ML20205A644
| ML20205A644 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/31/1984 |
| From: | Standerfer F GENERAL PUBLIC UTILITIES CORP. |
| To: | Murley T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| 0191A, 191A, 4410-85-L-0052, 4410-85-L-52, NUDOCS 8504250567 | |
| Download: ML20205A644 (8) | |
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l-FACILITY MODIFICATIONS iltems in'this section were performed without prior approval.of'the IRC' staff
.under the authority-of 10 CFR 50.59..The items listed below cover specific 5 activities and. include'several Engineering Change Memoranda'(ECM's). 'ECM's are a tracking mechanism for review, approval and documentation of specific
. plant' changes..ECM's selected for inclusion were those for which turn-over to
' Site Operations was completed during the calendar year 1984.
- ECM ll78,' Revision 0 - Feedwater Pump Removal This ECM doc ments removal of feedwater pump 1-B along with associa'ted components.
Safety Evaluation Summary The feedwater System.was no longer necessary. When.the pump was
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removed, the system was tagged out-of-service. Since the system was no longer necessary.for operations or accident mitigation, its removal was not considered to be an Unreviewed Safety Question.
ECM 1203 Revision 0; Removal of Core Flood Tank Instrumentation
. Cables _used for pressure and level transmitters on Core Flood Tank I were previously disconnected and used for head removal instrunentation (ECM 1080 Revision 0 and 1085 Revision 0). Removal of Core Flood Tank instruments CF-1-PT3 and CF-2-LT3 was performed for data acquisition.
Safety Evaluation Summary Removal'of the core flood tank instruments does not affect previous analysis or create new or different accident potential since the core
. flood tanks were removed from service shortly after the accident. The core flood tanks serve no operational function in the recovery mode.
Therefore, the change does not constitute an Unreviewed Safety Question.
ECM 1132 Revision 0 - Canal Seal Plate Modifications Modifications were made to the canal seal plate to accommodate hold down lugs f
-and to_ facilitate'the installation of a silicone sealant.
7 Safety Evaluation Summary These changes enhance the sealing ability of the canal seal plate by
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providing additional hold down force to ensure gasket contact and a secondary seal provided by a pourable silicon compound.
These modifications help to ensure that potential leakage from the IIF does i
not drain to the basement. It was determined that the modification did not constitute an Unreviewed Safety Question.
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ECM 122 Revision 0 - Service Air Compressor A service air compressor was added to Elevation 305 of the Auxiliary Building.
The compressor was installed to supplement and increase the capacity of the existing service air system.
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Safety Evaluation Summary Increasing the capacity of the service air system would not affect plant safety analyses as defined by 10 CFR 50.59.
The modification did not constitute an Unreviewed Safety Question.
ECM 1137 Revision 0 - Respirator Cleaning Facility This ECM docunents the safety evaluation for installation of a new respirator cleaning facility onsite. The new facility replaces and upgrades an existing facility and eliminates the use of contract services.
Safety Evaluation Summary The new facility is a closed loop system designed to handle 1000 respirators per month. No liquids will be discharged. Gaseous effluents are filtered and monitored prior to release. Monthly releases are well within Technical Specification limits and do not significantly increase discharges from the replaced facilities. Accidents are bounded by previous evaluations. The installation of the facility does not constitute an Unreviewed Safety Question.
ECM 1110 Revision 0 - Service Air System Portions of the existing service air lines were modified to eliminate highly contaminated areas of piping and prevent the spread of contamination in the plant. Other modifications were accomplished to allow use of the remainder of the system.
Safety Evaluation Summary This ECM did not affect over all system operation; only pipe routings were affected.
Thus, the change does not constitute an Unreviewed Safety Question.
ECM's 1209 Revision 0 and 1226 Revision 0 - Turbine Building Modifications These ECM's install three prefabricated structures in the Turbine Building.
These facilities include a command center for entry control, a counting lab and an office structure.
Safety Evaluation Summary These structures do not affect important-to-safety or nuclear safety-related systems or components. All tie-in's are to BOP, commercial or domestic lines.
