IR 05000320/1990003
| ML20043G083 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 06/14/1990 |
| From: | Young F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20043G081 | List: |
| References | |
| 50-320-90-03, 50-320-90-3, NUDOCS 9006190035 | |
| Download: ML20043G083 (7) | |
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V. S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-320/90-03 Docket No.
50-320 License No.
OPR-73 Priority Category C
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Licensee:
GPU Nuclear Corporation P. O. Box 48C Middletown, Pennsylvania 17057 Facility Name: Three Mile Island Nuclear Station, Unit 2 Inspection At: Middletown, Pennsylvania Inspection Conducted: March 3, 1990 - May 15, 1990 Inspectors:
F. Young, Senior Resident Inspector
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D. Johnson, Resident Inspector
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T. Moslak, Resident Inspector
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R. Brady, Resident Inspector L. Thonus, Project Manager M. Masnik, Senior Project Manager i
i Approved by:
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F. Young, Acting Section Ch4ef Date Reactor Projects Section 4B Division of Reactor Projects
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Inspection Summary:
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Areas Inspected:
Routine safety inspections were conducted by site inspectors of defueling and decontamination activities.
A review of your safe fuel mass limit analysis was conducted. The 1989 annual Summary of the Changes to TMI-2 Systems and Procedures-Report -and Licensee Event Report 90-01 were reviewed.
Results: The licensee safely completed vessel cleanup activities.
The; final shipment of core materials was shipped of fsite.to Department of' Energy facilities.
Inspection activities confirmed that removal of all canisters containing core material from the site i,cd occurred. Based on review of the licensee's final criticality study, inadvertet criticality from remaining core debris in the reactor vessel is precluded. The licensee has completed the transition from Mode 1 to Mode 3 in accordance sith Technical
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Specification (TS) requirements. One licensee identified violation was noted concerning Polar Crane Load handling.
L 9006190030 900614 PDR ADOCK 05000320 Q
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TABLE OF CONTENTS Page 1.0 Overview....................
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1.1 Licensee Activities.................
I 1.2 NRC Staff Activities (NIP 71707)*..........
1.3 Polar Crane Loading Handling Event.
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2.0 Facility Transition to Mode 3...............
3.0 Safe Fuel Mass Limit Analysis (37700)...........
4.0 Report Review,....
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4.1. Annual Summary of the Changes to TMI-2 Systems and Procedures Report (37700)
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4.2 Licensee Event Report Review (90712).........
5.0 Exit Meeting (NIP 30703).................
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- Indicates the NRC. inspection procedure used to inspect the area identified'
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DETAILS 1.0 Overvicw 1.1 Licensee Activities Following completion of the lower head sampling program, the final reactor vessel cleanup operation was conducted and was completed on March 13, 1990. A video inspection of the reactor vessel was con-ducted using an underwater color camera.
The defueling water cleanup system was shutdown and taken out of service on March 16. As of March 20, the reactor building was void of defueling canisters. The final shipment of fuel debris was shipped from TMI-2 on April 15 and TMI-2 transitioned from Mode 1 to Mode 3 on April 27, 1990.
1.2 NRC Staf f Activities Facility Transition to Mode 3 Inspections were conducted to verify the licensee compliance with Technical Specifications (TS) in regard to meeting conditions for transition from Mode 1 to Mode 3.
More detail regarding the licens-ing aspects of the mode changes are contained in Section 2,0 of this report, Evaporator During this inspection period, one successful surrogate test was completed.
Following the completion of the test, the licensee and subcontractor elected to perform an outage on the-evaporator unit.
The major scope of this outage is to perform major modifications to the solid waste packaging system, including the removal of'the pel-letizer unit and installation of a new waste packaging system. The evaporator outage is scheduled to be completed in early July.
The licensee plans to conduct two surrogate test-runs to verify oper-ability of the modified evaporator complex.
1.3 Polar Crane load Handling Event
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On February 19, during evolutions related to the lower head sampling program, Temporary Change Notice (4210-3255-90-022) was issued to TMI-2 Operating Procedure 4210-0PS-3555.04, " Withdrawing Incore Instrument Strings with Polar Crane".
The procedure-change allowed the polar crane to be moved over an exclusion zone with the main hoist energized. This violated SER 4700-3882-89-01 which only permitted polar crane movements with the main hoist deenergized.
The temporary change was immediately can-celled and corrective actions to preclude reoccurrence have been taken.
