IR 05000289/1982003
| ML20053A660 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/11/1982 |
| From: | Fasano A, Haverkamp D, Petrone C, Richards S, Young F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20053A656 | List: |
| References | |
| 50-289-82-03, 50-289-82-3, NUDOCS 8205270093 | |
| Download: ML20053A660 (29) | |
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U. S. NUCLEAR REGULATORY C0fiMISSION Region 1 Report No.
50-289/82-03 Docket No.
50-289 License No.
DPR-50 Priority Category C
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Licensee:
GPU Nuclear Corporation P.O. Box 480 Middletown, Pennsylvania 17057 Facility:
Three Mile Island Nuclear Station, Unit 1 Inspection at: Middletown, Pennsylvania Inspection conducted: March 5, 1982 - April 13, 1982 Inspectors:
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D.Haverkamp,SeniorpsidentInspector(TMI-1) d6te ' signed
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C. Petrone, Reactor I@spector date signed (April 8, 1982)*
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[M S. Richards, Reactor 6Aspector date signed v
(March 22-26, 1982 - April 7-9, 1982)*
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_5}es}igned 11 92 F.Yougl,ResigtInspector(TMI-1)
da Other NRC R. Jacobs, TMI-l Project Manager, Operating Reactors Personnel:
Branch 4, Division of Licensing, NRR (March 31, April 1, 1982)
D. Garner, Technical Assistant to Director of icensing, NRR (March 31,1982)
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Approved by:
_. Fasino, Chief, Three Mile Island Section da'te signed Projects Branch No. 2
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Inspection Summary:
Inspection on March 5,1982 - April 13,1982 (Report tiumber 50-289/82-03)
A_reas Inspected:
Routine safety inspection by resident and regional-based
inspectors (187 hours0.00216 days <br />0.0519 hours <br />3.091931e-4 weeks <br />7.11535e-5 months <br />) of plant operations during long term shutdown including a quarterly emergency preparedness drill, facility tours, and log and record reviews; TMI-l steam generator recovery program activities; TMI-l restart modifications - implementation; and, TMI-l restart modifications - task s:atus.
Results: fio items of noncompliance were identified.
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Details 1.
Persons Contacted
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GPU Nuclear Corporation R. Adamiak, Project Control Manager (TMI-1), Administration R. Ballard, Manager TMI Quality Assurance Modifications / Operations, Nuclear Assurance R. Barley, Lead Mechanical Engineer TMI-l J. Colitz, Plant Engineering Director TMI-l M. Beers, Engineering Assistant Senior II, Nuclear Assurance C. Davis, Modification Control Coordinator T. Faulkner, Planning and Scheduling Manager, Maintenance and Construction R. Fenti, Quality Control Manager, Nuclear Assurance J. Fritzen, Technical Functions TMI-l Site Supervisor
- J. Garrison, QA Engineer Assistant Senior III, Nuclear Assurance G. Giangi, Emergency Preparedness Manager, Nuclear Assurance T. Hawkins, Manager THI-l Startup and Test, Technical Functions
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- H. Henry, Engineer Assistant Senior II, Nuclear Assurance H. Hukill, Vice President and Director TMI-l i
- S. Levin, Maintenance and Construction Director (TMI-1)
Maintenance and Construction Division W. Miller, Nuclear Licensing Engineer, Technical Functions M. Ross, Manager Plant Operations TMI-l C. Smyth, Supervisor TMI-l Licensing, Technical Functions
- C, Stephenson, Nuclear Licensing Engineer, Technical Functions
- R. Toole, Operations and Maintenance Director TMI-1 E. Wallace, Manager PWR Licensing, Technical Functions
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S. Wilkerson, Lead Nuclear Engineer TMI-l The inspector also interviewed several other licensee employees during the inspection. They included control room operators, maintenance personnel, engireering staff personnel and general office personnel.
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Plant Operations During Long Term Shutdown a.
Plant Status During Inspection Period
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The plant has remained in the cold shutdown condition with Reactor i
Coolant System (RCS) temperature less than 200 F per NRC order of August 9, 1979. At the beginning of the inspection period on March 5, 1982, the RCS was at 90 F and 0 psig with cooling to the reactor supplied by the
"B" Decay Heat Removal (DHR) loop. The RCS was partially drained for eddy current testing and preparation for head removal.
Both Once Through Steam Generators (OTSG's) were partially drained on the secondary side.
The Borated Water Storage Tank (BWST) was on cleanup to remove sulfur contaminants.
During the course of the inspection the following major plant changes or evolutions occurred:
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-4-March 6, 1982:
Shifted from BWST to the Spent Fuel Pool on
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cleanup to remove sulfur contaminants.
March 8, 1982:
Placed "B" 0TSG in wet lay up on the secondary
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side.
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March 12, 1982:
Placed "A" 0TSG in wet lay up on the secondary side.
March 13, 1982:
Shifted from "B" DHR loop to "A" DHR loop.
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March 15, 1982: Spent Fuel Pool taken off cleanup.
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March 19, 1982:
Flushed Reactor Building Spray System lines interconnected to BWST to remove residue sulfur.
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March 23, 1982: Uncoupled control rod drive mechanisms from control rods in preparation for Reactor Vessel Head removal.
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March 24, 1982: Conducted quarterly emergency preparedness drill
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April 1, 1982: Shifted from "A" DHR loop to "B" DHR to perform maintenance on components in the "A" DHR loop.
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April 3, 1982: Shifted from "B" DHR loop to "A" DHR loop.
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April 7, 1982:
Placed Decay Heat Removal System on cleanup.
April 13, 1982:
Removed Reactor Vessel Head for vessel internal
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component inspection.
At the close of this inspection period the RCS was at 100 F and 0 psig with cooling to the reactor core from "A" DHR loop.
The RCS was partially drained with the Reactor Vessel Head removed.
Both OTSG's were in full wet lay up.
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Plant Logs and Operating Records The inspector reviewed selected portions of the following plant procedures to determine the licensee established requirements in this
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area in preparation for a review of selected logs and records.
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Administrative Procedure (AP) 1007, " Control of Records,"
Revision 4
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AP 1012, " Shift Relief and Log Entries," Revision 14
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AP 1033, " Operating Memos and Standing Orders," Revision 2
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AP 1037, " Control of Caution and [Do Not Operate] DN0 Tags,"
Revision 2
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AP 1044, " Event Review and Reporting Requirements," Revision 2
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-5-The inspector reviewed the following plant logs and 9perating records.
