ML20246P168
| ML20246P168 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/31/1988 |
| From: | Roche M GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 4410-89-L-0030, 4410-89-L-30, NUDOCS 8903280144 | |
| Download: ML20246P168 (13) | |
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ATTACHMENT 1 4410-89-L-0030 TMI-2 RECOVERY ACTIVITIES During 1988 a number of plant recovery activities were performed. Many of 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 program for Technical Evaluation Reports and System Descriptions. Updates to NRC-approved Safety Evaluation Reports are submitted on a "as needed" basis.
Since the documentation for the activities listed below was submitted to the NRC previously, the activities will not be discussed further in this report.
o Auxiliary and Fuel Handling Changes to this program are provided in Building Decontamination accordance with a cuarterly update program. Updates were submitted via GPU Nuclear letters 4410-88-L-0002 dated January 15, 1988; 4410-87-L-0062 dated April 14, 1988; 4410-87-L-0113 dated July 15, 1988; and 4410-87-L-0163 dated October 12, 1988.
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 4410-88-L-0094 dated June 22, 1988.
o Interim Solid Waste Changes to this program are covered by Staging Facility the annual update program for System Descriptions and Technical Evaluation Reports.
Update submitted via GPL Nuclear letter 4410-88-L-0080 dated September 15, 1988.
o Processed Water Storage Changes to this program are covered by and Recycle System the annual update program for SW tem i
Descriptions and Technical Eva:
_on Reports.
Update submitted via J
Nuclear letter 4410-88-L-0159 dated September 27, 1988.
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 submitted via GPU Nuclear letter 4410-88-L-0073 dated May 10, 1988. g~ww any i
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ATTACHMENT 1
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4410-89-L-0030 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 letter 4410-88-L-0155 dated September 27, 1988.
o Submerged Demineralized System Changes to this program are covered by the annual update program for System Descriptions and Technical Evaluation Reports. Update submitted via GPU 1
Nuclear letter 4410-88-L-0135 dated August 25, 1988.
o Containment Air Control Envelope Changes to this program are covered by the annual update program for System Descriptions and Technical Evaluation Reports. Update submitted via GPU Nuclear letter 4410-88-L-0194 dated December 29, 1988.
o Canister Handling and Changes to this program are covered by Preparation for Shipment System updates to the Safety Evaluation Report and other docketed correspondence.
Updated information submitted via GPU Nuclear letter 4410-88-L-0042 dated March 29, 1988.
o Core Region Defueling Changes to this program are covered by (Use of Polar Crane updates to the Safety Evaluation Report Auxiliary Hook) and other docketed correspondence.
Updated information submitted via GPU Nuclear letters 4410-88-L-0008 dated January 28, 1988; and 4410-88-L-0165 dated October 20, 1988.
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 4410-88-L-0090 dated July 29, 1988.
o Fuel Canister Storage Racks Changes to this program are covered by the annual update program for System Descriptions and Technical Evaluation Reports. Update submitted via GPU Nuclear letter 4410-88-L-0072 d&ted May 10, 1988. _ _ _.
ATTACHENT 1 4410-89-L-0030 o Defueling Canisters Changes to this program are covered by the annual update program for System J
Descriptions and Technical Evaluation Reports. Updated information submitted via GPU Nuclear letter 4410-88-L-0158 j
dated September 29, 1988.
o Defueling Canister Dewatering Changes to this program are covered by_
System the annual update program for System Descriptions and Technical Evaluation-Reports. Update submitted via GPU Nuclear letter 4410-88-L-0091 dated June 28, 1988.
o Waste Handling and Packaging Changes to this program are covered by Facility the annual update program for System j
Descriptions and Technical Evaluation Reports and other docketed correspondence. Updated information submitted via GPU Nuclear letter 4410-88-L-0039 dated March 15, 1988.
o Accident Generated Water Changes to this program are covered by the annual update program for System Descriptions and: Technical Evaluation Reports.
