ability
-5- Enclosure
to assess dose was not compromised. In addition, the finding had a cross-
cutting aspect in the work control component of the human performance area
because the licensee did not appropriately communicate, coordinate, and
cooperate with each other during the radiation protection pre-job briefing and
failed to keep personnel apprised of plant conditions that may affect work
activities to ensure radiological safety was maintained H.3(b)(Section 4.3).
- Green. The inspectors identified a noncited violation of Technical
Specification 5.4.1(a), for the failure to follow Radiation Procedure 9.EN-RP-141,
Job Coverage, Revision 8. Specifically, the radiation protection personnel were
monitoring workers pulling the intermediate range monitor shuttle tube from
under the reactor pressure vessel and failed to implement radiation protection job
coverage requirements that resulted in the workers being exposed to dose rates
as high as 39 rem per hour at 30 centimeters from the tip of the shuttle tube. The
licensee immediately evacuated and restricted access to the area. This finding
was documented in the licensees corrective action program as Condition
Reports CR-CNS-2011-04442, CR-CNS-2011-04255, CR-CNS-2011-04595,
CR-CNS-2011-05443, CR-CNS-2011-05444, CR-CNS-2011-05446,
CR-CNS-2011-05447, and CR-CNS-2011-05448.
The finding was more than minor because it is associated with the human
performance attribute of the Occupational Radiation Safety Cornerstone and
affected the cornerstone objective to ensure the adequate protection of the
worker health and safety from exposure to radiation from radioactive material
during routine civilian nuclear reactor operation because workers were exposed
to higher dose rates. The inspectors evaluated the significance of the finding
using NRC Inspection Manual 0609, Appendix C, Occupational Radiation Safety
Significance Determination Process, dated August 19, 2008. The inspectors
determined that the finding is of very low safety significance because it was not
associated with ALARA planning or work controls, there was no overexposure,
there was no substantial potential for an overexposure, and the licensees ability
to assess dose was not compromised. In addition, the finding has a cross-cutting
aspect in the work practices component of the human performance area because
the licensee failed to use human error prevention techniques such as self-
checking and peer-checking to ensure that job coverage procedures were
followed H.4(a)(Section 4.4).
B. Licensee-Identified Violations
None.
-6- Enclosure
REPORT DETAILS
4OA3 EVENT FOLLOW-UP
1.0 Special Inspection Scope
On April 2, 2011, while licensee workers were removing an intermediate range monitor
shuttle tube assembly from the reactor pressure vessel, they deviated from the written
work instructions. Workers under the vessel received dose rate alarms and exited the
area. The workers dosimeters measured dose rates of 1.35 Rem per hour, 14.3 rem
per hour, and 763 millirem per hour.
The inspection charter (refer to Attachment 2) required the team to: (1) develop a
timeline for the sequence of events, including actions taken prior to and post event, as
well as the associated decision-making process, (2) assess the licensees procedural
compliance during work order preparation and execution, including ALARA and
radiological work permit considerations, (3) characterize the area dose rates and dose
received by personnel, (4) review causal determination and short term corrective action
adequacy, and (5) review previous activity performance and compare to current activity
performance.
The team performed their reviews in accordance with NRC Inspection Procedure 93812,
"Special Inspection Procedure." The team used the requirements in 10 CFR
Parts 19, 20 and 50, the licensees technical specifications, and the licensees
procedures required by technical specifications as criteria for determining compliance.
The team reviewed licensee procedures, corrective action documents, as well as work
orders and radiological work permits for the maintenance activity. The team interviewed
station personnel regarding the events, compared this event to previously performed
evolutions, and assessed the adequacy of the licensees corrective actions. A list of
specific documents reviewed is provided in Attachment 1. The charter for the special
inspection is provided as Attachment 2.
2.0 Event Description and Chronology
2.1 Event Summary
On April 2, 2011, instrumentation and control technicians prepared to remove the source
range monitor B and the intermediate range monitor C shuttle and dry tube assemblies
from the top of the reactor vessel in accordance with Work Orders 4741006
and 4741002. Each intermediate range monitor and source range monitor detector is
contained inside a shuttle tube. This shuttle tube is fixed to a drive tube that a drive
mechanism moves up and down inside a dry tube. The entire tube assembly, along with
the detector was to be replaced. Arrangement of these components is depicted in
Attachment 3.
During the pre-job brief, the workers discussed the activity to be performed and the tools
needed. Nose cones were identified as a needed component, but the type of nose
cones (male or female threads) and their location was unknown. In particular, workers
discussed whether the shuttle tube could be removed from the bottom of the core, rather
than the top (as procedurally directed), however no resolution was achieved.
-7- Enclosure
Prior to beginning work, two nose cones were located, one male-threaded and one
female-threaded. The workers proceeded to conduct the work activity. Source range
monitor B drive tube was removed and the male-threaded nose cone was installed on
the bottom of the dry tube, sealing the shuttle tube inside. Intermediate range monitor C
drive tube was removed, but no other male-threaded nose cones were available to install
on the lower end of the dry tube prior to removal.
When the workers reported this issue to the outage control center, the outage control
center staff gave permission to the workers to remove the shuttle tube from the bottom
of the reactor vessel, rather than sealing it inside the dry tube for removal above the
reactor vessel (as originally planned). The licensee did not modify either the associated
work order or the corresponding radiation work permit to reflect this change.
As the workers removed the shuttle tube from the bottom of the vessel, the three
workers under the vessel received dose rate alarms. The workers then set the tip of the
tube on the floor at the 888 foot elevation and exited the area. The workers dosimeters
measured dose rates of 1.35 rem per hour, 14.3 rem per hour, and 0.763 rem per hour.
Surveys taken of the shuttle tube, during recovery operations, found that the tip of the
shuttle tube measured 3,226 rem per hour on contact and 39 rem per hour at
30 centimeters, and that the general area dose rate was 4.6 rem per hour at waist level,
increasing to 8.6 rem per hour at waist level near the tube.
2.2 Sequence of Events
December 2009 - The licensee identified the need to replace intermediate range
monitor C. Work Order 4741002 was generated to replace intermediate range monitor C
components. The work order contained, in part, the following actions:
- Remove dry tube, shuttle/drive tube, and detector from top of reactor
- Install new dry tube and shuttle/drive tube from top of reactor
- Install new detector
January 2010 - A planner was assigned for Work Order 4741002.
May 2010 - Planning for Work Order 4741002 begins.
November 2010 - The ALARA review of Work Order 4741002 was deemed satisfactory.
December 2010 - Instrumentation and control supervisory walkdown of Work
Order 4741002 was completed (not the same supervisor that performed the job).
February 2011 - The instrumentation and control lead technician completed shop
walkdown of Work Order 4741002. This completed planning of the work order. The
same technician later performed the job.
April 2, 2011 - (times approximated)
1730 - During maintenance supervisor turnover, the off-going supervisor
identified that the day-shift crew had pulled the detectors, and was ready to
remove the drive mechanisms and tube assemblies.
-8- Enclosure
1830 - The outage control center brief identifies that the source range monitor
and intermediate range monitor work was a priority. The backup plan to
remove the shuttle tube from under the vessel is not discussed at the brief.
2000 - Outage control center called instrumentation and control superintendent
to get status of locating nose cones for source range monitor and intermediate
range monitor work. Only one nose cone located. Outage control center
inquired about an alternate plan. Back-up plan was to pull the shuttle tube
manually from below. Instrumentation and control told outage control center that
the tube was made of titanium (easily bendable and nonirradiated). Outage
control center requested instrumentation and control to brief radiation protection
and ensure they understood and approved the back-up plan.
2030 - Inspection of top guide was complete, so refuel floor staff would be ready
for dry tube removal at midnight.
2130 - Instrumentation and control superintendent informed that one nose cone
had been located. Refuel floor manager looks for another nose cone.
2200 - The lead instrumentation and control technician conducted a shop brief
for upcoming under-vessel work on the source range and intermediate range
monitors. Attendees were the three technicians and instrumentation and control
supervisor. The feasibility of pulling the shuttle tube from below the vessel is
discussed, but no resolution was achieved.
2206 - Second nose cone located. However, the threads on this nose cone did
not match the threads on the other nose cone, so technicians head to the drywell
with one male-threaded and one female-threaded nose cone. They were not
sure which would be needed.
2230 - The radiation protection ALARA supervisor completed a brief for the
upcoming under-vessel work. Besides the supervisor providing the brief,
attendees were the three technicians who would work under-vessel and another
technician that would remain located outside the drywell to monitor radiological
conditions. This brief did not discuss the backup plan for shuttle tube removal
from the bottom of the vessel.
2300 - Source range monitor B work completed. Work on intermediate range
monitor C begins. Shortly afterward, technicians call instrumentation and control
superintendent to inform that intermediate range monitor C required male
threaded nose cone (like the ones used on source range monitor B), and
requested guidance on removal of shuttle tube without the proper nose cone.
