ML083660296
| ML083660296 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 12/31/2008 |
| From: | Christine Lipa Division of Nuclear Materials Safety III |
| To: | O'Connor T Northern States Power Co |
| References | |
| IR-08-003 | |
| Download: ML083660296 (28) | |
See also: IR 07200058/2008003
Text
December 31, 2008
Mr. Timothy J. OConnor
Site Vice President
Monticello Nuclear Generating Plant
Northern States Power Company, Minnesota
2807 West County Road 75
Monticello, MN 55362-9637
SUBJECT:
MONTICELLO NUCLEAR GENERATING PLANT
NRC INSPECTION REPORT 072-00058/2008-003(DNMS)
Dear Mr. OConnor:
On December 22, 2008, the U.S. Nuclear Regulatory Commission (NRC) completed its
inspection of the preoperational testing of an Independent Spent Fuel Storage Installation
(ISFSI) at the Monticello Nuclear Generating Plant. The inspection focused on the pre-
operational demonstrations and program reviews associated with preparations to load fuel as
well as the actual loading activities. The dry run inspection consisted of in-office review
beginning April 12, 2008, and concluded with an exit teleconference on December 22, 2008,
with onsite inspections June 30 through July 3, August 18 through 22, and September 8 through
September 11, 2008. The enclosed report presents the results of this inspection.
The inspection consisted of observations of the dry run activities utilizing the Transnuclear
NUHOMS 61 BT cask and its storage system and activities associated with loading, transfer,
and storage of dry fuel as they relate to safety and compliance with the Commissions rules and
regulations and with the conditions of the license. Areas examined during the inspection are
identified in the enclosed report. Within these areas, the inspection consisted of interviews with
licensee personnel, as well as a review of select procedures and programs.
Based on the results of this inspection, the NRC has determined that a Severity Level IV
violation of NRC requirements occurred. The violation was associated with a failure to establish
measures to ensure that applicable regulatory requirements and the design basis were correctly
translated into specifications, drawings, procedures, and instructions. This finding had a cross-
cutting aspect in the area of Human Performance, Resources, because the design control
process did not establish requirements necessary for complete, accurate, and up-to-date design
documentation.
Because the violation was of very low safety significance, was non-repetitive, and was entered
into the corrective action program, this violation is being treated as a Non-Cited Violation (NCV),
consistent with Section VI.A of the NRC Enforcement Policy. The NCV is described in the
subject inspection report. If you contest the violation or significance of this NCV, you should
provide a response within 30 days of the date of this inspection report, with the basis for your
denial, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC
20555-0001, with a copies to the Regional Administrator, Region III, the Director, Office of
Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001, and
the NRC Resident Inspector at the Monticello Nuclear Generating Plant.
T. OConnor
-2-
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its
enclosure((s), 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 NRCs
document system (ADAMS), accessible from the NRC Web site at
http://www.nrc.gov/readingrm/adams.html. To the extent possible, your response should not
include any personal privacy, proprietary, or safeguards information so that it can be made
available to the Public without redaction.
Sincerely,
/RA by J. Madera Acting for/
Christine A. Lipa, Chief
Materials Control, ISFSI, and
Decommissioning Branch
Docket No.72-058; 50-263
License No. DPR-22
Enclosure:
Inspection Report 072-00058/2008-003(DNMS)
cc w/encl:
D. Koehl, Chief Nuclear Officer
Manager, Nuclear Safety Assessment
P. Glass, Assistant General Counsel
Nuclear Asset Manager, Xcel Energy, Inc.
J. Stine, State Liaison Officer, Minnesota Department of Health
R. Nelson, President
Minnesota Environmental Control Citizens Association
Commissioner, Minnesota Pollution Control Agency
R. Hiivala, Auditor/Treasurer,
Wright County Government Center
Commissioner, Minnesota Department of Commerce
Manager - Environmental Protection Division
Minnesota Attorney Generals Office
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OFFICE
RIII
RIII
RIII
RIII
NAME
JENeurauter:jc*
SRBakhsh
CALipa
DATE
12/24/08
12/31/08
12/31/08
OFFICIAL RECORD COPY
Letter to Timothy OConnor from Christine A. Lipa dated December , 2008
DISTRIBUTION:
Mark Satorius
Steven Reynolds
Cynthia Pederson
Kenneth OBrien
Christopher Thomas
Silvia Brouillard
David Hills
Patricia Buckley
Nick Shah
William Snell
Enclosure
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket No.
072-00058
License No.
Report No.
072-00058/2008-003(DNMS)
Licensee:
Northern States Power Company
Facility:
Monticello Nuclear Generating Plant
Location:
2807 West County Road 75
Monticello, MN 55362-9637
Inspection Dates:
Onsite: June 30 through July 3, 2008; August 18 through
22, 2008; and September 8 through September 11, 2008.
In-office review completed on December 22, 2008
Exit Teleconference: December 22, 2008
Inspectors:
Sarah Bakhsh, Reactor Inspector
Matthew Learn, Reactor Engineer in training
Scott Atwater, Senior Project Inspector, Region II
John Bozga, Reactor Inspector,
James Neurauter, Senior Reactor Inspector
Jim Pearson, Senior Safety Inspector, Division of Spent
Fuel Storage and Transportation, Office of Nuclear
Material Safety and Safeguards
Approved by:
Christine A. Lipa, Chief
Materials Control, ISFSI, and Decommissioning Branch
Division of Nuclear Materials Safety
Enclosure
2
EXECUTIVE SUMMARY
Monticello Nuclear Generating Station
NRC Inspection Report 072-00058/2008-003(DNMS)
Preoperational Testing of an Independent Spent Fuel Storage Facility Installation at Operating
Plants (60854.1)
The inspectors observed the licensees dry run activities utilizing the Transnuclear NUHOMS
61 BT cask and its storage system and activities associated with loading, transfer, and
storage of dry fuel as they relate to safety and compliance with the Commissions rules and
regulations and with the conditions of the license.