Initially submitted via GPU Nuclear letters 4410-88-L-0012 dated February 16, 1988, and 4410-88-L-0168 dated October 17, 1988.
o Lower Head Defueling This program was described in GPU Nuclear letters 4410-88-L-0006 dated June 6, 1988; 4410-88-L-0100 dated June 27, 1988; and 4410-88-L-0137 dated September 9, 1988; and approved by NRC i
Letter dated December l', 1988.
o Lower Core Support Assembly This program was described in GPU Defueling Nuclear letters 4410-87-L-0014 dated February 6, 1987; 4410-87-L-0138 dated November 10, 1987; and 4410-88-L-0005 dated January. 18, 1988, and approved by l
NRC Letter dated April 1, 1988.
-i Further information was submitted via GPU Nuclear letters 4410-87-L-0139 i
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-0026 dated February 26,-1988; 4410-88-L-0044 dated March 16, 1988; 4410-88-L-0050 dated March 25, 1988; and 4410-88-L-0067 dated April 19, 1988. _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ - _ - _ _ - _ _ _ - _ - _ _..
m ATTACHMENT 1 4410-89-L-0030 o Criticality Safety Assessment This program was described in GPU For Use of the Plasma Arc Nuclear letters 4410-88-L-0110 dated Tourch to Cut the Upper Core August 11, 1988; and 4410-88-L-0192 Support Assembly Baffle Plates dated December 27, 1988; and was and the Core Support Shield approved by NRC Letter dated December 31, 1987.
o Upper Core Support This program is described in GPU Nuclear Assembly Defueling letter 4410-88-L-0138 dated September 9, 1988, which is currently pending NRC review and approval.
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ATTACHMENT 2 4410-89-L-0030 PROCEDURE CHANGES During the course of 1988, the requirement for procedures changed due to the defueling and fuel shipping efforts.
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. Cancelled procedures determined to have review significance underwent SRG review to determine the potential impact on safety prior to cancellation.
The remainder of the procedures changes were reviewed and it was determined 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 recommendations of various activity-related analyses. Typical system-oriented procedures receiving these types of changes are:
o Fire Protection System o
Service Air System o
Makeup Demineralized Resin Transfer Operations o
Filling Fuel Transfer Canal (FTC) with Processed Water from the Processed Water Storage Tank (PWSTs)
Transfer of FTC Water to the Reactor Building (RB) Cavity o
o Automated Cutting Eauipment System (ACES) Operation o
Transfer of Reactor Coolant Bleed Tank (RCBT) Water to the Miscellaneous Waste Holdup Tank (MWHT) or Neutralizer Tanks o
Recovery Quality Classification List o
In-Service Inspection System o
NUPAC 125-B Rail Cask Loading o
Filling of Borated Water Storage Tank (BWST) with Processed Water from the PWSTs Maintenance of RB Water Volume for "3" D-Ring Lifting and Handling o
o Shift and Daily Checks o
Control of Combustible Materials o
Operation of the Secondary Plant System o
Primary Plant Operations o
Filling Spent Fuel Pool (SFP)
"A" with Processed Water from the PWSTs o
Filling RCBTs with Processed Water from the PWSTs o
RB Canister Handling Bridge o
Response to Criticality Monitors Alarms o
Dropped EPICOR II Prefilter/ Demineralized Liner or Transporter Vehicle Accident Loss of Fluid Integrity and/or High Airborne Activity in EPICOR II o
o Rupture / Leak in a Water Gas Decay Tank o
Flood Response Procedure o
FTC Normal / Abnormal Operation o
Displacement Jet Pump System Operation o
RD Normal and Emergency Ventilation and Cooling o
Canister Transfer System. _ _ _
ATTACHMENT 2 441U 89-L-0030 o
Mechanical Draft Cooling Tower (Panel 331) o RB Entry o
Canister Positioning System Operation o
Nitrogen Supply Header Operation o
Shielded Work Platform Tool Slot Spray System Operation o
Instrument Air System o
Containment Integrity Verification o
Testing of RB Sump Recirculation System o
RB Cavity Boron Concentration Verification o
TMI-2 Core and Special Nuclear Material (SNM) Accountability Program These type of update changes to reflect current plant conditions were i
determined not to constitute an Unreviewed Safety Question.
