ML090410466
| ML090410466 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 02/10/2009 |
| From: | Stephanie West Division of Reactor Safety III |
| To: | Wadley M Northern States Power Co |
| References | |
| EA-08-349 IR-08-009 | |
| Download: ML090410466 (24) | |
See also: IR 05000282/2008009
Text
February 10, 2009
Mr. Michael D. Wadley
Site Vice President
Northern States Power - Minnesota
Prairie Island Nuclear Power Station
1717 Wakonade Drive East
Welch, MN 55089
SUBJECT: PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 1 AND 2
NRC INSPECTION REPORT 05000282/2008009; 05000306/2008009
PRELIMINARY YELLOW FINDING
Dear Mr. Wadley:
On January 21, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection
at your Prairie Island Nuclear Power Station. The enclosed report documents the inspection
findings, which were discussed on January 21, 2009, with members of your staff.
The inspection examined activities conducted under your license as they relate to safety and
compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
The enclosed report presents the results of this inspection including a finding that has
preliminarily been determined to be Yellow, a finding with substantial safety significance that
may require additional NRC inspections. As described in Section 2PS2 of this report, the
finding involves an October 29, 2008, radioactive material shipment from your facility, via an
exclusive-use open transport vehicle that did not conform to the applicable Department of
Transportation (DOT) regulatory requirements when it arrived at the shipping destination.
The NRC requires licensees to comply with DOT regulations. Specifically, upon receipt, the
external radiation levels on the surface of the affected package were determined to exceed
DOTs regulatory specification. The apparent cause was ineffective radiological
characterization and packaging of the package contents to assure that, under conditions
normally incident to transport, the package would conform with DOTs radiation level limits
specified in 49 CFR 173.441(a). Additionally, workers involved in preparing the package were
not trained as required by 49 CFR 172.704. After the finding was identified, the licensees staff
evaluated the radiological impact to the public to ensure there was no immediate safety concern
and implemented corrective actions, including suspension of all radioactive material shipments
to prevent future incidents.
This finding was assessed based on the best available information, using the applicable
Significance Determination Process (SDP). The final resolution of this finding will be conveyed
in separate correspondence. Preliminarily, we consider this a self-revealing finding having
substantial safety significance because the external package radiation level was greater than
M. Wadley -2-
five, but less than ten times the radiation level limitation specified in the DOT regulatory
requirement. Although the surface of the package with elevated radiation levels would not be
routinely accessible to a member of the public during transport, that aspect was fortuitous and
not the result of design nor package preparation by the licensee. Additionally, the condition
had the potential to adversely affect personnel who would normally receive the package or
respond to an incident involving the package with the reasonable expectation that the
package conformed to DOT radiation limitations. Accordingly, the finding is also an apparent
violation of NRC requirements specified by 10 CFR 71.5, which requires licensees to comply
with 49 CFR 172.704 [and 173.441(a)], and is being considered for escalated enforcement
action in accordance with the NRC Enforcement Policy. The current policy is on the NRCs
website at http://www.nrc.gov/reading-rm/doc-collections/enforcement.
In accordance with NRC Inspection Manual Chapter (IMC) 0609, we intend to complete our
evaluation using the best available information and issue our final determination of safety
significance within 90 days of the date of this letter. The significance determination process
encourages an open dialogue between the NRC staff and the licensee; however, the dialogue
should not impact the timeliness of the staffs final determination. Before the NRC makes a final
decision on this matter, we are providing you an opportunity to: (1) attend a Regulatory
Conference where you can present to the NRC your perspectives on the facts and assumptions
the NRC used to arrive at the finding and its significance; or (2) submit your position on the
finding to the NRC in writing. If you request a Regulatory Conference, it should be held within
30 days of the receipt of this letter and we encourage you to submit supporting documentation
at least one-week prior to the conference in an effort to make the conference more efficient and
effective. If a Regulatory Conference is held, it will be open for public observation and a press
release will be issued to announce it. If you decide to provide a written response in lieu of the
Regulatory Conference, the submission should be sent to the NRC within 30 days of the receipt
of this letter. If you decline to request a Regulatory Conference or to submit a written response,
you relinquish your right to appeal the final SDP determination, in that by not doing either, you
fail to meet the appeal requirements stated in the Prerequisite and Limitation Section of
Attachment 2 of Inspection Manual Chapter 0609.
Please contact Steven K. Orth at 630-829-9827 within 10 days of the date of this letter to notify
the NRC of your intentions. If we have not heard from you within 10 days, we will continue with
our significance determination and enforcement decision. The final resolution of this matter will
be conveyed in separate correspondence.
Since the NRC has not made a final determination in this matter, no Notice of Violation is being
issued for this inspection finding at this time. In addition, please be advised that the number
and characterization of apparent violations described in the enclosed inspection report may
change as a result of further NRC review.
Additionally, based on the results of this inspection, one NRC-identified finding of very low
safety significance was identified. The finding involved a violation of NRC requirements.
However, because of its very low safety significance, and because the issue was entered into
your corrective action program, the NRC is treating the issue as an Non-Cited Violation (NCV) in
accordance with Section VI.A.1 of the NRC Enforcement Policy.
M. Wadley -3-
If you contest the subject or severity of a NCV, you should provide a response within 30 days of
the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory
Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the
Regional Administrator, U.S. Nuclear Regulatory Commission - Region III, 2443 Warrenville
Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector Office at the
Prairie Island Nuclear Generating Plant.
