ML19351A271

Forwards Third Submittal of Defueling Completion Rept.Encl Provides Residual Fuel Quantities for Portions of Rcs.Page Changes Incorporating Minor Corrections to Previously Submitted Sections of Rept Encl

ML19351A271
Person / Time
Site:	Crane
Issue date:	10/09/1989
From:	Roche M GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
4410-89-L-0102, 4410-89-L-102, NUDOCS 8910190087
Download: ML19351A271 (40)

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4 OPU Nuclear Corporation gg gf Post Ofhce Box 480 Route 441 South Middletown, Pennsylve.nia 17057-0191 717 944 7621 L

TELEX 84 2386 Writer's Direct Dial Number:

(717) 948-8400 October 9, 1989 4410-89-L-0102/0477P US Nuclear Regulatory Commission Washington,'DC 20555 Attention: Document Control Desk Th m e Mile Island Nuclear Station, Unit 2 (TMI-2)

Operating License No. DPR-73 Docket No. 50 320 Defueling Completion Report, Third Submittal

Dear Sirs:

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Attached is the third submittal of the Defueling Completion Report (DCR).

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Previous submittals were provided by GPU Nuclear letters 4410-89-L-0070 dated JJ1y 5,1989, and 4410-89-L-0078 dated August 18, 1989.

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. The attachment provides residual fuel Quantities for portions of the Reactor Coolant. System (i.e., Section 5.3).

Fuel measurements of the remaining portions of the RCS,--which may impact or be impacted by ongoing defueling activities (reference Table 5-4), will be performed following the completion ofLReactor Vessel defueling. An update of Section 5.3, including the results of these measurements, will be provided as part of the final DCR submittal.

The attachment also includes Sections 4.4.3.3.8; 4.4.3.4, and 4.4.3.5, which describe fuel removal activities in the Lower Core Support Assembly (CSA),

Lower Head, and Upper CSA, and Section 4.3.4, "RCS Fuel Removal Assessme

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Page changes incorporating minor corrections to previously submitted s ons of the DCR are attached.

Changes are annotated by change bars in the m gin.

Changed pages and the new material submitted herein are identified a

. Revision 2 to the initial submittal of the DCR.

Sincerely, Yh K M. B. Roche Director, TMI-2 r s GPU Nuclear Corporation is a subsidiary of the General Pubhc Utihties Corporation 8910190087 891009 f

fDR ADOCK 0500 $ O j

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. Document Control' Desk

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October ' 9,1989.

g. e 4410-89-L-0102 L

Page 2.

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cc:

W. T; Russell - Regional Administrator, Region I J. F. Stolz - Director, Plant Directorate 1-4 L. H. Thonus - Project Manager, TMI Site F. I. Young - Senior Resident Inspector, TMI L

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'l DEFUELING COMPLETION REPORT

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List of Effective Pages PAGE REVISION PAGE REVISION f

Table of Contents 2-16 0

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_t DEFUELING COMPLETION REPORT:

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Appendix A 5-10 1

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l TABLE OF CONTENTS E

SECTION PAGE 4.0 FUEL REMOVAL ACTIVITIES 4-1 c

4.1 Auxiliary and Fuel Handling Buildings 4-1

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i 4.1.1 Cleanup Approach 4-1 4.1.2 Auxiliary and Fuel Handling Buildings Cleanup 4-3 Equipment and Techniques 4.1.3 Auxiliary and fuel Handling. Buildings 4-6 7

Cleanup Activities r.

4.1.4 Auxiliary and Fuel Handling Buildings fuel 4-8 Removal Assessment 4.2 Reactor Building Fuel Removal and Decontamination 4-9 l

Activities 4.2.1 Cleanup Approach 4-9 4.2.2 Reactor Building Cleanup Equipment and 4-9 j

Techniques 4.2.3 Major Reactor Building Cleanup Activities 4-9 4.2.4 Reactor Building Fuel Removal Assessment 4-11 4.3 Reactor Coolant System Defueling Operations 4-12 4.3.1 Reacter Coolant System Defueling Approach 4-12 l

4.3.2 Reactor Coolant System Defueling Equipment 4-12 and Techniques 4~.3.3 Reactor Coolant System Defueling Activities 4-12 l

4.3.4 Reactor Coolant System Fuel Remeval Assessment 4-13 l

4.4 Reactor Vessel 4-14 l

4.4.1 Reactor Vessel Defueling Approach 4-14 i

4.4.2 Reactor Vessel Defueling Equipment 4-14 i:

and Techniques l

l 4.4.3 Reactor Vessel Defueling Activities 4-14 4.4.4 Reactor Vessel Fuel Removal Assessment 4-20 i

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SECTION PAGE Ya

-5.0 RESIDUAL FUEL QUANTIFICATION AND CRITICALITY ASSESSMENT 5-1 p

h 5.1 Auxiliary and Fuel Handling Buildings 5-1 1

5.1.1 Auxiliary and Fuel Handling Buildings Cubicles 5-1 4

5.1. 2 '

Discussion of Areas Containing fuel in-5-2 the Auxiliary and Fuel Handling Buildings L

5.1.3 Summary 5-9 5.2 Reactor Building 5-10 h

5.2.1 Reactor Vessel Head Assembly 5-10 5'.2.2 Reactor Vessel Upper Plenum Assembly 5-11 5.2.3 Fuel Transfer Canal 5-12 5.2.4 Core Flood System 5-12 5.2.5 D-Rings 5-14 5.2.6 Upper Endfitting Storage Area 5-14 5.2.7 Reactor Coolant Drain Tank 5-15 5.2.8 Letdown Coolers 5-15 5.2.9 RB Basement and Sump 5-15 5.2.10 Miscellaneous Systems and Equipment 5-16' 5.2.11 Criticality Assessment 5-16 5.2.12 Summary 5-17 l

5.3 Reactor _ Coolant System L-5.3.1 Pressurizer 5-20 5.3.2 Decay Heat Drop Line 5-20 5.3.3 Once Through Steam Generator 5-21 5.3.4 Criticality Assessment 5-22 l

5.3.5 Summary 5-22 5.4 Reactor Vessel (To Be Published in Part 4 Submittal) l L

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SECTION PAGE~

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6;0 ASS'ESSMENT OF HAJOR' RESIDUAL FUEL DEPOSITS e.

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'(To Be Published in Part 4 Submittal) o b

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7.0' OCCUPATIONAL EXPOSURE

'(To Be Published in Part'4 Submittal)-

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f-8.0 C0NCLUSIONS

'(To Be Published in Part 4 Submittal) i 1

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[.4 TABLE OF CONTENTS l

.s, TABLES j.