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Shift Foreman Log and Control Room Log Book Shif t Turnover Checklist
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Temporary Change Notice (TCN) Log Book Active Tagging Application Book
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Do Not Operate and Caution Tag Log Plant logs and operating records were reviewed to verify the following items.
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Log keeping practices and log book reviews are conducted in accordance with established administrative controls.
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Log entries involving abnormal conditions provide sufficient detail to communicate equipment status, lockout status, correction and restoration.
Operating orders do not conflict with Technical Specifications
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(TS) requirements.
Tagging operations are conducted in conformance with established
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administrative controls.
No items of noncompliance were identified, c.
Facility Tours During the course of the inspection, the inspector conducted tours of the following plant areas.
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Control Room (daily)
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Auxiliary Building (March 9,17)
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Intermediate Building (March 18, 31, April 8)
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Vital Switchgear Ruoms (March 18, 23, 31, April 8)
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Diesel Generator Building (March 10, 22, 26, April 9)
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Yard Area (March 12, 24)
Site Perimeter (March 24)
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Reactor Building (March 9, 15, 16, 18, 23, 30, 31, April 1, 6,9,13)
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Operations Support Center, Technical Support Center, Emergency
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Offsite Facility, Alternate Emergency Offsite Facility, Environmental Assessment Command Center, and Emergency Assembly Area during licensee quarterly emergency preparedness drill (March 24)
The following observations / discussions / determinations were made.
Control room annunciators:
Selected lighted annunciators
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(Control Rod Drive Mechanism (CRDM) Failure, Fuel Transfer Canal HI/LO, Diesel Generation IB Blocked) were discussed with control room operators to verify that the reason for annunciation was understood and corrective action, if required, was being taken.
Control room manning: By frequent observation during the
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inspection, the inspector verified that control room manning requirements of 10 CFR 50.54(k) and the Technical Specifications were being met.
In addition, the inspector observed shift turnovers to verify that continuity of system status was maintained.
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Technical Specifications:
Through log review and direct observation during tours, the inspector verified compliance with Technical Specifications associated with Reactor Coolant System and Steam Generator Tube Inspection requirements during shutdown plant operation.
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Plant housekeeping conditions:
Unsatisfactory plcnt housekeeping conditions observed by the inspector had been identified previously by station personnel and corrective action had been initiated as necessary.
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Monitoring instrumentation:
The inspector verified that selected instruments (Steam Generator Wide Range level, Core
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Flood tank levels, BWST tank level and RCS pressure) were functional and indicated that parameters were within Technical Specification limits.
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Valve positions:
The inspector verified that selected valves were in the position or condition required by Technical Specifications for the applicable plant mode. This verification included control board indication and field observation of valve positions for the Spent Fuel Pool System and Reactor Coolant System.
Equipment tagging: The inspector selected plant components
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(Reactor Coolant System Valves) for which valid tagging requests were in effect and verified that the tags were in place and that i
the equipment was in the condition specified.
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Security: During the course of this inspection, observations relative to protected and vital area security requirements were made, including access controls, boundary integrity, search, escort, and badging.
No adverse conditions were noted.
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-7-Licensee meetings:
The inspector frequently attended the
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Plan-of-the-Day (POD) morning meetings and the daily status afternoon meetings held by licensee management and supervisory
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personnel.
The inspector observed the meetings to assess licensee evaluation of plant conditions, status and problems and to review the licensee's plans for conducting certain major plant operations and maintenance activities. The inspector also attended bi-weekly project status meetings to assess licensee progress and difficulties related to plant modifications required for restart.
Acceptance criteria for the above itens included inspector judgement
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and requirements of 10 CFR 50.54(k), Regulatory Guide 1.114,
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Technical Specifications, and the following procedures.
AP 1002, " Rules for the Protection of Employees Working on
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Electrical and Mechwical Apparatus," Revision 22 AP 1008, " Good Housekeeping," Revision 7
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AP 1037, " Control of Caution and DN0 Tags," Revision 2 No violations were identified.
3.
THI-l Steam Generator Recovery Program Activities a.
Background Repressurization of the Reactor Coolant System (RCS) in November 1981 revealed degradation of tubes in both Once Through Steam Generators (OTSG's).
In December 1981 a steam generator recovery program task force was established to coordinate and direct all actions regarding the invest'igation and repairs to the steam generator tube leaks.
During the period of December 1981 through March 1982, the licensee
i has conducted extensive eddy current testing and metallurgical analysis on removed sections of tubes. Metallurgical analysis has established that circumferential cracks, all initiated from the tube inside surface (primary side), have occurred mainly in the top
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portion of the tubes.
During March 1982, the licensee completed the j
first portion of their tube plugging which included tubes with removed sections and tubes that had observed leakage (see NRC Region i
1 Inspection Reports 50-289/81-32, 50-289/82-01, and 50-289/82-02).
To manage the various analysis, testing, inspection and repair tasks, the task force was expanded into seven task groups. The task groups I
are OTSG Tube Failure Analysis, 0TSG Long Term Tube Repair, 0TSG Tube Plugging Program, OTSG Tube Eddy Current Inspection, Plant /0TSG
Performance, Repair Criteria (OTSG Tube Plugging), and Primary System
"eview.
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(1) Failure Analysis - Task 1
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Failure analysis involves the integration and coordination of i
all activities aimed at identifying why the steam generator I
tubes cracked in the locations identified by eddy current
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-8-testing and the failure mechanism.
The task includes compilation and evaluation of manufacturing history, chemical history, and operating history to be used as input in determining the cause of the OTSG tube damage. This task involves the coordination of metallurgical and chemical analyses by Battelle-Columbus and Babcock and Wilcox.
Licensee studies that have been made of the steam generator's history, since the discovery of small leaks in TMI-l tubes last November, have developed a probable explanation for how the corrosion occurred. The licensee is proceeding on the basis of that explanation (which is still being verified by further testing) that a species of sulfur, most likely sodium thiosulfate, was the active corrosive agent. The licensee reported that the corrosion apparently occurred through a complex interaction of equipment test sequences, materials properties, stresses and chemical conditions that occurred during the latter part of the three-year period in which TMI-l i
has not been operating.