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ATTACHMENT 3 4410-89-L-0030 l
TESTS AND EXPERIMENTS A number of tests and experiments were performed during the year. The majority of these tests were covered by SER's provided for major recovery activities, as discussed previously in this report. The remainder of the tests or experiments were evaluated to determine 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 1988.
o RB Basement Block Wall Leaching Test o
Robotic Surveys for Sediment Sampling in the Basement o
Gamma and Neutron Measurements of the RCBTs o
Gamma Measurements of the Basement Wall, Transfer Pump and Valve Rooms, Liner Wall, and RCBTs "B" and "C" for Characterization o
Robotic Defueling and Sampling of the Pressurizer with the Mini-Submarine o
Cold-Leg and Reactor Coolant Pump Inspections o
SNM Measurements of the Makeup Tank, Makeup Pump Roans, Makeup Filters, and Seal Injection Filter o
Detector String Deployment in the Once-Through-Steam-Generator (OTSG) and J-Leg Piping to Quantify Remaining SNM o
An Activation Copper Foil and Coincidence Counting System Used for J-Legs, OTSG Lower Head, and on Top of the Tube Sheets o
In-situ Measurements Used to Measure Fuel in the "B" Steam Generator Tubes and Other Hard-To-Access Areas o
Alpha Probing Technioue Used for the OTSG Tube Bundle Characterization and for Various Ligament Sections and Baffle Plates o
Remote Assay Equipment Used as a Host and Slave Computer to Monitor and Control the HPGe and GM Detection Systems in the RB A Dual-Detector HPGe Detector System Using Various Gennanium Detectors o
and Electronics to Count Both Sides of the Grid Plate and Lower Core Support Assembly (LCSA) Forging Plates o
The He4 Detection System and the Neutron Interrogation System Development Which is Used to Detect Fast Neutrons as Opposed to Thermal Neutrons o
Gamma Radiation Measurements of Baffle Plates Gamma Scans of Removed Pieces of the Lower Grid Rib Section, Lower Grid o
Distributor Plate, and Lower Grid Forging o
Defined Conditions in Regions of the Reactor Vessel o
Core Debris Bed Inspection and Probing o
Video-Inspections of the Following:
Ex-Vessel Pressurizer Decay Heat Line "A" and "B" OTSG Lower Head and J-Legs
- RB Basement In-Vessel LCSA Periphery Inspection Lower Grid Forging, Lower Grid Plate, and Lower Grid Flow Distributor Plate Removal
- Plasma Arc Cutting Inspections
- Defueling Operations _ _ _ _ _ _ _ _ _ _ _ - _ _.
ATTACHENT 4 4410-89-L-0030 FACILITY MODIFICATIONS Activities included in this section were performed without prior approval of the NRC staff under the authority of 10 CFR 50.59. The items listed below cover specific activities performed under the authority of Engineering Change Authorizations (ECAs) and Mini-Mods. ECAs are tracking mechanisms for review, approval, and documentation of specific plant changes. Mini-Mods provide a process to implement an expedited modification process for a restricted class of modifications to the THI-2 plant. ECAs and Mini-Mods selected for inclusion were those for which turnover was completed during the calendar year 1988.
ECA 3233-84-0047, Revision 0 - Modification of WDS-T-2 Drain Line This ECA documents the installation of an additional piece of piping to the Waste Disposal Solid System drain line.
Safety Evaluation Summary This modification facilitates draining operations by extending the current "end" of the drain line. This change does not constitute an Unreviewed Safety Question.
ECA 3863-84-0112, Revision 0 - Additions of Maintenance Flanges and Valves to DF-P-2 Discharges This ECA documents the installation of flanges to the discharge of the diesel air start compressors.
Safety Evaluation Summary 1
The four (4) diesel air start compressors are difficult to maintain. The 1
implementation of this ECA should increase the availability of the Diesel Air Start System and thereby render the plant more safe. Furthermore, operability of the Emergency Diesel Generators is no longer required by the TMI-2 Technical Specifications. Thus, this change does not constitute an Unreviewed Safety Question.