The additions of these structures was evaluated as having no effect on components important to the safety of the plant as evaluated by 10 CFR 50.59.
Therefore, they do not constitute an Unreviewed Safety Question.
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ECM's 1065 Revision 0 and 1262 Revision 0 - Fire Alarm System These ECM's authorized addition of alarm capabilities to the fire alarm system. - ECM 1065 provides remote fire alarms for the Mobile Chemistry Lab trailer. ECM 1262 provides remote alarm capabilities for the Respirator Cleaning Facility.
Safety Evaluation Sunmary The addition of remote alarm capabilities enhances plant safety and the health and safety of the public by facilitating faster response to onsite fires.
The additional alarms do not jeopardize the existing system performance.
ECM 1297 and ECA 84-0027 - Air Chillers A Reactor Building air chiller system was installed to replace existing evaporative coolers. The new air chillers are more efficient in reducing temperatures in the Reactor Building.
Safety Evaluation Summary The new chiller system replaces the old evaporative coolers. Active mechanisms are external to all buildings. Additionally, safety is enhanced in that the chillers are a closed system as opposed to the evaporative coolers which could discharge unlimited river water in the event of a line break. Leakage from the new system as a result of a line break is limited; thus, nuclear safety is enhanced.
The chillers were evaluated as not constituting an Unreviewed Safety Question.
ECM 1188 Revisions 0, 1, 2 - Cable Removal Sections of the control and power cables for in board isolation valve SU-V54 were removed.
This valve is currently closed and need not be opened for any planned operation or emergency. Removal of cabling essentially places this valve in a locked closed position.
Safety Evaluation Summary Placing this valve in a " locked closed" position does not affect previous safety analyses or create a potential for a new or different accident.
Therefore, cable removal does not constitute an Unreviewed Safety Question.
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RECOVERY ACTIVITIES During 1984 a nmber of recovery plant activities were performed. Many of these activities combined modifications, procedural changes, and tests or experiments. All of these activities were subject to nmerous GPUNC reviews and approvals. Changes to previous activities are suomitted to the EC for information under the yearly update program for Technical Evaluation Reports, System Descriptions and Safety Evaluation Reports. In addition, certain activities were subject to NRC approval prior to implementation.
Those items receiving M C approval prior to implementation are listed below. Included is reference to the appropriate NRC correspondence approving the activity. Since all the activites listed below were previously submitted to the NRC, they will not be discussed further here.
Submerged Demineralizer Changes to this system are covered System (SDS) by the annual update program for the System Descriptions, Technical Evaluation Reports, or Safety Evaluation Reports.
EPICOR II Changes to this system are covered by the annual update program for the System Descriptions, Technical Evaluation Reports, and Safety Evaluation Reports.
Standby Pressure Control Changes to this system are covered by the annual update program for the System Descriptions, Technical Evaluation Reports, and Safety Evaluation Reports.
Once Through Steam Generator Changes to this system are covered by (OTSG) the annual update program for the System Descriptions, Technical Evaluation Reports, and Safety Evaluation Reports.
Containment Air Control Primary Information forwarded via Envelope GPU Nuclear Letter 4410-84-L-0175, dated October 30, 1984. Project approval via WC letter dated December 4, 1984.
First Pass and Final Stud Primary Information forwarded via Detensioning GPU Nuclear letter 4410-83-L-0222, dated September 29, 1983. Project approved via WC letters dated February 17, 1984 and June 21, 1984.
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Head Removrl Primary information forwarded via GPU Nuclear letter 4410-84-L-0014 dated March 9, 1984. Project approved via EC letter July 17, 1984.
Plenum Inspection and Removal Primary Information forwarded via GPU Nuclear letters 4410-84-L-0014 dated March 9, 1984,-4410-84-L-0032 dated June 18, 1984 and 4410-84-L-0164 dated November 2, 1984. Projects were approved via NRC letters dated July 13, 1984, E C/TMI 84-083 dated September 14, 1984 and November 27, 1984.