I The most likely manner in which small amounts of sodium thiosulfate entered the primary system water is by leakage through valves downstream of the sodium thiosulfate tank during testing of the TMI-l Containment Building spray pumps, and from there into the Borated Water Storage Tank (BWST) at TMI-1.
Makeup water from the BWST was injected into the Reactor Coolant System during September 1981 hot functional testing (HFT).
During HFT of the plant the chemical form of the sulfate compound was changed to one or more sulphur species contaminants, which.upon subsequent cooldown would be corrosive in combination with the "right" conditions of stress, metal structure and chemistry of the reactor system water.
(2) Long Term 0TSG Repair - Task 2 This task involves a technical cost and schedule evaluation of 0TSG sleeving, retubing, replacement, or other repair options.
The cost, schedule and program plan for implementing each of i
these long-term repairs is to be defined and documented. This task will consider information provided by the failure analysis and the eddy current examination in defining the scope of a possible sleeving program.
Information will be provided concerning long lead time material for each long term repair.
Both the steam generator retube and the steam generator replacement analysis will identify site related constraints such as accessibility and gross man-rem exposures for each long term program. There has been little work performed on this task as of April 13, 198.
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(3) Tube Plugging Program - Task 3 The scope of the plugging program includes qualification of removable plugs including any regulatory submittals and technical documentation required.
In addition, the implementation of a Phase I plugging program during which both removable plugs and permanent welded / explosive plugs are utilized is included in this task. The results from the repair criteria and eddy current tasks will define a Phase 2 plugging program that may include thousands of removable plugs.
Initial tube repair work on both OTSG's has been completed with 156 tubes taken out of service.
(4) Eddy Current Inspection - Task 4
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This task includes the work necessary to qualify the eddy current inspection program utilized at TMI-l and to define and perform the total scop. of the eddy current inspection that is required in order to coifirm the condition ~of the OTSG tubes.
The qualification program will determine the adequacy of the probes, gain settings, and other key parameters being utilized in the TMI-l eddy current examination. With regard to the overall eddy current inspection program, the task includes defining the inspection scope in terms of tube identification, tube length to be inspected, and eddy current probe to be used.
As part of this task, the licensee is in the process of l
qualifying a new 4 by 1 absolute probe. This equipment is being developed to resolve eddy current response limitations in the roll transition area when using standard probes.
Due to good correlation with metallography as seen by the 4 by 1 absolute probe, the licensee is in the process of performing eddy current testing of the first 4 inches of all tubes (area bounding the roll transition area).
In addition, eddy current testing using the standard differential probe is being performed from the top
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of the tube sheet down to the ninth support plate. Data from this test are still being collected and evaluated.
(5) OTSG Performance - Task 5 This task involves defining and technically justifying the acceptable reactor power based on plugging large numbers of tubes as defined by the repair criteria and eddy current examination tasks. This task is to be limited to the use of hand calculations and simple computer codes to provide for the definition of reactor power parametrically as a function of number of tubes plugged.
This task will consider the effect of tube plugging on such items as reactor coolant flow, steam generator heat transfer, exit quality, aspiration quality, steam generator shell temperature, hydraulic stability, emergency feedwater effectiveness, and licensing basis accidents.
Quality will be
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-10-considered in terms of its effect on the steam generator, steam piping, and other steam system components such as the turbine generator. There has been little work performed on this task as of April 13, 1982.
(6) Repair Criteria - Task 6 The scope of this task includes the integration of both j
licensing basis considerations and tube stress analysis in order to justify a tube plugging criteria that is different than the
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current Technical Specification requirements. The scope is limited to those eddy current indications that are within the upper tube sheet. This task will provide the technical justification for why tubes in this area are not subjected to burst considerations.
This task will justify that the plugging criteria for defects within the upper tube sheet can be based on the controlling tube axial loads for both operational events and licensing basis accidents.
This task also includes the integration of considerations such as crack growth during operation, uncertainties in the eddy current examination, and results from the 0TSG performance task into the plugging criteria definition. The output of this task will be an overall technical report.
(7) Primary System Review - Task 7 Based on the results of the failure analysis information, this task will involve an evaluation of the susceptibility of other materials in the Reactor Coolant System to the intergranular cracking which has been noted in the steam generator tubes.
The basic approach behind the testing program is to analyze a representative sample of the parts within the reactor system.
The testing process began on April 13, 1982, after the reactor head was removed. Tests will be conducted on 22 major areas of the reactor system over an anticipated 2 to 4 week period.
Various reactor internal components including two fuel assemblies will be removed and inspected.
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Review j
Due to the significance and severity of the OTSG tube degradation,
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the inspector has been closely monitoring the licensee's steam generator repair program task implementation and has been continuing a review to verify the below listed items.
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Accuracy of information related to the event submitted to NRC
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Stability of plant conditions, including provisions for decay heat removal
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Adequacy of licensee inspection (NDE) and maintenance activities
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Proper review and implementation of radiological controls including ALARA (As Low As Reasonably Achievable) considerations
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Selected sections of the following documents were reviewed.
Radiation Work Permits (RWP's) associated with entries into the
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OTSG's
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ALARA review conducted for reactor vessel head removal and eddy current testing of OTSG's Refueling Procedure (RP) 1505-3, "Special Precautions for Fuel
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Handling," Revision 1
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RP 1504-7, " Closure Head Removal," Revision 4
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RP 1504-8, " Reactor Upper Plenum Removc1," Revision 3
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AP 1106-16, "0TSG Secondary Fill, Drain, and Lay up,"
Revision 20 In addition to conducting interviews with responsible licensee personnel, the inspector attended several licensee Steam Generator Tube Leak Task Force meetings to assess the status of actions being taken.
The inspector also conducted field observations of eddy current testing on both OTSG's on several occasions, reactor vessel head removal and reactor vessel upper plenum removal.
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Findings The inspector held discussions with licensee management to review the planned inspection of two unirradiated fuel assemblies required by Task 7.
This inspection requires removal of two fuel assemblies without flooding the refueling transfer canal.
The inspector
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discussed his concerns of the possibility of inadvertent removal of an irradiated fuel assembly and potential personnel overexposure.
The licensee acknowledged the inspector's concerns and described the precautions which were planned to be included in fuel handling procedures. The inspector reviewed applicable Technical Speci-fications and identified no requirements that would prohibit the plr i fuel removal and inspection. The inspector had no further comments concerning this matter.