Instrumentation and control superintendent then calls outage control center
maintenance outage manager to request guidance on removing the shuttle tube
from under-vessel. Outage control center maintenance outage manager directed
instrumentation and control superintendent to proceed with removal from under-
vessel. Instrumentation and control superintendent relayed this direction to the
under-vessel technicians.
2400 - Instrumentation and control technicians call their superintendent again to
confirm removal of the shuttle tube from under-vessel. The technicians express
concern that removing the shuttle tube from under the vessel would require
-9- Enclosure
bending the tube and would therefore be irreversible. The instrumentation and
control superintendant requested and received confirmation from outage control
center maintenance outage manager and then related that confirmation to the
under-vessel technicians. Instrumentation and control technicians begin
removing shuttle tube from under-vessel.
April 3, 2011 - (times approximated)
0000 - Outage control center maintenance outage manager informed rest of
outage control center that shuttle tube would be pulled from below vessel. He
stated that the tube was assumed to be titanium, and therefore, would not
activate. However, this assumption was not verified and it turned out that the
shuttle tube was actually stainless steel. Outage control center radiation
protection representative challenged the assumption that titanium would not
activate. During the discussion, the three technicians working under the vessel
received dose rate alarms, immediately evacuated the under-vessel area and
told radiation protection personnel in the area that dose rates had significantly
increased.
0047 - The licensee entered their emergency procedure for elevated radiological
conditions inside the primary containment under-vessel area and drywell access
was restricted.
2.3 Immediate Actions Taken
Upon receiving the electronic dosimeter alarms, the workers immediately evacuated the
drywell. The licensee immediately evacuated all personnel from the drywell, restricted
access to the drywell, and entered Emergency Procedure 5.1RAD, Building Radiation
Trouble, Revision 15, due to unexpected elevated dose rates. The licensee
implemented radiological emergency procedures which identified the source as the
intermediate range monitor shuttle tube that was removed from the reactor pressure
vessel by the workers. The licensee implemented a recovery plan to isolate the source
of radiation and secure it in a shielded lead container. The recovery plan was executed
by three radiation protection technicians who were knowledgeable of the radiological
conditions. The plan included identifying the highest dose rates in the area, which was
the tip of the shuttle tube, and quickly cutting the stainless steel shuttle tube with metal
cutters and securing the approximately one foot piece of the shuttle tube. The remaining
tube was also cut up into approximately one foot pieces and secured in the shielded
container. The shielded container was then placed safely in the spent fuel pool.
3.0 Work Planning and Execution
3.1 Work Order Planning
a. Scope
The inspectors assessed the licensees performance while planning and preparing the
work package to replace source range monitor B and intermediate range monitor C. The
inspectors conducted interviews to assess the knowledge level and qualifications of
planners. The inspectors examined procedural guidance for work package creation to
determine adequacy and completeness. The inspectors also evaluated the licensees
ability to appropriately characterize and compensate for the risk associated with the
- 10 - Enclosure
maintenance activity. Interactions with other working groups, such as operations and
radiation protection, were similarly reviewed. The inspectors also evaluated the
licensees review process to ensure that work packages are complete and accurate.
b. Findings
.1 Introduction. The inspectors identified a Green noncited violation of Technical
Specification 5.4.1, for a failure to implement procedures described in Regulatory
Guide 1.33, Appendix A. Specifically, the licensee failed to implement procedures that
provide guidance on creating and reviewing clear, accurate work instructions. As a
result, the work instructions could not have been completed as written.
Description. On April 3, 2011, three instrumentation and control technicians were
performing Work Order 4741002, to remove intermediate range monitor C from the
underside of the reactor vessel and prepare the dry tube assembly for later removal from
above the reactor vessel. Work Order 4741002, Step 4, directed the technicians to
remove the drive tube per Procedure 14.2.9, SRM/IRM Detector and Drive Tube
Removal, Installation, Testing, and SRM/IRM Troubleshooting, Revision 26. Step 5.8 of
Procedure 14.2.9 directed the technicians to screw a nose cone onto the bottom of the
dry tube, enclosing the shuttle tube within the dry tube. Work Order 4741002, Step 5,
then directed the technicians to remove shuttle tube. However, if Procedure 14.2.9
had been correctly followed, Step 5 of the work order could not have been performed
because the shuttle tube would have been contained within the dry tube. Additionally,
no procedural reference is given for shuttle tube removal. When the inspectors asked
the work planner to clarify the intent of the unclear work instructions, he was unable to
provide any clarification.
Procedure 0.40.4, Planning, Revision 16, Attachment 1, included a checklist that was
to be used to ensure that work instructions were clear and concise. The inspectors
attempted to review this checklist since it was used to plan the work package. However
the checklist had been discarded. Use of the Attachment 1 checklist failed to identify
instructions that were not only unclear to the workers, but also to the work planner.
Additionally, Procedure 0.40.4, Step 5.2.10, required the work planner to ensure that all
specialized tools required to perform the work are identified and available. The nose
cone was not listed as a required part for work order execution and was not readily
available. If the nose cone had been made available (as required by procedure), the
technicians may not have attempted to execute an unclear instruction. This is evidenced
by the technicians performance of Work Order 4741006, removal of source range
monitor B, which contained the same unclear instructions. However, during this work,
the technicians had a male-threaded nose cone, so no attempt was made to remove the
shuttle tube from the bottom.
Another potential barrier to prevent the unclear work order instructions from reaching the
field was provided in Procedure 0.40, Work Control Program, Section 6.1, which
directed an instrumentation and control shop walkdown of the work instructions in
accordance with Procedure 0.40.4, Attachment 7. This attachment contained another
checklist for verifying work instruction. Inspectors attempted to review this checklist to
assess its performance; however this checklist was also discarded. The inspectors
concluded that despite the licensees assurance that the checklist was correctly utilized,
use of the checklist during the shop walkdown failed to identify the unclear work
instructions and lack of necessary parts.
- 11 - Enclosure
Although each individual requirement, if correctly performed, may or may not have
singularly prevented the confusing and incomplete work package from reaching final
approval, together they provide defense-in-depth; a set of guidelines that are intended to
provide multiple opportunities to detect and correct poor work instructions prior to field
execution. The failure of all three of these steps allowed poor work instructions to be
approved.
These unclear work instructions and lack of a needed part contributed to the decision to
remove shuttle tube from the bottom, despite lack of adequate procedural guidance.
As a result, three instrumentation and control technicians received an unexpected
radiation dose.
In interviews with station personnel, the inspectors encountered indications of a
widespread attitude among workers that failures to follow procedures were acceptable if
they achieved the desired outcomes. In those interviews, the inspectors found no
evidence that the licensee had effectively communicated their expectations regarding
procedural compliance. Also, the licensees root cause evaluation, documented as
Condition Report CR-CNS-2011-03763, determined that one root cause of this finding
was a work culture, supported by institutional reinforcement, that unacceptable
behaviors are acceptable as long as the outcome is good.
Analysis. The performance deficiency is that the licensee did not follow Procedure 0.40,
Work Control Program, and Procedure 0.40.4, Planning, when preparing Work Order 4741002. As a result, the work order could not be performed as written. This
deficiency was reasonable for the licensee to foresee and prevent occurrence. The
finding is more than minor because it is associated with the human performance attribute
of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective
to ensure the adequate protection of the worker health and safety from exposure to
radiation from radioactive material during routine civilian nuclear reactor operation, in
that the finding resulted in three technicians receiving an unexpected radiation dose.
The inspectors evaluated the significance of the finding using NRC Inspection
Manual 0609, Appendix C, Occupational Radiation Safety Significance Determination
Process, dated August 19, 2008. The inspectors determined that the finding is of very
low safety significance (Green) because it is not a finding related to ALARA planning or
work controls, it did not result in an overexposure, there was no substantial potential for
overexposure, and the licensees ability to assess dose was not compromised. The
inspectors determined that the apparent cause of this finding was the licensees failure
to correct the attitude among workers that failures to follow procedures were acceptable
if they achieved the desired outcomes. Therefore, the finding has a cross-cutting aspect
in the work practices component of the human performance area because the licensee
did not effectively communicate expectations regarding procedural compliance [H.4.(b)].
Enforcement. Technical Specification 5.4.1 requires the licensee to establish,
implement, and maintain procedures described in Regulatory Guide 1.33, Appendix A.
Appendix A, Section 9, requires, in part, that maintenance that can affect the
performance of safety related equipment should be properly preplanned in accordance
with written procedures appropriate to the circumstances. Licensee Procedures 0.40
and 0.40.4 are similar to those described in Section 9.
