ML20147F182
| ML20147F182 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/31/1987 |
| From: | Standerfer F GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| 4410-88-L-0022, 4410-88-L-22, NUDOCS 8803070261 | |
| Download: ML20147F182 (11) | |
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j GPU Nuclear Corporation Nuclear
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Middletown, Pennsylvania 17057-0191 717 944-7621 TELEX 84-2386 Writer's Direct Dial Number:
(717) 948-8461 March 2, 1988 4410-88-L-0022/0069P Document Control Desk US Nuclear Regulatory Comission Washington, DC 20555
Dear Sirs:
Three Mile Island Nuclear Station, Unit 2 (TMI-2)
Operating Liuense No. DPR-73 Docket No. 50-320 10 CFR 50.59 Report for 1987 in accordance with the requirements of 10 CFR 50.59, "Cha@es, Tests, and Experiments," forwarded herein is a description of changes to facility systems and procedures described in the THI-2 Final Safety Analysis Report (FSAR) which were effected during 1987. Also included is a summary of tests and experiments perfomed that are not described in the FSAR.
Sincere y, F. R. Standerfer Director, TMI-2 CJD/eml Attachments cc: NRC Resident Inspector - R. J. Conte Regional Administrator - Region 1, Dr. W. T. Ru3 sell Director - TMI-2 Cleanup Project Directorate, Dr. W. D. Travers 8003070261 871231 PDR ADOCK 05000320 6
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GPU Nuclear Corporation is a subsidiary of the General Pubhc Utilities Corporation i
ATTACHENT 1 4410-88-L-0022 RECOVERY ACTIVITIES During 1987 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 infomation 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 quarterly update program. Updates were submitted via GPU Nuclear letters 4410-87-L-0005 dated January 15, 1987; 4410-87-L-0055 dated April 14, 1987; 4410-87-L-0107 dated July 14, 1987; and 4410-87-L-0153 dated October 1S, 1987.
o EPICOR II Changes to this program are covered by the annual update program for System Descriptions and Technical Evaluation Reports. Update sutmitted via GPU Nuclear letter 4410-87-L-0075 dated May 14, 1987.
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 GPU Nuclear letter 4410-87-L-0074 dated May 12, 1987.
o Processed Water Storage Changes to this program are covered by and Recycle System the annual update program for System Descriptions and Technical Evaluation Reports. Update sutmitted via GPU Nuclear letter 4410-87-L-0129 dated 1
l August 26, 1987.
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 sutmitted via CPU Nuclear letter 4410-87-L-0085 dated May 12, 1987.
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ATTACHKNT 1 4410-88-L-0022 o Solid Waste Staging Facility Changes to this program are covered by the annual update program for System Descriptions and' Technic 31 Evaluation Reports. Update submitted via GPU Nuclear letter 4410-87-L-Oll4 dated July 30, 1987.
o Submerged 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 4410-87-L-0125 dated September 28, 1987.
o Sediment Transfer System Changes to this program are covered by updates to the Safety Evaluation Report. Updcted infomation was submitted via GPU Nuclear letters 4410-87-L-0022 dated February 2,1987; and 4410-87-L-0163 dated November 25, 1987.
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 4410-87-L-0177 dated December 3, 1987.
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 infor.v.ation was provided via CPU Nuclear letters 4410-87-L-0037 dated April 2, 1987; 4410-87-L-0078 dated May 26, 1987; 4410-87-L-0127 dated October 21, 1987; and 4410-87-L-0180 dated December 2, 1987.
o Core Region Defueling Changes to this program are covered by updates to the Safety Evaluation Report and other docketed correspondence.
Updated information was provided via GPU Nuclear letters 4410-87-L-0034 dated March 3,1987; and 4410-87-L-0165 dated November 4,1987.
ATTACHENT 1 4410- 88-L-0022 o Defuell'ng 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. Updates were submitted via GPU3 Nuclear. letters 4410-07-L-0051' dated April 29, 1987; 4410-87-L-0065 dated May 21, 1987; and 4410-87-L-0095 dated June 12, 1987, o Fuel Canister Storage Racks Changes to.this program are covered by'.
the annual update program for System Oescriptions and Technical Evaluation
( T.eports. Update subb.tted via GPU Nuclear letter 4410-87-L-0079 dated May 14, 1987.