The inspector also reviewed selected portions of licensee documents and observed portions of the OTSG repair program to verify that the requirements stated above were met oy the licensee.
The inspector determined that the licensee was exercising adequate control in preparation and conduct for this portion of their program.
During an April 7,1982, meeting with the NRC staff in Bethesda, fiaryland, licensee management updated the status of the work of each of its task groups.
During subsequent onsite meetings with licensee management, various repair methods were discussed.
The repair may involve either a new mechanical roll or a hydraulic roll to isolate
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-12-the affected area. Other repair options are being considered. At the close of this inspection the tube failure and repair program was still being developed. The licensee's continuing efforts to determine the cause of the tube degradation and complete adequate repairs will be reviewed during subsequent NRC inspections.
4.
TMI-l Restart Modifications - Implementation a.
General The inspector reviewed three facility modifications which are required to be completed prior to TMI-1 restart to verify that the new designs are provided consistent with the following items.
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licensee commitments stated in the TMI-l restart report, " Report in Response to NRC Staff - Recommended Requirements for Restart of Three Mile Island Nuclear Station Unit 1"
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requirements delineated in NUREG 0680 (and supplements), "TMI-l Restart Evaluation Report," to comply with NRC Commission Order of August 9, 1979
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industry codes and standards to which the facility was constructed (or as approved by NRR)
applicable regulatory guides
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the licensee's QA program
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applicable licensee procedures for modification design, procurement, installation, and construction testing For each of the listed modifications the inspector verified that the
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modification task w.s installed in accordance with the approved design based upon (1 observationofcompletedwork,(2) examination of installation records, (3) review of NDE and/or other inspection
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records, and (4) review of related modification task documentation.
Specific modification task observations and records reviewed by the i
inspector are identified below.
b.
Modification Task RM-13D, Emergency Feedwater (Alternate Manual Control)
(1) References
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TMI-l Restart Report Section 2.1.1.7
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NUREG 0680 Item lb, Emergency Feedwater Independent of Integrated Control System (ICS)
(2) Modification Description Task RM-13D modified the Emergency Feedwater (EFW) System by providing a separate manual EFW control station independent of ICS, for each EFW control valve in the control room, powered from redundant emergency sources. This was accomplished by a
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-13-Bailey Type T640, for each EFW control valve. The control staticn is located in the control room and allows the operator to manually set a 10 volt control signal into the voltage /
pneumatic converter mode in order to control the position of the EFW control valve. The manual control station is powered from a 115 volt, 60 hertz supply, independent of the ICS power supply.
The plant configuration changes associated with Task RM-130 included locating and wiring two manual loader stations, two associated selector switches, and routing / wiring eight electrical circuits.
(3) Documentation Reviewed /0bservations The inspector reviewed the following documentation related to Task RM-13D.
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System Design Description (SDD) 424A, Revision 0, Division II, TMI-l Emergency Feedwater System
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Engineering Change Modification (ECM) S-077, Revisions 0-2, Electrical, Add control selector switching and manual loader stations for EF-V-30A and EF-V-30B
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QC inspection reports
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Cable pull / termination sheets
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Work authorization notices
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Design change notices for design verification
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TMI Tie-In Authorization, ECM S-077, Revisions 0-1
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Engineering Management Procedure (EMP)-019, " Plant Modifications Engineered by Plant Engineering"
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Maintenance Procedure (MP) 1420-FB-1, Revision 6, " Fire Barrier Penetration Seal Repair / Installation"
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MP 1420-EL-2, "Preoperational Startup Testing of Electrical Equipment"
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Construction test data associated with Test Procedure (TP) 250/1.1, Verification of Operability of Valves EF-V-30A and EF-V-30B with Backup Manual Loaders and TP 250/2, Electrical Test
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GPUNC Modification Turnover Package 077, accepted December 11, 198.
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-14-In addition to the above documentation review, the inspector conducted a system walkdown of the plant modifications associated with Task RM-13D.
The inspector observed the installed equipment and verified the component location and installation was as described in applicable modification documentation.
l (4) Findings (a) Based on the modification documents reviewed by the inspector and the observation of installed equipment and components, the inspector determined that Task RM-13D was satisfactorily completed in conformance with the above referenced commitments and requirements.
(b) There are no significant incomplete work list items related to Task RM-13D, which require correction prior to TMI-l restart.
c.
Modification Task RM-8, Relocate Steam Generator _ Level Transmitters (1) Reference
Partial Initial Decision (PID),Section II, Subsection T,
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paragraph 1174 (2) Modification Description Task RM-8 modified the OTSG level transmitters by replacing the carbon steel sensing lines with stainless steel lines and by
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relocating the transmitters at a higher level.
The sensing lines were replaced to minimize potential problems due to corrosion products in the lines. The level transmitters were relocated at a higher position to ensure the transmitters would be above the flood level for a small break LOCA and thus ensure their ability to function.
The plant configuration changes associated with Task RM-8 included repositioning of the OTSG level transmitters on an existing rack, replacement of carbon steel sensing lines with stainless steel lines, electrical modifications to accommodate
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the repositioning of the transmitters, and improvements of support for one transmitter rack.
(3) Documentation Reviewed /0bservations The inspector reviewed the following documentation related to Task RM-8.
Engineering Change Modification (ECM) S-010, Revision 0,
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Relocate OTSG Level Transmitters
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ECM S-029, Revision 0, Replace OTSG Level Tubing with SS Tubing
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-15-ECM S-159, Revisions 0-1, Add Hold Down Plates to Base on
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Electrical Rack
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Purchase requisitions / orders (PR) including PR 86526 QC inspection reports
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Field questionnaires
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Design change notices for design verification
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Nuclear Safety / Environmental Impact rivaluation
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Work Authorization Notice (WAN) 015, Revisions 0-5 WAN 231, Revision 0
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Surveillance Procedure 1302-5.26, Revision 2, Data Sheets (check calibration of level transmitters)
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Anchor Installation Document (AID) - TR 7-A/B
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Installation drawings and "as-built" piping and instrumentation drawings
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Construction test data associated with Test Procedure S/G Level Hydro / Functional Test GPUNC Modification Turnover Package RM-8, accepted
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August 7, 1981.