Contrary to the above, on April 3, 2011, the licensee did not correctly implement the
above procedures by not properly preplanning maintenance in accordance with written
- 12 - Enclosure
procedures appropriate to the circumstances. Specifically, despite the guidance outlined
in Procedures 0.40 and 0.40.4, the licensee developed a work instruction that did not list
the needed tools and could not be followed as written. As a result, three instrumentation
and control technicians received an unexpected radiation dose. A site stand-down was
held to discuss the lessons learned from the event. Because this was of very low safety
significance and it was entered into the corrective action program as Condition
Reports CR-CNS-2011-4431, CR-CNS-2011-4581, CR-CNS-2011-4582,
CR-CNS-2011-4583, CR-CNS-2011-4584, and CR-CNS-2011-4585, this violation is
being treated as a noncited violation, consistent with Section 2.3.2 of the Enforcement
Policy: NCV 05000298/2011008-01, Unclear Work Instructions.
.2 Introduction. The inspectors identified a Green noncited violation of Technical
Specification 5.4.1, for a failure to implement procedures described in Regulatory
Guide 1.33, Appendix A. Specifically, the licensee failed to implement procedures that
provide guidance on recognizing risk associated with a maintenance activity and
properly accounting for that risk.
Description. On April 3, 2011, three instrumentation and control technicians executed
Work Order 4741002, intended to remove the intermediate range monitor C assembly
from the underside of the reactor vessel. The work order then directed the dry tube
assembly replacement. During this activity, the dry tube assembly should have been
pulled out from above the reactor vessel, creating a hole under the vessel, so a water
seal cap was to be installed under the reactor vessel to prevent reactor coolant from
draining out. This water seal cap becomes the new reactor coolant system pressure
boundary. Correct installation of this cap is critical, since any installation error could
induce a reactor coolant leak under the vessel and create a potential to drain the reactor
vessel.
The activities performed in Work Order 4741002 introduced a high level of risk to the
safe operation of the plant. Procedure 0.40, Work Control Program, Revision 75,
Section 5.7, provides examples of when a work package should be characterized as a
detailed (Level 1) work order. Because of the risk introduced, the procedure required
that Work Order 4741002 be considered a detailed work order, requiring a peer review
by both planning and engineering departments and including operational experience.
However, it was incorrectly characterized as a simple (Level 2) work order, so no
additional reviews were completed and site-specific operating experience was not
included in the work package.
Additionally, Procedure 0.40.4, Planning, Revision 16, Section 5.2.19, required the
plant impact to be determined by using Attachment 3, which includes a checklist. The
overall risk to the plant is then documented in the work package. The inspectors
attempted to review this checklist to assess its performance, however the copy of the
checklist had been discarded. The inspectors concluded that the checklist failed to
correctly categorize the risk associated with the work activity. The resulting plant impact
statement not only incorrectly stated that this work does not introduce unusual hazards
or risks and has no impact on the plant, but also incorrectly stated that the work could
be performed in Mode 4 or 5. This work can only be performed in Mode 5.
As a result of the unrecognized risk, additional barriers to ensure a quality work package
were bypassed. The resulting work package contained unclear work instructions that
could not be performed as written, did not contain a complete listing of parts needed to
- 13 - Enclosure
perform the activity, and did not contain appropriate operating experience. These
deficiencies contributed to the decision to remove the shuttle tube from under the vessel,
despite lack of actual procedural guidance. As a result, three instrumentation and
control technicians received an unexpected radiation dose. A site stand-down was held
to discuss the lessons learned from the event.
Analysis. The performance deficiency is that the licensee did not follow Procedure 0.40,
Work Control Program, and Procedure 0.40.4, Planning, when preparing Work Order 4741002 to determine the risk associated with the maintenance activity. The
resulting failure to recognize the associated risk led to the package being incorrectly
characterized as a simple work order, rather than a detailed work order, and the work
order was not given the appropriate level of attention or review. This deficiency was
reasonable for the licensee to foresee and prevent occurrence. The finding is more than
minor because it is associated with the human performance attribute of the Occupational
Radiation Safety Cornerstone and affected the cornerstone objective to ensure the
adequate protection of the worker health and safety from exposure to radiation from
radioactive material during routine civilian nuclear reactor operation, in that the finding
resulted in three technicians receiving an unexpected radiation dose. The inspectors
evaluated the significance of the finding using NRC Inspection Manual 0609,
Appendix C, Occupational Radiation Safety Significance Determination Process, dated
August 19, 2008. The inspectors determined that the finding is of very low safety
significance (Green) because it is not a finding related to ALARA planning or work
controls, it did not result in an overexposure, there was no substantial potential for
overexposure, and the licensees ability to assess dose was not compromised. The
inspectors determined that the apparent cause of this finding was the licensees failure
to ensure workers recognize the value of incorporating risk insights into plans for
maintenance activities. Therefore, the finding has a cross-cutting aspect in the work
control component of the human performance area because the licensee did not plan
work activities by incorporating risk insights H.3(a).
Enforcement. Technical Specification 5.4.1 requires the licensee to establish,
implement, and maintain procedures described in Regulatory Guide 1.33, Appendix A.
Appendix A, Section 9 requires, in part, that maintenance that can affect the
performance of safety related equipment should be properly preplanned in accordance
with written procedures appropriate to the circumstances. Licensee Procedure 0.40 and
Procedure 0.40.4 are similar to those described in Section 9. Contrary to the above, on
April 3, 2011, the licensee did not correctly implement the above procedures.
Specifically, despite the guidance contained in Procedures 0.40 and 0.40.4, the licensee
developed a work instruction that failed to recognize the risk associated with the activity
and failed to develop risk mitigation strategies. This activity had the potential to drain the
reactor vessel. Because this was of very low safety significance and it was entered into
the corrective action program as Condition Reports CR-CNS-2011-4435 and
CR-CNS-2011-4436, this violation is being treated as a noncited violation, consistent
with Section 2.3.2 of the Enforcement Policy: NCV 05000298/2011008-02, Failure to
Recognize Work Order Risk.
- 14 - Enclosure
3.2 Job Preparation
a. Scope
The inspectors assessed the licensees preparations to perform the work to replace
source range monitor B and intermediate range monitor C. This included conducting
interviews with all personnel present at the pre-job brief to determine the workers level
of understanding of the job to be performed, as well as determine the workers
procedural compliance. Procedural guidance was reviewed for adequacy and
completeness. Interactions with other working groups, such as operations and radiation
protection, were similarly reviewed. Operating experience was also reviewed to
determine the licensees efforts to incorporate and institutionalize the information.
b. Findings
Introduction. The inspectors identified a Green finding for a failure to implement human
performance procedures. Specifically, the licensee failed to implement procedures that
provided guidance on conducting pre-job briefs, preparing work in the field, and
informing technicians on what to do when the workers encountered a problem. As a
result, workers were uncertain how to proceed, especially when needed parts were not
available.
Description. On April 3, 2011, instrumentation and control technicians prepared to
perform Work Order 4741002 by conducting a pre-job brief. The brief was conducted by
the lead technician, with two other technicians and the supervisor present. Neither
technicians nor supervisor had previously performed this activity. The supervisor had
glanced at the work package, but was not familiar with it. Procedure 0-HU-Tools,
Human Performance Tools, Revision 17, Attachment 8, provides guidance on how to
conduct pre-job briefs. Attachment 8, the section entitled How To Do It, lists seven
steps for conducting the brief. Several of those steps were not adequately completed as
follows:
- Step 1 expected the briefer to have a thorough understanding of every aspect of
the activity, however the lead technician conducting the pre-job brief was not
sure which nose cones were needed, whether the correct nose cones were
readily available, and how the activity would proceed if the correct nose cones
could not be located. The pre-job brief was completed with these questions still
unanswered. The technicians believed they would figure it out after the work
began.
- Step 2 expected that the pre-job brief include all individuals participating in the
activity and anyone significantly impacted by the activity. The work activity
affected instrumentation and control technicians, radiation protection personnel,
and the outage control center staff. Additional work in the same package also
affected a contractor work group. Representatives from those other work groups
were not present at the instrumentation and control shop pre-job brief. A
separate pre-job brief was held with radiation protection personnel, but the level
of detail and focus of the discussion was different from that of the shop pre-job
brief.
- Step 3 expected the licensee to review operation experience during pre-job
briefs. One example of operating experience from another site was discussed,
- 15 - Enclosure
but relevant site specific operating experience from 1993 and 1994 was omitted.
The omitted operating experience described how workers received higher-than-
expected doses when a shuttle tube was removed from under the vessel in 1993.
Since the possibility of actually removing the shuttle tube from under the vessel
was discussed during the pre-job, this operating experience may have provided a
prompt to alert the technicians that shuttle tube removal from under the vessel
would elevate dose level and potentially dissuade them from working outside the
procedure.
