IR 05000289/1996007
| ML20135F197 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 11/18/1996 |
| From: | Marilyn Evans, Hansell S, King E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20135F140 | List: |
| References | |
| 50-289-96-07, 50-289-96-7, NUDOCS 9612120433 | |
| Download: ML20135F197 (26) | |
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U. S. NUCLEAR REGULATORY COMMISSION f
REGION I
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Docket No.
50-289 i
License No.
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i Report No.
96-07 Licensee:
GPU Nuclear Corporation Facility:
Three Mile Island Station, Unit 1
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Location:
P.O. Box 480 Middletown, PA 17057
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Dates:
September 29,1996 - November 18,1996 Inspectors:
Michele G. Evans, Senior Resident inspector Samuel L. Hansell, Resident inspector Edward B. King, Physical Security inspector Joseph L. Nick, Radiation Specialist Approved by:
Peter W. Eselgroth, Chief Reactor Projects Section No. 7 l
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9612120433 961206 l
PDR ADOCK 05000289
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EXECUTIVE SUMMARY Three Mile Island Nuclear Power Station Report No. 50-289/96-07 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a seven week period of resident inspection; in addition, it includes the results of an announced inspection in the area of physical security for unit 1.
Plant Operations The auxiliary operators performed the required tour and instrument readings as e
required by the operation management's program and written expectations. The auxiliary operators displayed a methodical, questioning and alert awareness on the plant tours (Section 01.1).
e The licensee's actions in response to a generic seismic concern related to Westinghouse 4160 Volt electrical circuit breakers were excellent. Engineering's use of the nuclear network to find out about the issue was good. The one hour report to the NRC was prompt and well coordinated between operations and engineering personnel. Immediate actions were taken by operations to relocate and secure the three susceptible breakers at an approved seismic location and to alert personnel to the concern. The licensee's additional long term actions were either planned or had been implemented and were appropriate (Section 01.2).
The borated water storage tank (BWST) was properly aligned per the plant's e
procedures to perform its design safety function. The system documentation and equipment material condition were good. The chemistry technician's sample techniques and procedure use were excellent for the BWST boron analysis (Section O 2.1 ).
e A thorough nuclear safety assessment audit of the operations department revealed a missed emergency feedwater surveillance test. Once discovered, the test was performed satisfactorily within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as required by Technical Specifications (Section 07.1).
Maintenance e
The surveillance test activities observed during this inspection were performed satisfactorily and demonstrated that the associated systems could perform their design safety functions. Control room operators consistently used the new " touch the tag" self checking method before surveillance test actions (Section M1.1).
e The seismic monitor installation, operation, and procedure documentation reflected that the equipment met the design requirements contained in the updated final safety analysis report (Section M1.1).
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The rnake-up system motor operator valve work was performed at a controlled pace
and all work activities were completed on time as planned, which minimized the time that the safety system was out of service (Section M1.1).
Security equipment repairs were being completed in a timely manner and the use of
compensatory measures was found to be appropriate and minimal (Sections M1.1 and S2.3).
- The inspectors identified two examples where the scaffold construction and inspection written procedure was not properly implemented for two safety related work activities. This issue is considered a violation of the Unit 1 Technical Specifications. In both cases immediate actions were taken to correct the problems (Section M3.1).
Enain**rina
The excellent questioning attitude of the make-up system engineer revealed a potential design vulnerability for a small break loss of coolant accident (SBLOCA)
coincident with the loss of the DC power resulting in make-up (MU) valve MU-V-18 failing open. Engineering initiated a quality deficiency report to document the safety concern and ensure proper resolution of the issue. A plant review group evaluation concluded that the condition was a missed surveillance; MU-V-17, the valve in series with MU-V-18, was stroked closed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to address the SBLOCA concern (Section E1.1).
- The environmental and radiological control engineering review of the tritium leak associated with the borated water storage tank was comprehensive and detailed (Section O2.1).