Changes to this program'are. covered by o Defueling Canisters f
the annual update program for System Descriptions and Technical Evaluation s
Reports. Updated information was submitted via GPU Nuclear letters 4410-87-L 'X152 dated May 7,1987; and 4410-87-L-Oll6 dated September 22, 1987.
o Defueling Canister Dewatering Changed to this program are covered by System the annual update program for System Descriptions and Technical Evaluation
/5eports. Update submitted via GPU Nuclear letter 4410-87-L-0120 dated August 17, 1987.
o Pressurizer Spray Line Changes to this program are covered by Defueling System updates to the Safety Evaluation Report. Updated infomation was provided via GPU Nuclear letter 4410-87-L-00ll dated January 15, 1987.
o Pressurizer Defueling Changes to this program are covered by updates to the Safety Evaluation Report. Updated infomation was provided via GPU Nuclear letters 4410-87-L-0033 dated March 11, 1987; 4410-87-L-0108 dated July 21, 1987; and 4410-87-L-0131 dated August 28, 1987.
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o'(asteHandlingandPackaging Changes to this program are covered by
'sFacility 5M the. annual update program for System s
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Reports and other-docketed
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Descriptions and Technical Evaluation.
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was submitted via GPU Nuclear letters 1
4410-87-L-0016 dated February 2,.1987; and 4410-87-L-0062 dated July 24, 1987.
o' Plasma Arc Cutting Changes to this program are covered by updates to the Safety Evaluation Report and other docketed. correspondence.
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Updated information was provided via GPU Nuclear letters 4410-87-L-0012 dated January 20, 1987; 4410-87-L-0067 dated May.7, 1987; and 4410-87-L-0091
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dated June 25, 1987.
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AT T ACHbENT 2 a'
4410-88-L-0022 r
PROCEDUPZ CHgES -
During the course of 1967, pm:edural requiremeds changed due to the defueling and fuel shipping erforts. Additionally,-in support of the recovery effort, a number of procedure changes were made and nc's proced tren were issued. Some of these recovery-related procedures received NRC.aview c.nd i
approval prior to hplementation as required by Technical Specif n.aion Section 6.8.2.
Additionally, many of these procedures perfomed activities in accordance with NRC-approved Safety Evaluation Reports.
Since these procedures have received NRC approval, they will not be discussed furtner in this report. Procedures whose scope of activity was completed during 1987 were cancelled. Cancelled procedures determined to have review significance i,.h underwent SRG review to determine the potential impact on safety prior to.
M cancellation. Due to the subject matter, some of these procedules received
.NRC review prior to cancellation.
A number of procedural changes were made to convert existing procedures into the new format being used at THI-2. These format' changos did not change the technical content. of procedures; tM.:efore,. these changes are not applicable to the 10 CFR 50.59 report, The remainder of the changes were reviewed and it was detemined tN ': there were no changes which specifically ccostituted a FSAR change as defined by 10 CFR 50.59 However, there were a nuTh3r of changes made to FSAR-type procedures. These changes.were maoe to' reflect changing plant conditions or to implement the recommendations of various activity-related analyses.
Typical system-oriented procedures receiving these types of charcas are:
o Transfer of Orated Water Storage Tank (BWST) Water to the Reactor Coolant Bleed Tanks (RCBTs) o Loss of Reactor Coolant System (RCS) Level Indication o
Operation of the Seco1dary Plant System o
Use of the Ds:ay Heat Removal System for errergency injectian of BWST water to the RCS o
Hydrogen Peroxide Acdition to the Reactor Vessel (RV) o Transfer of Misce]laneous Waste Holdup Tank (MWHT) water to the RCBTs o
Solid Face Bit Drilling Operations o
Canister Positioning System Operations o
Cavitating Water Jet Operations o
Use of Nitrogen for Nuclear and Radioactive Waste System o
Trar5fer of Defueling Tools to arrj Frcm the RV/ Tool Racks o
Flush of MWHT o
Oraining Makeup Tank MU-T-1 to t% RC6Ts o
Transfer of Reactor Building (RB) Suma Sludge to the Auxiliary Building
/AB) Sump Tank o
Canister Transfer System o
Demineralized Service Water Operation o
Recircu?.ation, Sampling, and Chemical Addition to the Spent Resin Storage Tanks o
Inspection of Dikes
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4410-68-L-0022 o
. Fire System Water Source Level Check o.
Cable Room and Transformer Room Halon System Functional Test o-
-Fuel Handling Building (FHB) Crane Operations o
Lif ting and Handling Inside the RB of Five (5) Tons and Lesser Loads o'
Containment Integrity Verification o
Fire' Pump. Capacity Testing o
Polar Crane Operation Procedures receivin0 this type of update change to reflect current plant conditions were determined to not constitute an Unreviewed Safety Question, i
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ATTACHENT 3 4410-88-L-0022 TESTS AND EXPERIENTS 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 wem 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 L which is representative of those performed during 1987.