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 any), will be available electronically for public inspection in
the NRC Public Document Room or from the Publicly Available Records System (PARS)
component of NRC's Agencywide Documents Access and Management System (ADAMS),
accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public
Electronic Reading Room).
Sincerely,
/RA/
Steven West, Director
Division of Reactor Safety
Docket Nos. 50-282; 50-306
Enclosure: Inspection Report 05000282/2008009; 05000306/2008009
w/Attachment: Supplemental Information
cc w/encl: D. Koehl, Chief Nuclear Officer
Regulatory Affairs Manager
P. Glass, Assistant General Counsel
Nuclear Asset Manager
J. Stine, State Liaison Officer, Minnesota Department of Health
Tribal Council, Prairie Island Indian Community
Administrator, Goodhue County Courthouse
Commissioner, Minnesota Department
of Commerce
Manager, Environmental Protection Division
Office of the Attorney General of Minnesota
Emergency Preparedness Coordinator, Dakota
County Law Enforcement Center
M. Wadley -3-
If you contest the subject or severity of a NCV, you should provide a response within 30 days of the date
of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission,
ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator,
U.S. Nuclear Regulatory Commission - Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-
4352; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001; and the Resident Inspector Office at the Prairie Island Nuclear Generating Plant.
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 any), will be available electronically for public inspection in the NRC Public Document
Room or from the Publicly Available Records System (PARS) component of NRC's Agencywide
Documents Access and Management System (ADAMS), accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
RA/
Steven West, Director
Division of Reactor Safety
Docket Nos. 50-282; 50-306
Enclosure: Inspection Report 05000282/2008009; 05000306/2008009
w/Attachment: Supplemental Information
cc w/encl: D. Koehl, Chief Nuclear Officer
Regulatory Affairs Manager
P. Glass, Assistant General Counsel
Nuclear Asset Manager
J. Stine, State Liaison Officer, Minnesota Department of Health
Tribal Council, Prairie Island Indian Community
Administrator, Goodhue County Courthouse
Commissioner, Minnesota Department
of Commerce
Manager, Environmental Protection Division
Office of the Attorney General of Minnesota
Emergency Preparedness Coordinator, Dakota
County Law Enforcement Center
See Previous Concurrence
DOCUMENT NAME: G:\DRS/Workinprogress/Prairie Island 2008 009 DRS MJP.doc
G Publicly Available G Non-Publicly Available G Sensitive G Non-Sensitive
To receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl "E" = Copy with attach/encl "N" = No copy
OFFICE RIII RIII RIII RIII
NAME SOrth:ls JGiesner KOBrien SWest
DATE 02/04/09 02/04/09 02/06/09 02/10/09
OFFICIAL RECORD COPY
Letter to Mr. Michael Wadley from Mr. Steven West dated February 10, 2009.
SUBJECT: PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 1 AND 2
NRC INSPECTION REPORT 05000282/2008009; 05000306/2008009
PRELIMINARY YELLOW FINDING
DISTRIBUTION:
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Patricia Buckley
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OEMAIL Resource
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket No.: 50-282; 50-306
Report No.: 05000282/2008009; 05000306/2008009(DRS)
Licensee: Northern States Power - Minnesota
Facility: Prairie Island Nuclear Power Station
Location: Welch, MN
Dates: November 17, 2008 through January 21, 2009
Inspectors: Martin J. Phalen, Health Physicist, DRS
Mark W. Mitchell, Health Physicist, DRS
Peter J. Lee, PhD., CHP, Senior Health Physicist, DNMS
Approved by: Steven K. Orth, Chief
Plant Support Team
Division of Reactor Safety
Enclosure
SUMMARY OF FINDINGS
IR 05000282/2008009; 05000306/2008009; 11/17/08 - 01/21/09; Prairie Island Nuclear Power
Station; Radioactive Material Processing and Transportation.
This report covered a one-week period of on-site inspection followed by in-office reviews of
licensee documents and records by regional health physics inspectors. One potential Yellow
finding and one Green finding were identified. The significance of most findings is indicated
by their color (Green, White, Yellow, and Red) using inspection Manual Chapter (IMC) 0609,
Significance Determination Process (SDP). Findings for which the SDP does not apply
may be Green or be assigned a severity level after NRC management review. The 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: Public Radiation Safety
- Preliminary Yellow: A self-revealing finding with an apparent violation of regulatory
requirements was identified involving a failure of the licensee to properly
radiologically characterize, prepare, and ship a package containing radioactive
material in a manner that assured, under conditions normally incident to transport,
conformance with Department of Transportation (DOT) radiation level limitations
specified by 49 CFR 173.441(a), (i.e., 200 millirem per hour (mrem/h)) on any external
surface of the package as required by 10 CFR 71.5 [and 49 CFR 173.441(a)].
Additionally, the licensee did not provide nor ensure that the individuals involved in
preparing this shipment were trained and qualified for the task as specified by
49 CFR 172.704, Training Requirements. The finding involved an October 29, 2008,
radioactive material shipment, via an exclusive-use open transport vehicle that was
determined to have radiation levels of 1630 mrem/h on the external surface of a
package upon receipt at the shipping destination. As immediate corrective actions, the
licensee suspended all radioactive shipment activities. The licensee entered this
performance deficiency in their corrective action program; initiated a root cause
evaluation; and initiated corrective measures, including various process improvements
to prevent recurrence.