Plant Suncort
The areas toured by the inspectors were well maintained and radiological housekeeping was generally very good. Some minor concerns were brought to management's attention for resolution and correction (Section R2.1).
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Very good radiological controls by all workers involved in the removal of equipment from the spent fuel pool were observed (R4.1).
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The TMI response during the annual medical response exercise was very good.
j Appropriate contamination controls were used during preparation and transport of the potentially contaminated, injured individual to the local hospital (Section P1.1).
- TMI has maintained an effective physical security program. Two previously j
identified items, both involving the licensee's failure to provide adequate compensatory measures during maintenance activities were closed (Sections S1, S2.1, S2.2 and S8).
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Security Management support is ongoing as evidenced by the procurement and
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installation of three new metal detectors, procurement of additional training aids to add realism during firearms train!ng, and the hiring of five additional site protection
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l officers. Management controls for identifying, resolving, and preventing programmatic problems were effective (Sections S2.1, SS, S6, S7.1 and S7.2).
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TABLE OF CONTENTS EX EC U TIV E S U M M A RY............................................. ii TABLE O F CO NT ENT S.............................................
v 1. Operations
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Conduct of Operations.................................... 1 01.2 Seismic Qualifications for 4160 Volt Circuit Breakers......... 1
Operational Status of Facilities und Equipment................... 3 O2.1 Engineered Safety Feature System Walkdown.............. 3
Quality Assurance in Operations
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07.1 Nuclear Safety Assessment Annual Audit and (Updated)
Unresolved item 50-289/96-08-01
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ll. Maintenance
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M1 Conduct of Maintenance
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M3 Maintenance Procedures and Documentation.................... 7 M3.1 Scaffold Construction for Safety Related Equipment.......... 7 111. E n g i n e e ri n g................................................... 9 El Conduct of Engineering
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E1.1 Make-up Valve Small Break Loss of Coolant Accident Concern and (Updated) Unresolved item 50-28 9/9 6-08-01............ 9 j
IV. Plant Support
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R2 Status of RP&C Facilities and Equipment...................... 10
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R4 Staff Knowledge and Performance in RP&C....
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R6 RP&C Organization and Administration
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P1 Conduct of EP Activitie s.................................. 12 S1 Conduct of Security and Safeguards Activities.................. 13 S2 Status of Security Facilities and Equipment
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SS Security and Safeguards Staff Training and Qualification
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S6 Security Organization and Administration...................... 16 S7 Quality Assurance in Security and Safeguards Activities........... 16 S8 Miscellaneous Security and Safeguards issues.................. 17 V.
M a n a g e m e nt M e e ting s.......................................... 18 X1 Exit Meeting Summary................................... 18 X2 TMl Systematic Assessment of Licensee (SALP) Management Meeting
PARTIAL LIST OF PERSONS CONTACTED............................... 19 ITEMS OPENED, CLOSED, AND DISCUSSED............................. 20 LIST OF ACRONYMS USED
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Report Details Summary of Plant Status Unit 1 remained at 100% power throughout the inspection period.
l. Operations
Conduct of Operations (71707)'
01.1 General Comments Using inspection Procedure 71707, " Plant Operations," the inspectors conducted frequent reviews of ongoing plant operations. In general, the conduct of operations was professional and safety-conscious; specific events and noteworthy observations are detailed in the sections below. Plant management ensured that potential generic safety issues were reviewed and addressed in a prompt timeframe. The most recent positive example was noted for the evaluation of a generic seismic concern related to Westinghouse electrical breakers.
The inspectors accompanied three auxi!iary operators (AOs) during the daily shift tour and equipment log readings. The AOs performed the required tour and instrument readings as required by the operation management's program and written expectations. The AOs displayed a methodical, questioning and alert awareness on the plant tours.
01.2 Seismic Qualifications for 4160 Volt Circuit Breakers a.
Inspection Scoce (40500 and 92700)
The inspectors reviewed a generic seismic concern related to Westinghouse 4160 Volt electrical circuit breakers. TMI received the information from the nuclear network and evaluated the data to determine if the possibility existed that the TMI breakers were l
susceptible to the generic seismic problem.
b.
