IR 05000344/1999005

Insp Repts 50-344/99-05 & 72-0017/99-04 on 990419-22.No Violations Noted.Major Areas Inspected:Decommissioning Performance & Status Review

ML20206J873
Person / Time
Site:	Trojan  File:Portland General Electric icon.png
Issue date:	05/07/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20206J872	List: IR 05000344/1999005 ML20206J868
References
50-344-99-05, 50-344-99-5, 72-0017-99-04, 72-17-99-4, NUDOCS 9905120369
Download: ML20206J873 (8)

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b ENCLOSURE U.S. NUCLEAR REGULATORY. COMMISSION

REGION IV

Docket No.: .50-344;72 17 License No.: NPF-1; SNM-2509 -

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- Report No.: 50-344/99-05; 72-17/99-04 Licensee: Portland General Electric Company (PGE)

Facility: Trojan Nuclear Plant l i

Location: 121 S. W. Salmon Street, TB-17 Portland, Oregon

' Dates: April 19 - 22,1999 Inspector: J. V. Everett, Senior Health Physics inspector Fuel Cycle & Decommissioning Branch Division of Nuclear Materials Safety ,

Approved By: D. Blair Spitzberg, Ph.D., Chief Fuel Cycle & Decommissioning Branch Division of Nuclear Materials Safety Attachment: Supplemental Information

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9905120369 990507

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. EXECUTIVE SUMMARY

- Trojan Nuclear Plant ..

NRC Inspection Report 50-344/99-05; 72-17/99-04 .I

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Significant progress had been made toward the decommissioning and dismantlement of the Trojan nuclear facility. The reactor vessel removal project took another major step forward with

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l the lifting of the reactor vessel. The reactor vessel will be lowered from containment in mid May -

. to be transported to Hanford, Washington, by the end of July.-

Preparations continue for the movement of spent fuel from the spent fuel pool to dry cask -

storage at the onsite independent spent fuel storage installation (ISFSI). The first basket is

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expected to arrive onsite in early May. During modifications to the lifting yoke, cracks were -

~ found under several welds. The lifting yoke is used to lift and move the transfer cask loaded

- with spent fuel. lThe licensee initiated an evaluation of the cracking proble Decommissionina Performance and Status Review

1 A substantial amount of dismantlement work had been completed over the past several months.~ Tours of work areas found radiological controls, fire loading, housekeeping, and general area work conditions to be acceptable (Section 1).

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Work relate' d to the lifting of the reactor vessel was observed. Work was being conducted safely related to positioning the vessel for removal from containment in

mid-May, The reactor vessel weighed approximately 820 tons. Most of the work areas associated with the heavy lift had been decontaminated and covered._ This will simplify the release of the contractor's equipment at the end of the project (Section .1).

Pre-Ooerational Testino of an ISFSI a'

The licensee had performed an extensive inspection of the 780 fuel assemblies to document the status of each assembly. Twenty special failed fuel cans will be needed for dry cask storage of damaged fuel assemblies and fuel debris from the fuel debris processing project (Section 2).

- Procedures for handling unexpected or abnormal events that could occur during the fuel loading process were reviewed and found adequate. Procedures had been developed

.for a' number of situations involving fuel movement in the spent fuel pool, movement of

. the transfer cask, and loss of offsite power (Section 2).

During a modification to the lifting yoke to be used to lift the transfer casket and basket-loaded with spent fuel, cracks were found under several welds. The welds were not part of the structural design of the yoke. The licensee initiated an investigation of the

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condition (Section 2).'

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Self Assessments. Auditino. and Corrective Actions '

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Review and acceptance of Revision 22 of the Quality Assurance (QA) Plan was j documented in the Safety Evaluation Report issued with the ISFSI license on March 31, 1999. Acceptance of the OA plan for the 10 CFR Part 72 license meets the requirement for approval of Revision 22 of the OA plan for use under the 10 CFR Part 50 license l (Section 3).

• Activities related to the Independent Review and Audit Committee were reviewed and -

found to comply with the requirements of Technical Specification 5.5.2. (Section 3).

