ML20042D754
| ML20042D754 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/31/1989 |
| From: | Roche M GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 4410-89-L-0021, 4410-89-L-21, NUDOCS 9004050329 | |
| Download: ML20042D754 (10) | |
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OPU Nucloor Corporation i
Nuclear
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Middletown, Pennsylvania 17057 0191 717 944 7621 TELEX 84 2386 Writer's Direct Dial Number:
(717) 948-8461 March 29, 1990 4410-89-L-0021/0069P I
i Document Control Desk US Nuclear Regulatory Commission Washington, DC 20555
Dear Sirs:
i Three Mile Island Nuclear Station, Unit 2 (TMI-2) l Operating License No. DPR-73 Docket No. 50-320 10 CFR 50.59 Report for 1989 In accordance with the reouirements of 10 CFR 50.59, "Cnanges, Tests, and Experiments," forwarded is a oescription of changes to facility systems and procedures oescribed in the TMI-2 Final Safety Analysis Report (FSAR) which were accomplished during 1989. Also included is a summary of tests and i
experiments performeo that are not described in the FSAR.
Sircerely,
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EDS/mkk Attachments cc:
W. T. Russell - Regional Administrator, Region I J. F. Stolz - Director, Plant Directorate I-4 l
L. H. Thonus - Project Manager, TMI Site F. I. Young - Senior Resident Inspector
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ATTACHMENT 1 4410-90-L-0021 TMI-2 RECOVERY ACTIVITIES During 1989 a number of plant recovery activities were performed. Many of i
these activities combined modifications, procedural changes, and tests or experiments.
All of these activities were subject to numerous GPU Nuclear reviews and approvals.
In addition, certain activities were subject to NRC review and approval prior to implementation.
Changes to previously approved activities are submitted to the NRC for information under the yearly update tions I
program for Technical Evaluation Reports and System Descrip'as nee. Ugdatesto NRC-approved Safety Evaluation Reports are submitted on a ded ba si s.
Since the documentation for the activities listed below was submitted to the NRC previously, the activities will not be discussed further in this report.
i o EPICOR II Changes to this program are covered by the annual update program for System Descriptions and Technical Evaluation Reports.
Update submitted via GPU Nuclear letter 4410-89-L-0060 dated July 24,1989, o Interim Solid Waste Changes to this program are covered by Staging Facility the annual update program for System Desoriptions and Technical Evaluation Reports.
Update submitted via GPU Nuclear letter 4410-89-L-0057 dated May 17,1989.
o Processed Water Disposal Changes to this program are covered by System the annual update program for System Descriptions and Technical Evaluation Reports.
Update submitted via GPU Nuclear letters 4410-89-L-0038 dated April 17,1989, 4410-89-L-0067 dated June 7,1989, 4410-89-L-0103 dated October 2,1989, and 4410-89-L-0099 dated October 19, 1989, o Reactor Building Sump Changes to this program are covered by Recirculation System the annual update program for System Descriptions and Technical Evaluation Reports.
Update submittei via GPU Nuclear letter 4410-89-L-0064 dated June 7,1989, o Solid Waste Staging Facility Changes to this program are covered by the annual update program for System Descriptions and Technical Evaluation Reports.
Update submitted via GPU Nuclear letters 4410-89-L-0084 dated August 8,1989 and 4410-89-L-0094 dated August 21, 1989. -
ATTACHMENT 1 4410-90-L 002) o $vbmer9ed Demineralizer System Changes to this program are covered by the annual update program for System Descriptions and Technical Evaluation Reports.
Update submitted via GPU Nuclear letter 4410-89-L-0090 dated September 22, 1989, o fuel Canister Storage Racts Changes to this program are covered by the annual update program for System Descriptions and Technical Evaluation Reports.
Update subn'itted via GPU Wuclear letter 4410-89-L-0058 dated May 24, 1989, o Defueling Canisters Changes to this program are covered by the annual update program for System Descriptions and Technical Evaluation Reports.
Update submitted via CPU Nuclear letter 4410-89-L-0100 dated September 27, 1989, o Defueling Canister Dewatering Changes to this program are covered b,'
System the annuel update program for System Descriptions and Technical Evaluation Reports.