Observations and Findinas l
On November 11,1996 the licensee was informed through the industr/ network that when certain circuit breakers were in the " racked-out" position the switchgear was no longer seismically qualified. Following licensee engineering department investigation of the issue, they identified three affected breakers in their Class 1E 4160 volt AC switchgear. The 'B'
high pressure injection pump breaker and a spare breaker on the '1D' bus and a spare breaker on the '1E' bus are normally in the " racked-out" position. These circuit breakers
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are Westinghouse model 50-DHP-250. The Westinghouse breakers that were not directly connected to the electrical bus had the potential to move during a postulated seismic event
' Topical headings such as 01, M8, etc., are used in accordance with the NRC standardized reactor inspection report outline. Individual reports are not expected to address all outline topic.
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because they were on metal wheels and not secured inside the breaker cubicle. The licensee's Plant Review Group (PRG) conducted a meeting on November 12,1996 to discuss the issue and determined that the condition was reportable to the NRC as a condition outside the plant design basis, since the plant design requires the Class 1E 4160 volt AC switchgear to be seismically qualified. The licensee reported this condition to the NRC via the Emergency Notification System on November 12,1996 and physically removed the " racked-out" breakers to restore seismic qualification to the affected switchgear. Two of the breakers were secured to seismic anchors in the electrical bus room and the third breaker was moved and anchored to the wallin a non seismic area of the plant.
The inspectors reviewed the li':ensee's corrective actions with regard to this seismic qualification concern. Following the PRG meeting, operations management made an entry in the Night Order Book on November 12,1996 to alert the operators to the seismic concern and the need to remove from the cubicles and store properly any breakers on the
'1D' and '1E' buses which were not fully racked-in. During the '1 A' building spray system outage performed the week of November 18,1996, the inspectors verified that the building spray pump breaker was appropriately removed from the cubicle on the '1D' bus and stored in a seismic location. Maintenance personnel performing maintenance and testing on the pump breaker were aware of the seismic concern.
The licensee intends to implement a modification (Work request #785715) to seismically qualify the breakers in the reS d-out position during the next refuel outage in the Fall of 1997. The inspector questioned engineering regarding how they would ensure that the information would be retained between now and the refuel outage. The engineer stated that he would issue a Temporary Change Notice and a Procedure Change Request to Administrative Procedure (AP) 1002, " Rules for Protection of Employees Working on Electrical and Mechanical Apparatus," to document the issue and provide guidance for breaker removal from service, c.
Conclusions The inspectors concluded that the licensee's actions in response to this recent industry concern were excellent. Engineering's use of the nuclear network to find out about the issue was good. The one hour report to the NRC was prompt and well coordinated between operations and engineering personnel. Immediate actions were taken by operations to relocate and secure the three susceptible breakers at an approved seismic location and to ciert personnel to the concern. In addition, the inspectors found that the licensee's additional long term actions were either planned or had been implemented and were appropriat.
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O2 Operational Status of Facilities and Equipment
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O 2.1 Enaineered Safetv Feature System Walkdown (71707)
a.
Inspection Scope The inspectors used Inspection Procedure 71707 to walkdown accessible portions of the borated water storage tank (BWST) and associated safety system interconnections. The inspection also included a review of the tritium leak from the BWST to the ground water and the observation of the weekly baron sample and analysis.
b.
Observations and Findinos l
Desian Verification
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The inspectors reviewed the updated final safety analysis report (UFSAR) section 9.5.2,
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" Decay Heat Removal /BWST Design Bases," Technical Specification (TS) sections 3.3,
" Emergency Core Cooling, Reactor Building Emergency Cooling and Reactor Building Spray Systems, " and 3.4, " Decay Heat Removal," and the associated emergency core cooling system (ECCS) operating procedures. The BWST parameters, procedures and design bases
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information contained in the UFSAR were up to date and no discrepancies were noted.