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TITLE PAGE 1-1 Facility Modes 1-4 I

' l -2 Acronyms 1-5

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' Post-Accident Estimated Ex-Vessel Core 2-12 L

Material Distributic,n b":

3-1 Fuel Measurement Selection 3-6 r

5-1 Auxiliary and Fuel Handling Buildings Cubicles 5-24

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Which Contain No Residual Fuel ir 5-2 Auxiliary and Ft.el Handling Buildings Cubicles 5-27 Which Potentially Contain Residual Fuel

'E 5-3 Residual' Fuel Quantification in the Reactor Building 5-30 5-4 Residual Fuel quantification in the Reactor Coolant 5-31 System FIGURES i

FIGURE NO.

TITLE PAGE 2-1 Hypothesized Core Damage Progression 2-13 2-2 Post-Accident Estimated Core Material Distribution 2-14 2-3 TMI-2 Core End-Strte Configuration 2-15 2-4 Reactor Coolant Sy; tam Components 2-16 2-5 Reactor Building Basement Floor Plan 2-17 4-1 Reactor Building Basement Floor Plan (Desludged) 4-21 4-2 Block Wall Face Identification 4-22 4-3 THI-2 Defueling Progress 4-23 4-4 Core Bore Machine 4-24 4-5 Lower Core Support Assembly 4-25 l

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TABLE OF C0i1 TENTS y,>

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' FIGURES (Cont'd) 2.g a

FIGURE'NO.

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5-1.

Auxiliary Building 280'-6" Elevation 5-32 5-2 Auxiliary Building 305' Elevation 5-33 5-3 Auxiliary Building 328' Elevation 5-34 5-4 Auxiliary Building 347'-6" Elevation 5-35 i

5-5 THI-2 Reactor - Upper Half 5-36 5-6 Leadscrew and LS Support Tube 5-37 i

APPENDICES Appendix A References

' Appendix B Safe Fuel Mass Limit Appendix C Inspection Plan (To Be Provided Later.

Appendix D Video Document References (To Be Proviced Later)

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TABLE l-2

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L ACRONYMS

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AB Aux 1?lary Building ACES Automated Cutting Equipment System AFHB Aux 1)lary and fuel Handling Buildings ALARA As Low As Is Reasonably Achievable 1

CBM Core Bore Machine CFT Core Flooo Tank CRA Contral Rod Assembly CSA Core. Support Assembly CWST.

Concentrated Waste Storage Tank DCR Defueling Completion Report DF Decontamination Factor DHR Decay Heat Removal DOE Department of Energy DHCS Defueling' Water Cleanup System i

ECCS Emergency Core Cooling System FHB Fuel Handling Building FTC-Fuel Transfer Canal GM Geiger-Mueller Counter HEPA High-Efficiency Particulate Absolute ~

HPGe High-Purity Germanium HPI High Pressure Injection IIGT Incore Instrument Guide Tube l

INEL Idaho National Engineering Laboratory L5 Lead Screw LCSA Lower Core Support Assembly MDL Minimum Detectable Level l

MeV Million Electron Volts l

MU Makeup MULP Makeup and Purification l

MWHT Miscellaneous Waste Holdup Tank l-l L

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+4 TABLE 1-2 (Cont'd)

ACRONYMS

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NaI(TI)

Thalluim Drifted Sodium Iodide NRC Nuclear Regulatory Commission

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OTSG Once-Through Steam Generator PORV Pilot Operated Relief Valve RB Reactor Building RCBT Reactor Coolant Bleed Tank RCDT Reactor Coolant Drain Tank RCP Reactor Coolant Pump-RCS Reactor Coolant System RV Reactor Vessel SDS Submerged Demineralizer System SER Safety Evaluation Report SFML Safe fuel Mass Limit SIVR Seal Injection Valve Room SNM Special Nuclear Material SRST Spent Resin Storage Tank SSTRs Solid-State Track Recorders Si(Li)

Lithium Drifted Silicon TMI-2 Three Mile Island, Unit 2 TRVFS Temporary Reactor Vessel Filtration System UCSA-Upper Core Support Assembly HDL Waste Disposal Liquid 1-6 Rev. 2/0461P

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4.3.3.3, Once-Through Steam Generators and Hot Legs (References 4.22

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and 4.23) l Pick-and-place and vacuuming techniques were used to defuel i

the "A" and "B" OTSG upper tube sheets.

Long-handled gripping tools were used to lift large pieces of debris into canisters c

and a vacuum system removed the smaller debris. While this process est,entially succeeded in defueling the "A" OTSG tubesheet, a crust of tightly adherent debris remained on the surface of the "B" OTSG tubesheet.. Despite extensive efforts to remove this crust or to t.ollect a sample for analysis by scraping, no further progress was achieved.

It has been concluded that no further defueling of the "B" OTSG tubesheet

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is necessary or appropriate because the small amount of remaining fuel is tightly adherent and unlikely to be transported elsewhere in the system.in the future due to a L

lack of a motive force and our demonstrated inability to t

remove it with dynamic defueling techniques.

The OTSG tubes were surveyed to detect blockages and adherent fuel-bearing films. GM counters and alpha detectors were c

used.

The lower head of the OTSGs and the J-Legs were surveyed using GM counters and activation foils.

No further defueling efforts are planned.

The hot legs were defueled using a combination scraper / vacuuming tool and the Westinghouse vacuum system.

Residual fuel in the "B" hot leg was scraped, flushed, and vacuumed into defueling canisters as part of RV defueling (Section 4.4).

Further assessment of the dynamic defueling techniques applied in attempting to remove the tightly adherent residual fuel in r

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the "B" OTSG upper tubesheets is provided in Section 6.0.

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4.3.3.4 Decay Heat Drop Line (Reference 4.23)

The in-vessel vacuum system was used to defuel the Decay Heat Drop Line. A deployment tool was developed to guide the vacuum hose into the Decay Heat Drop Line from the RCS "B" hot i

leg. All loose debris in the vertical portion of the Decay I

Heat Drop Line was vacuumed.

Below the vacuumable loose debris, a hard compacted region of debris was encountered. A drain cleaning machine was used to penetrate this hard debris 3

and size it so vacuuming could continue.

The material was l

airlifted into the "B" hot leg and was removed, as described in the above section, as part of the RV defueling.

4.3.4 RCS Fuel Removal Assessment Extensive defueiing operations were performed in the RCS with the l'

goal of removing the majority of fuel transported to the RCS as a result of the accident. These activities were successful.

For example, defueling operations removed approximately 97% of the fuel in the Pressurizer, approximately 70% of the fuel in the OTSG l

L Upper Tube Sheets, and approximately 95% of the fuel in the decay heat drop line.

The residual fuel quantity in the RCS components is discussed in Section 5.3.

4-13 Rev. 2/0461P

V re-installed and cutting of a large center section from the incore guide support plate was. begun.

By the end of December, g

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1988, the plate was sectioned into four, roughly pie-shaped

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pieces. All 25 cuts, including recuts required to section-this plate, were cleaned and verified.

