IR 07200001/1989001
| ML20244B737 | |
| Person / Time | |
|---|---|
| Site: | 07200001, 07001308 |
| Issue date: | 04/06/1989 |
| From: | France G, Gibbons D, Sreniawski D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20244B734 | List: |
| References | |
| 72-0001-89-01, 72-1-89-1, NUDOCS 8904190372 | |
| Download: ML20244B737 (9) | |
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S.l NUCLEAR REGULATORY. COMMISSION
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REGION III-
<a TReport.No.i72-001/89001(DRSS)
Docket No.72-001
. License No..SNM-2500 Licensee:
General' Electric Company 175 Curtner Avenue San Jose, CA 95125 Facility;Name:
Morris Oper6 tion LInspection At:
Morris Operation, Morris, Illinois.
- Inspecticn Condticted:
February 27 through. March 3, 1989 h'111 ivi& @
-Inspectors:
GI M.l France,'III.
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D.J.qIawski, Chief t
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/4/A f App' roved By:
Nuclear Materials Section 2 Date
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i InspectionSummary[
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~ Inspection on February 27'through March 3, 1989 (Report No. 72-001/89001(DRSS))
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Areas Inspected:. Routine, unannounced safety insnection, including:
management and organization controls-(IP 88005); radiation protection program (IP 83822);
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operations review (IP 88020); criticality safety (IP 88015); maintenance-surveillance (IP-88025); and transportation activities (IP 86740).
Res'ults:
0ne violation was identified.
The licensee was cited for shipping two spent fuel casks with surface contamination levels exceeding the DOT limit.
Two unusual events-involving loss of efficiency in the exit portal monitorandadroppedfuelbundleinthepoolbasingavenoindicatjonthat'
any radioactive material was released to the environment.
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DETAILS
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Persons Contacted (IP 30730).
l L.LL. Denio, Manager, Plant Services-J. W. Doman,' Manager, Operations-Program
- T. E. Ingels, Morris. Operation Manger
- J. D. Kesman, Manager, Plant Operations and Maintenance
- S. P. Schmid, Senior. Specialist Licensing and Safeguards
- R..T. Smith, Maintenance Engineer-R. Wright,' Service-Technician (part time employee)
NRC
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J. F. Schneider, USNRC HQ NMSS (Project Manager)
'* Denotes those present at the exit meeting on March 2, 1989.
2.
. General This inspection was conducted to examine licensee activities under Materials License SNM-2500.
The inspector examined the licensee's transportation prog' ram with emphasis on.the removal of cask surface contamination to meet regulatory requirements.
The licensee's performance in the areas of criticality safety, management controls,
. radiation protection,.operat' ions, and maintenance surveillance was viewed as adequate to meet the regulatory requirements.
Also examined were the actions taken by the licensee to recover a dropped fuel bundle i
and to correct the loss of efficiency in the exit portal monitor.
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3.
Management Organization and Controls (IP 88005)
The inspector reviewed the license's management organization and controls for radiation protection and operations, including changes in the organizational structure.
A review of safety committee activities was also conducted.
a.
Organization A recent cut in the staffing eliminated one of the Safety Technician positions.
Upon termination from General Electric, Morris Operations (GE:M0), this individual was hired by Pacific Nuclear which has an office on site and reassigned part-time to perform radiation protection services for Morris Operations.
As previously reported, remaining Safety Technicians and Service Technicians have either been cross-trained or scheduled for training and reclassified as Senior Technician Plant Services.
Now that receipt of spent fuel shipments are complete, workers reclassified as Senior Technician Plant Services will cover radiological health and safety, and plant services required by GE:MO.
The inspector determined that cross-training of Service
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a Technician in matters of radiation protection appears adequate.to support Health Physics' requirements.
Nevertheless, the. licensee's progress in this area will be monitored during future inspections.
The licensee stated that effective this inspection, the GE:M0 staff is comprised of 26 members.
b.
Safety Committee Effective August 11, 1988, Material License No. SNM-2500 was amended to enable the Safety Committee to conduct meetings with four members present instead of all six.
The committee consists of the following members:
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(a) Manager - Morru Operations (b) Manager - Plei. Operations and Maintenance (c) Manager - Plant Services (d) Operations - Program Manager (e) Maintenance Engineer (f) Senior Specialist - Licensing and Safeguards As required by Appendix A, Technical Specifications for Safety, License No. SNM-2500, tM plant Safety Committee conducted monthly meetings during the July through December 1988 operating period.