The BWST boron concentration was greater than the UFSAR value of 2500 parts per million (ppm), the BWST water level was above the TS minimum value of 350,000 gallons, j
and the BWST/ sodium hydroxide tank differential pressure was within the allowable TS band. All accessi' le valves were in the proper position for the current plant conditions.
o Overall, the system material condition was good with the exception of a small flange leak on the spent fuel pool recirculation line.
l Tritium Leak On September 11,1996, a meeting was held by radiological and environmental engineering to discuss the status of the tritium levels in the site ground water. As a result of this j
meeting, several actions were initiated to determine if the BWST was contributing to the site well water tritium contamination. As a result of these actions, detailed radiological control surveys revealed low levels of contamination on a small area of the BW5i ped and in nearby soil samples. A radiological control inspection revealed a gasket leak located at a pipe flange connected to the BWST.
The leak was approximately 150 milliliters per day from the flange on an eight inch recirculation pipe. The tank and leaky flange are located outside in the yard area which would result in rain water washing the tritium water into the adjacent ground.
Environmental engineering initiated an event or near miss capture form (ENMCF) on September 27,1996, to document the tritium leak from the BWST and determine the plant corrective actions.
Because the BWST tritium levels were about 100,000,000 pico curies per liter (pci/l), it was concluded that the leak could significantly influence ground water tritium levels. The leak was immediately contained and the small contaminated area was cleaned u.
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1 Additional visual inspections of the BWST did not indicate any more leaks.
On October 2,1996, another meeting was held to discuss the additional findings relative to the BWST leakage. While no additional leaks from the BWST had been identified, the initial sample results from three new service wells that were drilled to supply plant makeup water were found to have tritium levels of approximately 8,000 pCi/l. The wells which are located southwest of the BWST area, the direction considered down-field for the water table, provided further evidence that the BWST was the likely source of the tritium levels in the ground water west of the plant. As a result of this meeting, environmental engineering was tasked with providing specific recommendations relative to corrective actions to be taken to address the ground water tritium levels.
The environmental engineering department made the following recommendations in an attempt to resolve the ground water tritium associated with the BWST: 1) inspect all above ground penetrations of the BWST; 2) continue to perform well water samples on a weekly basis to determine changes related to the BWST leak termination and groundwater dynamics; 3) evaluate the installation of a monitoring well adjacent to the BWST to perform routine samples to detect and identify any future tritium problems; and 4)
continue pumping the new service water wells to reduce the tritium concentrations.
The inspectors monitored the plant response related to the identification, trending, and followup actions to monitor the BWST leak and associated tritium leveis in the ground water. The tritium levels were just above the site background levels and below the drinking water and regulatory limits. As of November 20,1996, the tritium levels in the
'A' and 'B' service water wells had dropped to approximately 3,000 to 5,000 pCi/l. The
'C' well has trended higher from an initial value of 3,700 pCi/l to approximately 6,500 pCi/l. The unmonitored release contamination levels were below the NRC reportability values. An NRC regional health physics specialist reviewed the tritium issue to determine if the TMI actions were acceptable. The specialist inspector concluded that the site personnel were taking appropriate actions to resolve and trend the tritium ground water concern. The BWST system engineer performed a detailed inspection of all mechanical joints associated with the storage tank. No additional leaks were identified and the original leak was routed to a permanent drain system that was routed to a permanent plant sump.
Environmental engineering established a comprehensive sampling program to monitor and j
trend the site well water and ground locations.
Boron Samole The inspectors observed a licensee chemistry technician perform a sample collection and boron analysis from the BWST. The technician used procedure N1904.1, " Determination of Boron Using METROHM Model 682 Autotitration System," to perform the analysis. The sample was performed to identify reactivity anomalies greater than 1 % as required by TS section 4.10. The inspectors observed good use of radiological controls and good industry practices during the sample collection. The sample was prepared in accordance with the appropriate laboratory protocols to prevent cross-contamination of samples.