Update submitted via GPU Nuclear letter 4410-89-L-0080 dated l
July 19,1989.
o Sediment Transfer and Changes to this program are covered by Frocessing Operations the annual update program for System Descriptions and Technical Evaluation Reports.
Update submitted via GPU Nuclear letter 4410-88-L-0183 dated January 10, 1989, o Waste Handling and Packaging Changes to this program are covered by Facility the annual update program for System l
Descriptions and Technical Evaluation Reports and other docketed correspondence.
Update submitted via GPU Nuclear letter 4410-89-L-0035 dated April 17,1989, o Defueling Water Cleanup Changes to this program are covered by System the annual update program for System Descriptions and Technical Evaluation Reports and other docketed correspondence.
Update submitted via GPU Nuclear letter 4410-89-L-0040 dated May 12,1989.
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ATTACHMENT 1 4410-90-L-0021 o Core Region Defueling Changes to this program are covered by (Use of Polar Crane updates to the Safety Evaluation Report Q
Auxiliary Hook) and other docketed corrtspondence.
Updated information submitted via GPU Nuclear letters 4410-89 L-0044 dated May 3,1989, and 4410-89-L 0110 dated October 26, 1989, o Lower Core Support Assembly This program is described in GPU Defueling Nuclear letters 4410-87-L 0014 dated February 6,1987; 4410-87 L-0138 dated November 10,1987; 4410-87-L-0139 dated November 30,1987; 4410-87-L-0160 dated December 3,1987; 4410 87-L-0189 dated December 28,1987; 4410-87-L-0192 dated December 31,1987; 4410-88-L-0005 dated January 18,1988; 4410-88-L-0026 dated February 26,1988; 4410-88-L-0044 dated March 16,1988; 4410-88-L-0050 dated March 25,1988; and approved by NRC Letter dated April 1,1988.
Further infomation was submitted via GPU Nuclear letters 4410-88-L-0067 dated April 19,1988; 4410-88-t-0006 dated February 8,1989; 4410-89-L-0085 dated August 18,1989; and 4410-89-L-0107 dated October 20, 1989, o Upper Core Support This program is described in GPU Nuclear Assembly Defueling letter 4410-88-L-0138 dated September 9,1988; and was approved by NRC letter dated April 4,1989, o Criticality Safety Evaluation This analysis was initially submitted for Increasing the TMI-2 Safe via GPU Nuclear letter 4410-89-L-0013 Fuel Mass Limit dated February 13, 1989.
o End Fitting Storage Primary infomaticn forwarded via GPU Nuclear letters 4410-86-L-0132 dated August 16,1986 and 4410-86-L-0160 dated September 9,1986; and was approved by NRC letter dated September 10, 1986.
Additional information submitted via GPU Nuclear letter 4410-89-L-0041 dated May 10, 1989.
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ATTACHMENT 2 4410 90-L-0021 PROCEDURE CHANGES During the course of 1989, the requirement for procedures changed as defueling and fuel shipping efforts progressed. A number of procedure changes were made and new procedures were issued.
Many of these procedures performed activities in accordance with NRC approved Safety Evaluation Reports; thus, they will not be discussed further in this report.
Procedures whose scope of activity was completed during 1987 were cancelled; procedures previously detemined to have review significance underwent SRG review to detemine the potential impact on safety prior to cancell.ation.
The remainder of the procedures changes were reviewed and it was detemined that there were no changes which specifically constituted a FSAR change as defined by 10 CFR 50.59.
However, there were a number of changes made to FSAR-type procedures.
These changes were made to reflect changing plant conditions or to implement the reconenendations of various activity-related analyses.