The committee responded to several items regarding radiological health and safety practices, specifically:
Protective Clothing - The committee stated that proper procedures for clothing selection, use, disposal and personal survey are documented in the Plant Safety Manual.
Hot Particles - The committee established a policy for examining elevated contamination levels for discrete particles that have the potential to chose high local exposure.
Internal Dose Projection
.The committee established,a procedure to examine whole body and bioassay data and determine if additional actions cre warranted, particularly for cobalt-60 arid cesium-137 j
levels.
ALARA - The plant committee enlisted the aid of Dr. K. Eger, former j
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Nuclear Licensing and Radiation Specialist at GE:M0 to serve as an independent consultant on matters pertaining to radiological health and safety.
Dr. Eger attended several plant committee meetings during the year 1968.
The inspector concluded that the licensee has responded adequately to the radiological health and safety concerns i antioned during a previous inspection.
c.
Quality Assurance (QA)
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During this inspection two representatives of USNRC HQ, NMSS, j
Messrs. Leu and Schneider reviewed the licensee's QA program b
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L and the events concerning the dropped fuel bundle.
Their findings
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are not included in this inspection report.
Mr. Leu will provide a separate report.
No. violations or deviations were identified.
4.
Radiation Protection (IP 83822)
The inspector reviewed the licensee's internal and external exposure control programs including the required records, reports, and notifications.
a.
Internal Exposure Control Whole body count results of operations and maintenance personnel were reviewed for fission and activation products.
Results for each individual indicated that the higliest reported observations (10 pCi - cesium-137) was less than one percent of the maximum permissible body burder Fourth quarter (1988) urinalysis results for persons performing spent fuel storage operations and/or cask maintenance duties disclosed trace quantities of cesium-137 in two workers.
Apparently, efforts to control exposure due to internal deposition of material continue to be successful.
b.
External Exposure Control The basin filter room remains a high radiation area where exposure levels exceed 100 mR/hr.
This area is locked for limited access.
TLD badge records for the 4th quarter.1988, indicate that exposure from cask maintenance and other basin activities was less than 500 mR/hr to any one individual.
Visitors to the fuel basin are limited to a maximum of 25 mrem per day.
Visitors must be accompanied by a GE:MO employee equipped with a self-reader dosimeter.
The inspector took independent measurements along the basin walkway which also serves as the visitors pathway and determined that direct readings along the basin wall ranged from 10 to 13 mrem /hr.
No problems were noted.
. Licensee personnel wore film badges supplied by the Landauer Company and are exchanged monthly.
Radiation safety personnel review the monthly reports and record dosage in a log book.
The log book is reviewed at the end of each year for subsequent preparation of the annual report.
Ultimately, the report is submitted to the Commission as required by 10 CFR 20.407(a)(1).
Licensee personnel also wear TLD and self-reader dosimeters that are recorded daily.
Records indicate that the September 1988 badges were lost while in transit to Landauer's (vendor) labo'ratory.
The
licensee correlated the data from the September 1988 daily records
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of self-reader dosimeters and used the results to assign exposure dosage for the September 1988 operating period.
To date, there has been no indication that the badges were recovered.
The inspector
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determined that the licensee's method of estimating the monthly i
exposure was acceptable.
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c.
Airborne Releases Records'of air sampling data were reviewed for the-July-1988 through January'1989 operating period.
The concentration of l
airborne radioactivity measured in the. fuel basin area continues L
to be less than the MPC value of IE-08 uCi/ml.
The highest
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concentration was reported.to be 2.6E-10 uCi/ml.
q d.
Surveys and Contamination Control l
The l'icensee limits. contamination in onsite areas to less than'
2200 dpm/100 cm2 beta gamma and 220 dpm/100'cm2 alpha. 'The inspector performed a beta gamma survey (for removable contamination)
in.the basin area and the first aid room. -The contamination level of_the highest smear was less than 100 dpm/cm2,.
The licensee indicated that workers are instructed to perform a l
.self-survey with a GM pancake probe (frisker) prior to leaving the restricted area.
A high efficiency portal monitor (Eberline PM-6)
is located at the plant exit' guard station.