This finding is more than minor since it was associated with the Public Radiation Safety
Cornerstone program and process attribute and affected the cornerstone objective to
ensure adequate protection of the public from exposure to radioactive materials given
that package radiation levels were elevated. Preliminarily, the significance of this finding
is considered as having a substantial safety significance (Yellow), since the radiation
level was greater than five times the limit (1000 mrem/h) but less than ten times the limit
(2000 mrem/h) specified by the DOT regulatory requirement. Although the surface of
the package with elevated radiation levels would not be routinely accessible to a
member of the public during transport, that aspect was fortuitous and not the result of
design nor package preparation by the licensee. The condition had the potential to
adversely affect personnel who would normally receive the package or respond to an
incident involving the package, with a reasonable expectation that the package
conformed to DOT radiation limitations.
1 Enclosure
Additionally, the cause of this finding had a cross-cutting aspect in the area of Human
Performance. Specifically, the licensee failed to appropriately plan the work activity by
incorporating risk insights and job site conditions, including conditions which may impact
radiological safety (H.3 (a)). This finding is documented within the licensees corrective
action system as RCE 1157726. (Section 2PS2)
Cornerstone: Occupational Radiation Safety
- Green. An NRC-identified finding of very low safety significance with an associated
Non-Cited Violation (NCV) of Technical Specification 5.4.1 was identified in the area of
occupational radiation safety associated with the licensees failure to perform adequate
job planning to evaluate the radiological hazards, as required by station procedures.
Specifically, the licensee failed to properly assess the radiological hazards to workers
associated with the decontamination, demobilization and packaging of fuel sipping
equipment on the refuel floor. This issue has been entered into the licensees corrective
action program and implemented corrective actions that include changes to procedures
to include a holistic risk-based review of radiologically significant work.
The finding is more than minor because, given the radiological uncertainty of working
with fuel handling equipment, if left uncorrected the finding could become a more
significant safety concern. The finding was determined to be of very low safety
significance because it did not involve unintended collective dose (ALARA planning);
there was no overexposure, nor potential for overexposure; and the licensees ability
to assess dose was not compromised. Additionally, the cause of this finding had a
cross-cutting aspect in the area of Human Performance. Specifically, the licensee failed
to appropriately plan the work activity by incorporating risk insights and job site
conditions, including conditions which may impact radiological safety (H.3 (a)).
(Section 2OS2)
B. Licensee-Identified Violations
No findings of significance were identified.
2 Enclosure
REPORT DETAILS
2. RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
2OS1 Access Control to Radiologically Significant Areas (71121.01)
.1 Job-In-Progress Reviews
a. Inspection Scope
Radiological work in high radiation work areas having significant dose rate gradients was
reviewed to evaluate the application of dosimetry to effectively monitor exposure to
personnel and to assess the adequacy of licensee controls.
Specifically, the inspectors reviewed the circumstances involving the decontamination,
demobilization, and preparation for shipment of fuel sipping equipment on October 23
and 24, 2008. The inspection was performed both on-site and through in-office reviews
of documents generated by the licensee. This review included discussion with various
members of the licensee staff, both in person and by teleconference, which provided a
common understanding of the events as they occurred. Additionally, select data
provided by the licensee was independently reviewed by a technical expert on the NRC
staff.
This inspection supplements the sample reported in Inspection Report 05000282/2008002; 05000306/2008002.
b. Findings
One finding of significance was identified.
Introduction: A Green NRC-identified finding of very low safety significance and
associated Non-Cited Violation (NCV) of Technical Specification 5.4.1 was identified in
the area of occupational radiation safety associated with the licensees failure to
perform adequate job planning to evaluate the radiological hazards, as required by
station procedures. Specifically, the licensee failed to properly assess the radiological
hazards to workers associated with the decontamination, demobilization and packaging
of fuel sipping equipment on the refuel floor.
Description: During a Unit 2 refueling outage in the fall of 2008, potentially degraded
fuel assemblies were tested for cladding integrity with vendor fuel sipping equipment.
The equipment was provided from the vendor as new, clean equipment. The tests were
performed under water in the spent fuel pool. After the outage, the equipment was
decontaminated, demobilized and packaged for shipment back to the vendor.
On October 23 and 24, 2008, the equipment was decontaminated at Prairie Island by
underwater hydro-lancing, demobilized by personnel, and prepared for shipment back to
the vendor. Personal accounts and record review of the work activity indicated that the
fuel sipping equipment was rinsed with water and surveyed for gross contamination and
3 Enclosure
radiological hazards when removed from the spent fuel pool. The staff used hand tools
for equipment disassembly. The fuel sipping equipment was then wrapped in plastic and
then placed into a designated laydown area near the pool. The radiation protection staff
initially posted and controlled the area as a Radiation Area and Contamination Area.
No extremity dosimetry nor additional controls were in place for controlling radiological
exposure during the hands-on demobilization of the equipment. After the fuel sipping
equipment was in the laydown area, the work group left the area. The next day the work
group loaded the fuel sipping equipment into a vendor supplied container in preparation
for shipment.
Station procedures, including FP-RP-JPP-01, RP Job Planning, Revision 04, required
formal job planning when removing items from the spent fuel pool. In the absence of
such a pre-job review, the potential exist for not performing the radiological surveys
necessary to properly assess the entire scope of radiological hazards present and,
subsequently to prescribe the appropriate supplemental dosimetry and/or dosimetry
placement requirements to workers associated with the decontamination, demobilization,
and preparation of the fuel sipping equipment.