In addition to the above documentation review, the inspector conducted a system walkdown of the plant modifications associated with Task RM-8.
The inspector observed the installed equipment and verified the component location and installation was as described in applicable modification documentation.
(4) Findings Based on the modification documents reviewed by the inspector and the observation of installed equipment and components, the inspector determined that Task RM-8 was satisfactorily completed in conformance with the above referenced commitments and requirements, except as described below.
During a walk-through inspection inside the primary containment, the inspector noted a section of piping that appeared not to be supported as required by Interim Drawing B-308-804 notes. The drawing note required transmitter sensing lines to be supported every 5 feet.
From the drawings attached to the ECM package, the inspector was unable to determine whether the section of piping in question fell under the scope of the work package or how piping outside the scope of the task are to be supported.
The inspector also noted two valves which were not supported in
-16-accordance with Interim Drawing B-308-886, Revision lA-0, notes.
The drawing note implied that all valves are to be supported within 2 inches of the valve itself. The note did not discuss the situation where the valve piping run was less than 2 inches from a properly supported transmitter.
The inspector was also unable to determine from the other documentation whether this case was acceptable. The inspector identified these findings as unresolved items which will be subject to further review during a subsequent NRC Region Inspection (50-289/82-03-01).
d.
Modification Task NM-39, Radwaste Building Foundation for Hittman Solidification System (1) References
--
TMI-l Restart Report Station 7.2.4, 7.3.1.3.2 NUREG 0680 Item 4, Separation of TMI-l Operations from
--
TMI-2 (Solid Radwaste Treatment System)
(2) Modification Description Task NM-39 provided a concrete slab to support operation of the Hittman Solidification System.
The Hittman System will treat solid radwaste from TMI-1. The system is necessary to separate TMI-l operations from TMI-2 operations.
The plant configuration changes associated with Task NM-39 included the addition of a concrete slab on the south side of the Auxiliary Building between two existing 1.oading docks. The slab is designed to support the Hittman System equipment and the building in which the equipment is installed.
(3) Documentation Reviewed / Observations The inspector reviewed the following documentation related to Task NM-39.
,
--
System Design Description (SDD) 151A, Revision 0, Radwaste Building Foundation for Hittman Solidification System
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Engineering Change Modification (ECM) 189, Revision 0, Structural - provide a raft / mat foundation for proposed Hittman System located adjacent to Auxiliary Building loading docks
--
Purchase requisitions / orders (PR) including PR 83580
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QC inspection report data sheet
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Work authorization notices
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Work Request #C44t;
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-17-
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Inter-office memorandum Concrete placement checklist sheet
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Concrete compression test data sheet
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TMINS Quality Assurance Project Procedure - TMI-lS-03,
--
Important to Safety Material Nonconformance Report Installation drawings
--
GPUNC Modification Turnover Package ECM-189, accepted
--
October 2, 1981.
In addition to the above documentation review, the inspector observed the installed modification associated with Task NM-39 and verified that the location and installation was as described in applicable modification documentation.
(4) Findings (a) The inspector noted that the modification was redesignated as "Important to Safety (ITS)" after the construction of the modification was complete. On the turnover package incomplete work list (IWL), the quality control organization noted that the proper QC coverage had not been performed during installation.
The lack of QC coverage was due to the modification not being ITS at the time of installation.
An engineering evaluation was performed by the responsible engineer of the data available from samples and tests performed. The engineer concluded that based on the available information, the installation was satisfactory for ITS.
The engineer's evaluatic.1 was reviewed and accepted by the quality assurance and quality control organizations. The engineering evaluation was documented only by a memorandum stating that the evaluation was done and the installation satisfactory. The inspector questioned the engineer concerning the scope cf the evaluation and the data utilized. The depth of the
!
evaluation appeared to be adequate for the task. The
!
inspector had no further questions in this area.
l Based on the modification documents reviewed by the inspector, the observation of installed equipment and components and satisfactory resolution of the above concern, the inspector determined that Task NM-39 was satisfactorily completed in conformance with the above
,
referenced commitments and requirements.
l (b) There are no significant incomplete work list items related l
to Task NM-39 which require correction prior to TMI-l restart.
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j 5.
TMI-l Restart Modifications - Task Status
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The inspector held discussions with licensee management representatives and reviewed licensee scheduling / status documentation for facility
'
modifications, which are required to be completed prior to or following i
TMI-l restart.
The purpose of the review was to determine that
'
modification task completion status (completed /in progress / scheduled, etc.) was consistent with commitments and requirements delineated in the
following documents.
TMI Restart Report, " Report in Response to NRC Staff - Recommended
--
Requirements for Restart of Three Mile Island Nuclear Station Unit 1" NUREG 0680 (and Supplements 1, 2 and 3) "TMI-l Restart Evaluation
--
,
Report," to comply with NRC Commission Order of August 9,1979
'
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ASLB Partial Initial Decision dated August 27, 1981, " Procedural Background and Management Issues" ASLB Partial Initial Decision dated December 14, 1981, " Plant Design
--
and Procedures and Separation Issues"
'
NUREG 0746 (and Supplement 1)
--
--
ASLB Partial Initial Decision dated December 14, 1981, " Emergency Planning Issues"
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NRC letter to Met-Ed dated April 22, 1981, " Safety Evaluation Reports for Items Contained in NUREG 0694"
--
NRC letter to Met-Ed dated April 22, 1981, " Safety Evaluation Reports for Items Contained in Enclosure 1 to NUREG 0737"
--
NUREG 0752 (and Supplement 1), " Control Room Design Review Report for TMI-1"
--
selected IE Bulletins and related licensee responses selected licensee event reports and related Technical Specifications
--
--
selected License Orders and Technical Specifications Amendments
The findirgs of the modifications task status review are summarized in the attached TMI-l Modifications Cross Reference Listing - Task to Requirement. Modification status for each task is identified on that listing by indicating the modification completion progress according to the following categories: engineering, installation, QC inspection, testing, incomplete work list (IWL) items, at. plant, and accepted.
Modification inspection status is also summarized in the attachment. As of April 13,1982,19 of the 86 modification tasks currently required for TMI-l restart have been completed.
The attachment will be updated approximately bi-monthly, and modification task status information will be provided in applicable NRC Region I inspection reports.
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6.
