ML083660296

IR 072-00058-08-003; Northern States Power Company; on 06/30/2008 - 07/03/2008 - 08/18-22/2008 09/08-11/2008; Monticello Nuclear Generating Plant (Dnms), NRC Inspection Report

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:

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Christopher Thomas

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Carole Ariano

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Patricia Buckley

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Matthew Learn

Lionel Rodriguez

Enclosure

U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Docket No.

072-00058

License No.

DPR-22

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

AR 01148733.

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

AR 01148282.

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

ISFSI.

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

ISFSI.

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)

1. T. Blake, Regulatory Affairs Manager

1. 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

1. T. J. OConnor, Site Vice President

1. S. Quiggle, ISFSI Project Manager

G. Ridder, ISFSI Project Engineer Nathan French - Operations Support Manager

• 1. L. Samson, Manager, Spent Nuclear Fuel Storage

1. B. Sawatzke, Plant Manager

• 1. K. Shriver, ISFSI Project Support

• Indicates individuals present at the August 22, 2008 debrief

1. 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

IP 60854.1

Preoperational Testing Of An Independent Spent Fuel Storage Facility

Installation (ISFSI) At Operating Plants

IP 60856

Review of 10 CFR 72.212(b) Evaluations (60856)

Attachment

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

07200058/2008-003-01

NCV

Rail Car Shelter Not Evaluated for Effects

Due to Design Basis Tornado

Closed

07200058/2008-003-01

NCV

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

AR 01029594

H-2 Missing Reactor Building Crane Runway

Rail Clips

May 11, 2006

AR 01033069

H-2 Trolley Rails do not Lay Flat on the Crane

Girders

May 31, 2006

AR 01035555

Potential Reactor Building Crane Bridge Bus-

Bar Issue

June 14, 2006

AR 01035947

H-2 Crane Main and Aux Hoist do not Operate

During 1131 Procedure

June 17, 2006

AR 01035961

RX Bldg Crane (H-2) Trolley North Stop Limit

Switch Failed

June 18, 2006

AR 01035962

H-2 Main Hoist Up Limit Switch (Geared

Switch) Failed to Act

June 18, 2006

AR 01047058

Drum Capture Plates

August 29, 2006

AR 01054379

RB Crane Equalizer Sheave Bearing Seat

Deformed

October 10, 2006

Attachment

DRAWINGS

Number

Description or Title

Date or Revision

AR 01059718

Crane h-2 Small Mark on the Sister Hook from

Load Test

November 3, 2006

AR 01065117

Main Hoist Line Shaft Coupling out of

Tolerance

December 2, 2006

AR 01065868

Discrepancies from H-2 Crane PM

December 12, 2006

AR 01067235

RB Crane Aux. Hoist Motor is not Functioning

During Tests

December 12, 2006

AR 01068103

Main Hoist Over Speed Switch Failed Function

December 16, 2006

AR 01068114

Condition of H-2 Crane During PM's Requires

Resolution

December 16, 2006

AR 01068939

H2 Main Hoist Tripped During 125 percent Test

December 21, 2006

AR 01070508

H-2 Rx. Bldg. Crane Overload Switch Tripped

During 125 percent Test

January 8, 2007

AR 01127967

Inspection of RB Crane Bridge End Truck

Welds

February 19, 2008

AR 01127972

Incorporation Risk Assessment of Heavy Load

in Site Procedure

February 19, 2008

AR 01134872

Dry Storage Canister Outer Packaging

Damaged During Shipping

April 17, 2008

AR 01137048

Flowable Grout Placed on ISFSI Pad Has

Flaked Off

May 7, 2008

AR 01138313

TN Supplied Weld Machine Does Not Meet

Expectations

May 20, 2008

AR 01139429

Crane H-2 Preoperational Testing Delayed Due

to Equipment & Wiring Issues

May 30, 2008

AR 01141164

Water Is Accumulating in Outer DSC

Packaging

June 17, 2008

AR 01141400

Surface of ISFSI Asphalt Apron is Being

Damaged

June 19, 2008

AR 01141418

Moisture in ISFSI Electrical Equipment on Pad

June 19, 2008

AR 01141785

RX Bld Crane 5 Year PM Revealed a Few

Issues

June 23, 2008

AR 01141786

Intermittent Failures of the Reactor Building

Crane Remote Control

June 23, 2008

AR 01142079

ISFSI Procedures Incorrectly Identify

Classification of Safety Related

June 25, 2008

AR 01142790

Evaluation of Rail Car Shelter Was Incomplete

July 1, 2008

AR 01142801

Failed to Demonstrate Anchor Bolt Adequacy

for SSE

July 1, 2008

AR 01143094

ISFSI Human Factor Errors Identified

July 2, 2008

AR 01143127

Electrical Discrepancies Discovered During

ISFSI Walkdown

July 3, 2008

AR 01143398