The inspectors identified one violation of 10 Code of Federal Regulations (CFR) 72.146,
Design Control. Specifically, the licensee failed to establish measures to ensure that
applicable regulatory requirements and the design basis were correctly translated into
specifications, drawings, procedures, and instructions. This finding is being treated as a
Non-Cited Violation, consistent with section VI.A of the NRC Enforcement Policy. The
finding has a cross-cutting aspect in the area of Human Performance, Resources, because
the licensees design control process did not establish requirements necessary for complete,
accurate, and up-to-date design documentation. H.2(c) (Section 1.0)
Review of 10 CFR 72.212(b) Evaluations (60856)
The inspectors reviewed the licensees 10 CFR 72.212 evaluation and determined that it
was in compliance with conditions set forth in the Certificate of Compliance, Final Safety
Analysis Report, and 10 CFR Part 72 requirements in regards to the NUHOMS 61BT cask
system. (Section 2.0)
Enclosure
3
REPORT DETAILS
1.0
Preoperational Testing Of an Independent Spent Fuel Storage Facility Installation
(ISFSI) at Operating Plants (60854.1)
a. Inspection Scope
The inspectors evaluated the licensees readiness to load spent fuel. The inspectors
observed the licensees dry run activities utilizing the Transnuclear NUHOMS 61 BT
cask and its storage system and activities associated with loading, transfer, and storage
of dry fuel as they relate to safety and compliance with the Commissions rules and
regulations and with the conditions of the license. The licensee faced several
challenges and the NRC identified several issues during the dry run inspection phase,
and these issues were subsequently resolved satisfactorily prior to loading spent fuel.
b.
Observations and Findings
Heavy Loads
The inspectors reviewed the licensees crane and heavy loads program with regards to
ISFSI operations. The inspectors reviewed topics associated with the reactor building
cranes hoisting system, wire rope, bridge and trolley, controls, crane inspection and
maintenance, load testing, limit switches, operation, and safe load paths. The inspection
consisted of documentation review, interviews with staff, and an inspection of the reactor
building crane.
The inspectors reviewed that the reactor building crane had been static loaded to
approximately 125 percent of the 105-ton maximum critical load on its main hook.
The inspectors verified that a nondestructive examination of the welds, whose
failure could result in the drop of a critical load, was performed following the 125 percent
cold-proof testing. After the 125 percent load test, the crane was given a full
performance test with approximately 100 percent of the maximum critical load attached.
The inspectors verified that the default minimum crane operating temperature was
defined as 70 degrees Fahrenheit in loading procedures. A 200 percent static load test
had been performed for each load-attaching hook. The hook load testing was followed
by a nondestructive examination and geometric measurements to verify the soundness
of fabrication and ensure integrity of the hook. All limiting and safety control devices
were tested.
The inspectors reviewed the cranes hoist brake system and observed the variable
frequency power control braking system and three holding brakes. Holding brakes
were tested to automatically apply the full holding position when power is off, and under
overspeed and overload conditions. The inspectors verified the cask height during
movement was sufficiently high to allow for engaging of the brakes during an
uncontrolled descent before the load would impact the floor and reviewed the licensees
procedure for emergency positioning of the crane and lowering the load.
The cranes reeving system consisted of two drums with quadruple reeving of four
wire ropes using sheave equalizers. The hoisting system had two mechanical load
switches installed in the equalizer sheave that were used to de-energize the hoist
drive motor and the main power supply under a load hang-up condition, but would still
Enclosure
4
allow a controlled lowering of the load. The Monticello Nuclear Generating Plant
(MNGP) reactor building crane employs a system of three independent upper travel limit
switches to prevent two-blocking (lower block coming in contact with the drum). The
inspectors also observed the lower limit switch and verified that a sufficient amount of
wraps around the drum were present at the lower limit. These devices de-energize the
hoist drive motor and the main power supply. The hoist drum was equipped with drum
capture plates put in place to limit drum drop during a shaft or bearing failure.
The inspectors reviewed the latest annual preventive maintenance program and crane
inspection. The annual inspection also replaces and installs recently calibrated
mechanical load switches used to prevent load hang-up. During ISFSI operations, the
MNGP crane was categorized as being under normal service. This categorization
required a frequent check on a monthly basis. The inspectors reviewed the cranes
daily inspection list.
The inspectors observed the licensee test electrical interlocks that permit only one
control station to be operated at a time. The inspectors reviewed the operators
qualifications; the licensee qualified the ISFSI crane operators based on a review of their
previous training, education, experience, and medical records. The inspectors observed
the emergency stop features in the cab, on the refuel floor and on the remote control
unit. The inspector reviewed the safe load paths defined for the movement of heavy
loads.
Dry Run Demonstrations
Inspectors observed the licensees NRC dry run activities in preparations to load fuel at
the MNGP August 18, 2008, through August 22, 2008. Additional operations, in
particular the welding demonstration by TriVis, were observed by inspectors prior to the
NRC dry run at the contractors facility in Pelham, Alabama and the inspection results are
documented in inspection report 072-00058/2008-002(DNMS). The licensee faced
challenges with several canisters received from the manufacturer, Transnuclear (TN),
due to fabrication tolerances, i.e., build-up of the specified fabrication tolerances resulted
in improper alignment of the outer top cover plate with the canister shell weld
preparation. Due to this misalignment, the weld configuration had to be modified from a
dual to a single bevel for the affected canisters. Transnuclear evaluated the issue and
concluded the affected DSCs could be placed into service without any additional repairs,
rework, testing, or weld demonstrations. The licensee documented this issue in the
corrective action program as Action Request (AR) 01144172.
The inspectors reviewed the loading and unloading procedures to ensure that they
contained commitments and requirements specified in the license, the Technical
Specifications, the Final Safety Analysis Report (FSAR), and Title 10 Code of Federal
Regulations (CFR), Part 72. The inspectors observed the licensees pre-job briefings.
The licensee conducted these meetings in a professional manner where the necessary
items to enhance safety were discussed. Radiation protection staff attended pre-job
briefs and gave insight into working conditions and As-Low-As-Is-Reasonably-
Achievable (ALARA) practices. The staff was interactive and questions were addressed,
as well as suggestions considered by supervisors to gain additional insight.
Enclosure
5
The inspectors observed licensee personnel perform a number of activities associated
with dry fuel storage to demonstrate their readiness to safely load spent fuel from the
spent fuel pool (SFP) into the dry cask storage system. The inspectors observed the
loading and unloading of dummy fuel bundles into the storage canister basket. The
licensee demonstrated removal of dummy fuel assemblies from the SFP storage racks,
placed them into the canister, and returned them from the canister to the SFP racks.
The licensee demonstrated alignment of the hold down ring and the shield plug.