Unresolved Items
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I Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations, or deviations. An unresolved item disclosed during the inspection is discussed in paragraph 4.c(4).
7.
Exit Interview
Meetings were held with senior facility management periodically during the course of the inspection to discuss the inspection scope and findings.
The inspectors met with the licensee representatives (denoted in
.
paragraph 1) at the conclusion of the inspection on April 13, 1982, and summarized the purpose and scope of the inspection and the findings. The licensee representatives acknowledged the findings.
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THI-l MODIFICATIONS Revision 2 CROSS REFERENCE LISTING April 13,1982 (TASK TO REQUIREMENT)
Page 1 of 10
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NRC MODIFICATION
. INSPECTION
[ifASK N0. B/A NO.
DESCRIPTION REQUIREMENT STATUS STATUS EM 7 -
412069 Evaluate and Modify TMI-l Block Walls IE Bulletin QC Inspection /
Partially
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80-11 At Plant Compl eted 0 LM-l/
412044 TSAT Meter Indicator and Alarm NUREG-0680 At Plant Partially (8-2.1.3.b)
Compl eted LM-38'
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o LM-2 412044 RCS Wide' Range Temperature NUREG-0680 At Plant Partially (8-2.1.3.b)
Compl eted
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LM-7A 412073 RB Spray System Modifications TSCR No.107/
IWL Items None TS A No. (TBD)
O LM-8 C 412047 RB Sump and Flood Level Instrumentation NUREG-0680 QC Inspection /
None (LT. Add'l #2)
IWL Items o LM-9 120002 Raise Pressurizer Level Transmitter PID (IIT, Accepted None 11174)
O LM-13 A 412024 Emergency Feedwater Cavitating Venturies NUREG-0680 At Plant Completed except (3-2.1.7.a)
IWL/ Testing o LM-13B 412024 Remove FW Latch from EF-V30A/B PID'(IIQ, Engineering None 1 1064)
+ LM-13X 412024'
EFW - Redundant Control and Block Valves NUREG-0680 Engineering N/A (8-2.1.7.a)
+ LM-13X 412024 EFW Auto Start - Low OTSG Leve.l.
NUREG-0680 Engineering N/A (8-2.1.7.a)
+ LM-13X 412155 Isolation of Main Feedwater on OTSG Overfill PID (IIQ, Engineering
- N/A.
1 1037)
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+ LM-13X 412024 Upgrade Main Steam Rupture Detection System PID (IIQ, Engineering N/A
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1 1037)
J NOTE See last page for explana ion of modi fication status symbols and categories.
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Revision 2 TH1-1 MODIFICATIONS _
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(TASK--10 REQUIREMENT)
Page 2 of 10
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NRC MODIFICATION INSPECTION
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TASK NO. B/A NO.
_
DESCRIPTIO:1.
REQUIREMENT
.
STATUS STATUS N
o LM-16 ', 412001 Fuel Handling Building Isolation /HVAC Chase NUREG-0680 (4)
IWL Items /
None Accepted
"+
,y.,
o LM-16A 412001 Fuel Handling Building Isolaf. ion Damper Control NUREG-0680 s(4)
Test /IWL Items None NUREG-0680[4)
Engineering N/A
+ LM-16B 412336 Fuel %A1ing Building ESF Ventilation System
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s Pressurizer NUREG-0680 b Installation /
None s
o LM-21A 412021 ( RCS Hi'M Aint Vents 7
- (LT Add'l #4)
QC Inspection t.
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+ LM-21B 412011 RCS Mgh rc ir,; Vents - RB Elect Penetration N REG-0680 On Hold N/A~
(LT Add'l #4)
+ LM-21B 412021 RCS High Point Vents - Hot Legs /RV Head NUREG-0680 On Hold N/A (LT Add'l #4)'-
' NUREG-0680 Installation Partially.
- LM-23 412013 High Range Containment Monitoring (8-2.1.8.b)
Completed
'A LM-24A 4 2032 Post Accident RCS Sampling (Pzr. Sample Bomb)
NUREG-0680 On Hold N/A.
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(8-2.1.8.a)
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LM-24A 412032 Post Accident RCS Sampling (Shielding)
NUREG-0680 Installation /
- Partially ( 8-2.1. 6. b)
QC Inspection Comple ted
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- lM-24B 412032 Post Accident RB Atmosphere Sampling NUREG-0680 Installation None
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(8-2.1.8.a)
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Post Accident Effluedt.Manitoring
' Gaseous NUREG-0680 QC Inspection None oLM55A 412013 (8-2.1.8.b)
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o LM-25B 412013 Post Accident Effluent. Monitoring - Iodine /Part.
NUREG-0680 IWL Items None_
(8-2.1.8.b)
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TMI-l MODIFICATIONS Revision 2
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CROSS REFERENCE LISTING April 13,1982 (TASK TO REQUIREMENT)
Page 3 o f 10 -
NRC MODIFICATION INSPECTION TASK NO.
B/A NO.
DESCRIPTION REQUIREMENT STATUS STATUS
- LM-26 A 412013 Post Accident Containment Pressure Indication NUREG-0680 Installation /
None
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(LT Add'l #1)
QC Inspection
- LM-26B 412013 Post Accident Monitoring - Containment.
NUREG-0680 Engineering /
None Hydrogen (LT Add'l #3)
In" =11 a tion
- LM-29 412064 High Rad. Monitoring Alarm System NUREG-0746
.esting None Improvements (II. H)
LM-30 412087 Automatic Reactor Coolant Pump Trip NUREG-0680 Engineering N/A
(79-05C LT #1)
- LM-32 412059 Loads of Safeguard (IE) Buses Tech. Specs.
Accepted None (LER 80-01 )
- LM-32A 412059 1B E.S. MCC Feed Breaker Upgrade Tech. Specs.
Installation None (LER 80-01)
- LM-38 412017 Remote Shutdown Panels NUREG-0680 Test /IWL Items None ( LT-1 )
+ LM-38 412017 Remote Shutdown Panels Appendix R
[ngineering N/A
- LM-42 412008 Decay Heat Check Valve Leakage Monitor License Order IWL Items /
None (4/20/81)
At Plant
- LM-43A 412028 ICS/NNI Loss of Power Indicator & Annunciator NUREG-Obo0 At Plant Partially (LT-1)
Completed
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- LM-4 3B 412028 Atm.. Dump / Pressurizer Spray / Turbine Bypass IE Bulletin IWL Items /
None Valves 79-27 At Plant
- LM-43C/ 412028 Indication of Critical Plant Parameters IE Bulletin IWL Items /
None 79-27 Testing LM-36
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TM1-1 MODIFICATIONS Revision 2 CROSS REFERENCE LISTING April 13,1932
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(TASK TO REQUIREMENT)
Page 4 of 10 NRC MODIFICATION INSPECTION TASK NO.