In early January 1989, the cut quadrants of the incore guide tube support plate were lifted from the LCSA, flushed, and transferred to Core Flood Tank "A" for storage.

4.4.3.3.7 Flow Distributor Plate Removal Following completion of the incore guide support plate removal, loose debris and small pieces of fuel rods were vacuumed from above and below the flow distributor plate.

Long-handled tools were used to pick-and-place larger pieces of debris, much of which had originated in the core region and had accumulated on the flow distributor plate as the result of defueling operations.

In late February 1989, the cutting of the flow distributor began.

The plasma arc torch made 104 cuts, with numerous recuttings needed to ensure severance.

The flow distributor was cut into 26 pieces.

By the end of March, the cutting was complete.

The sections of the flow distributor plate that did not contain incore guide tubes were removed from the RV and placed inside Core Flood Tank "A".

The sections of the plate o

l' that contained incore guide tubes were bagged and stored c

inside the "A" D-Ring.

4.4.3.3.8 Lower Core Support Assembly Remnant Defueling Following completion of LCSA plate removal, LCSA remnant defueling began.

This consisted of removing the loose and resolidified debris that remained on the plate remnants..The primary defueling approach utilized high volume, low pressure water flush and low volume, high pressure cavitating water jet flush. Much of this work was done under conditions of poor to zero visibility due to the suspension of loose debris and was accomplished by indexing positioning tools to LCSA remnants to access specific target areas.

High volume, low pressure water

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flush tools were used first to flush the loose debris off the remnants and into the lower head.

The newly exposed resolidified debris was then dislodged with the cavitating i

l water jet.

This displaced material was then removed from the lower head as part of lower head defueling using airlifting and vacuuming as well as pick and place activities, i

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'f 4'4.3.4 Lower. Head Defueling l

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Lower head'defueling commenced following the removal of the l

flow distributor plate which provided a large access hole to

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the lower head.

Lower head dafueling included the removal of

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the accident generated monolith and. loose core debris on the lower head as well as post.-accident' generated debris that relocated to the lower head during'the defueling of the other e,

areas within the vessel.

l This evolution involved sizing and conditioning of the resolidified material in the monolith with the impact hammer

.and the cavitating water jet; pick and place of the rods and large debris; and airlifting and vacuuming of loose core debris.

4.4.3.4.1 Loose. Debris Defueling in the Lower Head Prior to removal of the flow distributor plate, a large quantity of material was airlifted from the lower head.to

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facilitate cutting and removal of that plate. When the final LCSA plate was removed, airlifting of the lower head was again performed to remove additional debris.

The airlifting activity removed the bulk of the loose debris; pick and place activities removed the remaining loose debris.

These activities uncovered a monolith of resolidified debris in the lower head.

Following the conditioning and sizing of the monolith, airlif tirm was repeated in order to remove the remainder of the core debris.

4.4.3.4.2 Monolith Defueling in the Lower Head The accident resulted in formation of a resolidified mass in i

the lower head which was irregular in shape varying in depth to less than two (2) feet in the center.

This resolidified debris was sized and conditioned successfully using two (2) tools.

The first, an impact hammer, was used to break up the central region where there was ready access from above.

The monolith was broken up in much the same way as one would approach the demolition of a concrete slab, starting from the outside edges and working inward.

The cavitating water jet was used to break up the remaining resolidified debris on the l

lower head which was located under the LCSA remnants and was inaccessible to the impact hammer.

Pick and place and airlifting then removed the conditioned debris.

4.4.3.4.3 Vacuuming in the Lower Head following the completion of pick and place activities and airlifting in the lower head, the lower head was vacuumed to minimize the relocation of core debris to other surfaces in the vessel during use of the airlift and to improve visibility.

The in-vessel vacuum system, a modified l

application of the in-vessel filtration system utilizing a knockout canister and filter canister in series, was used for this evolution.

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4l4.3.5 Upper Core Support Assembly Defueling

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UCSA defueling encompasses removing the fuel debris located

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L between the baffle plates and the core barrel (i.e.', the core l

former region).

Resolidified debris formed in this.' region m

during the accident.

Loose debris also was deposited during D

the accident and subsequent defueling operations elsewhere in the RV.

The scope of this defueling effort Includes gaining access to the core former region through the removal of the J

'btffle plates and removal of the resolidified and loose debris.

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4.4.3.5.1 Gaining Access to the Upper Core Support Assembly To gain access to the UCSA required removal of the baffle plates.

This was accomplished by cutting the baffle plates into eight (8) sections using the plasma are torch.

Then, the bolts and screws that fastened the baffle plates to the former plates were removed.

Bolt removal required use of a hydraullc untorquing tool and a drill tool.

The drill tool was used when the untorquing tool either failed to remove the bolt or the untorquing tool could not be used. A total of 864 bolts and screws were removed. A third operation involved clearing the kerf and recutting or drilling the baffle plate cuts previously made by the plasma torch.

'4.4.3.5.2 Baffle Plate Handling Baffle plate handling exposes the UCSA for defueling of the core former area.

Two (2) of the eight (8) baffle plate sections.were removed and hung from vent valve seats.

The exposed area was defueled before removal of the next plate section.

Handling of the plates essentially rotates each plate 90* from its original location to its final location.

4.4.3.5.3 Defueling of Upper Core Support Assembly l

Defueling the UCSA includes brushing, vacuuming, conditioning resolidified debris, and a second vacuuming.

l' The inboard and outboard surfaces of the baffle plates, the j

top and bottom surfaces of the former plates, and the inboard I

surface of the core barrel, which contained visible fuel, are I

brushed.

The task is accomplished using hydraulically-powered counter-rotating brushes mounted on a pivoting deployment end L

effector.

Loose debris was vacuumed from the core former plates after I

removal of the baffle plates and again af ter conditioning the resolidified debris and brushing the plate surfaces.

The in-vsassel vacuum system was used for this task.

Conditioning

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I' the resolidified debris in the UCSA was accomplished using mechanical methods and the cavitating water jet system.

The I

cavijet was directed to the flow holes in the periphery of the l

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.O' grid rib section outboard'of the baffle plates, which 1

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contained core debris, while.the plates were being removed,

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permitting access to these flow holes.

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4.4.4 Reactor Vessel Fuel Removal Assessment Figure 2-2 indicates that there was approximately 133,000 kg of core debris in the.RV following the accident.. The extensive

-j defueling efforts described in the above sections have been very successful in removing this core debris.

Currently, it is c.

estimated that defueling efforts will result in the removal ~of 199% of the fuel (U0 ) and nearly all of.the post-accident core i

2 debris from the'RV. A summary of the efforts in defueling the various components of the RV is provided below.

The following sections-will be updated as part of the final DCR submittal.

4.4.4.1 Core Region Figure 2-2 indicates that there was a total amount of 104,000 s

kg of core debris in the upper debris bed, resolidified mass,.

and intact fuel assemblies. Of this total 99.9% was removed as a result of defueling activities.