The PM-6 is the final surveillance instrument for persons exiting the plant.
On December 10, 1988, a loss in detection efficiency occurred in the PM-6 counting system causing the foot and head detectors to be inoperative over a period of four operating shifts.
Apparently, there~was insufficient gas flow.
To prevent recurrence, the licensee installed an audible gas pressure alarm. Workers were also instructed to perform self-surveys with a portable survey
<7 meter whenever the instrument trouble light is illuminated.
An
'. investigation by the Plant Safety Committee indicated that there were no employees assigned to contaminated areas during the weekend of December 10 and 11, 1988.
Hence, it is unlikely that any employee exited the plant while wearing contaminated clothing.
e.
Sealed Sources
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Leak tests performed on sealed sources were completed during the May 1988 through January 1989 operating period.
No leaks were identified in any of the sources.
No violations or deviations were identified.
5.
Criticality Safety (IP 88015)
The inspector examined the licensee's nuclear safety records and confirmed that the licensee maintains a detailed description of the physical, thermal, and radiological characteristic of spent fuel received from the Nebraska Public Power District.
The fuel parameters such as uranium-235 enrichment, rod lattice limits, and/or pellet diameter, fall within the range defined in Appendix A, Material License SNM-2500.
Licensee work sheets indicated that Nebraska Cooper had'provided the fuel bundle acceptability list that certifies l
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the limits on enrichment'for cold fuel.
These limits were used by the licensee to compare the maximum reactivity of cold fuel with the
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post-irradiation data.
No discrepancies were.'oted in the' licensees-work sheets.
No violations or' deviations were identified.
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Operations Review-(IP 88020)
a.
Cask Operations On January 26, 1989, the licensee received the.last shipment of spent fuel scheduled for storage at GE:M0..Two spent fuel casks, one of. which is empty, are presently located at GE:MO.
The empty cask was being serviced in preparation for shipment to Carolina Light and Power.
The filled cask contained the last'of the spent fuel bundles received from Nebraska Public Power District.
The licensee stated.that radiation surveillance is being maintained on the filled cask.
The licensee agreed to inform Region III prior to transferring the last of the spent fuel bundles to basin storage.
b.
Unusual Event (BWR Fuel Grapple Malfunction) (See Operations Alert 88-8, Attached to this Report)
On November 20, 1988, during an underwater fuel unloading operation a-fuel bundle disengaged from the grapple and subsequently dropped about one foot to the bottom of the storage basket.
Disengagement occurred while transferring the bundle from the cask to the underwater
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storage basket.
A weak spring in the grapple jaw either allowed the-bundle to disengage or failed to open the grapple jaw to allow appropriate engagement.
In order to continue the cask unloading operation, a washer was inserted to increase spring tension.
The fuel unloading operation proceeded by using the field expedient correction (washer) until ultimately, the spring was replaced.
There was no increase'in airborne radioactivity as determined by examining the basin air sampler data.- Nor did the basin water quality exceed the specifications required by the license.
During this inspection, the inspector reviewed a videotape which was prepared by the licensee of the damage that occurred to the bundle and compared the configuration to an empty bundle that was intact and being used for training purposes.
The videotape showed a slight deformity to the lower base bundle tie plate (bundle inlet nozzle).
The inlet nozzle was pushed inward from the impact of the'one foot fall.
The inspector held discussions with the licensee ar.d reviewed Operation Alert 88-8 (attached to this report).
Checking the latch spring tension has been added to the preventive maintenance schedule.
In addition to routine service checklists, fuel handling equipment such as the basin crane and its auxiliary equipment (yoke, guide rack, grappling device) are serviced during annual maintenance.
The maintenance and testing as performed by the licensee is verified
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he under the licensee's Quality Assurance Plan (No. NED0-20776).
Testing the latch spring tension by routine service and operations checklists along with annual-and QA review appears to be adequate for ensuring that conditions adverse to safaty are promptly identified and corrected.
The Region III staff will continue to monitor the licensee's program for maintaining fuel handling equipment during future inspection.
The inspector concluded that the licensee took appropriate measures to assess the damage of the dropped fuel bundle, to implement steps for corrective and/or preventive action, and to evaluate the affect
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that the incident had on the radiological health and safety of GE:M0
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workers.
c.