- Steps 4 and 5 directed that Procedure 2.01.1, Conduct of Infrequently
Performed Tests or Evolutions, Revision 5, be used. However, this procedure
was not used. Additionally, the pre-job brief checklist directed attention to
potential error traps, such as time pressure and task unfamiliarity, but checklist
identification of these traps failed to prevent an error from occurring.
Work began after the brief was complete. During performance of Work Order 4741002,
the technicians determined that not all the needed parts were present, so a step in the
procedure could not be performed. The technicians stopped work and spoke with the
instrumentation and control supervision, who gave the workers verbal direction. This
direction included marking the procedural step as a discrepancy and continuing work
via an undocumented, unapproved back-up plan discussed at the pre-job brief.
In interviews with station personnel, the inspectors encountered indications of a
widespread attitude among workers that failures to follow procedures were acceptable if
they achieved the desired outcomes. In those interviews, the inspectors found no
evidence that the licensee effectively communicated their expectation regarding
procedural compliance. Also, as documented in Condition Report CR-CNS-2011-03763,
the licensees root cause evaluation determined that one root cause of this event was a
work culture, supported by institutional reinforcement, that unacceptable behaviors are
acceptable, as long as the outcome was good.
Analysis. The performance deficiency is that the licensee did not follow
Procedure 0-HU-Tools while preparing for and executing Work Order 4741002. As a
result, the technicians incorrectly continued work when the needed parts were not
available, rather than stopping work. This deficiency was reasonable for the licensee to
foresee and prevent occurrence. The finding is more than minor because it is
associated with the human performance attribute of the Occupational Radiation Safety
Cornerstone and affected the cornerstone objective to ensure the adequate protection of
the worker health and safety from exposure to radiation from radioactive material during
routine civilian nuclear reactor operation, in that the finding resulted in three technicians
receiving an unexpected radiation dose. The inspectors evaluated the significance of
the finding using NRC Inspection Manual 0609, Appendix C, Occupational Radiation
Safety Significance Determination Process, dated August 19, 2008. The inspectors
determined that the finding is of very low safety significance (Green) because it is not a
finding related to ALARA planning or work controls, it did not result in an overexposure,
there was no substantial potential for overexposure, and the licensees ability to assess
dose was not compromised. The inspectors determined that the apparent cause of this
finding was the licensees failure to promote the use of human performance tools to
ensure job tasks were properly completed. Therefore, this finding has a cross-cutting
aspect in the work practices component of the human performance area because the
- 16 - Enclosure
licensee did not adequately communicate human error prevention techniques such that
work activities are completed safely H.4(a).
Enforcement. This finding does not involve enforcement action because no regulatory
requirement was violated: FIN 05000298/2011008-03,Failure to Implement Human
Performance Procedure.
3.3 Work Execution
a. Scope
The inspectors assessed the licensees execution of the work to replace source range
monitor B and intermediate range monitor C. This included reviewing the work order,
procedures, and conducting interviews with all personnel present at the job site, as well
as the decision-makers in the outage control center. The inspectors assessed the
workers and managers level of understanding of the job activity and any contingency
plans or abort criteria. The inspectors reviewed procedural guidance for adequacy and
completeness, and assessed the licensees in-field procedural compliance.
Maintenance practices demonstrated by in-field workers were compared to the
licensees expectations for maintenance activities.
b. Findings
Introduction. The inspectors identified a Green noncited violation of Technical
Specification 5.4.1, for a failure to comply with procedures described in Regulatory
Guide 1.33, Appendix A. Specifically, the licensee failed to implement procedures and a
work order instruction that required the work order to be returned to work planners and
revised if the original work scope is changed or a problem is encountered.
Description. On April 3, 2011, three instrumentation and control technicians were
implementing Work Order 4741002, to remove intermediate range monitor C from the
underside of the reactor vessel and prepare the dry tube assembly for later removal from
above the reactor vessel. Work Order 4741002, Step 4, directed the technicians to
remove the drive tube per Procedure 14.2.9, SRM/IRM Detector and Drive Tube
Removal, Installation, Testing, and SRM/IRM Troubleshooting, Revision 26. Step 5.8 of
Procedure 14.2.9 directed the technicians to install a male-threaded nose cone that is
screwed onto the bottom of the dry tube, enclosing the shuttle tube within the dry tube.
Without this nose cone, the shuttle tube would fall out the bottom of the dry tube and
remain in the reactor vessel when the dry tube is removed from the top. While
performing this task, the technicians determined that the nose cone was not available.
The technicians discussed the inability to continue following the work instructions with
their supervisor. A nonconservative decision was made to pull the shuttle tube out of the
core from the bottom of the vessel rather than enclosing it in the dry tube assembly as
originally directed by the work package. This nonconservative decision was based on
unvalidated assumptions, such as shuttle tube material, expected dose rates, and
instrumentation and control familiarity with the plan change. The technicians pulled the
shuttle tube from the bottom and exposed a 3,226 rem per hour source. The
technicians dosimetry alarmed and they exited the area. As a result, three
instrumentation and control technicians received an unexpected radiation dose. A site
stand-down was held to discuss the lessons learned from the event.
- 17 - Enclosure
Work Order 4741002, Prerequisite 2, stated in part, that if during the performance of the
work order, should problems arise, workers should stop work and return the work
package to planning for revision before proceeding with work.
This guidance is congruent with two other site procedures governing procedural
compliance. Site Procedure 0.40, Work Control Program, Revision 75, Step 7.4.7,
states that if work cannot be performed as written, the worker shall stop work and
contact the supervisor, who assesses the type of change needed in accordance with
Procedure 0.40.4, Planning. Procedure 0.40.4, Revision 16, Step 5.4.1, required that a
work order revision was required if the work scope changes. In this case, the work
scope could not be completed as stated and the licensee made the decision to change
the work scope by pulling the shuttle tube from the bottom, rather than remaining within
the dry tube assembly. Additionally, Procedure 7.0.4, Conduct of Maintenance,
Revision 32, Step 10.2.3, also states that changes in intent of work activities performed
should not be made without changes to the original controlling document (work order).
Despite similar procedural guidance located in different locations, the nonconservative
decision was made to pull the shuttle tube from the bottom of the vessel, rather than
revising the work package as procedurally directed.
Analysis. The performance deficiency is that the licensee did not follow Procedure 0.40,
Work Control Program, and Procedure 7.0.4, Conduct of Maintenance, when Work Order 4741002 could not be performed as written. Work Order 4741002 also included
instructions that required the work package to be sent back to planning to be revised, if
problems arose during work order performance. This deficiency was reasonable for the
licensee to foresee and prevent occurrence. The finding is more than minor because it
is associated with the human performance attribute of the Occupational Radiation Safety
Cornerstone and affected the cornerstone objective to ensure the adequate protection of
the worker health and safety from exposure to radiation from radioactive material during
routine civilian nuclear reactor operation, in that the finding resulted in three technicians
receiving an unexpected radiation dose. The inspectors evaluated the significance of
the finding using NRC Inspection Manual 0609, Appendix C, Occupational Radiation
Safety Significance Determination Process, dated August 19, 2008. The inspectors
determined that the finding is of very low safety significance (Green) because it is not a
finding related to ALARA planning or work controls, it did not result in an overexposure,
there was no substantial potential for overexposure, and the licensees ability to assess
dose was not compromised.
The inspectors determined that the apparent cause of this finding was the licensees
failure to ensure that risk-significant changes to the work orders were made only through
established processes. Therefore, this finding has a cross-cutting aspect in the decision
making component of the human performance area because the licensee did not use a
systematic process to make the risk-significant decision to deviate from work
instructions H.1(b).
Enforcement. Technical Specification 5.4.1 requires the licensee to establish,
implement, and maintain procedures described in Regulatory Guide 1.33, Appendix A.
Appendix A, Section 9 requires, in part, that maintenance that can affect the
performance of safety related equipment should be properly preplanned in accordance
with written procedures appropriate to the circumstances. Licensee Procedure 0.40,
Procedure 7.0.4, and maintenance Work Order 4741002 are similar to those described
- 18 - Enclosure
in Section 9, in that, they required work orders that could not be performed as written to
be returned to planning for revision. Contrary to the above, on April 3, 2011, the
licensee did not correctly implement the above procedures. Specifically, the licensee
failed to return the work package to planning for a revision when the work order could
not be performed as written and when workers changed the intended work scope. As a
result, three instrumentation and control technicians received an unexpected radiation
dose. A site stand-down was held to discuss the lessons learned from the event.
Because this was of very low safety significance and it was entered into the corrective
action program as Condition Reports CR-CNS-2011-4428, CR-CNS-2011-4581,
CR-CNS-2011-4582, CR-CNS-2011-4583, CR-CNS-2011-4585, CR-CNS-2011-4591,
and CR-CNS-2011-4592, this violation is being treated as a noncited violation,
consistent with Section 2.3.2 of the Enforcement Policy: NCV 05000298/2011008-04,
Failure to Revise Unclear Work Instructions.