The inspectors observed crane operation to ensure that heavy loads could be safely
lifted and transferred. Down ending of the transfer cask containing a storage canister
filled with dummy assemblies from the refueling floor to the transfer trailer was observed
as well as lifts from the transfer trailer to the refueling floor. Due to space limitations
during the down ending evolution, the licensee had to move the crane and transfer trailer
simultaneously to properly lower the transfer cask. The inspectors observed the
licensees response to overspeed trips of the trolley during the down ending due to the
trolley being positioned in front of the load without sufficient lowering. The licensee
determined that this occurred when the trolley control was returned to neutral, and the
trolley positioned itself above the load. As a contingency the licensee moved the
transfer trailer and main hoist to complete the demonstration. For future down ending,
the licensee decided to maximize the transfer trailer motion and minimize the trolley
motion, which proved to be successful. The licensee documented this issue in
The inspectors also observed a lift of the transfer cask out of the spent fuel pool and
onto the cask preparation area. Inspectors verified that lifts were performed in
accordance with appropriate industry standards and followed the designated safe haul
path.
Inspectors observed the installation of the transfer cask lid, as well as removal of the lid
at the Horizontal Storage Module (HSM). The inspectors observed the successful
transfer of the storage canister to the ISFSI. During the licensees internal
demonstrations for dry shielded canister (DSC) insertion and retraction from an HSM,
the DSC shell was damaged due to misalignment with the HSM. Canister DSC-001 was
sent to TN for repair prior to use for storage of spent fuel. During the NRC dry run
demonstration the inspectors observed both successful insertion and retraction of the
storage canister from the HSM. The licensee documented this issue in AR 01145084.
Proper controls were in place during the transfer of the canister from the reactor building
to the HSM on the ISFSI. These controls included health physics coverage, adherence
to the heavy haul path, and appropriate security oversight. The inspectors verified
adequate communication and team work between departments and adherence to
procedures.
Fuel Selection
The inspectors reviewed the licensees processes and methods associated with fuel
characterization and selection. The inspectors reviewed a completed fuel selection
package for the first cask to be loaded during the campaign to verify that the licensee
used the criteria specified in the Technical Specifications to verify the acceptability of
assemblies to be loaded in a cask. The inspectors observed the licensees methods to
independently verify and document fuel assemblies. The licensee did not plan to load
any damaged fuel assemblies during this campaign.
Enclosure
6
Radiation Protection
The inspectors evaluated the licensees radiation protection program pertaining to the
operation of the ISFSI. The inspectors reviewed the licensees procedures describing
the methods and techniques used when performing dose rate and surface contamination
surveys and verified that they ensured dose rate limits and surveillance requirements of
the Technical Specifications were met. The inspectors interviewed the licensees
personnel to verify their knowledge regarding the scope of the work and the radiological
hazards associated with transfer and storage of spent fuel.
Training
The inspectors reviewed the licensees training program which consisted of classroom
and on-the-job training to ensure involved staff was adequately trained for the job they
were responsible to perform. The licensees contractor prepared a dry fuel storage
qualification matrix which documented each workers training courses completed.
The inspectors reviewed the training material, including the content of the manuals.
Training material topics were consistent with TN Technical Specifications. The
inspectors independently verified satisfactory completion of training by applicable staff
by comparing training documentation in the contractors qualification matrix to the
licensees Learning Management System. The inspectors interviewed select individuals
who were responsible for performance of specific tasks during loading to evaluate their
knowledge regarding the campaign activities, the cask loading process, and use of the
equipment.
The inspectors reviewed training records of welders and other personnel who the
licensee authorized to perform the non-destructive examination inspections to ensure
that these individuals training was current.
Quality Assurance
The inspectors reviewed the licensees Quality Assurance program, as it applied to
the ISFSI. The inspectors also reviewed procedures pertaining to the receipt inspection
of dry shielded canisters, transfer trailer, transfers cask, and auxiliary equipment. The
inspectors observed that gauges were within their calibration date, and that the use of
99.999 percent pure helium was used during backfilling.
Emergency Preparedness and Fire Plan
The inspectors reviewed the licensees emergency preparedness plan required by
10 CFR Part 50.47 for conformance with 10 CFR 72.32(c). The inspectors verified
that the licensee incorporated Emergency Action Levels to the plant emergency plan
to address the possible emergency scenarios, their classification, and recovery actions
associated with the ISFSI. The inspectors interviews with staff revealed confusion
regarding protected area and plant protected area, which the licensee clarified with
staff and made enhancements to the definitions to clarify the two terms for their use in
EAL classifications. In response to this NRC-identified issue, the licensee initiated
Enclosure
7
The inspectors reviewed the licensees Fire Hazard Analysis (FHA) at the ISFSI for
compliance with the regulations in 10 CFR Part 72 and the Certificate of Compliance
(CoC). The inspectors identified inconsistencies in the evaluation regarding the
minimum separation distance for vehicles and addition of a control on transient
combustibles. In response to the NRC identified issues with the FHA, the licensee
initiated AR 01146176.
Structural Modifications and Associated Design Documentation
The inspectors reviewed plant design documentation, design calculations, safety
evaluations, and resultant structural modifications that demonstrated the fuel cask could
be safely placed into the SFP and loaded with spent fuel, lifted from the SFP and placed
on the designated laydown areas, transferred to the transport vehicle, and transported to
the ISFSI. The inspectors verified that the fuel cask loading, unloading, and transfer
activities met MNGP site specific commitments and requirements with respect to the
Specifically, the inspectors reviewed the licensees structural calculations associated
with the reactor building superstructure, the structural integrity of the Rail Car Shelter
(RS), Reactor Building cask laydown areas, and the spent fuel pool structure to support
the 105-ton cask load. The inspectors also reviewed the licensees structural calculation
associated with the buried utilities along the haul path to support the 105-ton cask load.
Lastly, the inspectors reviewed the licensees structural calculation associated with the
transfer cask hazard for a postulated RS collapse during a design basis tornado (DBT)
event.
The inspectors identified technical errors in Calculation CA-08-135, Transfer Cask
Hazard from Rail Car Shelter Collapse, Revision 0, which resulted in exceeding the
acceptance criteria of the calculation. In response to the NRC identified technical errors,
the licensee initiated AR 01149709. The licensee removed conservative assumptions in
the calculation, and revised Calculation CA-08-135, Transfer Cask Hazard from Rail
Car Shelter Collapse. Revision 1 of CA-08-135 was reviewed by the inspectors and no
technical issues were identified. Therefore, the NRC identified errors were determined
to be minor using Inspection Manual Chapter (IMC) 0612, Appendix E, Examples of
Minor Issues.