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Report Details

- Summarv of Plant Status Dismantlement work continued to show significent progress. The licensee had removed over 190 per cent of the radioactive equipment and piping in the containment and turbine building and over.70 per cent in the fuel / auxiliary building. Numerous areas posted as contaminated zones -

during the January 1999 inspection had been cleaned and no longer required protective '

clothing for entry. Most large tanks, pumps, motors, and equipment from contaminated areas :

have been removed from the plan The reactor removal project continued to progress safely. Plans to transport the reactor vessel i to Hanford, Washington, by the end of July,1999, are still on schedule. During this inspection, the reactor vessel was lifted from it's normal location and moved near the roll-up door of the -

containment.' The vessel will be removed from containment in mid-May. When the reactor vessel is removed from the site and the spent fuel is moved to the ISFSI,99 percent of the

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. radioactive material will be removed from PGE's Part 50 licens Decommissioning Performance and Status Review (71801) Insoection Scooe -

Significant decommissioning work was underway at Trojan related to dismantlement of .

the facility and removal of the reactor vessel. The work activities related to these two major projects were observed to verify activities were being performed safel .2 - . Observation and Findinos-l A tour of the plant was completed to ovaluate general plant conditions.- Fire loading, safety hezards, progress of dismantlement work, contamination control, radiological posting, housekeeping, condition of the spent fuel pool concrete wall, and general area work conditions were observed. A substantial amount of work had been completed concerning removal of piping, equipment, cabling, and attachments to walls. The number of areas posted as radiation area or contaminated area were notably reduced since the January 1999 inspection. Contaminated material stored in work areas was

, also considerably reduced in volume. No fire hazarde safety concerns were identified during the tours. Ventilation systems were operatior . . .ind postings were observed in the appropriate areas. Ste . off pads were properly placed and protective clothing was available in adequate quar :ity for workers. Dismantlement work that had been

performed around the sput fuel pool wall had not created any degraded structural i conditions that affected ' ye wall or created a new path for water to leak from the poo l The mctor vessel was lifted vertically on April 21,1999 and positioned to be laid on its sidt me following week. The reactor vessel weighed approximately 820 tons. The work activities associated with the lifting of the vessel were observed by the NRC. Work activities were conducted safely. No incidents or unexpected events occurred during the !

lifting process.' Most work areas associated with the equ:pment needed for the reactor I

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5-lifting had been decontaminated and cleaned. The areas were covered with plastic or other material to reduce the potential for contaminating the lifting equipment and thereby

- reducing the effort to release the material back to the contractor. This also allowed for much of the work to be performed without wearing protective clothing. Radiological levels associated with the lifting effort were low, typically around 1-2 mR/hr. These low radiation levels were due to the shielding placed on the reactor vesse L The vessel will be lowered out of containment onto a transport trailer in mid-March, then barged to Hanford, Washington, for burial at the end of July. A Safety Analysis Report (SAR) had been submitted to the NRC by PGE. The latest version of the Safety Analysis Report was September 23,1998. The Safety Analysis Report provided a detailed description of the plans for removal of the reactor vessel. The Safety Analysis

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Report was reviewed by the NRC and on October 29,1998, the NRC issued a letter to PGE approving the removal and shipment of the reactor vesse ' Conclusion A substantial amount of dismantlement work had been completed over the past several months. Tours of work areas found radiological controls, fire loading, housekeeping,

' and general area work conditions to be acceptabl Work related to the lifting of the reactor vessel was observed. Work was being

. conducted safely related to pcsitioning the vessel for removal from containment in mid May. Most of the work areas associated with the heavy lift had been decontaminated and covered. This will simplify the release of the contractor's equipment at the end of the projec .-

2 : Pre-Operational Testing of an ISFSI (60854)

2.1: Insoection Scooe The licensee planned to begin fuel loading of spent fuel into casks for storage at the ISFSI beginning in May. This inspection reviewed the effort completed by the licensee to categorize the fuel in accordance with criteria established by the Department of -

Energy for standard and failed fue .