Instrument preparation involved excellent quality controls and was performed according to the procedure. The associated instrumentation was maintained in good working order.
Sample analysis was performed in accordance with the procedure and included appropriate
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use of the instrumentation and equipment. The inspector observed the recording of data and verification that the reactivity was not greater than 1% difference from the previous analysis and previous averages. Boron activity was recorded on a surveillance sheet that required appropriate level of management review. Data was entered into a computer tracking system and the activity was also posted in the control room on a data board.
c.
Conclusions The inspectors concluded that the BWST was properly aligned per the plant's procedures to perform its design safety function. The system documentation and equipment material
condition observed were good. The chemistry technician's sample techniques and procedure use were excellent for the BWST boron analysis. In addition, the environmental
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and radiological control engineering review of the tritium leak associated with the BWST i
was comprehensive and detailed.
Quality Assurance in Operations 07.1 Nuclear Safety Assessment Annual Audit and (Updated) Unresolved item 50-289/96-08-01
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Inspection Scope l
i The inspectors reviewed a quality deficiency report (QDR) and portions of the nuclear safety assessment's (NSAs) routine audit of the operations department. In addition, the inspectors discussed the findings of the audit with the engineer who performed the evaluation.
b.
Observations and Findinas The NSA audit was thorough and included a detailed look at the inservice surveillance test (IST) procedures and program related to the operations department. The diligent work of the NSA evaluator revealed a missed emergency feedwater (EFW) surveillance test related to auxiliary steam (AS) valve AS-V-4 leak test. Valve AS-V-4 is an auxiliary steam stop check valve that isolates AS to the steam driven EFW pump. Normally, main steam supplies the motive force for the EFW turbine and auxiliary steam is isolated to prevent backflow from the main steam to the auxiliary steam supply piping. The frequency of the AS-V-4 leak tight check was erroneously changed from a quarterly test to once a year during a September 1995 procedure revision. NSA initiated ODR No. 962016 to document the problem and ensure adequate corrective actions were completed. An event or near miss capture form (ENMCF) was also completed to track the QDR resolution.
When the missed surveillance was recognized, plant operators completed the test satisfactorily within the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allotted by TSs. The inspector observed the satisfactory performance of the missed test. A surveillance deficiency report was completed by the Shift Supervisor (SS) to document the missed test and ensure management was aware of the problem and associated corrective actions. The SS also initiated a procedure change request to perform the AS-V-4 portion of the EFW test every quarter as required. This licensee identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Polic _
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Conclusions A thorough NSA audit of the operations department revealed a missed EFW surveillance test related to AS-V-4 leak test. Once discovered, the test was performed satisfactorily
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within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as required by TSs.
II. Maintenance M1 Conduct of Maintenance (62707,61726)
M 1.1 General Comments a.
Inspection Scone i
The inspectors observed all or portions of the following maintenance and surveillance work
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activities:
Job Order No. 123823, "'A' Emergency Feedwater Pump Oil Change."
- Job Order No. 123824, "'A' Emergency Feedwater Pump Motor and Electrical Cable insulation Checks."
- Job Order No. 121024, " Sodium Hydroxide Tank Recirculation Pump Motor and Pump Alignment."
- Job Order Nos. 125780 and 125511, "MU-V-14A Motor Operator Valve Clean and inspect Internals and Replace the Grease."
- Surveillance Procedure 1301-1, " Shift and Daily Checks."
Surveillance Procedure 1300-3G, " Steam Driven Emergency Feedwater Pump
and Valves Test."
- Operations Procedure 1104-45R, " Inspect and Test DC Emergency Lights."
- Surveillance Procedure 1303-5.2A, " Emergency Safeguards Train 'A' Logic Channel / Component Test."
- Surveillance Procedure 1303 3.1, " Control Rod Movement."
- Operations Procedure 1107-9, " Station Blackout Diesel Test Run."
b.