The inspectors identified a violation of 10 CFR 72.146, Design Control, involving the
licensees design control process performed for the RS for the ISFSI transfer operations.
Specifically, the inspectors identified a failure to assure and verify structural integrity of
the RS due to the effects of a DBT event in accordance with ISFSI licensing
requirements. This licensing issue was identified during review of calculation
CA-05-104, Design Adequacy of the Rail Car Shelter @ Elevations 935-0, for the
ISFSI Transfer Operations, which analyzed the RS for design basis loading conditions
associated with the ISFSI during transfer operations.
The inspectors reviewed calculation CA-05-104, which evaluated the RS structural
integrity to withstand a design basis earthquake to demonstrate no collapse onto the
transfer cask. This calculation provided the basis for storing the transfer cask in the RS
during ISFSI transfer operations The licensee was required by 10 CFR 72.92(a) to
identify the natural phenomena that could occur in the region and to assess their
potential effects on the safe operation of the ISFSI or Monitored Retrievable Storage
Enclosure
8
Installation (MRS). The important natural phenomena that affect the ISFSI or MRS
design must be identified. According to the Monticello Updated Safety Analysis Report,
Section 2.3.5, tornadoes were a natural phenomena that could occur at the Monticello
site. The licensee was further required by 10 CFR 72.122(b)(2)(i) to design its IFSFI
structures, systems, and components to withstand the effects of natural phenomena
such as earthquakes, tornadoes, lightning, hurricanes, floods, tsunami, and seiches,
without impairing their capability to perform their intended design functions.
The inspectors noted that for the ISFSI transfer operation, as implemented, the licensee
failed to assure and verify the integrity of the fuel cask system for a potential collapse of
the RS during a DBT event. Specifically, since the licensee failed to evaluate the effects
of a DBT on the RS structure, the licensee did not demonstrate that a collapse of the RS
structure onto the fuel cask system during a DBT event would not have invalidated the
licensing basis requirement of the fuel cask system to withstand tornado effects (wind
force, missiles, and differential pressure) as described in Table 3.2-1 of NUH-003,
Revision 10, NUH003.0103, Updated Final Safety Analysis Report for the Standardized
NUHOMS Horizontal Modular Storage Systems for Irradiated Nuclear Fuel. In
response to this issue, the licensee initiated AR 01142790.
In response to AR 01142790, the licensee performed additional analysis that provided
reasonable assurance the integrity of the fuel cask system would be maintained during a
DBT event while inside the RS.
The inspectors noted that the licensees failure to evaluate the RS for the effects of a
DBT event warranted a significance evaluation. The inspectors determined the
performance deficiency was within the licensees ability to foresee and correct because
the error could have been identified during the independent review.
Because this issue was related to an ISFSI license, it was dispositioned using the
traditional enforcement process per Supplement I of the Enforcement Policy.
In accordance with Inspection IMC 0610, Nuclear Material Safety and Safeguards
Inspection Reports, the inspectors determined that the deficiency was more than minor
in accordance with IMC 0610, Nuclear Material Safety and Safeguards Inspection
Reports, Section 06, which references the use of IMC 0612, Power Reactor Inspection
Reports Appendix E. The deficiency was determined to be more than minor using
IMC 0612, Appendix E, Examples of Minor Issues, Example 3k, in that the design
package did not assure cask integrity during a DBT and additional calculations were
required to evaluate the effects of the DBT during transfer operations through the RS in
accordance with the ISFSI licensing/design basis analysis requirements.
The finding was determined to be a Severity Level IV Violation per Enforcement Policy,
Supplement I, example D.3, a failure to meet regulatory requirements that have more
than a minor safety or environmental significance. Specifically, Calculation CA-08-135,
Transfer Cask Hazard from Rail Car Shelter Collapse, when updated by Revision 1,
demonstrated that the integrity of the fuel cask system was in accordance with licensing
requirements even if a collapse of the RS were to occur during a design basis tornado
event.
Enclosure
9
This finding has a cross-cutting aspect in the area of Human Performance, Resources,
because the licensees design control process did not establish requirements necessary
for complete, accurate, and up-to-date design documentation. Specifically, the
appropriate ISFSI design and licensing basis requirements related to a DBT were not
established for all structures and components that could affect the transfer cask during
ISFSI transfer operations. H.2(c)
Title 10 CFR Part 72.146, Design Control, required, in part, that the licensee, applicant
for a license, certificate holder, and applicant for a CoC shall establish measures to
ensure that applicable regulatory requirements and the design basis, as specified in the
license or CoC application for those structures, systems, and components to which this
section applies, are correctly translated into specifications, drawings, procedures, and
instructions. Further, it required that the design control measures must provide for
verifying or checking the adequacy of design by methods such as design reviews,
alternate or simplified calculation methods, or by a suitable testing program.
Title 10 CFR Part 72.92(a), Design Bases External Natural Events, requires, in part,
that natural phenomena that may exist or that can occur in the region of a proposed site
must be identified and assessed according to their potential on the safe operation of the
Title 10 CFR 72.122(b)(2)(i) requires that structures, systems, and components
important to safety must be designed to withstand the effects of natural phenomena
such as earthquakes, tornadoes, lightning, hurricanes, floods, tsunami, and seiches,
without impairing their capability to perform their intended design functions.
Contrary to the above, as of May 30, 2008, the licensee failed to establish measures to
ensure that applicable regulatory requirements and the design basis, as specified in the
license or CoC application for those structures, systems, and components to which this
section applies, were correctly translated into specifications, drawings, procedures, and
instructions. Specifically, the licensee failed to establish measures to ensure that the
tornado design bases accident analyses were correctly translated into specifications,
drawings, procedures, and instructions. Licensee Calculation CA-05-104, Design
Adequacy of the Rail Car Shelter at Elevation 935-0 for the ISFSI Transfer Operations
did not evaluate the adequacy of the Rail Car Shelter, a structure important to safety, for
tornado conditions, an applicable regulatory requirement.
Because this violation was of very low safety significance, was non-repetitive, and was
entered into the corrective action program (AR 01157276), it is being treated as a
Non-Cited Violation (NCV), consistent with Section VI.A of the NRC Enforcement Policy
(NCV 07200058/2008-003-01).
c. Conclusion
The inspectors observed the licensees dry run activities utilizing the Transnuclear
NUHOMS 61 BT cask and its storage system and activities associated with loading,
transfer, and storage of dry fuel as they relate to safety and compliance with the
Commissions rules and regulations and with the conditions of the license.