Review of this event discovered that the polar crane had
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also been moved over the exclusion zone with the main hoist energized
on January 4,1990. The root cause on both cases was pers_onnel error i
on part of the supervisors who misinterpreted the_ operating procedure l
limits and precautions.
Although an approved SER was violated, no notice of violation will be issued based on 10 CFR 2, Appendix C V.G.
The event was licensee identified, and is classified as a Level IV. violation per 10 CFR 2, Appendix C, Supplement I.D.3.
The Licensee has met the reporting-t requirements as specified in 10 CFR 50.73.
The inspector reviewed the licensee's corrective actions and concluded they are adequate.
The inspector noted this was an isolated occurrence and, therefore, could not have been prevented by corrective actions implemented due to a previous violation.
For administrative purposes, this item will be tracked as a licensee identified violation.
This item is considered closed (50-320/
90-03-01).
2.0 Facilities Transition to Mode 3 Section 1.3 and Table 1.1 of the Technical Specification, Three Mile Island - Unit 2 require the following-three conditions for the licensee to transition from Mode 1 to Mode 2:
t The Reactor Vessel and Reactor Coolant System are defueled to the
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extent reasonably achievable.
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The possibility of c iticality in the Reactor Building is precluded.
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There are no canisters containing core material in the Reactor
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Building.
- The additional requirement for transition to' Mode 3 requires that no can-isters containing core material be stored on the THI-2 site.
Mode 1 was a post-accident defueling and cleanup mode. Mode 2 recognized the completion of defueling; licensed operators are no longer required to supervise core alterations and many systems related to core monitoring, cooling, and shutdown are no longer required to be operable. Mode 3 recog-
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nizes the offsite shipment of all canisters of core material.
In Mode 3, bora i s ind monitoring of the spent fuel pool is no longer required.
The NRC staff implemented a program to verify that the licensee had met
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all the criteria for the different mode changes._ The program included a review of the ccmpletion of defueling, possibility of criticality and physical verification of the absence of fuel canisters on the -TMI-2 site.
On March 23, 1990, an inspection of the reactor building was conducted.
The inspectors verified that no containers containing core material remained in the reactor building.
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On April 20, 1990, an inspection was conducted in the auxiliary building, fuel handling building, truck bay, and rail siding.
In the fuel handling i
building, spent fuel storage pool "A" was observed to contain three l
" dummy" canisters and one empty filter canister. The' inspectors verified J
that removal of canisters containing core debris material had been accom-l plished in these areas and there are no canisters containing core material
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stored on the TMI-2 site.
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The site staff (resident inspectors and the onsite NRR project manager)
reviewed the licensee's Defueling Completion Report dated February 20,
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1990, and additional documentation supplied in a licensee letter dated April 12, 1990. -Following the completion of defueling, the licensee
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videotaped the inside of the reactor vessel and portions of the reactor coolant system (RCS).
The licensee made external radiation measurements
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to determine the residual fuel content in the reactor coolant and on the
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inside surfaces of the RCS.
The residual fuel was primarily in the form l
of a thin film coating on the inside of RCS piping and pump volutes.
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NRC staf f reviewed selected video tapes and plant drawings and performed j
independent calculations.
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Based on this review, the staff made the following conclusions and obser-i vations:
i The' licensee had identified all locations containing significant amounts
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(greater than 4 Kg) of residual fuel. The licensee's assumptions j
(density, three dimensional geometry model) in determining residual fuel
were conservative.
The licensee's calculations were accurate.
The remain-
ing fuel was in locations which were dif ficult to access and further de-i fueling would be person rem intensive. The TMI-2 reactor, RCS'and auxil-l iary systems had been defueled to the maximum extent practicable.
h The NRC staff, assisted by consultants from Batte11e's Pacific Northwest
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Laboratory (PNL), reviewed the licensee's criticality analysis for the 609 kg of fuel remaining in the reactor vessel and examined the assumptions and calculations performed by the licenset.
The licensee's assumptions were' highly conservative and their calculations using the KENO V.a com-puter code were conservative and acceptable.
The conservative calcula-tions indicated a worst case K-ef fective of less than 0.95 for the resi-dual fuel.
The inspectors performed a similar examination of a licensee
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criticality model in which fuel accumulates in the reactor vessel lower head.
The results showed the same conservative modeling with K-effective of less than 0.95.
The staff concluded that criticality could not be attained provided that the remaining core debris maintained the_ geometries specified in the Defueling Completion Report.
The NRC staff concluded that the possibility of criticality in the reactor vessel was preclude.