Contingency procedures for unexpected events during the handling of the fuel assemblies were also reviewed; During this inspection, the licensee informed the NRC inspector that an examination of the lifting yoke for the transfer cask had resulted in the discovery of cracks in the bea .2 - Observations and Findinas An ' examination was completed of all fuel assemblies to classify the fuel in accordance with DOE criteria for spent nuclear fuel. Prior to loading the spent fuel into dry cask storages a final inspection will be conducted as the fuel is loaded into the basket. The

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6-initial evaluation of the fue1 was conducted in January and February,1997. The examinations were conducted using Procedure FHP-15," Irradiated Nuclear Fuel inspection / Classification," Revision 14. The procedure established the process for conducting the examinations and for classifying the fuel. Over 60 video tapes were generated to document the condition of the fuel. Fuel examination consisted of moving the fuel past four undarwater cameras. All 780 spent fuel assemblies were examine The criteria for failed fuel defined in the ISFSI Technical Specifications, Section described damaged fuel as fuel assemblies which can be handled by normal means but

~ have known 'or suspected cladding defects greater than pinhole leaks or hairline cracks or have missing fuel rods that are not replaced with dummy fuel rods. Damaged fuel will be placed in special failed fuel cans inside the basket. The licensee plans to use 20

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failed fuel cans for damaged fuel and fuel debris from the fuel debris processing project completed in late 199 The licensee also used the subcategories for failed fuel classification defined in DOE Contract DE-CR01-83NE 44406, " Contract for Disposal of Spent Nuclear Fuel and/or High Level Radioactive Waste." Failed fuel was subdivided into three types. Fuel with visual failure or damage was categorized as F 1. Fuel with small defects that could result in radioactive leakage was classified as F-2. Fuel that required encapsulation was categorized as F-3. In the early 1980 time frame, an incident occurred at the Trojan facility in which coolant flow in the reactor vessel jetted through a gap in the baffle plate and damaged a number of fuel assemblies. This problem resulted in several fuel elements being damaged to the point that fuel pellets were dislodged. For some of these damaged fuel assemblies, the good fuel elements were removed and placed into l

.a new bundle'. The failed elements were left in the original bundle and will be placed in a failed fuel can. Eight fuel assemblies met the criteria of the F-1 failed fuel. This 1 included several assemblies having breached and broken fuel elements with fuel pellets lodged in the matrix of the fuel assembly.- All of these fuel assemblies will be placed into failed fuel cans. . In addition, fuel assemblies F05 from the F-2 list and fuel assemblyi ;

A45 will be placed in failed fuel cans. Fuel assembly A45 was not a complete assembl Fuel elements had been removed from A45 and used to fill spaces in other fuel assemblies. < l

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Seven fuel assemblies were classified as F-2 failed fuel. These had pin hole leaks or -

other defects observed during the visual examination. One had failed a fuel sipping examination and two had failed previous ultrasonic tests. Six of the seven assemblies were structurally intact and will not require special handling or storage. One fuel rod storage rack, FRS1,' consisting of 23 failed or suspected fuel rods taken from damaged ,

fuel assemblies will be stored in a failed fuel can. The remaining failed fuel cans will

, contain fuel debris from the fuel debris processing projec Selected segments of one of the videos recorded by the licensee of the fuel examinations were reviewed during this inspection. The video had good clarity sufficient to verify the adequacy of the visual examinations. The fuel assemblies that were damaged or contained lodged debris were readily recognized. The licensee generated a OA record for each fuel assembly examined using Attachment 2, " Fuel Assembly Inspection Sheet." from Procedure FHP-15. Selected records were reviewed and found J

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-7 to be complete including the necessary signatures and dates of examination. The records provided the assembly number, classification assigned, associated video tape documenting the visual examination, and specific criteria that the fuel was examined agains Contingency procedures for unexpected events during the handling of the spent fuel assemblies were reviewed. Procedures FHP-13, " Fuel Handling Emergency Procedures," Revision 22; ONI 50-04, " Response to Operational Events," Revision 0; j and FHP-14, " Limitations and Precautions for Handling Fuel Assemblies," Revision 13, l were reviewed.- ' Procedure FHP-13 provided instructions for response to a fuel handling accident or a situation that could lead to fuel damage. The procedure considered fuel j