Observations and Findinas The scheduled maintenance outage for the 'A' make-up (MU) pump was performed and controlled as planned with good coordination by maintenance, operations, radiological control and engineering. The electrical maintenance technicians were knowledgeable about l
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the motor operated valve (MOV) work associated with MU-V-14A. The MOV work was performed at a controlled pace and when problems arose during valve disassembly, the technicians used a methodical approach to solve the problems. The work was completed on time as planned and minimized the time that the safety system was out of service.
The seismic monitor surveillance was performed satisfactorily and all equipment was installed by design. The system was operated as designed and documented in the UFSAR.
The inspectors' plant walkdown of all seismic components noted that all equipment was in a good working condition.
Control room operators (CROs) consistently used the new " touch the tag" self checking method before writing the equipment parameter on the log sheet for the daily shift surveillances and scheduled surveillance tests.
The physical security inspection noted that security equipment repairs were being completed in a timely manner and the use of compensatory measures was found to be appropriate and minimal.
c.
Conclusions The surveillance test activities observed during this inspection were performed satisfactorily and demonstrated that the associated systems could perform their design safety functions. The MU system MOV work was performed at a controlled pace and all work activities were completed on time as planned which minimized the time that the safety system was out of seivice. The physical security inspection noted that security equipment repairs were being completed in a timely manner and the use of compensatory measures was found to be appropriate and minimal.
M3 Maintenance Procedures and Documentation M 3.1 Scaffold Construction for Safety Related Eauioment a.
Insoection Scoce The inspectors reviewed scaffold construction and inspection for safety related work activities. Deficiencies were identified with the scaffold construction and permit tag No.96-196 associated with the MU system MOV work for MU-V-14A, and a scaffold, without a tracking number on the permit tag, at the river water screenhouse for preservation work, as discussed below.
b.
Observations and Findinas The inspectors reviewed TMI corrective maintenance procedure 1440-Y-3, " Scaffold Construction / Inspection and Use of Extension Ladders," which provides the written program requirements. The scaffolds are approved initially by operations; constructed by the maintenance utility department; inspected and verified safe by the utility supervisor; and inspected by operations to ensure the scaffold is stable, strong, and does not endanger or interfere with emergency safeguards systems. After the worker inspects and signs the scaffold inspection tag, the scaffold is available for us.
On September 26,1996, the inspectors identified that a scaffold in the screenhouse lower bay was improperly attached to a nuclear river water system pipe support and was not questioned by the Senior Reactor Operator (SRO) review of the scaffold construction.
Specifically, procedure 1440-Y-3, section 5.8 states, in part, that " scaffolding shall not be tied off or supported by pipe hangers or pipe supports without prior engineering evaluation and approval." This scaffold was tied-off and supported by a safety related pipe support without prior engineering evaluation and approval. The inspectors notified the shift SRO about the problem. The SRO contacted the utility department immediately to evaluate the inspector's concern. The utility mechanic recognized the error and removed the scaffold support pipe that was attached to the nuclear river pipe support. The inspector and plant SRO reviewed the satisfactory change to the screenhouse scaffold.
On November 5,1996, the inspectors identified that work was started on MOV MU-V-14A before the operations department was notified to inspect the scaffold for final approval.
The scaffold was constructed near the MU valve gallery in the Auxiliary Building, inside the protected area. Procedure 1440-Y-3, section 8.14.2 states, in part, that "if scaffold is constructed inside the protected areas of the plant, have operations department sign the scaffold inspection tag verifying that the scaffold will not endanger emergency safeguard system equipment or operation." This scaffold was used during the lift of the motor off of valve MU-V-14A before the operations department reviewed and approved that the scaffold's finalinstallation would not endanger emergency safeguards equipment. The inspectors notified the control room SRO about the scaffold concern. The plant SRO reviewed the scaffold lifting rig and signed the scaffold inspection form to document that the scaffold was constructed properly and that it did not endanger emergency safeguards equipment.