In this case, the licensee planned hands-on work with fuel sipping equipment that had
a high potential for individuals to come in contact with discrete radioactive particles and
high contamination from the fuel sipping equipment. The licensee recognized that when
in use underwater, the fuel sipping equipment had detectable underwater contact dose
rates up to 40 rem/h on the inside of the fuel sipping canisters and up to 9 rem/h at
contact on the outside of the fuel sipping canisters. Radiological surveys taken in the
work area on the refuel floor detected discrete radioactive particles ranging from 800 to
35,000 disintegrations per minute (dpm) per 100 cm2. However, the inspectors identified
that the licensee had not performed adequate planning/evaluation to assess the hazard
and the potential radiological impact to the workers. As a result of this failure, the
licensee had not prescribed the appropriate dosimetry and monitoring commensurate
with the hazard, and the licensee had to perform a post-job dose evaluation to fully
assess the workers dose.
Several days later, after the fuel sipping equipment had been shipped off-site, more
detailed follow-up surveys of the equipment identified the presence of additional discrete
radioactive particles with gamma dose rates of up to 11 rem/h at contact. Isotopic
analysis of the particles determined that cobalt-60 was the dominant radionuclide of
interest. Once informed of the elevated dose rates of discrete radioactive particles, the
licensee performed extremity and whole body dose assessments to determine the actual
dose received by the workers that handled the equipment in order to confirm that no
significant radiological exposures had occurred.
Analysis: The inspectors identified a performance deficiency, in that, prior to the start of
work, the licensee failed to perform adequate radiological evaluations necessary to
properly assess the radiological hazards and to prescribe appropriate radiological
controls necessary to evaluate and minimize dose to workers. In accordance with
Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports,
Appendix B, Issue Screening, the inspectors determined that the issue was more than
minor because given the radiological uncertainty of working with fuel handling
equipment, if left uncorrected the finding could become a more significant safety
concern.
4 Enclosure
The finding does not involve the application of traditional enforcement, because it did not
result in actual safety consequences or the potential to impact the NRCs regulatory
function and because it was not the result of willful actions. The finding was evaluated
using the Significance Determination Process (SDP) in accordance with IMC 0609,
Appendix C, for the Occupational Radiation Safety cornerstone. The finding was
determined to be of very low safety significance (GREEN) because the finding did not
involve unintended collective dose (ALARA planning), there was no overexposure, nor
potential for overexposure, and the licensees ability to assess worker dose was not
compromised.
This finding was caused by inadequate planning of radiological work activities.
Consequently, the cause of this deficiency had a cross-cutting aspect in the area of
Human Performance. Specifically, the licensee failed to appropriately plan the work
activity by incorporating risk insights and job site conditions which may impact
radiological safety H.3(a).
Enforcement: Prairie Island Technical Specification 5.4.1 requires 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, Revision 2, Appendix A, Section 7, requires
procedures for the control of radioactivity, including limiting personnel exposure. Station
procedure FP-RP-JPP-01, RP Job Planning, Revision 04, Step 5.5.7, requires formal
job planning when removing items from the spent fuel pool.
Contrary to the above, the licensee failed to perform formal job planning when removing
items from the spent fuel pool. Specifically, on October 23 and 24, 2008, the licensee
disassembled, decontaminated, removed from the spent fuel pool, and packaged fuel
sipping equipment without a formal job plan to evaluate the radiological hazards
associated with these activities. The licensee documented this condition in its corrective
action program (AR 1162343) and instituted corrective measures including changes to
procedures to include a holistic risk-based review of radiological significant work.
(NCV 05000282/2008009-02; 05000306/2008009-02)
2OS2 As-Low-As-Reasonably-Achievable Planning And Controls (71121.02)
.1 Inspection Planning
a. Inspection Scope
The inspectors reviewed procedures associated with maintaining occupational
exposures as-low-as-is-reasonably-achievable (ALARA) and processes used to estimate
and track work activity specific exposures.
This inspection constituted one required sample as defined in IP 71121.02-5.
b. Findings
No findings of significance were identified.
5 Enclosure
.2 Radiological Work Planning
a. Inspection Scope
The inspectors reviewed the ALARA work activity evaluations for fuel sipping equipment
decontamination and shipping preparation activities. Specifically, the inspectors
reviewed the exposure estimates and exposure mitigation requirements in order to verify
that the licensee had established procedures and engineering and work controls that
were based on sound radiation protection principles in order to achieve occupational
exposures that were ALARA. The inspectors also determined if the licensee had
reasonably grouped the radiological work into work activities, based on historical
precedence, industry norms, and/or special circumstances.
This inspection constituted a partial sample as defined in IP 71121.02-5.
b. Findings
One finding of significance was identified, which is documented in Section 2OS1.1.
Cornerstone: Public Radiation Safety
2PS2 Radioactive Material Processing and Transportation (71122.02)
.1 Radioactive Waste System
a. Inspection Scope
The inspectors reviewed the liquid and solid radioactive waste system description in the
Updated Final Safety Analysis Report (UFSAR) for information on the types and
amounts of radioactive waste (radwaste) generated and disposed. The inspectors
reviewed the scope of the licensees audit program with regard to radioactive material
processing and transportation programs to verify that it met the requirements of
This inspection constituted one sample as defined in IP 71122.02-5.
b. Findings
No findings of significance were identified.
.2 Radioactive Waste System Walkdowns
a. Inspection Scope
The inspectors performed walkdowns of the liquid and solid radwaste processing
systems to verify, that the systems agreed with the descriptions in the UFSAR, and the
Process Control Program, and to assess the material condition and operability of the
systems. The inspectors reviewed the status of radwaste processing equipment that
was not operational and/or was abandoned in place. The inspectors reviewed the
licensees administrative and physical controls to ensure that the equipment would not
6 Enclosure
contribute to an unmonitored release path nor be a source of unnecessary personnel
exposure.