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handling accidents including a fuel assembly being dropped or damaged during movement or an object dropped onto a fuel assembly causing damage and abnormal radiation levels. Procedure FHP-14 provided information concerning the construction and limitations of the fuel assemblies that the operator needed to be aware of to safely handle the fuel assemblies. This included information and precautions on vertical and horizontal movement of the fuel assemblies, storage, axial and lateral load limitations, loading of assemblies into the PWR baskets, and spent fuel bridge crane handling operations. For situations involving the failure of the spent fuel bridge crane, the spent fuel assembly would be placed in as safe a configuration as possible, such as over an empty storage cell, and the shift manager notified. The shift manager and engineering would then evaluate the situation and develop a recovery pla Procedure ONI 50-04, " Response to Operational Events," provided information for response to a problem involving an ISFSI component. This included a drop, tipover, or mishandling of an ISFSI component inside the fuel building, failure of the fuel building crane, and loss of offsite powe During the inspection at Trojan, the licensee informed the NRC inspector that cracks

, were discovered in the beams of the lifting yoke to be used to lift the transfer cask from the fuel pool and lower the transfer cask to ground level. The transfer cask is used to move a basket loaded with spent fuel. During a modification of the lifting yoke, weld material was removed from the beam at the location where two small metal pieces, called keepers,'were welded in place to limit the movement of the hooks on the bea Upon removal of the welds and grinding the locations to remove all weld material, an

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examination of the base metal found cracks in all four locations where welding had been performed to place the keepers onto the beams. The design of the transfer yoke consisted of two beams which hold the two hooks that connect onto the transfer cask trunnions.L There are a total of 8 keepers,2 keepers welded on the outside of each hook -

and 2 keepers welded on the inside of each hook. The welded metal keepers do not provide any structural purpose, but simply limit the movement of the hooks on the beams.~ The licensee issued Corrective Action Report C 99022 identifying the problem with the cracks. The licensee initiated an evaluation of the problem and planned to remove all 8 keepers to determine the extent of cracking. Followup on the yoke

. cracking issue will be conducted upon completion of the licensee's assessment of the

' cracks. This will be tracked as an inspection followup item (IFl 72-13/9904 01).

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, -8-1 Conclusion -

-.The' licensee had performed an extensive inspection of the 780 fuel assemblies to document the status of each assembly. Twenty special failed fuel cans will be needed for dry cask storage of damaged fuel assemblies and fuel debris from the fuel debris processing projec Procedures for handling unexpected or abnormal events that could occur during the fuel

- loading process were reviewed and found adequate., Procedures had been developed

' for a number of situations involving fuel movement in the spent fuel pool, movement of.-

the transfer cask, and loss of offsite powe During a modification to the lifting yoke to be used to lift the transfer casket and basket loaded with spent fuel, cracks were found under several welds. The welds were not par of the structural design of the yoke. The licensee initiated an investigation of the

' conditio :3 Self-Assessments, Auditing, and Corrective Actions (40801,40500) Insoection Scooe I The licensee submitted their 10 CFR Part 50 QA Plan to the NRC to meet the  ;

requirements of 10 CFR 72, Subpart G for the ISFSt. The NRC included the QA Plan 1 in the 10 CFR Part 72 license application revie The activities of the independent review and audit committee were reviewed to evaluate

. the type of issues that were being reviewed by the committee and the level of review being performe .2 - Observation and Findinas On October 9,1995,' the licensee informed the NRC of their intent to apply the Trojan 10 CFR Part 50, Appendix B, OA program to the activities associated with the ISFS Revision 19 of the OA Plan was' submitted to the NRC for review under 10 CFR 72,

. Subpart G. During the review process, Revisions 20,21, and 22 were submitted and reviewed by the NRC. On March 31,1999, the NRC issued License SNM-2509 to PGE i

. for the ISFSI. Issued with the license was the Safety Evaluation Report which accepted ;

Revision 22 of the OA Plan. The acceptance of Revision 22 of the OA Plan is j considered applicable to both 10 CFR Part 72 and 10 CFR Part 50, Appendix '

Technical Specification 5.5.2 established the requirements for the Independent Review

.and Audit Committee. The committee was responsible for reviewing and advising the Plant General Manager on matters related to the safe storage of irradiated fuel. The

committee was composed of five members. A quorum required attendance of at least three members or their alternatives. Quarterly meetings were require _

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Meeting minutes were reviewed for the past 12 months. Meetings were conducted more