Typical system-oriented procedures receiving these types of changes are:
o Recovery Quality Classification List o
Polar Crane Operation o
Canister Positioning System Operation o
Maintenance of Reactor Building (RB) Water Volume for "B" D-Ring Lif ting and Handling o
Liquid Waste Disposal System Operations o
EPICOR !! Operation o
Core Bore Machine Operation o
Primary Plant Operations o
Baffle Plate Cutting and Removal Operations o
Transfer of Defueling Tools in the Reactor Building o
In-vessel Filtration System Operations o
Solid Waste Staging Facility Sump Pump Operations o
Defueling Water Cleanup System Operations o
Withdrawing Incore Instrument Strings with the Polar Crane o
Radiation System Monitoring Setpoints o
Leakage Containment System Sampling j
o NUPAC 125-B Rail Cask Maintenance o
Spent Resin Transfer Operations l
o Hydrogen Peroxide Addition to the Reactor Yessel (RV) and the Spent Fuel i
Pool o
Instrument Air System o
Domestic Water System o
Fire Protection System o
Automated Cutting Equipment System (ACES) Operation o
Long-handled Tool Operation o
Power Supply System o
Shif t. Daily and Weekly Surveillance Checks o
Containment Integrity Verification o
Reactor Coolant System Chemistry Yerification o
Submerged Demineralizer System Operation o
Response to Criticality Monitors Alarms o
Periodic Inspection of the Polar Crane These changes to accomodate current plant conditions were detemined not to constitute an Unreviewed Safety Question, i
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ATTACHMENT 3 4410 90-L-0021 TESTS AND EXPERIMENTS A number of tests and experiments were perfomed during the year.
The majority of these activities were covered by SERs provided for major recovery l
activities, as discussed previously in this report.
The remainder of the 6
tests or experiments were evaluated to detemine if they constituted an Unreviewed Safety Question or a significant risk to the health and safety of the public or workers.
In no case was there a determination of an Unreviewed Safety Question or significant risk.
Below is a list of tests or experiments which is representative of those performed during 1989, o
An activation copper foil and coincidence counting system used for Once-Through steam Generatgrs (OTSG) "A" and "B" Upper Tubesheets o
Characterization of the He9 Detection System and tie Neutron Interrogation System which is used to detect fast neutrons as opposed to thermal neutrons o
Defined Conditions in Regions of the Reactor Yessel o
Core debris bed inspection and probing o
Neutron measurements of the Upper Endfitting Storage Containers o
Gamma Scan of an Incore Instrument Guide Tube o
Sample analysis of R-6 and lower RV debris sam)1es o
SNM measurements of the reactor coolant bleed 1oldup tanks o
Gamma spectrometry measurements on two of the four sections of the incore guide support plate removed f rom the RV o
Video-Inspections of the Following:
Ex-Yessel Cold Legs Hot legs Decay Heat Drop Line
B" OTSG In-Vessel Defueling Operations l
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T ATTACHMENT 7 4410-90-L-0021 FACILITY MODIFICATIONS Activities included in this section were perfomed without prior approval of the NRC staff under the authority of 10 CFR 50.59 The items listed below cover specific activities perfomed under the authority of Engineering Change Authorizations (ECAs) and Mini-Mods.
Er,As are tracking mechanisms for review, approval, and documentation of specific plant changes.
Mini-Mads provide a process to implement an expedited modification process for a restricted class i
of modifications to the TMI-2 plant.
ECAs and Mini-Mods selected for i
inclusion were those for which turnover was completed during the calendar year 1989.
ECA 3153-85-0172, Revision 0 - Install Monorail Under Reactor Building l
ITevation 347'-6" structural steel This ECA documents the installation of a 43-foot long monorail and its supports.
Safety Evaluation Summary This modification facilitated entry / exit of material into the RB via the equipment hatch.
This change does not constitute an Unreviewed Safety Question.
ECA 3154-86-0338, Revision 1 - AX004_ Cubicle Floor Concreting This ECA documents the addition of concrete to prepare the AX004 cubicle for PDMS.
Safety Evaluation Summary The intent of the approximately 6-inch thick concrete fill slab is to stablize radionuclides and to provide shielding to reduce dose rates within the cubicle.
This concrete addition is in keeping with the ALARA principle and does not constitute an Unreviewed Safety Question.
ECA 3251-86-0373, Revision 0 - Canal Fill and Tool Flushing Flowmeter Installation This ECA documents the installation of a flowmeter in the Fuel Transfer Canal fill piping.
Safety Evaluation Summary The installation of a flowmeter on the canal fill and tool flushing line allows for monitoring of the borated water added to the Fuel Transfer Canal.
This change does not constitute an Unreviewed Safety Question.