The inspectors reviewed changes to the waste processing system to verify that the
changes were reviewed and documented in accordance with 10 CFR 50.59 and to
assess the impact of the changes on radiation dose to members of the public. The
inspectors reviewed the current processes for transferring waste resin into shipping
containers to determine if appropriate waste stream mixing and/or sampling procedures
were utilized. The inspectors also reviewed the licensees methods for waste
concentration averaging to determine if representative samples of the waste product
were provided for the purposes of waste classification, as required by 10 CFR 61.55.
This inspection constituted one sample as defined in IP 71122.02-5.
b. Findings
No findings of significance were identified.
.3 Waste Characterization and Classification
a. Inspection Scope
The inspectors reviewed the licensees radiochemical sample analysis results for each of
the licensees waste streams, including dry active waste (DAW), spent resins, and filters.
The inspectors also reviewed the licensees use of scaling factors to quantify difficult-to-
measure radionuclides (e.g., pure alpha or beta emitting radionuclides). The reviews
were conducted to verify that the licensees program assured compliance with
10 CFR 61.55 and 10 CFR 61.56, as required by Appendix G of 10 CFR Part 20. The
inspectors also reviewed the licensees waste characterization and classification
program to ensure that the waste stream composition data accounted for changing
operational parameters and thus remained valid between the annual sample analysis
updates.
This inspection constituted one sample as defined in IP 71122.02-5.
b. Findings
No findings of significance were identified.
.4 Shipment Preparation and Shipping Manifests
a. Inspection Scope
The inspectors reviewed the documentation of shipment packaging, radiation surveys,
package labeling and marking, vehicle inspections and placarding, emergency
instructions, determination of waste classification/isotopic identification, and licensee
verification of shipment readiness for five non-excepted material and radwaste
shipments made from 2006 to 2008. The shipment documentation reviewed consisted
of:
7 Enclosure
- One LSA-I Shipment of Resins to a Radwaste Vendor; and
- Four SCO-II Shipments to a Fuel Analysis Vendor.
For each shipment, the inspectors determined if the requirements of 10 CFR Parts 20
and 61 and those of the Department of Transportation (DOT) in 49 CFR Parts 170-189
were met. Specifically, records were reviewed and staff involved in shipment activities
was interviewed to determine if packages were labeled and marked properly, if package
and transport vehicle surveys were performed with appropriate instrumentation, if
radiation survey results satisfied DOT requirements, and if the quantity and type of
radionuclides in each shipment were determined accurately. The inspectors also
determined whether shipment manifests were completed in accordance with DOT and
NRC requirements, if they included the required emergency response information, if the
recipient was authorized to receive the shipment, and if shipments were tracked as
required by 10 CFR Part 20, Appendix G.
This inspection constitutes one sample as defined by IP 71122.02-5.
Selected staff involved in shipment activities were observed and interviewed by the
inspectors to determine if they had adequate skills to accomplish shipment related
tasks and to determine if the shippers were knowledgeable of the applicable regulations
to satisfy package preparation requirements for public transport with respect to NRC
Bulletin 79-19, Packaging of Low-Level Radioactive Waste for Transport and Burial,
and 49 CFR Part 172 Subpart H.
This inspection constitutes one sample as defined by IP 71122.02-5.
b. Findings
One finding of significance was identified.
Introduction: A preliminary Yellow self-revealing finding of substantial safety significance
with associated apparent violation was identified involving the licensees failure to
properly prepare and ship a package containing radioactive material in a manner that
assured, under conditions normally incident to transport, conformance with Department
of Transportation (DOT) radiation level limitations specified by 49 CFR 173.441(a), (i.e.,
200 millirem per hour (mrem/h)) on any external surface of the package.
Description: On October 29, 2008, a Surface Contaminated Object (SCO) shipment of
fuel sipping equipment was surveyed, radiologically characterized, prepared, packaged
and shipped from the licensees site to Westinghouse Corporation in Waltz Mill,
Pennsylvania. The shipment was made on an exclusive use open transport vehicle with
no over-pack, barrier, or barricade restricting access to the radioactive package
(transport box). Prairie Islands final shipment surveys indicated that all applicable
survey parameters were below required DOT regulatory limits, with the maximum
radiation level on any package surface at 170 mrem/h (telepole measurement).
However, upon receipt at Waltz Mill, Pennsylvania, the package containing the fuel
sipping equipment was surveyed and found to have surface radiation levels on the
package underside (bottom) in excess of DOT regulatory limits. Westinghouse
Corporation notified the licensee and the State of Pennsylvania.
8 Enclosure
On November 3-4 and 12-14, 2008, Prairie Island Nuclear Generating Plant personnel
were on-site at the Waltz Mill facility to facilitate incident investigation. Members of the
Prairie Island and corporate health physics staff detected similar radiological conditions
as those initially reported by Westinghouse Corporation surveys. Final survey results
determined that the package surface radiation levels were 1630 mrem/h (telepole
measurement). Under controlled conditions at the Waltz Mill facility and under the
observation of Prairie Island radiation protection personnel, qualified individuals opened
the package containing the fuel sipping equipment to determine the source of the
elevated radiation levels. Upon examination, the licensees staff identified that a small
radioactive particle was embedded into the umbilical cable to the lid of the fuel sipping
canister. The fuel sipping equipment (lid and umbilical cable) was found to be not
properly braced, nor secured as required; apparently, the lid and the umbilical cable
shifted from the time of the Prairie Island shipping package departure survey to the time
of the Westinghouse Waltz Mill shipping package receipt survey (i.e., during transport).