ECA 3524-85-0259, Revision 0 - Relocation Sampling Points for PW-T-1 and PW-T-2 This ECA documents the relocatinn of the sampling points for the Processed Water (PW) Storage Tanks Safety Evaluation Summary This modification allows sampling of the PW tanks when their level is below 8 feet. The PW System is not important to safety. This change does not constitute an Unreviewed Safety Question.
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ATTACHMENT 4 4410-89-L-0030 ECA 3231-85-0264, Revision 0 - Addition of Hold-down Brackets to WDG-R-1485 Housing This ECA documents the addition of brackets to the pump within the WDG-R-1485 assembly Safety Evaluation Summary i
During operation of the pump, vibration causes the existing housing to impinge upon the pump. Adding brackets to secure the housing will i
prevent the pump from being damaged. The reliability of the pump will be enhanced by the addition of these hold-down brackets. This change does not constitute a Unreviewed Safety Question.
ECA 3228-86-315, Revision 0 - Removal of RTO and Associated Thermowell from RC-H-18 Hot Leg This ECA documents the removal of two (2) RTDs and their cominon thermowell from the hot leg of steam generator 1B.
l Safety Evaluation Report These RTDs are not reauired to maintain the plant in a safe shutdown-condition. The thermowell will be replaced with a cap to re-establish the Reactor Coolant System Pressure Boundary. This change does not constitute an Unreviewed Safety Question.
ECA 3255-86-332, Revision 0 - Reactor Building Fuel Transfer Canal Rail Removal This ECA documents the removal of the southwest portion Reactor Building Fuel Transfer Canal railing.
Safety Evaluation Report Removal of the rail was was accomplished to support decontamination activities. Rail stops have been added to the rail to prevent travel of the Canister Handling Bridge over the section where ihe rail was removed. This change does not constitute an Unreviewed Safety Question.
ECA 3153-86-0342, Revision 2 - Stairway #1 Removal - 282'-8" - 319'-2" Reactor Building This ECA documents the removal of stairway #1 in the Reactor Building from elevation 282'-8" to elevation 319'-2".
Safety Evaluation Summary Removal of this stairway will provide access to the Reactor Building i
basement for decontamination by robotics. This change does not constitute an Unreviewed Safety Question.
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m ATTACHMENT 4 4410-89-L-0030 ECA 3820-86-0349, Revision 1 - Modification to Service Building Soiled Exhaust This ECA dccuments the installation of a larger fan motor and associated control components in the Service Building exhaust ventilation.
Safety Evaluation Summary This modification increases the monitoring and filtering of exhaust ventilation for contaminated work activities in the Service Building.
Plant safety is enhanced and an Unreviewed Safety Question does not exist.
ECA 3211-86-0372, Revision 0 - Removal of Block Orifice (MU-1-FE) and Associated Piping This ECA documents the removal of the letdown system block orifice and associated piping in order to achieve Phase III endpoint decontamination criteria.
Safety Evaluation Summary The purpose of this change is to eliminate high dose rates in cubicle FH101 and to provide a replacement piping assembly containing isolation and a system flush connection. This change does not constitute an Unreviewed Safety Question.
ECA 3830-86-0392. Revision 0 - Fuel Handling Building Exhaust Isolation Damper Modification This ECA documents the addition of ductwork to the Fuel Handling Building exhaust isolation dampers Safety Evaluation Summary This modification provides more complete isolation of the Fuel Handling
,i Building Ventilation System and better differentiation of effluents from Unit 1 and Unit 2.
This change does not constitute an Unreviewed Safety Question.
ECA 3257-86-0400, Revision 0 - Canister Loading Decon (CLD) System This ECA documents the installation of piping to the CLD System to allow system operation at 1800F.
Safety Evaluation Summary The installation of this piping changed the path of the CLD System water and not the system function, water source, or outlet. This changes does not constitute an Unreviewed Safety Question. l l
m-ATTACHMENT 4 4410-89-L-0030 i
ECA 3510-87-0469, Revision 0 - Removal of CW-P-2B Motor i
This ECA documents the removal of a Circulating Water (CW) System pump motor.