Future Needs for Calculations05-101 and

July 9, 2008

Attachment

DRAWINGS

Number

Description or Title

Date or Revision 05-103 Not Tracked by AR

AR 01143567

Inadequate Conclusion Stated in Calculation 05-104

July 9, 2008

AR 01143601

Arc Strikes Noted on Interior of DSC #002

July 9, 2008

AR 01143643

DSC Cover Plate Weld Preps Possible

Undersized

July 9, 2008

AR 01144172

Lid Fit Up Issues Discovered on DSC-001

July 15, 2008

AR 01144276

USAR 12.2 Description Inadequate re: SFP

Structure Design Criteria

July 15, 2008

AR 01144280

Calc 05-01 Enhancements Needed

July 15, 2008

AR 01144452

Spurious Alarms of the ISFSI UPS Battery

Discharge

July 17, 2008

AR 01144664

Wrong Method Submitted in LAR

July 18, 2008

AR 01144861

Strong Diesel Fumes During ISFSI Dry Run

July 21, 2008

AR 01144920

Procedure Changed in Field Without Required

Review / Approval

July 22, 2008

AR 01145012

Revised Weld Specification Not Reviewed by

Site Weld Representative

July 23, 2008

AR 01145052

Small Piece of Concrete from HSM 1A Broke

Loose

July 23, 2008

AR 01145084

DSC Shell Deformation from Dry Run Insert /

Retrieve

July 23, 2008

AR 01145347

NRC Inspectors Concerns of ISFSI 72.212

July 25, 2008

AR 01145347

NRC Inspection of ISFSI 10 CFR 72.212

Report

July 25, 2008

AR 01145916

HSM Rail Alignment

July 31, 2008

AR 01146174

Revise MNGP 72.212 Report to Incorporate

Additional Information

August 1, 2008

AR 01146176

Revise MNGP Fire Hazards Report to

Incorporate Site Identified Corrections

August 1, 2008

AR 01146570

Procedure Not In Compliance with

4 AWI-02.03.13

August 5, 2008

AR 01146826

In Pool Interference Interrupts ISFSI Dry Run

August 7, 2008

AR 01147364

ISFSI Battery Discharge Trouble Alarm

August 13, 2008

AR 01147693

Spent Fuel Cask Lid Weld Procedure Revisions

August 15, 2008

AR 01147693

Spent Fuel Cask Lid Weld Procedure Revisions

August 15, 2008

AR 01148282

Enhancement to EAL Protected Area Clarity

August 22, 2008

AR 01148601

Contamination Identified on Cask Transport

Trailer

August 26, 2008

Attachment

DRAWINGS

Number

Description or Title

Date or Revision

AR 01148733

H-2 Crane Trolley Over Speed Trip During

ISFSI Dry Run During Downending

August 27, 2008

AR 01149709

Error Identified by NRC in Vendor Calculation

September 5, 2008

AR 01150005

ISFSI Cask Loading Started with Operations

Approval

September 9, 2008

AR 01150088

DSC #4 Inner Lid Weld Problem Requires

Repair

September 10, 2008

AR 01150191

ISFSI Hydrogen Nuisance Alarm

September 10, 2008

AR 01150233

TN UFSAR Appendix C.5 is vague re Tornado

Missile

September 11, 2008

AR 01157276

Proposed NRC Violation - ISFSI Calculation

Error

December 22, 2008

50.59/72.48 SCREENINGS

Number

Description or Title

Date or Revision

SCR-05-0487;

10 CFR 50.59

Screening

Modification 04Q162 Related Documents

Revision 0

SCR-07-0123;

10 CFR 50.59

Screening

Calculation CA-05-101 Revision 3, Evaluation of

Reactor Steel Superstructure for 105 Ton

Reactor Building Crane

Revision 0

SCR-08-0291;

10 CFR 72.48

Screening

Calculation 08-135, Transfer Cask Hazard from

Rail Car Shelter Collapse

Revision 0

August 19, 2008

SCR-08-0291;

10 CFR 72.48

Screening

Calculation 08-135, Transfer Cask Hazard from

Rail Car Shelter Collapse

Revision 1

September 3, 2008

SCR-08-0291;

10 CFR 72.48

Screening

Calculation 08-135, Transfer Cask Hazard from

Rail Car Shelter Collapse;

Revision 2

September 8, 2008

SCR-08-0315;

10 CFR 50.59

Screening

Calculation 08-135, Transfer Cask Hazard from

Rail Car Shelter Collapse

Revision 0

September 3, 2008

SCR-08-0315;

10 CFR 50.59

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-

FC-003

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

---

ISFSI Crew LMS Reports

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

EC-783

MNGP ISFSI 50.59 Screening

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)

EP-6

Emergency Plan

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

EC-785

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

ALARA

As Low As Reasonably Achievable

AR

Action Request

CoC

Certificate of Compliance

CFR

Code of Federal Regulations

DBT

Design Basis Tornado

DSC

Dry Shielded Canister

FHA

Fire Hazard Analysis

FSAR

Final Safety Analysis Report

HSM

Horizontal Storage Modules

IMC

Inspection Manual Chapter

ISFSI

Independent Spent Fuel Storage Installation

MNGP

Monticello Nuclear Generating Plant

MRS

Monitored Retrieval Storage Installation

NCV

Non-Cited Violation

NRC

Nuclear Regulatory Commission

RS

Rail Car Shelter

SFP

Spent Fuel Pool

TN

Transnuclear