Additionally, two other discrete radioactive particles were detected inside the shipping
box. Radiological surveys indicated that the particle embedded into the lids umbilical
cable exhibited a radiation level of about 11,000 mrem/h; and the two other discrete
radioactive particles exhibited radiation levels at a nominal 2000 mrem/h. Isotopic
analyses of the discrete radioactive particles identified cobalt-60 as the sole radioactive
isotope. The licensees staff found the fuel sipping canister lid and associated umbilical
cable in a location in the shipping package, that coincided with the elevated radiation
levels on the external surfaces of the package, as identified on the Westinghouse Waltz
Mill receipt radiation survey.
The detection of the highly radioactive fuel sipping canister lid umbilical cable and the
associated discrete radioactive particles indicated that Prairie Islands on-site efforts to
decontaminate and radiologically characterize the fuel sipping equipment prior to
shipment was not completely successful. The on-site radiological surveys were not
sufficient for detecting highly radioactive small particles; and that the fuel sipping
equipment was not properly braced nor secured in its package for shipment under
conditions normally incident to transport. Additionally, the phenomena of the
redistribution of highly radioactive particles and the potential for loads shifting during the
transportation of radioactively contaminated equipment are neither uncommon or
unknown to the industry.
The inspectors determined that, in the configuration used to transport the package, the
elevated radiation levels on the bottom external surface of the package would not be
routinely accessible to a member of the public. However, the licensee determined
through event reconstruction, that there were elevated radiation levels on the underside
of the transport trailer, that were unanticipated by the licensee at time of transport.
Further, the NRC concluded, that the elevated radiation levels, although on the
underside of the package, had the potential to adversely affect personnel who would
normally receive the package and/or respond to an incident involving the package with
the reasonable expectation that the package conformed to DOT radiation limitations.
Redistribution of the canister lid and associated umbilical cable likely occurred as a
result of conditions normally incident to transport. The inspectors concluded that it was
fortuitous and not the result of design nor package preparation, that the material was
deposited in such manner that effectively limited the potential for public exposure.
9 Enclosure
The inspectors identified issues associated with the surveying, preparation, and
packaging of the fuel sipping equipment. In addition to the ineffective radiological
characterization and packaging of the fuel sipping equipment, the inspectors determined
that the licensee did not provide nor ensure that the individuals involved in packaging the
shipment were trained and qualified for the task as required by 49 CFR 172.704
Training Requirements. None of the licensees staff involved in the loading and
preparation were trained, and only some of the Westinghouse personnel involved in
loading, preparing, and packaging of the fuel sipping equipment for transport were
trained and qualified for the task. Specifically, nine of the thirteen people involved in
preparing this package for radioactive shipment and transport had not received the
required function specific training. Two of the remaining four individuals had not
received the required recurrent training within the last three years.
As a follow-up to this issue, the licensee performed a dose assessment to evaluate
potential effective dose to a member of the public for a person reasonably close to the
radioactive point source. This review concluded that the potential for an exposure above
the NRC annual limit for a member of the public was minimal given the radiation
sources location and an individuals necessary orientation to come in sustained contact
with the radiation field. The inspectors reviewed the licensees data and concluded that
an overexposure to a member of the public would not be plausible based on the location
of the radiation levels and the corresponding dose rates. However, the inspectors
concluded that the licensees assessment did not provide any information that applied to
the applicable regulatory requirement concerning radiation levels on the surface of the
package.
Analysis: The inspectors concluded that the issues concerning the shipment constituted
a performance deficiency. Specifically, the licensee failed to characterize the distribution
of radioactive material on the equipment and to package the equipment in such a
manner that radioactive material would not shift nor redistribute within the package
under normal transportation conditions as required by 10 CFR 71.5 and 49 CFR 173.441
for the transportation of radioactive licensed materials. Additionally, the licensee did not
provide nor ensure that the individuals involved in this shipment were trained and
qualified for the task as required by 49 CFR 172.704.
The inspectors concluded that the failures were within the licensees ability to foresee,
correct, and should have been prevented, particularly, since the potential for loads to
shift and the redistribution of highly radioactive particles during transportation of
radioactively contaminated equipment is not uncommon and is known to the industry.
However, the matter had no actual safety consequence (i.e., no overexposure to any
member of the public) or impact on the NRCs ability to perform its regulatory function,
and there were no willful aspects associated with this finding. Also, after the finding was
identified, the licensees staff evaluated the radiological impact to the public for the
event, and implemented corrective actions including the suspension of all radioactive
material shipments to ensure that the finding did not present an immediate safety
concern. This finding was considered more than minor since it was associated with the
Public Radiation Safety Cornerstone program and process attribute relative to DOT
package radiation limits and affected the cornerstone objective to ensure adequate
protection of the public from exposure to radioactive materials released into the public
domain in that package radiation levels were elevated. Application of Manual Chapter
0609, Appendix D, the Public Radiation Safety Significance Determination Process is
10 Enclosure
applicable since the finding involved an occurrence in the licensees radioactive material
transportation program that was contrary to DOT regulations, (i.e., 49 CFR 173.441(a)).
The cause of this finding had a cross-cutting aspect in the area of Human Performance.
Specifically, the licensee failed to appropriately plan the work activity by incorporating
risk insights and job site conditions, including conditions which may impact radiological
safety (H.3 (a)). This finding is documented within the licensees corrective action
system as RCE 01157726.