4.0 Radiation Protection Performance
4.1 ALARA Planning
a. Scope
The inspectors assessed the licensees performance while developing the ALARA work
package to replace source range monitor B and intermediate range monitor C. The
inspectors conducted interviews to assess the knowledge level and qualifications of
ALARA planners. The inspectors examined the adequacy and completeness of
procedural guidance for developing ALARA work packages. Interactions between
radiation protection, maintenance, and operations were reviewed to determine if ALARA
planning was performed with appropriate coordination and communication. The
inspectors also evaluated the licensees review process to ensure that work packages
are complete and accurate.
b. Findings
No findings were identified. The inspectors determined that the ALARA planning for the
job was completed adequately for removing the source range monitor and intermediate
range monitor through the top of the reactor vessel. However, since the workers
changed plans during the execution of the plan and did not seek a revision to the ALARA
plan, a finding was identified in the area of work execution (see Section 3.3).
4.2 Radiation Work Permit Adequacy
a. Scope
The inspectors assessed the licensees performance with respect to maintaining
occupational individual and collective radiation exposures ALARA. The inspectors used
the requirements in 10 CFR Part 20, the technical specifications, and the licensees
procedures required by technical specifications as criteria for determining compliance.
The inspectors reviewed the licensees previous experience with similar jobs, historical
information regarding doses received, and historical and current survey data used to
establish the radiological conditions of the radiation work permit including dose and dose
rate alarm setpoints.
- 19 - Enclosure
b. Findings
No findings were identified. The inspectors determined that the radiation work permit
was adequate for the original plan to remove the source range monitor and intermediate
range monitor through the top of the reactor vessel. However, since the workers
changed plans and did not seek a revision to the radiation work permit, a finding was
identified in the area of work execution (see Section 3.3).
4.3 ALARA Briefing
a. Scope
The inspectors assessed the licensees ALARA briefing of workers preparing to enter the
drywell to perform the work to replace source range monitor B and intermediate range
monitor C. The inspection included conducting interviews with personnel in attendance
at the pre-job ALARA briefing to determine the workers level of understanding of the job
to be performed, as well as, determine if the workers were appropriately briefed per high
radiation area technical specifications and licensee procedures. The inspectors
reviewed the licensees radiation work permit and high radiation area briefing sheets to
determine if the licensee had adequately assessed the scope of the job to be performed.
The inspectors reviewed the licensees implementation of the requirements of 10 CFR
Parts 19 and 20.
b. Findings
Introduction. The inspectors identified a Green noncited violation of Technical
Specification 5.7.2, for the failure to adequately brief radiation workers entering a locked
high radiation area. Specifically, on April 2, 2011, the radiation protection pre-job
briefing failed to discuss radiation dose rates that may be encountered when pulling the
intermediate range monitor shuttle tube from under the reactor pressure vessel and did
not identify any scope change or breach of the under-vessel nuclear instrument system.
Description. On April 2, 2011, three instrumentation and control technicians were
provided with an ALARA pre-job briefing by radiation protection personnel for entry into a
high radiation area to perform work on special (radiation) Work Permit 2011-422. The
job scope included removing intermediate range monitor C shuttle tube from the bottom
of the reactor pressure vessel. The shuttle tube was highly radioactive because it had
been in the reactor core. The ALARA briefing provided information to the workers about
general area dose rates and electronic dosimetry alarm setpoints. However, the ALARA
briefing did not provide dose rates that would be encountered when removing the shuttle
tube because the radiation protection personnel providing the ALARA briefing did not
have an understanding of the full scope of the job and did not ask any questions to
clarify or confirm the full scope of the job. Therefore, the ALARA briefing did not make
workers knowledgeable about the dose rates they would encounter during the job. As a
result, when the workers removed the shuttle tube from the bottom of the vessel,
radiation levels of 3,226 rem per hour on contact with the tip of the shuttle tube and
39 rem per hour at 30 centimeters, as measured later by an AMP-200 detector, were
encountered. The workers electronic dosimetry alarmed and they immediately left the
area and contacted radiation protection personnel.
The inspectors interviewed radiation protection personnel, the three workers, and other
site personnel involved in the event. The inspectors reviewed the special work permit
- 20 - Enclosure
requirements, surveys used during the ALARA briefing, and the radiation protection
briefing form used for the ALARA briefing. The inspectors determined that the ALARA
briefing form indicated no system breach was to be performed during this job, however,
that was not true because the workers planned to breach the incore nuclear instrument
system. The ALARA briefing did not cover a system breach of the nuclear instrument
system, even though it was originally planned. The ALARA briefer lacked a questioning
attitude with respect to gaining an understanding of the full scope of the work activity that
the technicians were about to perform. The briefer did not question the special work
permit dose setpoints that were set at 300 and 600 millirem/hr even though the ALARA
briefing form indicated dose rates in the area of 80-120 millirem/hr. Additionally, there
was no discussion or review of relevant Cooper Nuclear Station operating experience,
which would have identified that high dose rates would be encountered during the
performance of this work activity.
The inspectors determined that the pre-job ALARA briefing was inadequate because the
workers were not made knowledgeable of the dose rates in a high radiation area while
performing the activities they had planned as required by Technical Specification 5.7.2.
The inspectors also determined that the licensee failed to appropriately communicate,
coordinate, and cooperate with each other during the ALARA pre-job briefing and to
keep personnel apprised of plant conditions that may affect work activities to ensure
radiological safety was maintained.
Analysis. The failure to perform an adequate ALARA briefing to make workers
knowledgeable of the dose rates in the work area is a performance deficiency. The
finding is more than minor because it is associated with the human performance attribute
of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective
to ensure the adequate protection of the worker health and safety from exposure to
radiation from radioactive material during routine civilian nuclear reactor operation, in
that the finding resulted in three technicians receiving an unexpected radiation dose.
The inspectors evaluated the significance of the finding using NRC Inspection
Manual 0609, Appendix C, Occupational Radiation Safety Significance Determination
Process, dated August 19, 2008. The inspectors determined that the finding is of very
low safety significance (Green) because it is a finding related to ALARA planning or work
controls, but the licensees three year rolling average for collective dose is less than
240 person-rem. The inspectors determined that the apparent cause of this finding was
that the licensee had not encouraged interdepartmental communication and coordination
between workers to ensure that workers were properly prepared to begin work activities.
Therefore, this finding has a cross-cutting aspect in the work control component of the
human performance area because the licensee did not incorporate actions to address
the need for work groups to communicate, coordinate, and cooperate with each other
during activities in which interdepartmental coordination is necessary to assure human
performance, in that the licensee did not address the need for work groups to
communicate, coordinate, and cooperate with each other during the ALARA pre-job
briefing, which was an activity in which interdepartmental coordination is necessary to
assure human performance H.3(b).
Enforcement. Technical Specification 5.7.2 states that, in addition to the requirements of
Specification 5.7.1, entry into high radiation areas accessible to personnel with dose
rates such that a major portion of the whole body could receive in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> a deep dose
equivalent in excess of 1000 millirem shall be provided with locked doors except during
periods of access by personnel under an approved special work permit which shall
- 21 - Enclosure
specify the dose rates in the area. Technical Specification 5.7.1(b) states, in part, that
individuals permitted to enter high radiation areas shall be provided with a monitoring
device that continuously integrates the radiation dose and alarms when a preset dose is
received. Entry into such areas may be made after the dose rates in the area have been
established and personnel have been made knowledgeable of them. Contrary to this
requirement, on April 2, 2011, the licensee failed to adequately brief the dose rates in
the immediate work area and make workers knowledgeable of the dose rates within the
high radiation area before allowing entry into the area. Because this violation was of
very low safety significance and it was entered into the corrective action program as
Condition Report CR-CNS-2011-04441, this violation is being treated as a noncited
violation, consistent with Section 2.3.2 of the Enforcement Policy:
NCV 05000298/2011008-05, Failure to Perform an Adequate High Radiation Area
Briefing.
4.4 Job Coverage
a. Scope
The inspectors reviewed the licensees actions with respect to providing radiation
protection coverage of workers entering a locked high radiation area to perform work
during the shuttle tube event. The inspectors used the requirements in 10 CFR Part 20,
the technical specifications, and the licensees procedures required by technical
specifications as criteria for determining compliance. During the inspection, the
inspectors interviewed the radiation protection manager, radiation protection
supervisors, radiation protection technicians, and radiation workers. The inspectors
performed tours of the plant to understand scope of the job during the shuttle tube event.