Cesium Elution Primary Information forwarded via GPU Nuclear letters 4410-84-L-0094 dated July 19, 1984 and 4410-84-L-0136 dated August 29, 1984. Project was approved via NRC letter NRC/TMI 84-069 dated September 19, 1984.
Internals Indexing Fixture Primary Information forwarded via GPU Nuclear letters 4410-84-L-0082 dated May 31, 1984 and 4410-84-L-0108 dated July 16, 1984. Project was approved via NRC letter NRC/TMI-84-051 dated July 24, 1984.
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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 determination of an Unreviewed Safety Question or significant risk. Below is a list of tests or experiments
~which is representative of those performed during 1984..
Core Debris Sampling o
Core Debris Sample Analysis Core Debris Bed Probing o
o Detailed Core Void Video Mapping Plenun Inspection o
Reactor Building Characterization o
Basement Characterization Source Term Identification Air Cooler Characterization RCS RTD Removal and Analysis o
o Ex-Vessel Fuel Location Cable Sample Removals o
Reactor Coolant Drain Tank Sample Removal and Analysis o
o Reactor Building Sunp Sample Removal and Analysis i
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PROCEDURE CHANGES With the issuance of the Interim Recovery _ Technical Specifications, many procedures issued for surveillance under the Operating Technical Specifications became unnecessary. It was not possible to perform a large percentage of these procedures due to inaccessability. Other procedures were unnecessary due to the current operational mode. These procedures were then cancelled and, where necessary, alternate surveillance procedures were issued under the guidance..of; Recovery Technical Specification Sections 6.8.1 and
.6.8.2. -All cancelled proceduresiunderwent SRG review to_ determine impact on safety prior to cancellation. Due to the subject matter, some of these
_ procedures received EC review prior to cancellation.
Additionally, in stoport of the recovery effort,.a number of procedure changes were made and new procedures were issued. These recovery related procedures
-received E C review and approval prior to' implementation, if required by Technical Specifications, Section 6.8.2.
Additionally, many of these
. procedures performed activities described in SER's submitted to the NRC on-the-docket. _Since these procedures have E C approval, they will not be discussed further in this report.
A number of procedural changes were made to convert existing procedures into3 the new format being used at TMI-2.
These format changes did not change the' technical content of the procedure, therefore, these changes are not applicable to the 10 CFR 50.59 report.
The remainder of the changes were reviewed and it was determined that there
. ere no changes which specifically constituted a FSAR change as defined by 10 w
CFR 50.59.- However, there were a number of changes made to FSAR-type procedures.
These changes were made to reflect chanaina plant conditions or to implement the recommendations of various boron dilution analyses. Typical system-oriented procedures receiving these types of changes are:
Temporary Nuclear Sampling System Reactor Building Purge Decay Heat Removal System RCS Operating Procedures Containment Integrity Verification Source Range Instruments Control Rod Drive System All procedures receiving this t'ype of change were determined to not constitute an Unreviewed Safety Question.
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F-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 4410-85-L-0052 Document ID 0191A March 15, 1985 Office of Inspection and Enforcement Attn: Dr. T. E. Murley Regional Administrator US Nuclear Regulatory Commission Region I 631 Park Avenue King of Prussia, PA 19406
Dear Dr. Murley:
Three Mile Island Nuclear Station, Unit 2 (TMI-2)
Operating License No. DPR-73 Docket No. 50-320 10 CFR 50.59 Report for 1984 In accordance with the requirements of 10 CFR 50.59, enclosed for your information are one (1) signed original and thirty-nine (39) copies of a description of changes to facility systems and procedures in TMI-2 for 1984 as described in the TMI-2 Final Safety Analysis Report (FSAR). Also included is a sumiary of tests and experiments performed that were not described in the FSAR.
Sincerely, F. R. Standerfer Vice President / Director, TMI-2 FRS/RBS/vjf Attachments Program Director - TMI Program Office, Dr. B. J. Snyder cc:
Deputy Program Director - TM1 Program Office, Dr. W. D. Travers 0191A GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation hW 1qi