The inspectors identified one violation of 10 CFR 72.146, Design Control. Specifically,
the licensee failed to establish measures to ensure that applicable regulatory
Enclosure
10
requirements and the design basis were correctly translated into specifications,
drawings, procedures, and instructions. This finding is being treated as an NCV,
consistent with Section VI.A of the NRC Enforcement Policy. The finding has a cross-
cutting aspect in the area of Human Performance, Resources, because the licensees
design control process did not establish requirements necessary for complete, accurate,
and up-to-date design documentation. H.2(c)
2.0
Review of 10 CFR 72.212(b) Evaluations (60856)
a. Inspection Scope
The inspectors reviewed the licensees 10 CFR 72.212 evaluation to determine its
acceptability and compliance with conditions set forth in the CoC, the FSAR, and
10 CFR Part 72 requirements in regards to the NUHOMS 61BT cask system.
b.
Observations and Findings
The inspectors reviewed portions of select documents referenced in the evaluation,
including but not limited to radiological evaluations, fire hazard analysis, quality
assurance topical report, records management procedure, and documentation of
subsurface profiles.
The inspectors identified needed enhancements and weaknesses in the level of
information in the evaluation. In particular, the inspectors determined that the licensee
needed to add specific language to their 10 CFR 72.212 response to 72.212(b)(2)(c) in
addressing 72.104(c) which requires that operational limits be established for radioactive
materials in effluents and direct radiation levels associated with the ISFSI. The
evaluation, in regard to satisfying 72.104(c), did not include Monticellos specific
approach to establishment of operational limits.
The licensee also needed to address how it would store all quality records in the
appropriate records management system. The inspectors noted that that the final record
location for many documents was not fixed, as many documents were not yet transferred
from a working location to the recognized records management system for each of the
documents. The team discussed this situation with the ISFSI Project representative and
indicated that all records should become resident in the proper system prior to loading
fuel. In response to the NRC identified issues, the licensee initiated AR 01145347 and
01146174.
c. Conclusion
The inspectors reviewed the licensees 10 CFR 72.212 evaluation and determined that it
was in compliance with conditions set forth in the CoC, the FSAR, and 10 CFR Part 72
requirements in regards to the NUHOMS 61BT cask system.
Enclosure
11
3.0
Exit Meeting Summary
Interim debriefs regarding heavy loads were conducted on July 3, 2008,
August 22, 2008, and September 11, 2008. An exit meeting for inspection procedure
60854.1 was held on December 22, 2008. The inspectors presented the inspection
results to members of the licensee management and staff. Licensee personnel
acknowledged the information presented. The inspectors asked licensee personnel
whether any materials examined during the inspection and requested to be taken offsite
should be considered proprietary. No proprietary information was identified.
Attachment: Supplemental Information
Attachment
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
B. Brown, ISFSI Project Support
N. French, Operations Support Manager
S. Quiggle, ISFSI Project Manager
L. Samson, Manager, Spent Nuclear Fuel Storage
K. Shriver, ISFSI Project Support
Licensee and Contractor Employees in Attendance at July 3, 2008 Interim Debrief on ISFSI
R. Baumer, Compliance Engr Analyst (Regulatory Affairs)
- T. Blake, Regulatory Affairs Manager
- B. Brown, ISFSI Project Support
D. Crofoot, Nuclear Oversight (NOS) Supervisor
J. Gitzen, Cranes and Heavy Loads System Engineer
J. Grubb, Director Site Engineering
R. Lindberg, Sargent and Lundy Project Manager
- T. J. OConnor, Site Vice President
- S. Quiggle, ISFSI Project Manager
G. Ridder, ISFSI Project Engineer Nathan French - Operations Support Manager
- L. Samson, Manager, Spent Nuclear Fuel Storage
- B. Sawatzke, Plant Manager
- K. Shriver, ISFSI Project Support
- Indicates individuals present at the August 22, 2008 debrief
- Indicates individuals present at the September 11, 2008 debrief
Licensee and Contractor Employees in Attendance at the Exit Teleconference on December 22,
2008
T. Blake, Regulatory Affairs Manager
K. Shriver, ISFSI Project Support
INSPECTION PROCEDURES USED
Preoperational Testing Of An Independent Spent Fuel Storage Facility
Installation (ISFSI) At Operating Plants
Review of 10 CFR 72.212(b) Evaluations (60856)
Attachment
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
07200058/2008-003-01
Rail Car Shelter Not Evaluated for Effects
Due to Design Basis Tornado
Closed
07200058/2008-003-01
Rail Car Shelter Not Evaluated for Effects
Due to Design Basis Tornado
Discussed
None
Attachment
LIST OF DOCUMENTS REVIEWED
CALCULATIONS
Number
Description or Title
Date or Revision
Job No. 5828
Civil-Structural Design Criteria for The Monticello
Nuclear Generating Plant - Unit 1
Revision 1
Calculation
CA-05-076
Documentation of Subsurface Profiles at the
ISFSI Site
Revision 0
Calculation
CA-05-099
Evaluation of Reactor Building Elevation 1027-8
Cask Laydown Area for 100 Ton Cask
Revision 1
Calculation
CA-05-100
Design Adequacy of the Reactor Building Rail
Car Bay @ Elevation 935-0 for the Independent
Spent Fuel Storage Installation (ISFSI) Transfer
Operations
Revision 1
Calculation
CA-05-101
Evaluation of Reactor Steel Superstructure for
105 Ton Reactor Building Crane
Revision 3
Calculation
CA-05-102
Evaluation of Spent Fuel Pool for 100 Ton Cask
Laydown Load
Revision 0
Calculation
CA-05-103
Reactor Building Superstructure Seismic
Response Analysis with 105 Ton Crane
Revision 0
Calculation
CA-05-103
Reactor Building Superstructure Seismic
Response Analysis with 105 Ton Crane
Revision 0A
Calculation
CA-05-104
Design Adequacy of the Rail Car Shelter @
Elevation 935-0 for the ISFSI Transfer
Operations
Revision 0
Calculation
CA-05-106
Monticello Upgrade Trolley Calculations
February 24, 2006
Calculation No.
CA-06-112
Evaluation of Buried Equipment for 100-Ton
Cask Transfer Trailer Load. (for utilities inside the
Plant Protected Area)
Revision 1
Calculation No.