The inspectors reviewed radiological hazards control and work coverage, including the
adequacy of surveys, radiation work permits, radiation protection job coverage, and
contamination controls. The inspectors reviewed radiation worker and radiation
protection technician performance during the shuttle tube event.
b. Findings
Introduction. The inspectors identified a Green noncited violation of Technical
Specification 5.4.1.a for the failure to follow radiation protection Procedure 9.EN-RP-141,
Job Coverage, Revision 8. Specifically, during the nightshift on April 2, 2011, radiation
protection personnel were monitoring workers pulling the intermediate range monitor
shuttle tube from under the reactor pressure vessel and failed to adequately implement
several requirements of the job coverage procedure which resulted in workers being
exposed to unexpected high dose rates up to 39 rem per hour at 30 centimeters from the
shuttle tube.
Description. On April 2, 2011, three workers entered the drywell, which was a posted
locked high radiation area, to perform work to remove the intermediate range monitor C
shuttle tube. Prior to entering the drywell, the workers donned protective clothing and
respiratory protection (powered air purifying respirators). The radiation protection
technicians at the drywell entry point assisted the workers with donning the respirators
and were responsible for monitoring the workers radiation dose. For entries into locked
high radiation areas, radiation protection technicians were required to monitor work
activities remotely or at the job site. For this activity, the licensee determined that
remote monitoring using teledosimetry (radiation dose transmitted from electronic
- 22 - Enclosure
dosimeters to a remote monitoring station) and continuous communications (via a site
cell phone) was sufficient to provide adequate radiation protection job coverage for the
workers. The required actions for remote monitoring job coverage activities in a locked
high radiation area prior to, and during, the performance of the work were described in
Station Procedure 9.EN-RP-141, Job Coverage, Revision 8.
The inspectors interviewed station personnel, toured the drywell entry point, and
reviewed station procedures. The inspectors identified that radiation protection
personnel had failed to adequately implement several job coverage procedural
requirements, which either resulted in, or contributed to, the workers being exposed to
higher than expected dose rates. The failures are described below:
1. The remote-monitoring technician did not attend the ALARA pre-job briefing for
the work. Attachment 2, Section 2, Responsibilities, Procedure 9.EN-RP-141,
states that the radiation protection technician providing job coverage is
responsible for attending the pre-job briefing. This failure resulted in the remote
monitoring technician not having a full understanding of the work scope, and
therefore, the remote monitoring technician was not able to identify when the
scope changed. The remote monitoring technician believed that the work scope
was limited to an inspection activity only, and not a maintenance activity to
remove the intermediate range monitor C shuttle tube from beneath the reactor
pressure vessel.
2. Radiation protection technicians providing job coverage failed to establish a
method of communication. For this activity, it was decided that site cell phones
would be used to communicate with the workers. Step 2.13 of
Procedure 9.EN-RP-141 required that when using site cell phones as a
communication device during continuous job coverage, it is required to have the
keypad locked. Locking the phone ensured that communication from radiation
protection to the workers was maintained during remote job coverage activities.
(Step 2.13 was added to the procedure as a corrective action to a 2009 NRC
violation because the site cell phone used during job coverage activities in 2009
had been inadvertently turned off and communication with workers was lost.
That issue was documented in NRC Inspection Report 05000298/2009005.)
However, the procedure did not make clear whose responsibility it was for
locking the cell phone. When the inspectors interviewed station personnel to
determine whose responsibility it was to lock the cell phone keypad, the
inspectors received mixed answers, with some personnel stating it was the users
responsibility, others stated it was radiation protection technicians responsibility,
while others stated it was workers responsibility to lock the keypad but radiation
protection personnel had to verify that the cell phone keypad was locked. The
inspectors determined this lack of clarity about whose responsibility it was to
have the phone locked contributed to the failure to ensure the phone was locked
and stayed locked except when needed to establish communications with
radiation protection.
3. The remote monitoring technician failed to review the applicable special
(radiation) work permit as required by Step 5.5.1 of Procedure 9.EN-RP-141.
This requirement ensures that the remote monitoring technician becomes
knowledgeable of the work scope, such that if the scope changes the remote
monitoring technician can take the appropriate actions when necessary. For this
- 23 - Enclosure
event, the appropriate action would have been to stop the job, have the workers
leave the work site, and prepare a revised radiation work permit.
4. The remote monitoring technician providing job coverage failed to communicate
with workers to inform the workers of the radiological hazards associated with the
nuclear instrument system, potential changes that would occur during the course
of activities, understand the details of the work activity, and in particular any job
steps that could impact radiological conditions as required by Step 5.5.4.2 of
Procedure 9.EN-RP-141. The remote monitoring technician did not discuss the
details of the work activity with the workers, and therefore, was not able to
communicate the hazards that were associated with the work activity. The
remote monitoring technician believed the workers were only going to perform an
inspection under the reactor pressure vessel. The remote monitoring technician
assumed that the ALARA pre-job briefing covered all radiological aspects of the
work activity and did not believe the work activity would breach any systems or
remove any parts. This assumption was not verified or validated.
5. Step 6.4.5 of Procedure 9.EN-RP-141 required that communication devices are
verified operational between the remote monitoring station and the work location.
Neither the workers nor the remote monitoring technician attempted to make
contact with each other during the work activity.
6. Workers used the dedicated radiation protection cell phone to contact the outage
control center to discuss the work activity with maintenance personnel. The
remote monitoring technician could view the workers on the video monitor and
see that the site cell phone designated for radiation protection coverage was in
use and was not locked in accordance with Step 2.13 of the procedure. While
the site cell phone is in use, it cannot be called. There is no call waiting. There
is only a busy signal. Step 6.4.7 of Procedure 9.EN-RP-141 required that if
communication is lost then it should be re-established in accordance with the
procedure, or work activities suspended and personnel cleared from the area.
No attempt was made to perform these requirements while communication was
lost. In addition, the inspectors identified that the licensees dayshift remote
monitoring technicians used radios for communications and nightshift used cell
phones. This inconsistency between dayshift and nightshift contributed to the
loss of communications during this activity. The licensee corrected this
discrepancy by requiring all remote monitoring technicians to use radios for
continuous coverage communications.
7. Workers lowered the shuttle tube to the floor of the drywell prior to receiving
permission to pull it all the way out of the reactor vessel. Step 5.5.4.2 of
Procedure 9.EN-RP-141 required the remote monitoring technician to monitor
the work location to determine if new sources of exposure are being generated
(e.g., trash or parts removed from the system). The shuttle tube is a part of the
nuclear instrument system and was beyond the scope of what the remote
monitoring technician believed to be the work activity (inspection only). Video
monitoring showed the part being lowered to the floor at which point the
technician should have called the workers and told them to stop the activity.
8. The remote monitoring technician failed to exercise the stop work authority.
Step 7.1 of Procedure 9.EN-RP-141 stated that radiation protection technicians
- 24 - Enclosure
have both the responsibility and authority to stop work if there is a change in
work scope or the continuance of work would result in a violation of good
radiological work practices, or a violation of radiological work permit or special
work permit requirements. When the workers changed scope during the
performance of the work activity from what was understood by the remote
monitoring technician, the work was required to be stopped.
The work was not stopped when the shuttle tube was initially pulled from the reactor
vessel. Therefore, the workers under the vessel pulled the entire 27-foot-long shuttle
tube out of the reactor vessel, and exposed themselves to the highly radioactive end of
the shuttle tube. The workers electronic dosimeters alarmed on high dose rate. The
workers immediately left the area under the vessel and informed a radiation protection
technician in the area that the dose rates had significantly increased. The licensee
entered their emergency procedures for unexpected radiation levels in the building,
cleared the drywell, and restricted access until the source of the radiation was identified.
The licensees immediate corrective actions were to restrict access to the drywell,
ensure that further work activities in the drywell had been reviewed and approved by the
radiation protection supervision, and pursue activities to recover the drywell area under
the reactor vessel by securing the shuttle tube.
During the recovery phase of the activity, radiation protection personnel measured
contact radiation dose rates as high as 3,226 rem per hour, and 39 rem per hour at
30 centimeters from the shuttle tube. Radiation protection technicians recovered the
drywell by placing the highly radioactive portion of the shuttle tube in a shielded
container.
While interviewing personnel involved in this event, the inspectors encountered no
indication that workers had used human error prevention techniques to ensure that they
followed procedures.
Analysis. The failure to follow radiation protection job coverage procedures is a
performance deficiency. The finding is more than minor because it could be viewed as,
both, a precursor to a significant event, and if left uncorrected, could have led to a more
safety significant concern. It is also associated with the human performance attribute of
the Occupational Radiation Safety Cornerstone and affected the cornerstone objective to
ensure the adequate protection of the worker health and safety from exposure to
radiation from radioactive material during routine civilian nuclear reactor operation
because it resulted in workers receiving higher than expected doses. The inspectors
evaluated the significance of the finding using NRC Inspection Manual 0609,
Appendix C, Occupational Radiation Safety Significance Determination Process, dated
August 19, 2008. The inspectors determined that the finding is of very low safety
significance (Green) because the finding did not involve ALARA planning and work
controls, did not result in an overexposure, did not involve a substantial potential for
overexposure, and did not compromise the licensees ability to assess dose.