Safety Evaluation Summary The CW System is currently out of service; Unit 1 recuires the use of the pump motor. Removal of this pump motor does not constitute an Unreviewed
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I Safety Question.
i ECA 3213-88-0505, Revision 0 - Core Flood Line A Isolation This ECA documents the installation of a barrier in the core flood line A.
4 Safety Evaluation Sunmary Tank A was opened to store core support assembly pieces. A barrier was added in the core flood line to prevent the potential for underborated water injection in to the Reactor Coolant System. This provides a second barrier in the flow path so that the core flood line is in compliance with the Boron Hazards Analysis. This change does not constitute an Unreviewed Safety Question.
ECM 0856, Revision 0 - SDS Spent Ion Exchanger Storage Racks This ECM documents a modification to the upper SDS spent ion exchanger storage racks.
Safety Evaluation Summary This modification raises the upper moveable storage racks to accommodate flexible hoses. This change does not constitute an Unreviewed Safety Question.
MMA 3558-88-0058, Revisions O and 1 - Reactor Vessel Cavity Flooding and Sampling System This mini-mod documents the modification to the 14 inch vent duct in the Reactor Vessel Cavity.
Safety Evaluation Summary Removal of the blind flange on the 14 inch vent duct in the Reactor Vessel Cavity and installation of a submersible pump with a 1/2" discharge hose was accomplished to verify the boron concentration of the water in the Reactor Vessel Cavity. Plant safety is enhanced and an Unreviewed Safety cuestion does not exist. - _ _ _ _ _ _ _ _ _ -
m ATTACHMEifi 4 4410-89-L-0030 MMB 3523-88-00'73, Revision 0 - Defueling Water Tie-In to Demineralized Service Water System This mini-mod documents the modification to the Demineralized Service Water System.
Safety Evaluation Summary This modification provides a location to draw demineralized water from the Service Building for use in the Reactor Building. There is double isolation provided in the tie-in line. This change does not constitute an Unreviewed Safety Question.
HMA 3233-88-0092, Revision 0 - Air Sparge for SRSTs This mini-mod documents the modification to the sludge processing system.
Safety Evaluation Report Modifying the sludge processing system line to provide service air for sparging the Spent Resin Storage Tanks (SRST) provides agitation at the SRST outlets to facilitate transferring solids. This change does not constitute an Unreviewed Safety Question.
MMB 3737-88-0115, Revision 0 - Power for Environmental Chemistry Lab in Turbine Building El. 331' This mini-mod documents the source of the electrical power supply for the Environmental Chemistry Laboratory located in the Turbine Building.
Safety Evaluation Summary The source of power is non-Class lE and is totally contained in the Turbine Building. This mini-mod does not constitute an Unreviewed Safety Question. i
1 GPU Nuclear Corporation GT#U luclear ggonia=48o i
Middletown, Pennsylvania 17057-0191 717 944 7621 TELEX 84 2306 Writer's Direct Dial Nurnber:
I (717) 948-8461 March 20,1989 4410-89-L-0030/0069P Document Control Desk US Nuclear Reguletory Commission Washington, DC 20555
Dear Sirs:
Three Mile Island Nuclear Station, Unit 2 (TMI-2)
Operating License No. DPR-73 Docket No. 50-320 10 CFR 50.59 Report for 1988 In accordance with the requirements of 10 CFR 50.59, " Changes, Tests, and Experiments," forwarded is a description of changes to facility systems and procedures described in the TMI-2 Final Safety Analysis Report (FSAR) which were accomplished during 1988.
Also included is a summary of tests and experiments performed that are not described in the FSAR.
Sincerely,
)6 Lee-.
M. B. Roche Director, TMI-2 RDW/ emf Attachments cc:
T. A. Moslak - Acting Senior Resident Inspector W. T. Russell - Regional Administrator, Region I J. F. Stolz - Director, Plant Directorate I-4 L. H. Thonus - Project Manager, TMI Site s
I GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation
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