The small percentage of core debris remaining in this region is essentially in the R-6 incore location.

Following extensive defueling in this area, some of this resolidified mass remains.

4.4.4.2 Lower Core Support Assembly figure 2-2 indicates that there was approximately 6000 kg of core debris contained in LCSA components following the accident. Approximately 85% of this core debris was removed during the extensive LCSA removal phase and LCSA remnant defueling.

4.4.4.3 Lower Head l

Figure 2-2 indicates that approximately 19,000 kg of core l'

debris (i.e., 12,000 kg of loose debris and 7,000 kg of l

resolidified mass) existed in the RV lower head due to the accident.

The defueling efforts in the lower head region, described in Section 4.4.3.4, has removed approximately 93% of this debris.

Final vacuuming of the lower head after UCSA defueling is completed is expected to further reduce the L

amount of core debris in thic region.

4.4.4.4 Upper Core Support Assembly Video inspections of the UCSA indicated that there was approximately 5000 kg of core debris in this region, primarily l

behind the baffle plates.

Following removal of the baffle plates, the core debris in this area was accessible for defueling activities (e.g., flushing, vacuuming, cavijet).

Thus, the quantity of core debris in the region is being reduced by ongoing defueling activities.

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I This estirate is.a total of the fuel measured in the seven (7)'

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cubicles before decontamination and resin transfer.

Following these measurements, the cleanup filters, cleanup demineralizers, and 85% of the combined total of the A and B MU.demineralizers resin were removed.

Therefore, a reasonable estimate of the residual fuel content is 710 grams.

For bounding purposes, 800 grams is used in Table 5-2.

t 5.1.2.3 Cubicle AX021 - Reactor Coolant Bleed Tank 1A The RCBT 1A cubicle contains one of the three (3) 80,000 gallon tanks that are used as a reservoir'for reactor d

coolant.

RCBT.lA was drained and decontaminated after the THI-2 accident but has subsequently been returned to service

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as a drain tank for reactor coolant.

The residual fuel in RCBT IA is less than 1 kg (reference 5.3.1).

5.1.2.4 Cubicle AX102 - Reactor Building Sump Pump Filter Room The.RB sump pump filters (WDL-F-8A, 88), filter housings, and associated piping are located in the AX102 cubicle.

The RB sump filters were used during the THI-2 accident to filter the water from.the flooded RB basement as it was pumped to the Auxiliary Building. Post-accident sampling of the sludge in the RB basement found it contained a small quantity of fuel.

Therefore, some fuel may have been transferred from the RB basementtand deposited in AX102 during the accident as a result d f the water transfer.

Since the THI-2 accident, there has been no transfer of water from the RB to the Auxiliary Building sump via the RB sump filters.

The RB sump filters that were installed during the accident were removed during 1980 and disposed as radioactive L

waste. Subsequent to the accident, the RB sump filters have I

been used routinely to filter water transferred from the Auxiliary Building sump to the_MHHT.

During the time from 1980 to the present, there have been over 30 filter changeouts of the RB sump filters.

The residual fuel content of AX102 has not been measured because the system is still in use.

The residual' fuel content will be measured after the defueling program is completed.

l A bounding estimate of the residual fuel content of AX102 is 300 grams.

This estimate is conservative since any fuel deposited in the RB sump filters and piping as a result of the accident is believed to have been flushed into the filters and removed as part of the multiple (over 30) filter changeouts or by being flushed to the NHHT.

The major use of the RB sump b

filters during the post-accident period has been to filter 5-4 Rev. 2/0496P

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transport.

Storage of.the LCSA components outside but in e

proximity to the RV (e.g., in the "A" CFT) was deemed necessary to permit continuous progress in the RV defueling activities.

Prior to removal from the R'/, the.LCSA segements were flushed and brushed to remove fuel.

The segments were then video inspected to ensure that no visible fuel was present.

Sample sections of each plate were measured by gamma spectroscopy and/or alpha

. measurements to determine the quantity of residual fuel.

Extrapolation of fuel content in other sections was determined based on the fuel quantity of the measured sections.

For example, two (2) of the four (4) quadrants of the lower grid distributor plate were measured for fuel content and determined to contain a total residual fuel quantity of 163 grams. These measurements were extrapolated for the other two (2) quadrants and an MDL value of less than or equal to 320 grams of residual fuel was assigned to the lower grid distributor plate (Reference 5.10).

Likewise, one (1) of the 11 pieces of the flow. distributor plate was measured for fuel content (Reference 5.13).

Its residual fuel

'value (i.e., 10 grams) was deemed to be representative of the remaining segments and a total residuhl fuel quantity of 110 grams was assigned for the flow distributor plate, Based on the above approach, the "A" CFT, which contains the LCSA components, has baen assigned a total of approximately 2.4 kg (References 5.9 through 5.13) of residual fuel, distributed as follows:

Components Fuel (ko)

Lower Grid Rib Section

<0.1 Lower Grid Distributor Plate

<0.3*

Lower Grid Forging 1.7 Incore Guide Support Plate

<0.2*

Flow Distributor Plate 0.1

[

l

. TOTAL 2.4

• = MDL value The portion of the "B" core flood line between the CFT and the check valve was measured for fuel debris using both a directional L

gamma probe and a cadmium telluride' gamma spectrometer.

This measurement determined a maximum residual fuel quantity of 130 grams (Reference 2.12).

L Measurement of the residual fuel in the "B" CFT and the "A" core flood line are planned and will be provided in a subsequent DCR l

submittal.

Based on the residual fuel content in the "B" core flood line, the residual fuel quantity in these areas is not l

expected to substantially increase the current core flood system L

estimate.

There are no post-defueling plans to remove the LCSA components stored in the "A" CFT due to the relattvely small 1

quantity of residual fuel involved.

l L

l-l 5-13 Rev. 2/0496P 1

l

g spray. tater, decontamination water, condensation, and additional f

leakage-from the RCS.

The basement remained flooded for approximately two (2) years.

During this period, sediment and fuel fines settled into a sludge on the basement floor.

As discussed in Section 4.2, a significant portion of this sludge was removed during cleanup operations in the RB basement.

The sludge remaining af ter desludging operations war analyzed by sampling and gamma spectroscopy methods. Uranium concentrations measured in three (3) samples were combined with estimates of residual sediment volume to calculate the total residual fuel on-the basement floor excluding the RCDT discharge area.

A gamma scan was performed in the RCDT area since the maximum amount of fuel was initially expected to be located in the RCDT.

The total fuel contained in the remaining basement sludge following cleanup operations is estimated to be approximately 1.1 kg.

Additionally, fuel particles from washdown of defueling tools was i

transported to the RB sump.

Reference 5.18 provides an initial I

estimate that 0.2 kg of fuel could have been added to the basement inventory from this activity.