Additionally, the inspectors determined that the apparent cause of this finding was the
licensees failure to encourage workers to use human error prevention techniques to
ensure that they followed procedures. Therefore, this finding has a crosscutting aspect
in the work practices component of the human performance area because the licensee
failed to use human error prevention techniques such as self-checking and peer-
checking to ensure that job coverage procedures were followed H.4(a).
- 25 - Enclosure
Enforcement. Technical Specifications 5.4.1 states in part, that written procedures shall
be established, implemented, and maintained covering the applicable procedures
recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.
Regulatory Guide 1.33, Section 7, required radiation protection procedures, including
access control to radiation areas. Licensee Procedure 9.EN-RP-141, Job Coverage,
Revision 8, in part, required the licensee to implement the following job coverage
activities :
(1) (Step 2.13) when site cell phones are used as a communication device during
continuous job coverage, the keypad must be locked,
(2) (Step 3.4) communicate with workers to tell them about radiological hazards
associated with the systems to be worked and potential changes that would
occur during the course of activities, and understand the details of the work
activity to be performed and job steps that could impact radiological conditions or
result in personnel contaminations,
(3) (Step 5.5.1) upon assignment, review the applicable special work permit to
determine the scope of work to be performed,
(4) (Step 5.5.4.2) monitor the work location to determine if new sources of exposure
are being generated (e.g., parts removed from the system),
(5) (Step 6.4.5) verify communication devices operate between the remote
monitoring technician station and the work location,
(6) (Step 6.4.7) if continuous coverage by remote monitoring is lost, then either
reestablish continuous job coverage by other means or suspend work activities
and clear personnel from the work area,
(7) (Step 7.1) stop work if there is a change in work scope or if the initiation of work
or the continuance of work would result in a violation of good radiological work
practices or a violation of radiation work permit/special work permit
requirements, and
(8) (Attachment 2, Section 2 responsibilities, Step 2.4) attend the pre-job briefing.
Contrary to the above, on April 3, 2011, the licensee failed to:
(1) lock the cell phone keypad,
(2) inform the workers of radiological hazards associated with the nuclear instrument
system,
(3) review the special work permit,
(4) monitor the work location to determine if new sources of exposure are being
generated,
(5) verify communication devices operation between the remote monitoring
technician and the work location,
- 26 - Enclosure
(6) suspend work activities and clear personnel from the area when communication
was lost,
(7) stop work when there was a change in work scope or the work would result in a
violation of the radiation work permit requirements, and
(8) attend the pre-job briefing.
Because this finding is of very low safety significance and has been entered into the
licensees corrective action program as Condition Reports CR-CNS-2011-04442,
CR-CNS-2011-04255, CR-CNS-2011-04595, CR-CNS-2011 -05443,
CR-CNS-2011-05444, CR-CNS-2011-05446, CR-CNS-2011-05447, and
CR-CNS-2011-05448, this violation is being treated as a noncited violation consistent
with Section 2.3.2 of the NRC Enforcement Policy: NCV 05000298/2011008-06, Failure
to Follow Radiation Protection Job Coverage Procedures.
4.5 Dose and Dose Rate Assessment
a. Scope
The NRC performed an independent assessment of the dose and dose rate information
using time motion studies to identify source of radiation, the exposure time, and the
distance between the source and the workers tissue. The source of the radiation was
the activated tip of the intermediate range monitor B shuttle tube. It was activated by the
nuclear reactor core because it is composed of stainless steel exposed to neutrons
during the operating cycle of the power reactor. Although the shuttle tube is in its
retracted position during the cycle and is only inserted into the core during startup and
shutdown operations, the end of the tube still becomes radioactive from long-term
exposure to neutrons. Approximately, the top one inch of the tube is activated
significantly more than the rest of the tube because of its retracted position, which is
about 24 inches below the bottom core plate. The NRC performed an independent
assessment of the skin dose to the hand of the worker who removed the shuttle tube
from the reactor pressure vessel. This assessment was performed by the NRCs senior
advisor for health physics using a software program called Monte Carlo N-particle. Dose
rate data measured by the licensee during the recovery phase of event was entered into
Monte Carlo N-particle. The data included AMP-200 Geiger Mueller detector and
optically stimulated luminescent dosimeters which are specifically designed to measure
shallow dose equivalent to human tissue. The AMP-200 data included 3,226 rem per
hour on contact and 39 rem per hour at 30 centimeters. The optically stimulated
luminescent data included 0.338 rem per second at one inch from the source. Based on
time motion studies conducted later with workers, the individual handling the shuttle tube
grasped the end of the shuttle tube for about 1.7 seconds.
Based on the time motion study that was reviewed by the inspectors and the
independent Monte Carlo N-particle calculation performed by the NRC, the estimated
skin dose to the hand of the worker who grasped the source was 2.9 rem. This dose is
well below the regulatory limit of 50 rem. The licensee employed a certified health
physicist to perform the dose calculation. The certified health physicist used manual
calculations and a combination of computer codes to determine the skin dose. The
licensees estimated skin dose was 3.1 rem. Although the licensee used a different
methodology than the NRC, the estimated skin doses are in relative agreement and
differ by only 8 percent. Both values are significantly below regulatory limits and
- 27 - Enclosure
therefore warrant no further analysis. The whole body dose assigned to the individual
was 0.040 rem based on the electronic dosimeter readings and the time motion studies.
The whole body dose is also below the annual regulatory limit of 5.0 rem.
b. Findings
No findings were identified.
5.0 Review of Previous Activity Performance
a. Scope
The inspectors reviewed previous intermediate range monitor and source range monitor
removal activities. The inspectors assessed the adequacy of prior work packages and
the execution of those work orders. Previous condition reports and past operating
experience were reviewed for lessons learned. The inspectors compared the previous
work orders to Work Order 4741002, to determine if this method (pulling the shuttle tube
from the bottom) had been used in the past.
b. Findings
No findings were identified. Operating experience showed that a shuttle tube had
previously been pulled from the bottom of the vessel, however this was a necessary
action resulting from a stuck detector. In this instance, the licensee also experienced
elevated radiation levels. The normal (proceduralized) method for replacing the tubing
assembly was to remove the assembly from the top of the core.
6.0 Review of Causal Determination and Corrective Actions
a. Scope
The inspectors reviewed the preliminary root cause evaluation report and corrective
actions identified to prevent recurrence of the root causes. The inspectors interviewed
members of the licensees root cause team and licensee management. At the end of the
inspection period, the inspectors did not have the opportunity to review the final version
of the root cause evaluation because the final report had not been completed and
reviewed by licensee management.
b. Findings
No findings were identified. Because the final root cause report had not been completed
at the time of this report, the inspectors were unable to evaluate its adequacy against the
licensees corrective action program procedures. Therefore, the final root cause report
will be subject to inspection at a future date. Notwithstanding the issuance of the final
root cause evaluation report, the inspectors noted that the licensees preliminary root
causes were consistent with the findings identified in this report. The licensees long
term corrective actions are still in the process of being developed, however, interim
actions have been taken to prevent recurrence of this event. These actions include work
order process procedure revisions to include identification of materials required to
perform maintenance, implementing a work order quality review panel, revising work
order risk assessment procedures, revising radiation protection briefing forms to ensure
full extent of job scope is discussed at the ALARA briefing, reinforcing requirement for
- 28 - Enclosure
radiation protection to attend all locked high radiation area briefings, and developing
specific expectations for supervisors to ensure procedure compliance is mandatory.
4OA6 MEETINGS
On April 15, 2011, the team presented the preliminary results of this inspection at the
end of the onsite week to Mr. D. Willis, General Manager Plant Operations, and other
members of the licensee staff who acknowledged the findings. The team returned all
proprietary information reviewed during the inspection prior to leaving the site.
On May 3, 2011, the team presented the final results of the inspection to
Mr. A. Zaremba, Director of Nuclear Safety Assurance, and other members of the
licensee staff via telephonic exit. The team obtained permission from the licensee to use
the diagrams and photographs in this report.
On June 9, 2011, the team re-exited and presented revised results of the inspection to
Mr. A. Zaremba, Director of Nuclear Safety Assurance, and other members of the
licensee staff via telephonic exit.