B/A NO.
DESCRIPTION REQUIREiiENT STATUS STATUS LM-51 A 412020 Shielding Design Review Modifications NUREG-0580 IWL Items /
None (8-2.1.6.b)
At Plant LS-6 412015 Control Room Control Console Arrangement NUREG-0152 (5)
Installation None
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JT-X X 120002 RPS High Pressure Trip Setpoint NUREG-0580 Accepted Completed (79-05B-3)
o JT-XX 113202 Hard Piping for Hittman Solidification System NUREG-0580 (4)
Engineering /
None Installation
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o UM-17 120002 IGSCC (Pipe Cracks)
IE Bulletin At Plant Partially 79-17 Compl eted o NM-31 412342 Anchor Bolt Inspections and Repairs IE Bulletin Installation /.
Partially 79-02 Accepted Completed o NM-34 120002 480V Bus Trip TSCR No. 60A/
Accepted None TSA No. 70 o HM-35 412342 Pipe Support Location Inspection and Repairs IE Bulletin Installation /
Partially (Safety Factor > 2)
79-14 Accepted Completed 0 NM-39 412191 Foundation for Hittman Solidification System NUREG-0680 (4)
Accepted Completed o NM-40 412041 Fire Protection TS Amendment Install ation'
Partially No. 50/ App. R to Accepted Compl e ted
+ NM-41 432009 Perimeter Security Modificat' ions Proposed Security Installation /
Hone Plan Rev.
Testing o NM-47 412086 Control Room Emergency Phone (Hot Line)
NUREG-0746 At Plant None (II.F)
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TMI-l MODIFICATIONS Revision 2
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CROSS REFERENCE LISTING April 13,1982
(TASK TO REQUIREMENT)
Page 5 of 10
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NRC MODIFICATION INSPECTION TASK NO.
B/A NO.
DESCRIPTION REQUIREMENT STATUS STATUS o NM-48 412010 Seismic Supports for Conduit NUREG-0680 Installation None (various)
to Accepted o NM-49 120002 Modify Internals:
Crane Tilting Disc Tech. Specs.
QC Inspection Partially Check Valves (LER 80-03)
Completed o HM-51 120002 Modi fy Internals:
Feedwater to OTSG Tech. Specs.
Accepted Partially Check Valves (LER 80-03)
Completed PM-l 412060 Replace Containment Isol. Valve RB-V7 Tech. Specs.
At Plant /
Compl eted (LER 80-07)
Accepted PM-3 412074 Purge and Vent Valves (0 pen Limits)
NUREG-0737 Installation None (SER II.E.4.2)
PM-4 412003 PORV/ Code Safety Tail pipe Temperature NUREG-0680 At Plant None (8-2.1.3.a)
PM-8A 412065 TSC Computer Operations /CRT/ Intercom NUREG-0680 IWL Items None (8-2.2.2.b)
+ PM-8C 432014 Near Site Emergency Operations Facility NUREG-0746 Engineering N/A (II.H)
+ PM-80 432014 Environmental Command Center Upgrade NUREG-0746 Engineering N/A (II.H)
o PM-10 412015 Control Room Human Factors Upgrade NUREG-0752 Installation /
None (various)
Testing o PM-ll 412093 Remote Shutdown Panel Area Communications NUREG-0752 (9)
At Plant None o PM-18 412086 Telephone Equipment Vital Power Feed NUREG-0746 At Plant None (II.F)
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TMI-l MODIFICATIONS Revision 2
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CROSS REFERENCE LISTING
- April 13,1982
)
(TASK TO REQUIREMENT)
Page 6 of 10
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I NRC MODIFICATION INS PECTION TASK NO.
B/A NO.
DESCRIPTION REQUIREMENT STATUS STATUS
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i PM-32 412070 Relay Room Catwalk Seismic Upgrade Tech. Specs.
Engineering /
Partially (LER 81-11)
QC Inspection Completed
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RM-1 412049 Heat Shrink Tubing / Butt Splicing Inside RB Bulletin Installation Partially 79-01 B Completed
'
i o RM-3B 412051 Reactor Trip on Loss of Feedwater NUREG-0680 At Plant Partially (79-058-5)
Completed O RM-4 A 412056 Connect Incore Thermocouples to Computer NUREG-0680 At Plant Partially (8-2.1.3.b)
Completed
!
i
! o RM-4B 412056 Incore Thermocouple Backup Display System NUREG-0680 Installation None (8-2.1.3.b)
,
0 RM-5B 412052 R.B. Isolation on High Radiation NUREG-0680 At Plant Partially (8-2.1.4)
Completed O RM-5C 412052 R.B. Isolation on Reactor Trip /High R.B.
NUREG-0680 IWL ltems Partially
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Pressure /HPI (8-2.1.4)
Completed
- RM-5D 412052 R.B. Isolation on NSCCC and ICCC Pipeline NdREG-0680 IWL Items None
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Break (8-2.1.4)
RM-7-412054 Computer Modification-High Speed Printers /CRT NUREG-0752 (2)
IWL Items None o RM-8 120002 Raise Steam Generator Level Transmitter PID (IIT, Accepted Partially 1 1174)
Completed
.
RM-9 120002 Raise Setpoint of Power Operated Relief Valve NUREG-0680 Accepted Partially (79-058-3)
Completed o RM-10/ 412042 PORV and Safety Valve Position Indication NUREG-0680 Testing /
Partially ( 8-2.1. 3. a )
At Plant Completed LM-8 C
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TMI-1 MODIFICATIONS Revision 2
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CROSS REFERENCE LISTING April 13,1982 (TASK TO REQUIREMENT)
Page 7 of 10 NRC MODIFICATION INSPECTION TASK NO.
B/A NO.