Thus, the total fuel in the RB basement is currently estimated to be 1.3 kg.

5.2.10 Miscellaneous Systems and Equipment In addition to the residual fuel quantities reported in Sections 5.2.1 through 5.2.9, residual fuel is expected to be contained in various systems / equipment located in the RB which were utilized during the defueling effort.

included are the DHCS, the Defueling Tool Rack which-contains the various long-handled tools used to defuel the RV,.the TRVFS, and the RB drain system.

Residual fuel contained in these operating cleanup systems / equipment is expected to amount to a very small fraction of the SFML and will pose no criticality concern.

For example, the NRC approved DHCS Technical Evaluation Report (TER) (Reference 5.19) states that the DWCS has been designed to prevent a poss'.ble critical configuration of fuel.

Further, the DHCS will be internally flushed and partially disassembled prior to being decommissioned.

This action will remove'a portion of the internal deposits.of residual fuel contained in'the DWCS. Additionally, as discussed in Section 5.2.9, defueling tools are generally flushed prior to removal from the RV in order to remove any loose residual fuel.

The estimate of residual fuel in these cleanup systems will be provided in a subsequent DCR submittal.

5.2.11 Criticality Assessment Table 5-3 lists the total quantity of residual fuel in the RB excluslve of the RCS and RV.

This table will be updated following the completion of remaining fuel measurements. As indicated, the total fuel mass remaining in the RB is well below the SFML of 140 kg presented in Appendix B.

Subcriticality is further enhanced since most of the residual fuel is tightly adhered to RV l

l 5-16 Rev. 2/0496P l

~

ff ~

+

. 7, 5.3 : React'or. Coolant System As described in Section 2.0, during the accident fuel was transported to the RCS as a result of the core degradation event and operation of the RCPs. Section 2.2 reported that approximately 230 kg of fuel was transported throughout the.RCS during the accident. 'Section 4.3 describes the defueling operations performed on these RCS components.

The following sections provide the current estimate of residual fuel in i

the RCS: excluding the RB (i.e., Section 5.2) and RV (i.e., Section 5.4).

i These estimates are based on fuel measurements and extensive evaluations of RCS components.

The residual fuel measurements.in the RCS hot legs, RCS cold' legs, RCPs..and core flood lines cannot be determined at this time as' measurement activities would interrupt ^ ongoing defueling operations in the RV and further defueling activities could affect the final. quantity of residual fuel in regions of interest. The residual fuel quantity for these areas will be provided at the end of defueling.

J The basis for each approach is provided within each section.

5.3.1-Pressurizer (Reference 5.20)

Following the completion of Pressurizer defueling operations in June 1988 (see Section 4.3.3.1), a small amount of core debris, consisting of small particles, remained in the Pressurizer. A video examination of the debris at the bottom of the Pressurizer was used to. determine the volume of core debris.

A 100 gram sample was removed from the Pressurizer in March 1988.

Neutron interrogation and gamma' spectrometry were used to analyze the sample. The neutron counts were compared to a natural uranium standard and the gamma counts were compared.to standard Ce-144 and Eu-154 sources.

From these comparisons, the uranium content of the sample was calculated. Scaling from the sample to the total quantity of residual debris in the Pressurizer yielded the total fuel in the Pressurizer.

From this analysis, it has been calculated that 0.3 kg of fuel remains in the pressurizer, As described in Section 4.3.3.2, debris in the Pressurizer spray line was flushed back into the Pressurizer.and was subsequently e

removed during defueling operations.

Therefore, there is no i

measurable quantity of residual fuel in the Pressurizer spray line.

5.3.2 Decay Heat Drop Line-(Reference 5.21)

A video inspection and gross gamma measurement of the decay heat drop line was performed after defueling of the decay heat drop line was completed in January 1989.

This video inspection and gamma probing data indicated that the radletion levels measured in the horizontal portion of the decay heat drop line corresponded to i

small amounts of debris on the bottom internal surface of the line.

t 5-20 Rev. 2/0496P

%,'~

~

1 p9 bi

-A' sample,of th'e decay htat line debris was analyzed by gamma spectrometry to determine the radionuclide distribution.

The line

(*-

c was then modeled with a-shielding code using the sample F

information as the source.

By matching the model to the measured gamma exposures, it is calculated that 1 5 kg of fuel remains in 1

the decay heat drop line.

5.3.3 Once_Through' Steam. Generators L

5.3.3.lT Tubesheets/ Upper Heads (References 4.22 and 5.22)

The estimate of fuel remaining on the "B" upper tubesheet was generated based on copper foil activation measurements performed in.lanuary 1989 (Reference 4.22).

Four (4) copper foils were placed inside the "B" OTSG above the tubesheet.

They were activated by exposure to the fuel in this environment and were measured with a coincidence counting e

sys tem.'.In addition, foils were positioned to measure o

background at the counting station and inside the "A" CiTSG upper head' The'"A" OTSG and background foils were activated to the same level, indicating an undetectable quantity of fuel on the "A" tubesheet using'this method. Using the background and'"A"tubesheet measurements as calibration data, the "B"

[

tubesheet was calculated to~contain 36 kg of fuel.

s The "A" OTSG upper tubesheet had less than one (1) liter of h,

debris on it following the accident.

Following defueling, the quantity of fuel on the tube sheet was so low as to be undetectable via copper foil activation coupons. An estimate of record of the residual fuel quantity in the "A" 0TSG upper tubesheet has not yet been performed.

However, based on an analysis of debris samples, it can be reasonably estimated that the residual fuel quantity in the "A" OTSG upper tubesheet is less than 1 kg (Reference 5.22).

5.3.3.2 Tube Bundles (Reference 5.23) l Fuel in the OTSG tube' bundles was measured using a gross gamma

, probing technique.

Preliminary shielding code work showed that the gamma detectors proposed for these measurements could 1

detect a tube plugged with fuel to a radius of B inches.

By probing a grid of 52 tubes, the whole OTSG tube bundle could be measured. The data was collected at 1-foot increments down the length of the 52 chosen tubes in each OTSG.

l Analysis of the probing data indicates that there are no l-significant radiation sources within the tube region that are attributable to large fuel blockages.

High radiation fields within the upper 6 feet of the "B" tube bundle are attributed to the upper tubesheet debris. Additionally, high dose rates were also associated with the water / air interface l;

l approximately half-way down the tube, possibly corresponding to a " bathtub ring" of boron and crud. Dose rates for all r

other areas were relatively uniform within the calculated L

deviation.