CA-07-015
Heavy Haul Road Design
Revision 0
Calculation No.
CA-07-016
ISFSI Pad and Approach Slab
Revision 0
Calculation
CA-08-135
Transfer Cask Hazard from Rail Car Shelter
Collapse
Revision 0
Calculation
CA-08-135
Transfer Cask Hazard from Rail Car Shelter
Collapse
Revision 1
Calculation
CA-82-769
Monticello Plant Unit 1 - Fuel Pool
Revision 2
Design Information
Transmittal
ISFSI-003
Reactor Building Structural Upgrades for ISFSI
(04Q162
November 18, 2004
Design Information
Transmittal
Reactor Building Structural Upgrades for ISFSI
(04Q162)
January 4, 2005
Attachment
CALCULATIONS
Number
Description or Title
Date or Revision
ISFSI-012
Design Information
Transmittal
ISFSI-014
Reactor Building Structural Upgrades for ISFSI
(04Q162)
January 13, 2005
Design Information
Transmittal
ISFSI-070
Independent Spent Fuel Storage Installation
August 26, 2008
Design Information
Transmittal
ISFSI-071
Independent Spent Fuel Storage Installation
September 2, 2008
MPS No. 0407
Specification for Installation and testing of
Concrete Expansion Bolts (P-503)
Revision 10
NUH-003,
NUH003.0103
Update Final Safety Analysis Report for the
Standardized NUHOMS Horizontal Modular
Storage Systems for Irradiated Nuclear Fuel
Revision 10
DRAWINGS
Number
Description or Title
Date or Revision
Drawing NF-36575
Reactor Building, Floor Framing, Plan at Elevation
1027-8, Sheet 1
Revision 6
Drawing NF-36578
Reactor Building, Truss Plan & Lower Chord Bracing
Details
Revision 76
Drawing NF-36579 Reactor Building, Craneway Plan & Details
Revision 77
Drawing NF-36580
Reactor Building, Framing Elevations & Details,
Base Plate & Anchor Bolt Details
Revision 76
Drawing
NGS-3483-S-001
Reactor Building, Partial Floor Framing, Plan
Elevation 1027-8
Revision 1
Drawing
NGS-3483-S-002-1
Reactor Building, Partial Floor Framing, Plan
Elevation 935
Revision 0
Drawing
NGS-3483-S-002-2
Reactor Building, Floor Framing Details, Plan
Elevation 935
Revision 0
Drawing
NGS-3483-S-003-1
Reactor Building, Craneway Plan & Details, Sheet 1
Revision 2
Drawing
Reactor Building, Craneway Plan & Details, Sheet 2
Revision 2
Attachment
DRAWINGS
Number
Description or Title
Date or Revision
NGS-3483-S-003-2
Drawing
NGS-3483-S-004
Reactor Building, Framing Elevation & Details
Revision 0
Drawing
NGS-3483-S-005
Reactor Building, Truss Lower Chord, Plan & Details
Revision 1
Drawing
NH-211482-1-1
Reactor Building, Craneway Plan & Details, Sheet 1
Revision 0
Drawing
NH-211482-1-2
Reactor Building, Craneway Plan & Details, Sheet 2
Revision 0
Drawing
NX-7865-11
Secondary Containment, Floor Loading
Revision 2
Drawing
NX-9324-22
Reactor Building, Truss Lower Chord Bracing, Plan
& Details
Revision A
Drawing
NX-9324-24
Reactor Building, Framing Elevations & Details,
Base Plate & Anchor Bolt Details
Revision A
Drawing
NX-9324-33
Reactor Building, Column Details
Revision 2
Drawing
NX-9324-35
Reactor Building, Column Details
Revision 2
CORRECTIVE ACTION PROGRAM DOCUMENTS
Number
Description or Title
Date or Revision
H-2 Missing Reactor Building Crane Runway
Rail Clips
May 11, 2006
H-2 Trolley Rails do not Lay Flat on the Crane
Girders
May 31, 2006
Potential Reactor Building Crane Bridge Bus-
Bar Issue
June 14, 2006
H-2 Crane Main and Aux Hoist do not Operate
During 1131 Procedure
June 17, 2006
RX Bldg Crane (H-2) Trolley North Stop Limit
Switch Failed
June 18, 2006
H-2 Main Hoist Up Limit Switch (Geared
Switch) Failed to Act
June 18, 2006
Drum Capture Plates
August 29, 2006
RB Crane Equalizer Sheave Bearing Seat
Deformed
October 10, 2006
Attachment
DRAWINGS
Number
Description or Title
Date or Revision
Crane h-2 Small Mark on the Sister Hook from
Load Test
November 3, 2006
Main Hoist Line Shaft Coupling out of
Tolerance
December 2, 2006
Discrepancies from H-2 Crane PM
December 12, 2006
RB Crane Aux. Hoist Motor is not Functioning
During Tests
December 12, 2006
Main Hoist Over Speed Switch Failed Function
December 16, 2006
Condition of H-2 Crane During PM's Requires
Resolution
December 16, 2006
H2 Main Hoist Tripped During 125 percent Test
December 21, 2006
H-2 Rx. Bldg. Crane Overload Switch Tripped
During 125 percent Test
January 8, 2007
Inspection of RB Crane Bridge End Truck
February 19, 2008
Incorporation Risk Assessment of Heavy Load
in Site Procedure
February 19, 2008
Dry Storage Canister Outer Packaging
Damaged During Shipping
April 17, 2008
Flowable Grout Placed on ISFSI Pad Has
Flaked Off
May 7, 2008
TN Supplied Weld Machine Does Not Meet
Expectations
May 20, 2008
Crane H-2 Preoperational Testing Delayed Due
to Equipment & Wiring Issues
May 30, 2008
Water Is Accumulating in Outer DSC
Packaging
June 17, 2008
Surface of ISFSI Asphalt Apron is Being
Damaged
June 19, 2008
Moisture in ISFSI Electrical Equipment on Pad
June 19, 2008
RX Bld Crane 5 Year PM Revealed a Few
Issues
June 23, 2008
Intermittent Failures of the Reactor Building
Crane Remote Control
June 23, 2008
ISFSI Procedures Incorrectly Identify
Classification of Safety Related
June 25, 2008
Evaluation of Rail Car Shelter Was Incomplete
July 1, 2008
Failed to Demonstrate Anchor Bolt Adequacy
for SSE
July 1, 2008
ISFSI Human Factor Errors Identified
July 2, 2008