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ATTACHMENT 7 4410-90-L-0021 ECA 3211-87-0501, Revision 0 - Makeup and Purification Demineralizers Discharge Piping Tie-In This ECA documents a modification to the' Makeup and Purification discharge piping.
Safety Evaluation Summary This modification provides access to the internals of the discharge piping of demineralizer MU-K-1B to remove an obstruction of flow that is presumed to exist in the vicinity of the demineralizer outlet.
Since materials with suitable design pressure and temperature rating were used, this change did not constitute an Unreviewed Safety Question.
ECA 3213-88-0507, Revision 0 - Core Flood Tank "A" Head Removal and Drain Connections This ECA documents a modification to the "A" Core Flood Tank (CFT).
Safety Evaluation Summary CFT "A" was opened to store the core support assembly pieces removed from the RV.
A relief valve and vent line were removed prior to removing the tank upper head; the line was capped with a drain connection to facilitate filling and draining of the tank.
This change does not constitute an Unreviewed Safety Question.
MMA 3252-87-0026, Revision 0 - FH - Canister liandling Bridge Defeat DS Slow Zone (Hoist-up extended zone only)
This mini-mod documents the modification of the user prog' defeat" or block the ram of the fuel handling - canister handling programmable controller to extended hoist-up slow zone from being activated while the fuel handling bridge is at the Dewatering / Cask Loading Station end of the "A" Spent Fuel Pool (i.e., the North end).
Safety Evaluation Summary The original design intent was to have the hoist raise slowly when there was a loaded canister on the grapple so that installed gamma detectors in the dewatering racks could perfom a gamma scan.
However, the need for the gamma detectors has been eliminated and they have been removed from the A" Spent Fuel Pool.
Thus, the defeat of the extended hoist-up slow zone resulted in ALARA improvements by decreasing the time spent in canister handling.
This change does not constitute an Unreviewed Safety Question.
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4410-90-L-0021 l
MMA 3233-87-0053, Revisions 0 and 1 - Differential Pressure Detector for Decant Filters.
This mini-mod documents the replacement of the existing differential pressure i
(DP) instrumentation with a device having an extended range to facilitate the increase in the administrative limit on DP across the sludge processing system j
decant filters.
Safety Evaluation Sumary Replacement of the DP instrumentation and the increase of the administrative limit improved overall system perfomance by extending l
decant operations.
Since materials with suitable design pressure ratings j
were used in construction, this change did not constitute an Unreviewed Safety Question, MMA 3232-88-0105, Revisions 0 and 1 - Install Additional Valve in WDL-F-4A/B Drain Line This mini-mod documents the installation of an additional isolation valve in i
the Waste Disposal Liquid System drain line piping, s
safety Evaluation Sumary This modification was necessary to prevent recurrence of radiological spills when drain line diaphragm valves fail.
Plant safety is enhanced and an Unreviewed Safety Question does not exist.
MMA 3233-88-0111, Revision 0 - Sludge Processing System Delumper Installation i
This mini-mod documents a modification to the sludge processing system.
Safety Evaluation Sumary Modifying the sludge processing system to provide an in-line slurry conditioner (i.e., "delumper") facilitates the transfer of solids and minimizes the likelihood of flow blockages.
This change does not constitute an Unreviewed Safety Question.
MMB 3528-88-0152, Revision 0 - Relocate the Domestic Water Sup)1y to the Respirator Cleaning Facility (RCF) and the Waste Handling and Packaging Facility (WHPF)
This mini-mod documents the relocation of Domestic Water Supply piping to the RCF and the WHPF.
Safety Evaluation Sumary This modification alleviated a low supply pressure condition caused by the length of small bore piping and also o)viated the need for heat tracing af ter TMI-2 is placed in Facility Mode 4 (Post-Defueling Monitored Storage).
This change does not constitute an Unreviewed Safety Question.
Y ATTACHMENT 7 4410 90-L-0021 m 3541 89-0156, Revision 0. NS Pipe Removal for PDMS This mini-mod documents a modification to the Nuclear Services (NS) Closed Cooling Water System, Safety Evaluation Summary This modification removes a one (1) foot section of pipe from the NS supply / return headers to/from the "A" D-ring to facilitate the pumpout of residual water. This change does not constitute a Unreviewed Safety Question.
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