1' E

5-21 Rev. 2/0496P l~

L

{

4. '

9

? y

-Based on the modeled steam generator fuel debris and i

m %.

corresponding' dose' rates, the residual fuel in'the "A" and "B" "t

K

'OTSG tube bundles were calculated to be 1.7 kg and 9.1 kg, respectively.

m

+v t

y; 5.3.3.3-Lower Heads /J-Legs (Reference 5.24) q n

The "A"'and "B" OTSG lower head and J-leg fuel measurements were performed using GM probe fuel measurement strings containing copper coupons which were inserted through the generator tube bundle to the lower head and associated J-legs. Miniature lights and videoprobes were also inserted through surrounding tubes and used to verify placement of the strings and location of debris. Fuel estimates based on r

in-situ exposure rates for the "A" and "B" lower heads and J-legs are 1.0 kg and 6.3 kg, respectively.

The copper coupons were removed from the strings and transferred to the DOE for subsequent independent fuel measurement assessment.

The DOE estimated residual fuel quantities in the "A" and "B" OTSG lower head and J-legs of l

~

5.2 kg and 5.4 kg, respectively (Reference 5.25).

For purposes of the DCR, the GPU Nuclear fuel estimates are reported. The GPU Nuclear fuel estimates are believed to be more representatives of the residual fuel in the "A" and "B" lower head and J-legs based on the location of the GM counters and the sensitivity of the fuel measurements (i.e., Reference 5.25-states that the DOE fuel measurements have an uncertainty by a factor of two).

5.3.4 Criticality Assessment Table 5-4 lists the total quantity of residual fuel in the RCS exclusive of the RB and RV.

This table will be updated following the completion of remaining fuel measurements; however, it is antic! pated that the totP.1 quantity of fuel in the RCS will be below the SFML of 140 kg. Subcriticality is further enhanced l

since most of the residual fuel is tightly adhered to RCS components.

Fuel in this configuration is significantly less reactive than in the optimum conditions assumed in Appendix B l

(i.e., fuel pellets, optimum moderation with unborated water, and I

spherical geometry). Additionally, the current configuration prevents any significant debris transport, thus minimizing any interactive effects of the various fuel accumulations.

The potential for fuel transport and interaction with the-RB components and RV will be described in the final DCR submittal, t

5.L 5 Summary Table 2-1 indicates that approximately 230 kg of fuel was deposited in RCS components as a result of the accident.

Subsequently, extensive defueling operations were performed in the r

RCS as described in Section 4.3.

RCS defueling operations were 1

l i

5-22 Rev. 2/0496P

,~

a 4

l i

, /6 performed in the Pressurizer, the Pressurizer spray line. the "A"

- and "B" OTSG Upper tube sheet, the RCS hot legs, and the decay J,

heat drop line. As a result of these defueling operations, the 5

jfk, residual fuel in RCS components has been reduced significantly and i

?~

does not pose a criticality concern. Currently, the largest measured quantity of residual fuel in the RCS is in the "B" OTSG i

upper tubesheet (i.e., 36 kg). A variety of defueling techniques have been used on the tubesheets (e.g., pick and place, vacuuming,.

scraping).

It has been determined that the remaining fuel-in the "B" OTSG upper tubesheet is a crust of tightly adherent debris not readily removable by available dynami: defueling techniques, j

Further assessment of the residual fuel in the "B" OTSG upper l

tubesheets is'provided in Section 6.0.

The total estimated quantity of fuel in those portions of the RCS, listed in Table 5-4, is less than the SFML.

There is no potential for transport of fuel within the RCS which could result in a

)

critical mass.

Thus, subcriticality is assured.

GPU Nuclear has 1

concluded that no further efforts to remove fuel from these portions of the RCS cre appropriate or necessary to preclude criticality or otherwise demonstrate that defueling has been completed to the extent reasonably achievable.

However, the RCS l

will be drained as part of post-defueling activities and this activity may result in the removal of additional small quantities of fuel.

l d

1 l

1 l

i L

l

(

1 5-23 Rev. 2/0496P

y Q

W

,r

(

j 0

TABLE 5.

'* ^

AFHB CUBICLES WHICH CONTAIN NO RESIDUAL FUEL l

DESIGNATION NAME EXPLANATION AX001 RB Emergency Pumps No fuel transport pathway

.AX002 Access Corridor No waste piping in area 1

AX003 Access Area No waste piping in area j

AX013.

Evaporator Condensate All pathways isolated prior to and Tank Pumps following the accident l

L AX022 North Stairwell No waste piping in area AX023 Elevator Shaft No waste piping in area AX027

~ South Stairwell No waste piping in area AX101' Radwaste Disposal Panel No waste piping in area AX103 MCC 2-11 EB No waste piping in area AX104 MCC 2-21 EB No waste piping in area AX105 Substation 2-11E No waste piping in area AX106-Substation 2-21E No waste piping in area AX107' MCC 2-11 EA No waste piping in area l

AX108 MCC 2-21 EA No waste piping in area AX109 Nuclear Service Coolers All pathways isolated since accident and Pump AX110 Intermediate Coolers All pathways isolated since accident AX111 Intermediate Cooling Pump All pathways isolated since accident AX113 Haste Gas Analyzer System design prevents fuel transport C

AX118 Spent Fuel Coolers All pathways isolated since accident AX120 Spent Fuel Filters All pathwgys isolated since accident AX121 Elevator Shaft No waste piping in area AX122 North Stairwell No waste piping in area AX123 Access Area No waste piping in area AX125 Haste Gas Decay TK-1B System design prevents fuel transport l

AX126 Haste Gas Filter Room System design prevents fuel transport I

'AX127 Waste Gas Decay TK-1A System design prevents fuel transport j

AX128 Valve and Instrument Room System design prevents fuel transport AX132 Corridor Between UI & U2 All pathways isolated since accident AX133 South Stairwell No waste piping in area 1'

5-24 Rev. 2/0496P 1

l

p, i

L t

TABLE 5-1 (Cont'd)

I

• AFHB CUBICLES WHICH CONTAIN NO RESIDUAL FUEL DESIGNATION NAME EXPLANATION AX135' Radwaste Disposal Control No waste piping in area Panel t

AX201.

North Stairwell No waste piping in area AX202 Elevator Shaft No waste piping in area AX203 4160 Switchgear 2-1E No waste piping in area AX204 4160 Switchgear 2-2E No waste piping in area AX205 RB Purge Air Supply System design prevents fuel transport AX206 RB Purge Exhaust - B System design prevents fuel transport AX207 RB Purge Exhaust - A System design prevents fuel transport AX208 AB Exhaust Unit B System design prevents fuel transport AX209 AB Exhaust Unit A System design prevents fuel transport AX210 FHB Exhaust Unit B System design prevents fuel transport AX211 FHB Exhaust Unit A System design prevents fuel transport AX212 Decay Heat Surge Tank No fuel transport pathway AX213

. Unit Substation No waste piping in area AX214 Decon Facility No fuel transport pathway

'AX215 FHB Supply Unit System design prevents fuel transport l

AX216 AB Supply Unit System design prevents fuel transport i

AX217 Access Area No waste piping in area AX219 Instrument Racks System design prevents fuel transport L