Electrical Discrepancies Discovered During
ISFSI Walkdown
July 3, 2008
Future Needs for Calculations05-101 and
July 9, 2008
Attachment
DRAWINGS
Number
Description or Title
Date or Revision 05-103 Not Tracked by AR
Inadequate Conclusion Stated in Calculation 05-104
July 9, 2008
Arc Strikes Noted on Interior of DSC #002
July 9, 2008
DSC Cover Plate Weld Preps Possible
Undersized
July 9, 2008
Lid Fit Up Issues Discovered on DSC-001
July 15, 2008
USAR 12.2 Description Inadequate re: SFP
Structure Design Criteria
July 15, 2008
Calc 05-01 Enhancements Needed
July 15, 2008
Spurious Alarms of the ISFSI UPS Battery
Discharge
July 17, 2008
Wrong Method Submitted in LAR
July 18, 2008
Strong Diesel Fumes During ISFSI Dry Run
July 21, 2008
Procedure Changed in Field Without Required
Review / Approval
July 22, 2008
Revised Weld Specification Not Reviewed by
Site Weld Representative
July 23, 2008
Small Piece of Concrete from HSM 1A Broke
Loose
July 23, 2008
DSC Shell Deformation from Dry Run Insert /
Retrieve
July 23, 2008
NRC Inspectors Concerns of ISFSI 72.212
July 25, 2008
NRC Inspection of ISFSI 10 CFR 72.212
Report
July 25, 2008
HSM Rail Alignment
July 31, 2008
Revise MNGP 72.212 Report to Incorporate
Additional Information
August 1, 2008
Revise MNGP Fire Hazards Report to
Incorporate Site Identified Corrections
August 1, 2008
Procedure Not In Compliance with
4 AWI-02.03.13
August 5, 2008
In Pool Interference Interrupts ISFSI Dry Run
August 7, 2008
ISFSI Battery Discharge Trouble Alarm
August 13, 2008
Spent Fuel Cask Lid Weld Procedure Revisions
August 15, 2008
Spent Fuel Cask Lid Weld Procedure Revisions
August 15, 2008
Enhancement to EAL Protected Area Clarity
August 22, 2008
Contamination Identified on Cask Transport
Trailer
August 26, 2008
Attachment
DRAWINGS
Number
Description or Title
Date or Revision
H-2 Crane Trolley Over Speed Trip During
ISFSI Dry Run During Downending
August 27, 2008
Error Identified by NRC in Vendor Calculation
September 5, 2008
ISFSI Cask Loading Started with Operations
Approval
September 9, 2008
DSC #4 Inner Lid Weld Problem Requires
Repair
September 10, 2008
September 10, 2008
TN UFSAR Appendix C.5 is vague re Tornado
Missile
September 11, 2008
Proposed NRC Violation - ISFSI Calculation
Error
December 22, 2008
50.59/72.48 SCREENINGS
Number
Description or Title
Date or Revision
SCR-05-0487;
Screening
Modification 04Q162 Related Documents
Revision 0
SCR-07-0123;
Screening
Calculation CA-05-101 Revision 3, Evaluation of
Reactor Steel Superstructure for 105 Ton
Reactor Building Crane
Revision 0
SCR-08-0291;
Screening
Calculation 08-135, Transfer Cask Hazard from
Rail Car Shelter Collapse
Revision 0
August 19, 2008
SCR-08-0291;
Screening
Calculation 08-135, Transfer Cask Hazard from
Rail Car Shelter Collapse
Revision 1
September 3, 2008
SCR-08-0291;
Screening
Calculation 08-135, Transfer Cask Hazard from
Rail Car Shelter Collapse;
Revision 2
September 8, 2008
SCR-08-0315;
Screening
Calculation 08-135, Transfer Cask Hazard from
Rail Car Shelter Collapse
Revision 0
September 3, 2008
SCR-08-0315;
Screening
Calculation 08-135, Transfer Cask Hazard from
Rail Car Shelter Collapse
Revision 1
September 10, 2008
Attachment
MODIFICATIONS
Number
Description or Title
Date or Revision
Modification
04Q162
Design Description: Reactor Building Structural
Upgrades for ISFSI
0
RPT-EC-785
Capacity Upgrade Modification and Safety
Evaluation for the Reactor Building Crane
System
1
PROCEDURES
Number
Description or Title
Date or Revision
Crane Daily Checks Placard
July 2, 2008
0000-H
Operations Daily Log - Part H
Revision 91
4 AWI-02.07.02
DFS UFSAR and Monticello 72.212 Report
Control
Revision 0
3832
ISFSI Fire Protection Change Review
Revision 0
4250-01-PM
Reactor Building Crane, Bridge Drive System
Revision 24
4250-02-PM
Reactor Building Crane, Trolley Drive System
Revision 22
4250-03-PM
Reactor Building Crane, Main Hoist System
Revision 21
4250-04-PM
Reactor Building Crane, Auxiliary Hoist System
Revision 22
4250-04-PM
Reactor Building Crane, Auxiliary Hoist System
Revision 20
4361-PM
Reactor Building Crane Inspection Checklist
Revision 5
8151
Heavy Load Movement Procedure
Revision 13
9009
Procedure for Moving Fuel Within the Fuel
Storage Pool
9501
Transfer Trailer Assembly, Receipt Inspection
and Pre-Operational Testing
Revision 0
9502
Transfer Cask Inspection and Pre-Job Brief
Revision 0
9503
Dry Shielded Canister Receipt Inspection and
Pre-Operational Testing
Revision 0
9504
Ancillary Equipment Receipt Inspection
Revision 0
9505
Preparations for Loading Dry Shielded Canister
Revision 1
9506
Dry Shielded Canister Sealing
Revision 1
9507
DSC Transport from Refueling Floor to ISFSI
Revision 1
9508
DSC Transfer from Transfer Cask to HSM
Revision 1
9513
HSM Equilibrium Temperature Monitoring
Revision 0
9514
Cask Registration Info
Revision 0
B.08.15-05
Reactor Building Crane Emergency Positioning
Revision 18
Attachment
PROCEDURES
Number
Description or Title
Date or Revision
and Manual Lowering of Load
D.2-05
Operations Manual D.2-05 Reactor and Core
Components Handling Equipment - Tool and
Equipment Operation
Revision 19
FP-PE-pAWS-I-II-