ATTACHMENT 1: SUPPLEMENTAL INFORMATION
ATTACHMENT 2: SPECIAL INSPECTION CHARTER
ATTACHMENT 3: PICTURES AND DIAGRAMS
- 29 - Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
J. Bednar, Supervisor, Radiation Protection
J. Corey, Manager, Radiation Protection
E. McCutchen, Senior Licensing Engineer, Licensing
H. A. Hawkins, Superintendent, Instrumentation and Control
D. Willis, Plant Manager
A. Zaremba, Director of Nuclear Safety Assurance
NRC Personnel
M. Chambers, Resident Inspector
B. Hagar, Senior Project Engineer
J. Josey, Senior Resident Inspector
R. Pedersen, Senior Health Physicist
S. Sherbini, Senior Level Advisor for Health Physics
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
050000298/2011008-01 NCV Unclear Work Instructions (Section 3.1)
050000298/2011008-02 NCV Failure to Recognize Work Order Risk (Section 3.1)
050000298/2011008-03 FIN Failure to Implement Human Performance Procedure
(Section 3.2)
050000298/2011008-04 NCV Failure to Revise Unclear Work Instructions (Section 3.3)
050000298/2011008-05 NCV Failure to Perform an Adequate High Radiation Area
Briefing (Section 4.3)
050000298/2011008-06 NCV Failure to Follow Radiation Protection Job Coverage
Procedures (Section 4.4)
A1-1 Attachment 1
DOCUMENTS REVIEWED
Section 4OA3: Event Follow-up
CONDITION REPORTS
94-1262 NCR 93-045
CR-CNS-2011-3769 CR-CNS-2011-4584 CR-CNS-2011-4588 CR-CNS-2011-3763
CR-CNS-2011-4255 CR-CNS-2011-4256 CR-CNS-2011-4258 CR-CNS-2011-4317
CR-CNS-2011-4428 CR-CNS-2011-4431 CR-CNS-2011-4432 CR-CNS-2011-4436
CR-CNS-2011-4429 CR-CNS-2011-4430 CR-CNS-2011-4438 CR-CNS-2011-4439
CR-CNS-2011-4440 CR-CNS-2011-4441 CR-CNS-2011-4442 CR-CNS-2011-4583
CR-CNS-2011-4581 CR-CNS-2011-4435 CR-CNS-2011-4433 CR-CNS-2011-4582
CR-CNS-2011-4591 CR-CNS-2011-4585 CR-CNS-2011-4586 CR-CNS-2011-3890
CR-CNS-2011-4592 CR-CNS-2011-4583 CR-CNS-2011-4587 CR-CNS-2011-4258
CR-CNS-2011-4593 CR-CNS-2011-4594 CR-CNS-2011-4595 CR-CNS-2011-4596
CR-CNS-2011-4597 CR-CNS-2011-4598 CR-CNS-2011-4599 CR-CNS-2011-4600
CR-CNS-2011-4601 CR-CNS-2011-5443 CR-CNS-2011-5444 CR-CNS-2011-5446
CR-CNS-2011-5447 CR-CNS-2011-5448 CR-CNS-2011-5450
WORK ORDERS
4741009 4741002 4741006 4491177
RADIATION/SPECIAL WORK PERMITS
2009-422 2011-422 2011-465
PROCEDURES/DOCUMENTS
NUMBER TITLE REVISION
14.2.19 SRM/IRM Detector and Drive Tube Removal, 26
Installation, Testing, and SRM/IRM Troubleshooting
14.2.19 SRM/IRM Detector and Drive Tube Removal, 27
Installation, Testing, and SRM/IRM Troubleshooting
0.40 Work Control Program 75
A1-2 Attachment 1
0.40.4 Planning 16
0.1 Procedure Use and Adherence 36
7.0.4 Conduct of Maintenance 32
0-HU-TOOLS Human Performance Tools 17
2.0.1.1 Conduct of Infrequently Performed Tests and 5
Evolutions
10.29 LPRM and SRM/IRM Dry Tube Removal and 29
Installation
IAC722-00-00, Fig. 12 Detector Drive Unit 0
IAC722-00-00, Fig. 9 Source Range and Intermediate Range Detector 0
Drive
9.EN-RP-141 Job Coverage 8
9.ALARA.4 Radiation Work Permits 14
9.ALARA.5 ALARA Planning and Controls 21
5.1RAD Building Radiation Trouble 15
A1-3 Attachment 1
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGI ON I V
612 EAST LAMAR BLVD, SUITE 400
ARLINGTON, TEXAS 76011-4125
April 7, 2011
MEMORANDUM TO: Dean Overland, Resident Inspector
Projects Branch E
Division of Reactor Projects
Binesh Tharakan, Resident Inspector
Projects Branch A
Division of Reactor Projects
FROM: Kriss Kennedy, Director /RA/
Division of Reactor Projects
SUBJECT: SPECIAL INSPECTION CHARTER TO EVALUATE UNEXPECTED
DOSES TO WORKERS AT COOPER NUCLEAR STATION
A Special Inspection Team is being chartered in response to a work activity that resulted in
unexpected doses to workers at the Cooper Nuclear Station on April 3, 2011. Dean Overland is
designated as the Special Inspection Team Lead with respect to work-control issues. Binesh
Tharakan is designated as the Special Inspection Team Lead with respect to radiological
issues.
A. Basis
On April 3, 2011, while licensee workers were preparing to remove the Intermediate
Range Monitor-C (IRM-C) drive mechanism shuttle tube from the top of the reactor
vessel, they discovered they did not have access to a waterproof nose cone that was to
be attached to the lower end of the tube prior to removal.
When the workers reported this issue to the Outage Control Center (OCC), the OCC
staff reportedly either instructed or gave permission to the workers to remove the shuttle
tube from the bottom of the reactor vessel, instead of from the top as originally planned.
The inspectors understand that the licensee did not modify either the associated work
order or the corresponding Radiation Work Permit (RWP) to reflect this change.
As the workers removed the tube from the bottom of the vessel, the three workers under
the vessel and one worker at the access point received dose-rate alarms. The workers
then set the tip of the tube on the floor at the 888 elevation and exited the area. The
workers dosimeters reportedly measured dose rates of 1.35 rem per hour, 14.3 rem per
hour, and 763 millirem/hr.
A2-1 Attachment 2
Surveys taken later found that the tip of the tube measured 3226 rem/hr on contact and
39 rem/hr at 30 cm, and that the general area dose rate was 4.6 rem/hr at waist level,
increasing to 8.6 rem/hr at waist level near the tube.
B. Scope
The inspection is expected to perform data gathering and fact-finding in order to address
the following:
1. Develop a sequence of events leading up to the event, actions taken upon receipt of
dose rate alarms, and actions taken to reduce the dose rates following the event.
2. Develop a timeline and assess the decision-making process used by licensee
personnel to deviate from the planned method to remove intermediate range
monitor C.
3. Assess licensee compliance with procedures and work orders in accomplishing the
evolution.
4. Compare and contrast performance of this activity on April 3, 2011 to the
performance of similar activities during the current outage.
5. Review history of the licensees conduct of this evolution to determine if they have
used this method of removal prior to April 3, 2011.
6. Characterize the dose rates during the event and the dose received by involved
personnel.
7. Assess as low as reasonably achievable (ALARA) planning for the evolution.
8. Assess adequacy of the radiation work permit and pre-job briefing for this activity.
9. Review any preliminary cause determination the licensee has completed and assess
adequacy of short term corrective actions.
10. Collect data necessary to support completion of the significance determination
process.
C. Guidance
Inspection Procedure 93812, ASpecial Inspection,@ provides additional guidance to be
used by the Special Inspection Team. Your duties will be as described in Inspection
Procedure 93812. The inspection should emphasize fact-finding in its review of the
circumstances surrounding the events. Safety concerns identified that are not directly
related to the event should be reported to the Region IV office for appropriate action.
The team will report to the site, conduct an entrance, and begin inspection no later than
April 11, 2011. While onsite, you will provide daily status briefings to Region IV
management, who will coordinate with the Office of Nuclear Reactor Regulation to
ensure that all other parties are kept informed. Depending on the outcome of the
inspection, inspection results will be documented in U. S. Nuclear Regulatory
Commission (NRC) Special Inspection Report No. 05000298/2011008. This report will
be issued within 45 days of the completion of the inspection.
A2-2 Attachment 2
This Charter may be modified should the team develop significant new information that
warrants review. Should you have any questions concerning this charter, please contact
Vince Gaddy or Bob Hagar.
R:\_Reactors\CNS 2011\SI Charter 110407.docx ADAMS ML
SUNSI Rev Compl. Yes No ADAMS Yes No Reviewer Initials VGG
Publicly Avail Yes No Sensitive Yes No Sens. Type Initials
VGG
RIV:C/DRP/PBC C/DRP/PBC D:DRP D:DRP
RHagar: jm VGaddy KKennedy
/RA/ /RA/ /RA/
04/ 6 /2011 04/ 7 /2011 04/ 7 /2011
OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
A2-3 Attachment 2
A3-1 Attachment 3
A3-2 Attachment 3