Preliminarily, the significance of this finding is considered as having a substantial safety
significance, since the radiation level was greater than five times the limit (1000
mrem/h), but less than ten times the limit (2000 mrem/h) specified by the DOT regulatory
requirement. This determination is also reinforced by the determination that, though the
potential for public exposure was limited during transportation, that aspect was
fortuitous, and not the result of design nor package preparation by the licensee, and that
the condition had the potential to adversely affect personnel who would normally receive
the package or respond to an incident involving the package with a reasonable
expectation, that the package conformed with DOT radiation limitations.
Enforcement: An apparent violation associated with the preliminary Yellow performance
deficiency in the public radiation safety cornerstone was identified. Title 10 CFR 71.5,
Transportation of Licensed Material, requires licensees to comply with the Department
of Transportation (DOT) regulations in Title 49 CFR Parts 170 through 189 relative to the
transportation of licensed material. Specifically:
(1) Title 49 CFR 173.441(a) requires that each package of radioactive material offered
for transportation must be designed and prepared for shipment, so that under
conditions normally incident to transportation, the radiation level does not exceed
2 mSv/h (200 mrem/h) at any point on the external surface of the package.
Contrary to these requirements, on October 29, 2008, Northern States Power -
Minnesota (Prairie Island) shipped a package containing radioactive material that
was not sufficiently designed nor prepared to assure that, under conditions normally
incident to transportation, the radiation level on the external surface of the package
would not exceed 200 mrem/hour. When received and surveyed at the shipping
destination (Westinghouse in Waltz Mill, Pennsylvania), on October 31, 2008, the
external surface of the package exhibited radiation levels of 1630 mrem/h [i.e.,
package radiation levels greater than five and less than ten times the regulatory
limit].
(2) Title 49 CFR 172.704 Training Requirements requires that individuals involved in
the transport of hazardous materials receive function specific training relative to
their specific tasks, and that these individuals receive recurrent training at least
once every three years.
Contrary to the above, nine of the thirteen people involved in preparing this
package for radioactive shipment and transport had not received the required
function specific training. Also, two of the remaining four individuals had not
received the required recurrent training within the last three years.
11 Enclosure
Following identification of the apparent violation, Prairie Island Nuclear Generating Plant
documented the condition and initiated a root cause review (RCE 01157726); and
instituted corrective measures, including the suspension of radioactive material
shipments that were susceptible to discrete radioactive particle contamination. Pending
determination of a final safety significance, this finding is identified as an apparent
violation, (AV)05000282/2008009-01; 05000306/2008009-01, Radioactive Material
Shipment Package Radiation Levels Exceeded.
.5 Identification and Resolution of Problems
a. Inspection Scope
The inspectors reviewed condition reports, audits and self-assessments that addressed
radioactive waste and radioactive materials shipping program deficiencies since the last
inspection, to verify that the licensee had effectively implemented the corrective action
program, and that problems were identified, characterized, prioritized and corrected.
The inspectors also verified that the licensee's self-assessment program was capable of
identifying repetitive deficiencies or significant individual deficiencies in problem
identification and resolution.
The inspectors reviewed corrective action reports from the radioactive material and
shipping programs since the previous inspection, interviewed staff and reviewed
documents to determine if the following activities were being conducted in an effective
and timely manner commensurate with their importance to safety and risk:
- Initial problem identification, characterization, and tracking;
- Disposition of operability/reportability issues;
- Evaluation of safety significance/risk and priority for resolution;
- Identification of repetitive problems;
- Identification of contributing causes;
- Identification and implementation of effective corrective actions;
- Resolution of NCVs tracked in the corrective action system; and
- Implementation/consideration of risk significant operational experience feedback.
This inspection constituted one sample as defined in IP 71122.02-5.
b. Findings
No findings of significance were identified.
12 Enclosure
4. OTHER ACTIVITIES
4OA6 Management Meetings
.1 Exit Meeting Summary
On January 21, 2009, via a telephone conference call, the inspectors presented the
inspection results to Mr. M. D. Wadley, Site Vice President, and other members of the
licensee staff. The licensee acknowledged the issues presented. The inspectors
confirmed that none of the potential report input discussed was considered proprietary.
ATTACHMENT: SUPPLEMENTAL INFORMATION
13 Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
S. Derleth, Radiation Protection Specialist
B. Hite, Station Radiation Protection/Chemistry Manager
M. Kent, Radiation Protection Supervisor
K. Kono, Radiation Protection Specialist
S. Nelson, Corporate Radiation Protection Manager
S. Rupp, Radiation Protection Specialist
C. Sweet, Radiation Protection Specialist
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
Opened
05000282/2008009-01; AV Radioactive Material Shipment Package Radiation Levels05000306/2008009-01 Exceeded. (Section 2PS2)05000282/2008009-02; NCV Failure to Perform Formal Job Planning to Evaluate the
05000306/2008009-02 Radiological Hazards. (Section 2PS2)
Closed
05000282/2008009-02; NCV Failure to Perform Formal Job Planning to Evaluate the
05000306/2008009-02 Radiological Hazards.
Discussed
None
1 Attachment
LIST OF DOCUMENTS REVIEWED
The following is a list of documents reviewed during the inspection. Inclusion on this list does not
imply that the NRC inspectors reviewed the documents in their entirety, but rather, that selected
sections of portions of the documents were evaluated as part of the overall inspection effort.
Inclusion of a document on this list does not imply NRC acceptance of the document or any part of it,
unless this is stated in the body of the inspection report.