Fleet Procedure: Groove & Fillets, Group I & II,
FCAW, without PWHT;
Revision 0
FP-PE-WLD-02
Fleet Procedure: General Welding Specification
Revision 2
FP-E-SE-03
10 CFR 50.59 And 72.48 Processes
Revision 1
FP-G-RM-01
Records Management
Revision 5
GWS-3
Spent Fuel Cask Welding - NUHOMS Canisters
Revision 5
NMC-1 QATR
Quality Assurance Topical Report
Revision 4
NUC-06.02
Selecting Fuel Bundles for ISFSI Storage
Revision 0
R.02.01
Dose Rate Surveys
Revision 19
R.02.02
Surface Contamination Surveys
Revision 24
REFERENCES AND MISCELLANEOUS DOCUMENTS
Number
Description or Title
Date or Revision
Table 1 Monticello Compliance Summary to
the Heavy Load Handling Criteria of NRC
Documents for Spent Fuel Transfer Cask
Handling with the Reactor Crane
February 2, 2008
August 18, 2008
Monticello Nuclear Generating Plant ISFSI 10
CFR 72.212 Evaluation Report
Revision 1
Response to Crane Load Testing Question
Page 17
July 16, 2008
Response to Crane Load Testing Question
Page 27
July 9, 2008
Response to NRC 72.212 Inspection #7 - #12
Questions
-
TriVis Dry Fuel Storage Training and
Qualification Matrix
August 19, 2008
4 AWI-01.03.01
Quality Assurance Program Boundary
Revision 16
4 AWI-05.05.02
Fuel Integrity and Failed Fuel Action Plan
Revision 9
4 AWI-8.04.01
Radiation Protection Plan
Revision 24
A.2-101
Classification of Emergencies
Revision 39
Attachment
REFERENCES AND MISCELLANEOUS DOCUMENTS
Number
Description or Title
Date or Revision
EC-1098/ECN-
9423
Reactor Crane Upgrade to 105T for ISFSI
Electrical Improvements
June 12, 2008
Revision 0
RPT-EC-785
Capacity Upgrade Modifications and Safety
Evaluation for Reactor Building Crane System
June 9, 2008
Common Book Final Document Package for
DSCs (Volumes 1-3)
Revision 30
Final Document Package for DSC-002
Final Document Package for DSC-003
MNGP 72.212 Evaluations Report
Revision 0
Monticello Nuclear Generating Plant ISFSI 10
CFR 72.212(b)(2)(i)(C) Radiological
Evaluation
Revision 0
Monticello Nuclear Generating Plant ISFSI
Fire Hazards Analysis
Revision 0
NMC letter L-HU-05-017, Notification of Intent
to Apply the NMC Quality Assurance Topical
Report (QATR), NMC-1, to ISFSI, Spent Fuel
Cask and Radioactive Waste Shipment
Activities at NMC Operated Plants
September 13, 2005
NMC Letter L-MT-08-010, 90-Day Notification
PORC Meeting 2594 Minutes (documents
72.212 report review)
QF-0528 72.212 Review Comments
QF-0528 ISFSI FHA Review Comments
Reactor Building Crane Upgrade for ISFSI
Revision 2
Technical Evaluation Report-Control of Heavy
Loads
January 30, 1984
ISFSI Loading Reports for 2008 Campaign
ISFSI Radiation Protection Work Plan
GNF Engineering Documents - Monticello
Plant Fuel Reliability History Review
February 2008
Casks 1- 10 Fuel Bundle Movement History
(Sipping and Discharge Information)
USAR Section 02.03
Revision 24
Attachment
REFERENCES AND MISCELLANEOUS DOCUMENTS
Number
Description or Title
Date or Revision
Westinghouse-Summary of Sipping Results
for Monticello 2008 Cask Sipping Campaign-
Assembly Cycles 10, 11, 12
June 17, 2008
VENDOR DOCUMENTS
Number
Description or Title
Date or Revision
Magnetek Certificate of Compliance
May 1, 2006
Monticello Reactor Building Crane 5 Year PM &
Refuel Bridge Support
June 27, 2008
Overhead / Gantry Crane Worksheet - Crane
Certification Co.
December 12, 2006
Use of OS197-1 Hydraulic Ram at MNGP
June 30, 2008
Washington Chain and Supply Certificate of
Compliance
April 19, 2006
70587723
Design Criteria Review Monticello Reactor
Building Crane Uprate From 85 Ton to 105
Tone Capacity - Par Nuclear
May 12, 2008
NUH-06-106M
Maintenance & Modification Procedure for the
NUHOMS OS197-1 Transfer Cask Lifting Yoke
and Other TN Owned Lifting Yokes
June 13, 2008
WCS-1051765
Certification of Test and Examination of Chains,
Rings, Hooks, Shackles, Swivels, and Blocks
October 13, 2006
Bechtel Report
12085
Monticello Nuclear Power Station Reactor
Building Seismic Evaluation of Spent Fuel Pool
Structure
January 1977
Revision 1
WORK DOCUMENTS
Number
Description or Title
Date or Revision
WO00142573 07 Modify Reactor Building Structural Steel for
Upgrade to Crane H-2, Gusset Weld
Confirmation at elevation 1064-2
March 6, 2006
WO00142573 08 Weld Control Record 142573-08-01
Weld Map Sketch WM-142573-01
March 1, 2006
WO00142580 02 Reactor Building Crane Load Test
July 1, 2008
WO00142583 16 Site Acceptance Test Main & Auxiliary Hoist
Control Panels &105 Ton Up-Rate
December 13, 2006
Attachment
WORK DOCUMENTS
Number
Description or Title
Date or Revision
WO00280440 01 PM 4250 (RX Building Crane H-2)
January 12, 2007
WO00331532 01 PM 4250 (RX Building Crane H-2)
January 4, 2008
Attachment
LIST OF ACRONYMS
As Low As Reasonably Achievable
Action Request
Certificate of Compliance
CFR
Code of Federal Regulations
Design Basis Tornado
Dry Shielded Canister
Fire Hazard Analysis
Final Safety Analysis Report
HSM
Horizontal Storage Modules
IMC
Inspection Manual Chapter
Independent Spent Fuel Storage Installation
Monticello Nuclear Generating Plant
Monitored Retrieval Storage Installation
Non-Cited Violation
NRC
Nuclear Regulatory Commission
RS
Rail Car Shelter
Spent Fuel Pool
TN
Transnuclear