o Core Debris Bed Inspection ~ and Probing o
Detemination of Stub Assembly Heights o
Gamma Measurements of the Basement Wall, the Waste Transfer Pump and Valve Room,. the Liner Wall, Reactor Coolant Bleed Tanks B and C for Characterization o'
Maintained Core Material Balance for Core Debri s in the Reactor Vessel o
Video Inspection and Sanpling of the Lower Head o
. Defined Conditions in Regions of the Reactor Vessel o
Verification of an Unfolding Code to Quantify Fuel from Fewer Measurement Locations than Deposition Sites o
Robotic Surveys for Sediment Sampling in the Basement a
Sampling and Estimating Sediment Voltane in the Reactor Coolant Pump Casings and Discharge Legs, the B Hot Leg and Attached Decay Heat Line, between the Core Former Baffle Plates and the Core Barrel, and the Deborating Demineralizers o
Gamma Measurements of the Seal Injection Filter Room and Makeup Pump Room B to Estimate Fuel Quantities o
Determination of Equilibrium Airborne Activity in the Contairunent During Post-Defueling Monitored Storage
ATTACHENT 4
'4410-88-L-0022 FACILITY MODIFICATIONS Items in this _section were perfornd without prior approval of the NRC staff under the authority of 10 CFR 50.59. The items listed below cover specific activiles performed under the authority of Engineering Change Authorizations (ECAs). FCAs are tracking mechanisms for review, approval, and documentation of spec: ic plant changes. ECAs selected for inclusion were those for which turnover was co:npleted during the calendar year 1987.
ECA 3851-85-0285, Revision 1 - Air Supply to Model Room Solidification Machine Filter Unit This ECA documents the installation of a dry air supply to the solidification machine in the model room.
Safety Evaluation Stsnmary Addition of a dry air supply to the filter in the solidification machine allows the dust to be cleaned from the filtor. The service air released from the machine will be controlled by plant ventilation. This change does not constitute an Unreviewed Safety Question.
.ECA 3820-86-0355, Revision 0 - Reactor Building Evaluation 282' Ventilation This ECA documents the addition of two (2) ventilation units within the Reactor Building to support decontamination of the Reactor Building 282' elevation.
Safety Evaluation Summary The Reactor Building ventilation system will be enhanced by the addition of the ventilation units designated for an area that is being actively decontaminated. Plant safety is enhanced and an Unreviewed Safety Question does not exist.
ECA 3851-86-0384, Revision 1 - Modify Decontamination Service Air Line to Reactor Building Penetration R561 This ECA documents the addition of a 3/4" service a.r line to Reactor Building Penetration R561.
Safety Evaluation Stanary The air line attaches to a 3/4" fitting already present in Penetration R561. This work does not affect the integrity of the penetration and, therefore, an Unreviewed Safety Question does not exist.
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ATTACHENT 4 4410-88-L-0022 ECA 3211-86-0385, Revision 0 - Modification to Reactor ~ Building Fenetration R333 t
This ECA documents mcdifications to Reactor Building Penetration R535 to I
pmvide for the penetration of a general purpose lower pressure fluid line.
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Safety Evaluation Summary
~ The integrity of Penetration R535 is not compromised by this modification. The modification does not alter the existing fluid head -
closure plate and no Unreviewed Safety Question exists.
ECA 3211-86-0389, Revision 0 - Nitrogen Manifold System Modification This ECA documents the removal of local sections of the nitrogen system which are highly contaminated internally.
Safety Evaluation Sumary Removal of local sections of the nitrogen system will reduce recontamination potential by eliminating potential leakage paths.
Removal of the sections of the nitmgen system specified in this ECA are not covered by the Technical Specifications. No Unreviewed Safety Question exists.
ECA 3211-86-0399, Revision 0 - Removal of Select Instruments of the N.'.trogen System This ECA documents the removal-of instruments that were rendered inoperable by the work of ECA 3211-86-0389 Safety Evaluation Report The removal of these instruments did not affect safety since they were inoperable when they were removed. No Unreviewed Safety Question exists.
ECA 3251-86-0427,. Revision 1 - Retum SF-P-2 to Serv 1Lce This ECA documents the testing of SF-P-2 after corrective maintenance.
Testing will require the fabrication of a spoolpiece.
Safety Evaluation Sumary The fuel transfer canal fill pump (FCC-P-2) must be removed from service for repairs. The Borated Water Reciruelation Pump (SF-P-2) will be returned to service prior to removing FCC-P-2.
SF-P-2 is capable of duplicating FCC-P-2 functions; there is no loss of safety and an Unreviewed Safety Question does not exist.
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ATTACHENT 4 4410-88-L-0022 ECA 4330-86-0437, Revision 0 - Decontamination Process Water Pumps
- This ECA documents the changing of pumps to increase the decontamination process water system capabilities to provide sufficient flushwater for sludge transfer and processing.
Safety Evaluation Sumary The decontamination process water system performed no safety. function and has' no interface with Inportant to Safety systems. Plant safety is not affected and no Unreviewed Safety Question exists.
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