2OS1 Access Control to Radiologically Significant Areas
AR 1123112; Various Boxes in Radiological Controlled Area Not Labeled; January 8, 2008
AR 1126185; 121 Spent Resin Tank Inlet Wont Open; February 5, 2008
AR 1144127; Two Gauges Found in Hot Instrument Shop without RAM Labels; July 14, 2008
AR 1127294; Safety and Radiation Issues Regarding Resin Sluices; February 14, 2008
AR 1131225; Accumulated Trash Results in Increased Dose Rates; March 15, 2008
AR 1131673; Adverse Trend - Resin Sluicing Issues, March 19; 2008
AR 1137541; Continued Issues with Waste Gas System; May 13, 2008
AR 1145069; Elevated Dose Rates on Resin Sluice Line; July 23, 2008
AR 1150471; Unexpected Dose Readings During Resin Sluice; September 12, 2008
AR 1150626; Iron-55 Elevated In Rad Waste Tanks; September 15, 2008
AR 1151006; Pressurized Drums of Floor Mop Slop; September 18, 2008
AR 1152117; 2R-25 Personnel Contamination Event During Steam Generator Manway Insert
Removal; September 25, 2008
AR 1152179; Radiation Protection Sign-Off Steps Missed in D67 Incore Instrumentation;
September 26, 2008
AR 1156041; Workers Entered High Radiation Area Without Required Brief; October 17, 2008
AR 1157246; Frisker Source Removed From Site; October 29, 2008
AR 1162343; ALARA Planning Not Performed for 2R25 Fuel Sipping; December 12, 2008
FP-RP-BP-01; Bioassay Program; Revision 05
FP-RP-JPP-01; RP Job Planning; Revision 04
FP-RP-RWP-01; Radiation Work Permit; Revision 06
FP-RP-SD-01; Special Dosimetry; Revision 04
2 Attachment
Fuel Sipper Lid Hot Particle Extremity and Whole Body Dose Evaluation; December 12, 2008
Fuel Sipping Removal Sentinel Exposure Tracking for 22-25 October 2008; dated
December 12, 2008
RPIP 1121; RWP Issue; Revision 23
RPIP 1122; Hot Particle Program; Revision 13
RPIP 1126; Contamination Monitor Alarm Response and Personnel Decontamination:
Revision 24
RPIP 1130; On the Job Dose Monitoring and Timekeeping; Revision 19
RPIP 1135; RWP Coverage; Revision 19
Qualification History Reports; December 08, 2008
Work Order Package 00367253; September 24, 2008
2PS2 Radioactive Material Processing and Transportation
Shipment Number 06-026; Refueling Equipment to Westinghouse; May 26, 2006
Shipment Number 06-028; Refueling Equipment to Westinghouse; June 5, 2006
Shipment Number 06-066; Refueling Equipment to Westinghouse; December 13, 2006
Shipment Number 070-011; Spent Resin to Studsvik; July 10, 2007
Shipment Number 08-069; Refueling Equipment to Westinghouse; October 29, 2008
AR 1119827; NRC Radioactive Materials Inspection: Process Control Program and Updated
Safety Analysis Report Do Not Accurately Reflect Current Plant Practice; December 3, 2007
AR 1121239; Effluent and Waste Disposal Annual Report; December 14, 2007
AR 1121243; D59, Revision 8 Process Control Program for Solidification; December 14, 2007
AR 1157726; PI Shipment Arrives at Consignee Above DOT Rad Limits; October 31, 2008
AR 1158879; Sea land Dose Rates; November 12, 2008
AR 1160011; Radiation Survey Not Complete; November 21, 2008
AR 1160060; NRC Observation - D11 Procedures Need Work; November 11, 2008
C21.1.3.7; Spent Resin; Revision 15
D11; Radioactive Material Shipment; Revision 16
3 Attachment
D11.4; Radioactive Material Shipment Greater Than Type A Quantities in Exclusive Use
Vehicle to Barnwell, SC Using RWE Nukem Cask and High Integrity Container Liner;
Revision 23
D11.7; Radioactive Material Shipment LSA/SCO/LDT Quantity to a Licensed Facility;
Revision 15
D20.12; Sluicing Resin From 11 Mixed Bed Ion Exchanger to 121 Spent Resin Tank;
Revision 19
D20.24; Sluicing Resin From 11 Steam Generator Blowdown Ion Exchanger to Barrels;
Revision 16
D20.17; Sluicing Resin From 11 Evaporator Feed Ion Exchanger to Low Level Resin Liner;
Revision 11
Radiological Survey Records; Various Dates
RPIP 1310; RadWaste Streams Scaling Factors; Revision 8
RPIP 1314; Solid Radioactive Waste Annual Report; Revision 9
RPIP 1319; Loading LSA Boxes/Sealand Containers; Revision 10
RPIP 1325; Shipping RAM using Radman/Radship Software; Revision 3
RPIP 1322; Radman for Windows to Generate Scaling Factors; Revision 5
RPIP 1721; Resin Sluice; Revision 17
4 Attachment
LIST OF ACRONYMS USED
ALARA As-Low-As-Is-Reasonably-Achievable
AR Action Request
AV Apparent Violation
CFR Code of Federal Regulation
DOT Department of Transportation
FP Fleet Procedure
IMC Inspection Manual Chapter
IP Inspection Procedure
NCV Non-Cited Violation
NRC Nuclear Regulatory Commission
RP Radiation Protection
RPIP Radiation Protection Implementing Procedure
RWP Radiation Work Permit
SDP Significance Determination Process
5 Attachment