DESCRIPTION REQUIREMENT STATUS STATUS
- RM-12 412043 Hydrogen Recombiner' Installation NUREG-0680 IWL Items /
Partially (8-2.1.5.a)
At Plant Compl eted
- RM-13A 412012 Backup Instrument Air. System NUREG-0680 Accepted Completed (ST la-2)
except Testing
- RM-13 412012 Emergency Feedwater Auto Start (Control Grade)
NUREG-0680 Accepted Compl eted
.
A/E (ST la-1, la-3)
- RM-13B 412012 Emergency Feedwater Flow Measurement NUREG-0680 IWL Items /
None (8-2.1.7.b)
At Plant
- RM-13C 412012 Emergency Feedwater Control Valves -
NUREG-0680 Accepted Compl eted
" Fail Open" (ST la-2)
- RM-13D 412012 Emergency Feedwater ( Alternate Manual Controls)
NUREG-0680 (lb)
Accepted Completed
- RM-13 412012 Emergency Feedwater Auto Start (Safety Grade)
NUREG-0680 Accepted Compl eted (8-2.1.7.a)
except Testing E/A
- RM-13G 412012 Emergency Feedwater C.R. Annunciator Panel NUREG-06:0 Accepted Completed (ST la-8)
except Testing
- RM-13H 412012 EFW Pump Turbine Safety Valves Lift Setpoints NUREG-0680 Accepted Compl eted (la-Add'1 6)
- RM-13H 412012 Two Hour Backup Air Supply NUREG- 0680 Installation /
None (la-Add'1 6)
Accepted
- RM-131 412012 Condensate Storage Tank Low-Low Level Alarm NUREG-0680 Testing Partially (la-Add'l 1)
Completed
+ RM-13I 412012 CST Low Level Alarm-Independent Power NUREG-0680 Engineering N/A Supplies (8-2.1.7.a)
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TMI-l MODIFICATIONS Revision 2
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CROSS REFERENCE LISTING April 13, 1982 (TASK TO REQUIREMENT)
Page 8 of 10
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NRC MODIFICATION INSPECTION TASK NO.
B/A NO.
DESCRIPTIO_N REQUIREMENT STATUS STATUS o Ril-13J/ 412012 OTSG Level in C.R. Ihdependent of ICS NUREG-0680 IWL Items None (8-2.1.7.a)
LM-38 o RM-14 412072 High Pressure Injection Cross Connect PID (II0, Accepted Partially 1 943)
Completed o RM-16 412063 Pressurizer Heater Emergency Power NUREG-0680 IWL Items Pa rtially (8-2.1.1 )
Completed -
l 0 RM-17 412029 Power Supply to ICS/NNI System NUREG-0680 Accepted Partially (LT-1 )
Completed 0 RM-21 412061 Relocation of Radiation Monitors NUREG-0G80 IWL Items /
Hone (8-2.1.8.b)
At Plant o TS-147 220010 Independent TMI-2 Sampling System NUREG-0680 (4)
Accepted None o WG-56 432061 Low Level Solid Waste Storage Facility NUREG-0680 (5)
Installation None 120011 Steam Generator Long Term Tube Repairs Tech. Specs.
Engineering None
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(LER 81-13)
0 --
120012 Steam Generator Tube Plugging Tech. Specs.
Engineering /
None (LER 81-13)
QC Inspection
120016 Primary System Review Tech. Specs.
Engineering None
--
(LER 81-13)
412019 Pressure Temperature Plot Part-Task Trainer PID (IIM, Accepted Completed
-
Condition 4)
+ --
412006 Replicate Simulator PID (IIM,
.
Engineering N/A Condition 5)
+ --
412019 Basic Principles Trainer PID (IIM, Engineering N/A Condition 6)
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THI-l MODIFICATIONS CROSS REFERENCE LISTING April 13,1982
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Page 9 of 10
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NRC MODIFICATION INSPECTION TASK NO. B/A NO.
DESCRIPTION REQUIREMENT STATUS STATUS
+
412023 Reactor Vessel Water Level Instrument NUREG-0680 Engineering N/A (8-2.1.3.b)
--
+ --
41203)
EFW Control Valves-Environmental Quali fication NUREG-0680 Engineering N/A (l a-Add'1 7 )
412031 Material Qualification Replacement Items PID (IIT, Engineering N/A
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TSA No. 56/
Engineering None o
412088 Overpressure Protection at Low RC Temperatrue TSCR No. 74
--
o 412068
. Control Room Habitability Upgrade TAP III.D.3.4 Engineering None
-
412244 Remote Shutdown Capability Appendix R Engineering N/A
+
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412250 Control Room Lighting & Glare Reduction NUREG-0752 Installation None NUREG-0752 Installation None o
412251 Alarm System Reformat
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+
432107 Protected Area Barrier Intruder Det. Sys.
Proposed Security Engineering N/A P1 n Rev.
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THI-l MODIFICATIONS Revision 2
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CROSS REFERENCE LISTING April 13, 1982 (TASK TO REQUIREMENT)
Page 10 of 10
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EXPLANATION OF MODIFICATION STATUS SYMBOLS AND CATEGORIES
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Task Requirement Status Symbol s:
.
o Task scheduled for pre-restart installation.
.
O Task scheduled for post-restart installation.
.
d Task pending NRR review of proposed design /SBLOCA analysis.
New Task / Engineering not issued.
Modi fication Status Categories _:
.
Engineering - Modification in approval process (funding / engineering design / Quality Control (QC) hold and witness review / safety evaluation review / work package planning)
Installation - Work package released for installation by Maintenance and Construction (M&C), Plant Maintenance, or Startup and Test organizations QC Inspection - Completed modification installation package to M&C Planning for release to QC/Startup and Test or package at QC for inspections Test / Testing - Work package at Startup and Test / Plant Engineering for testing, or M&C to provide final as installed drawing package IWL Items - Modification package includes incomplete work list (IWL) items preventing release to plant staff for acceptance At Plant - Modification package at plant for acceptance review or package in final turnover signoff process Accepted - Modification package accepted by plant NOTE: Summary and detailed modification status for the above tasks with regards to engineering, production (installation) planning and scheduling, production installation in progress, QC certification, modification testing, and plant acceptance is contained in the following licensee documents:
(1)
TMI-l Restart Modifications Status Report.
(2)
TMI-l M&C Schedule - Modifications in Production.
(3)
TMI-1 Restart Project Status Meeting Agenda.
.