AX220 Caustic Mixing Area All pathways isolated since accident l

AX221 Caustic Mixing Area All pathways isolated since accident L

AX222 South Stairwell No waste piping in area L,

AX223 Air Handling Units System design prevents fue) transport AX301 Elevator Shaft No waste piping in area AX302 North Stairwell No waste piping in area AX303 Elevator and Stairwell No waste piping in area Access AX401 Roof No waste piping in area AX402 Cooling Hater Storage Tanks No fuel transport pathway AX403 Damper Room System design prevents fuel transport FH002 Access Corridor No waste piping in area 5-25 Rev. 2/0496P

p

,; 3 j V

L+

6:

TABLE 5-1 (Cont'd)

~*

AFHB CUBICLES WHICH CONTAIN NO RESIDUAL FUEL 4

DESIGNATION NAME EXPLANATION FH004 West Valve Room All pathways isolated since accident "p

FH005 Mini Decay Heat Sevice All pathways isolated since accident M

Coolers p

l 1,

FH006 Decay Heat Service Coolers All pathways isolated since accident 8

FH007 Neutral 12er and Reclaimed All pathways isolated since accident Boric Acid FH010 Reclaimed Boric Acid Tank All pathways isolated since accident FH011 Reclaimed Boric Acid Pump All pathways isolated since accident FH013 Oil Drum Storage No waste piping in area FH102 East Corridor No waste piping in area FH103 Sample Room System flushtid periodically no deposits FH104 Hest Corridor No waste piping in area FH105

'Model Room A No watte piping in area FH107 Trash Compactor No watte piping in area FH108 Truck Bay No waste piping in area FHill fuel Cask Storage See Section 5.1.2.9 FH201 East Corridor No waste piping in area FH202 West Corridor No waste piping in area FH203 Surge Tank Area All pathways isolated since accident FH204 Standby Pressure Control System design prevents fuel transport i

Area FH302 SDS Operating Area See Section 5.1.2.9 FH303 Upper SPC Area System design prevents fuel transport l

FH305 Spent Fuel Pool Access System design prevents fuel transport l

l l

l 5-26 Rev. 2/0496P

e TABLE 5-2

-AFHB CUBICLES V.dICH POTENTIALLY CONTAIN RESIDUAL FU Q(I) (2)

J FUEL QUANTITY (ka)

DESIGNATION NAME REFERENCE O.13 AX004 Seal Injection Valve Room Section 5.1.2.1 l

<0.003**

AX005 Makeup Pump - IC TB SNM 89-03*

0.066-AXOO6 Makeup Pump - 1B TB SNM 87-02

<0.062**

AX007 Makeup. Pump - 1A Eng. Calculation 4550-3211-87-027 AX008 Spent Resin Storage TK-1B Section 5.1.2.2 AX009 Spent Resin Storage TK-1A Section 5.1.2.2 AX010 Spent Resin Storage Tank Pump Section 5.1.2.2 AX014 Reactor Coolant Evaporator Section 5.1.2.2 AX015a Cleanup filters Section 5.1.2.2 AX015b Cleanup Filters Section 5.1.2.2 0.8 AX016 Cleanup Demineralizer - 2A Section 5.1.2.2 AX017 Cleanup Demineralizer - 2B Section 5.1.2.2 AX114 MU&P Demin - 1A Section 5.1.2.2 AX115 MU&P Demin - IB Section 5.1.2.2 AX119 Spent Fuel Demineralizer Section 5.1.2.2 AX129 Deborating Demineralizer - IB Section 5.1.2.2 AX130 Deborating Demineralizer - 1A-Section 5.1.2.2 FH001 MU Suction Valves Section 5.1.2.2

<0.002**

AXO11 AB Sump Pump and Valve TB 86-28

<0.300**

AX012 AB Sump Pumps and Tank TB 86-28 (0.01**

AX018 Waste Transfer Pump TB 86-38

- TB refers to a THI-2 Technical Bulletin

- Denotes Minimum Detectable Level (1) - Based on current available data, this table will be updated as necessary as further data become available.

Wherever uncertainties exist as to the quantity of fuel, the upper bound estimate is used.

(2) - The predominant form of residual fuel identified in the AFHB is finely divided, small particle size, sediment material with minor amounts of fuel found as adherent films on metal oxide surfaces.

5-27 Rev. 2/0496P A-1

p:e TABLE 5-2 (Cont'd)

AFHB CUBICLES WHICH POTENTIALLY CONTAIN RESIDUAL FUEL _(1) (2)

FUEL QUANTITY (kg)

DESIGNATION NAME REFERENCE I

<0.005**

AX019 HDL Valves TB'86-38 14 AX020 RCBTs 18 and IC TB 87-12

<1 AX021 RCBT 1A Section 5.1.2.3 0.005 AX024 AB Sump Filters TB SNM 89-02

.l

<0.002**

AX026 Seal Injection Filters TB SNM 87-04 0.300 AX102 RB Sump Pump filters Section 5.1.2.4 0.292 AX112 Seal Return Coolers TB SNM 88-03 0.309 AX116 Makeup Tank Eng. Calculation 4550-3211-87-038 0.04 AX117 HU&P Filters TB 86-38 1

AX131 Miscellaneous Haste Holdup Tank Section 5.1.2.5 AX134 Miscellaneous Haste Tank Pumps Section 5.1.2.5 0.5 AX124 Concentrated Liquid Haste Pump Section 5.1.2.5 AX218 CHSTs Section 5.1.2.5 0.002 AX501 RB' Spray Pump - lA TB 86-47 0.002 AX502 RB Spray Pump - IB TB 86-47 0.002 AX503 DHR Cooler & Pump - 1A TB 86-47 0.002 AX504 DHR Cooler & Pump - IB TB 86-47 d

<0.008**

FH003a MU Discharge Valves TB 86-38 (0.060**

FH003b MU Discharge Valves TB 86-38 i

- TB refers to a TMI-2 Technical Bulletin

- Denotes Minimum Detectable Level (MDL)

(1) - Based on current available data, this table will be updated as necessary as

]

further data become available.

Wherever uncertainties exist is to the quantity of fuel, the upper bound estimate is used.

1 4

(2) - The predominant form of residual fuel identified in the AFHB is finely divided, small particle size, sediment material with minor amounts of fuel found as adherent films on metal oxide surfaces.