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3.0 Safe Fuel Mass Limit The safe fuel-mass limit is the amount of fuel that can safely be con-tained in a single mass in any geometry without the possibility of an inadvertent criticality.
The NRC staff and consultants from Battelle PNL reviewed the licensee's safe fuel mass limit analysis. With the exception of fuel enrichment, the licensee's assumptions were conservative; in some cases very conservative.
The licensee's analytical methods using the ORIGEN and KEN 0 V a computer codes were conservative and acceptable to the staff.
The principal conservatisms included using standard size pellets, neg-lecting the presence of dilutents and poisons, optimizing reflection and moderation.
Greater than ninety per cent of the remaining fuel at TMI-2 is in the form of surface films, fine granular debris and resolidified masses.
This size and shape is much less reactive than standard pellets.
The remaining fuel has incorporated and is interspersed with a variety of structural material, principally Inconel and stainless steel, which tends to dilute the theoretical fuel / moderator matrix.
A variety of poisons, including boron, fission products and portions of the silver-indium-cadmium control rods are also incorporated in and interspersed with the remaining fuel. The licensee's calculations also assumed that the theo-retical mass was optimally reflected with steel and optimally moderated with unborated water.
The staff found these assumptions to be conserva-tive and acceptable.
Based on this. analysis, the licensee established a safe fuel mass limit of 140 kg of core debris.
During review of the analysis, the staff questioned the licensee's assump,
tion of fuel enrichment, and fuel mixing.
The licensee assumed that the fuel was uniformly mixed with the average enrichment of the three regicns of the core after burnup of 2.24 percent.
However, fuel samples taken during the defueling process has shown localized areas where the enrich-ment has exceeded 2.24 percent.
Based on these empirical values, the NRC determined the appropriate safe fuel mass limit to be 93 kg of core debris in the reactor vessel.
The calculation that established this limit does not take credit for uniform mixing of the three fuel regions.
The staff concurred and accepted the licensee's safe fuel mass limit of 140 kg only for the out-of-reactor vessel region, and only for the fuel that was transported out of the reactor vessel during the March 28, 1979 accident and subsequent defueling activities.
If any fuel is removed from the reactor vessel in the future, the 93 kg limit will apply to that fuel.
If the fuel in the reactor vessel is rearranged outside the analyzed geo-metries used in the reactor vessel criticality analysis, the 93 kg limit will apply to the rearranged fuel.
These restrictions will apply until an additional safety analysis is approved by the NRC staff.
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4.0 Report Review 4.1 Annual Summary of the Changes to TMI-2 Systems and Procedures Report The inspectors reviewed the report dated March 29, 1990, submitted to the NRC by the licensee in accordance with the requirements of 10 CFR 50.59.
This report provided a short discussion of procedure changes and system modifications made to procedures and system described in the Final Safety Analysis Report (FSAR).
The inspector review of this report determined that the information was adequate and generally descriptive enough to allow the inspector j
to determine the significance of the particular change, or modifi-j cation.
The inspector _ verified that no modification or procedure change required a change to the unit technical specifications.
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' inspector concluded that the licensee had adequately fulfilled the
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requirements of 10.FR 50.59 in yearly reporting of changes and modifications.
l 4.2 Licensee Event Report (LER) Review The inspector reviewed the Licensee Event Report (LER) 90-01 which was submitted to the NRC Region I of fice pursuant to 10 CFR 50.'/3.
LER.90-001 dated March 21, 1990 addressed heavy load handling activity in the Containment Building outside the bounds of -
docketed Safety Evaluation Report (SER) approved by the NRC.
- Based on a review of the LER, the inspector determined that: the technical content was accurate, the corrective actions specified I
were appropriate and there were no generic issues.
In addition to the technical-adequacy of the LER, the compliance with.
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the requirements of 10 CFR 50.73 were reviewed.
There were no i
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deficiencies noted.
i 5.0 Exit Meeting
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i The inspector met the licensee representatives, denoted below, at the
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conclusion of the inspection on May 11, 1990. The inspectors summarized the purpose, scope, and findings of the inspection. At no time during the inspection did the inspector provide any written information to the-
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licensee.
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- J. Byrne, Manager, TMI-2 Licensing
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- W. Marshall, Manager, Plant Operations, TMI-2
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- E. Schrull, TMI-2 Licensing Engineer
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J. Thomas, Engineer, TMI-2 i
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M. Roche, Director, TMI-2 i
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R. Rogan, Director, Licensing & Nuclear Safety, TMI-2
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- Attended the final management meeting, d
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