Basin Integrity The Region III staff inquired about the longevity and integrity of the basin for spent fuel storage.
According to GE:M0 Consolidated Safety Analysis Report (SAR), the
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cask unloading basin is a reinforced concrete structure poured against a natural formation of bedrock.
Concrete samples were inspected for every pour of 100 yards to ensure ASTM cement forming specifications.
The concrete storage basin is equipped with a stainless steel liner.
Beneath the stainless steel liner the basin floor is protected from shock with a 1.75 inch thick steel absorber i
plate.
This basin design consisting of reinforced concrete along with the stainless steel liner has been estimated (contractors i
engineering estimate) to have a useful life of more than 100 years.
The licensee indicated that testing involves routine water quality j
and leak detection system surveillance.
In a nonroutine test,
. coupons made of basin liner material were subjected to corrosion tests.
Test results indicated that there was no evidence of pitting or cracking.
The licensee stated that the basin design also included consideration for earthquake and tornado conditions.
The licensee plans to examine the spent fuel storage basin and determine if additional testing for basin longevity and integrity is needed.
The Region III staff will monitor the licensee's program for basin integrity during future inspections.
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No violations or deviations were identified.
7.
Maintenance Surveillance Tests (IP 88025)
The inspector reviewed the results of the surveillance tests required by the technical specifications of Appendix A to License No. SNM-2500.
The required measurements of basin water quality, basin leak rate and operability, criticality monitors, and stack effluent air were made at the specified frequencies.
Off-standard conditions were either recorded on the operator log and/or was referenced in the licensee's unusual event
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file.
The most significant finding was a partially restricted dip-leg located in the basin leak detection system.
The restriction caused a false reading on the level indicator.
Consequently, the restriction was removed and the indicator was recalibrates.
No violations or deviations were identified.
8.
Transportation Activities (IP 86740)
The inspector reviewed the licensee's program for receipt and/or shipment of radioactive materials.
During the July 1988 through January 1989 operating period, five Region III inspectors reviewed the licensee's program for receipt and shipment of radioactive materials.
The review included inspection of seven incoming rail shipments transporting 13 casks filled with spent reactor fuel.
The inspectors determined that minor cask contamination problems are still being noted in the spent fuel shipment program.
On January 13, 1989, two empty IF-300 casks arrived at Nebraska Cooper Public Power District (NPPD).
The casks were shipped by GE:M0 on January 6, 1989.
The pre-shipment survey data (examined by the inspector) averaged less than 500 dpm/100 cm2 for each of the two casks.
Upon arrival at NPPD, the incoming survey indicated that 17 individual wipes exceeded 22,000 dpm/100 cm2 on the first cask and six individual wipes exceeded the 22,000 dpm/100 cm2 DOT limit on the second cask.
NPPD indicated that the highest individual wipe determined for each cask was
141,140 dpm/100 cm2 on the first cask and 77,125 dpm/100 cm2 on the
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second cask.
Although NPPD workers used other wipe sampling methods as allowed in 10 CFR 71.87 and clarified in IE Information Notice No. 88-46, to demonstrate greater efficiency, the removable contamination levels exceeded the 10 CFR 71.87 and 49 CFR 173.443 transportation criteria.
The inspector reviewed the radioactive contamination survey results obtained on the two casks, and concurred with NPPD findings that contamination levels exceeded the limits specified in 10 CFR 20.205(b)(2)
and transportation limits specified in 10 CFR 71.87(i) and 10 CFR 173.443.
One violation was identified.
9.
Exit Meeting The scope and findings of the inspection were discussed with licensee representatives (Section 1) at the close of the onsite inspection on March 2, 1989.
The following matters were discussed:
a.
It was noted that the licensee was in violation of certain requirements of 10 CFR 71, Transportation of Radioactive Materials.
(Two casks exceeded surface contamination limits, Section 8)
b.
The licensee agreed to notify Region III prior to transferring the remaining fuel bundles to basin storage.
(Section 6)
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The inspector acknowledged the licensee's actions in returning the i
grappling device to an operable condition (Section 6); and to I
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install an alarm device in.the portal monitor.
(Section 4)
d.
The licensee plans to review methods for testing basin integrity and estimating longevity.
(Section 6)'
During the course of the inspection and the exit meeting, the licensee did not identify any documents or inspector statements and references to specific processes as proprietary.
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