I 5-28 Rev. 2/0496P f

i l

I 3

TABLE 5-2 (Cont'd)

AFHS CUBICLES WHICH POTENTIALLY CONTAIN PESIDUAL FUEL (l) (2)

(UEL OUANTITY (ko)

DESIGNATION NAME REFERENCE l

1 FH008 Neutra11rer Tank Pump Section 5.1.2.5 l

FH009 Neutralizer Tank Section 5.1.2.5 FH012 Neutralizer Tank Filers Section 5.1.2.5

<1 FH014 Annulus Section 5.1.2.6 FH112 Annulus Section 5.1.2.6 TH205 Annulus Section 5.1.2.6 l

0.71 FH101 HU&P Valve Room TB 86-38/86-21 i

L 1

FH106 SDS Monitor Tanks Section 5.1.2.7 FH110 Spent Fuel Pool "B" (3)

FH109 Spent fuel Pool "A" Section 5.1.2.8

<23 kg - TOTAL F

(1) - Based on current available data, this table will be updated as necessary as further data become available. Wherever uncertainties exist as to the quantity of fuel, the upper bound estimate is used.

(2) - The predominant form of residual fuel identified in the AFHB is finely divided, small particle size, sediment material with minor amounts of fuel found as adherent flims on metal oxide surfaces.

(3) - No Value Assigned; See Section 5.1.2.8 for d1 tails.

5-29 Rev. 2/0496P

TABLE 5-3 6-RESIDUAL FUEL QUANTIFICATION IN THE REACTOR BUILDING (a)

RESIDUAL FUEL COMPONENT O'JANTITY (KG)

RV Head 1.4(b) i RV Plenum 2.1 L

Fuel Transfer Canal (b)

Core Flood System 2.4(b)

D-Rings 23 Upper Endfittings

< 54(b)

Reactor Coolant Drain Tank 0.1 4 (C)

Letdown Coolers RB Basement / Sump 1.3 Cleanup Systems / Equipment (b)

(e.o. DWCS. TRVFS. RB Drains)

TOTAL

< 88 kg(b) l (a) - Excluding the RV and RCS.

1 (b) - To Be Updated in a Subsequent DCR Submittal.

1..

l (c) - HDL 1'

l l

5-30 Rev. 2/0496P l

TABLE 5-4

5 RESIDUAL FUEL OVANTIFICATION IN THE RCS(a)

RESIDUAL FUEL COMPONENT OUANTITY (kg)

"A" Side Hot Leg (b)

OTSG Upper Tubesheet

<1.0 Tube Bundle 147 Lower Head and J-Legs 1.0 Reactor Coolant Pumps (b)

Cold Legs (b)

Core Flood Line(c)

(b)

"B" Side Hot Leg (b)

OTSG Upper Tubesheet 36 Tube Bundle 9.1 Lower Head and J-Legs 6.3 Reactor Coolant Pumps (b)

Cold Legs (b)

Core Flood Lines (C)

(b)

Pressurizer 0.3 Decay Heat Drop Line 1.5 TOTAL -

<57(o)

(a) - Excluding the RV.

(b) - To Be Provided (c) - Between the RV and First Check Valve (d) - To Be Updated 5-31 Rev. 2/0496P

7 FIGURE 5-1 AUXILIARY BUILDING 280'-6" ELEVATION sb

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'n FIGURE 5-3

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r 4.26 GPU Nuclear letter 4410-86-L-0049, Defueling Safety Evaluaticn s.

Report, Revid on 10 dated May 15, 1986.

4.27 Technical voiletin 86-40 Swiss Cheese Drilling of Core, Revision 0, GPU Nuclear, Middletown, PA, September 1986.

5.1 Technical Bulletin SNM 87-04, " Seal Injection Filter (MUF-4A/B)

Room /AXO26 SNM Accountability Summary," Revision 1, October 22, 1987.

5.2 Technical Bulletin 86-38, " Summary of Fuel Quantities External to the Reactor Vessel," Revision 3 January 24, 1989.

5.3 Technical Bulletin 86-48, " Cleanup filters (HDL-F6A & B and WDL-F9A & B) Fuel Quantification," Revision 0, dated December 8, 1986.

5.3.1 GPU Nuclear Calculation 4800-3232.89-077, " Reactor Coolant Bleed Holdup Tank 1A (WDL-T-1A)," Revision 0, August 21, 1989.

5.4 Technical Bulletin 86-28, " Aux 111ary Building Sump. Sump Tank and Valve Gallery Reactor Fuel Quantification," Revision O.

May 16, 1986.

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5.5 GPU Nuclear Caiculation No. 4550-3211-87-038, " Makeup Tank Room (AX116) SNM Accountability Calculation," Revision 1 February 1, 1988.

5.6 Technical Evaluation for Defuelino Canisters, 3527-016, Revision 4. September 22, 1987.

5.7 TMI-2 Post-Defueling Survey Report for the Reactor Vessel Head (Draft).

5.8 TMI-2 Post-Defueling Survey Report for the Reactor Vessel Plenum.

5.9 Technical Bulletin SNM 88-06, " Lower Core Support Assembly Grid Rib Section," Revision 0, October 12, 1988.

5.10 Technical Bulletin 88-16 " Removal of Lower Grid Distributor i

Plate," Revision 0, July 21, 1988.

5.11 Technical Bulletin SNM 89-01, " Lower Core Support Assembly Lower 7

Grid Forging SNM Accountability Summary," Revision 0, t

January 25, 1989.

5.12 Technical Bulletin SNM 89-04, "Incore Guide Support Plate SNM Accountability Summary," Revision 0, January 30, 1989.

5.13 Technical Bulletin 89-07, " Video Inspections of Incore l

Instrument Guide Tubes," Revision 0, May 23, 1989.

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'6 5.14 CPU Nuclear letter 4410-86-L-0132, "St; rage of Upper Endfittings," dated August 16, 1986.

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5.15 GPU Nuclear letter 4410-86-L-0160, "End Fitting Storage," dated September 9, 1986.

5.16 GPU Nuclear letter 4410-89-L-0041, " Program for Surveying the Endfttting Storage Containers," dated May 10, 1989.

5.17 TMI-2 Post-Defueling Survey Report for the letdown Coolers (Revision 1).

5.18' TMI-2 Post-Defueling Survey Report for the Reactor Building Basement PDSR Reactor Fuel.

5.19 Technical Evaluation Report for Defueling Water Cleanup System, 3525-015 Revision 12, dated May 12, 1989.

5.20 TMI-2 Post-Defueling Survey Report for the Pressurizer.

5.21 GPU Nuclear Calculation 4800-3212-89-010. " Decay Heat Line fuel Estimate," Revision 0, dated May 16, 1989.

5.22 Technical Bulletin 86-24. "0TSG-A Upper Tube Sheet Debrit Samples," Revision 0, April 25, 1986.

5.23 GPU Nuclear Calculation 4800-3224-89-006, "0TSG Tube Bundle Fuel Estimates," Revision 0, dated May 3, 1989.

5.24 Technical Bulletin 88-19, "A and B OTSG Lower Head and J-Leg Fuel Estimate," Revision 0, dated October 26, 1988.

5.25

" Neutron Measurement of the Fuel Remaining in the THI-2 OTSGs,"

Pacific Northwest Laboratory Operated for hte U.S. Department of Energy by Battelle Memorial Institute, PNL-6807, January 1989.

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