' frequently than the required quarterly time period. All meetings were attended by the required minimum quorum. Meeting minutes documented the various activities conducted by the committee and included areas such as review of License Change Application 247, " Trojan Plant License Termination Plan"; License Change Application 246, " Spent Fuel Cask Loading in the Fuel Building-Contingency Fuel Unloading to the, Fuel Pool"; Revision 3 to License Change Application 237, " Spent Fuel

Cask Loading," audit and surveillance schedules, and repair of a leak on the modular spent fuel pool cooling coil A considerable number of documents were reviewed by the committee including corrective action requests, NRC inspection reports, audits -

conducted by the OA department, license document change requests, NRC information notices, surveillances, and safety evaluation .3 Conclusion Review and acceptance of Revision 22 of the OA Plan was documented in the Safety Evaluation Report issued with the ISFSI license on March 31,1999. Acceptance of the OA Plan for the 10 CFR Part 72 license meets the requirement for approval of Revision 22 of the OA Plan for use under the 10 CFR Part 50 licens Activities related to the Independent Review and Audit Committee were reviewed and found to comply with the requirements of Technical Specification. 5. .

L4 Follow-up on Open items (92701) (Closed) IFl 50-344/9901-01: Movement of Concrete Cask Near the Edae of the ISFSI

- A walkdown of the ISFSI pad was completed. A groove had been cut around the pad of

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appropriate size to prevent the concrete storage casks from being moving beyond the edge of the pad before the air pads would deflat L 4.2 - (Closed) IFl 50-344/9901-02: Emeroency Suoolies Emergency equipment identified in

~ the emergency plan had been placed in the building on the ISFSI pad and in the access control facility for response to an emergency conditio .3 (Closed) IFl 72-017/9902-01: Deficiencies with Protected Area Barriers and Detection bida The licensee had. modified the detection. system at the location where successful penetration had been achieved through the protected area. Demonstration of an

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attempted entry was observed and confirmation was received that an alarm had

occurred at the security alarm station. The detectors that were in continuous alarm during the previous NRC inspection had been repaired and were operational. The modification to the fence gate had been complete l (Closed) IFl 72-017/9902-02

Deficiency with an Uninterruptible Fower Supolv System A test of the new battery was conducted for the uninterruptible power source that had-not met the required security criteria during the last NRC inspection. The battery exceeded the minimum criteria for providing back-up power, a

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g 15' Exit Meeting The inspector presented the irispection results to members of licensee management at the conclu'sion of the inspection on April 22,1999. The licensee acknowledged the findings presented. The licensee identified as proprietary two drawings provided to, or

reviewed.by, the inspector, which were not discussed in detail in this repor !

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l'* ATTACHMENT i

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PARTIAL LIST OF PERSONS CONTACTED

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Licensee K. Allison, ISFSI Project Manager A. Bowman, Radiation Protection Supervisor H. Caballero, Industrial Safety J. Cooper, Emergency Planning Engineer L. Dusek, Nuclear Regulatory Affairs Manager T. Meek, Radiation' Protection Manager J. Mehelich, Engineering Manager S. Nichols, Decommissioning Project Manager B. Shoemaker, Security Supervisor C. Storms, ISFSI Specialist G. Zimmerman, Licensing Engineer State of Oreaon A Bless, Resident inspector, Oregon Office of Energy INSPECTION PROCEDURES USED 40500 Effectiveness of Controic in Identifying, Resolving, and Preventing Problems 40801 Self-Assessments, Auditing, and Corrective Actions 60854 Pre-operational Testing of an ISFSI 71801 Decommissioning Performance and Status Review i 92701 Follow-up of Open Items '

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ITEMS OPENED, CLOSED, AND DISCUSSED:

Ooened 50-344/9905-01 IFl Cracks on the ISFSI Lifting Yoke

' Clgs9d 50-344/9901 01 IFl Movement of Concrete Cask near the Edge of the ISFSI Pad-50-344/9901 02 IFl Emergency Supplies Discussed None I

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l: 2 LIST OF ACRONYMS

'CFR Code of Federal Regulations FHP Fuel Handling Procedure 1Fl inspection Iollowup item ISFSI Independent Spent Fuel Storage Installation LER Licensee Event Report NRC Nuclear Regulatory Commission ONI Off NormalInstruction PGE' Portland General Electric Company SAR- Safety Analysis Report T. S. 1echnical Specification

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