The failure to follow the scaffold construction procedure 1440-Y-3 for safety related equipment work activities is considered a violation of the TMl Unit 1 TS, section 6.8.1.a.
(VIO 50-289/96-07-01).
c.
Conclusions The inspectors identified two examples where the scaffold construction and inspection written procedure was not properly implemented for two safety related work activities.
This issue is considered a violation of the Unit 1 Technical Specifications. In both cases immediate actions were taken to correct the problem. _ _
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111. Enaineerina
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E1 Conduct of Engineering (37751)
E1.1 Make-up Valve Small Break Loss of Coolant Accident Concern and (Uodated)
Unresolved item 50-289/96-08-01 a.
Insoection Scope f
The inspectors reviewed a potential small break loss of coolant accident (SBLOCA) design concern related to the failure of make-up (MU) valve MU-V-18. The review included the MU system SBLOCA response in the updated final safety analysis report (UFSAR) section
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6.1.3.1., the associated quality deficiency report (QDR) and event or near miss capture form (ENMCF).
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b.
Observations and Findinas
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MU-V-18 and MU-V-17 are the normal MU flow control valves, arranged in a series configuration, that provide a flow path from the MU pump discharge back to the RCS.
This normal MU injection line provides a flow path that is parallel to the four RCS emergency injection lines. The emergency injection lines are normally isolated by a motor operated valve (MU-V-16A/B/C & D) and each line was designed with a cavitating venturi that limits the flow through a line that experiences a postulated line break. With the postulated line break, the cavitating venturi will ensure that sufficient MU pump flow will reach the RCS through the other three injection lines. Also, if the normal MU flow control valves failed open during a postulated accident, the RCS make-up flow could be diverted away from the reactor because the line does not contain a cavitating flow venturi. Prior to a recent review by the high pressure injection (HPI) system engineer, TMI assumed that valve MU-V-18 failed closed for all conditions.
The questioning attitude of the HPI system engineer revealed a potential design vulnerability for a SBLOCA coincident with the loss of the 1M DC panel resulting in valve MU-V-18 failing open. If MU-V-18 failed open, then MU-V-17 is required to be closed to prevent the MU pump runout for a postulated MU line break downstream of the cavitating venturis. To ensure that the required flow is maintained to the RCS, the normal MU injection line (MU-V-17 and MU-V-18 in series) must be isolated. If the HPI line break occurred in conjunction with a loss of DC electrical power to MU-V-18, then MU-V-18 could fail open necessitating the closure of MU-V-17 to prevent MU pump runout and proper RCS flow.
Engineering initiated a quality deficiency report (QDR) to document and ensure proper resolution of the issue. A plant review group (PRG) evaluation concluded that the ability to verify that the normal MU line could be isolated during a postulated line break condition was a missed inservice test (IST) program surveillance; make-up valve MU-V-17, the valve in series with MU-V-18, was stroked within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to address the SBLOCA concern.
ENMCF No.96-312, was initiated to track the QDR resolution. The ENMCF was initiated because MU-V-17 is needed to mitigate the consequences of an accident as described in the updated final safety analysis report (UFSAR).
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The inspectors determined that the engineer's initiation of the ODR was warranted based
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on the significance of the potential safety issue. In particular at TMI, because the SBLOCA is the highest contributor (18.8%) toward core damage as noted in the individual plant examination (IPE). The decision to add MU-V-17 to the IST program and perform the
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surveillance test within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the issue was prudent. The ability to close the normal
make-up flowpath from the control room should ensure the plant operators that they could
combat the potential SBLOCA with a concurrent loss of DC electrical power to MU-V-18.
The engineering and operations departments are in the process of evaluating the make-up system configuration to determine the best long term resolution of the issue that will not
I rely on operator action.
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Conclusions l
j The excellent questioning attitude of the MU system engineer revealed a potential design
vulnerability for a SBLOCA coincident with the loss of the DC power resulting in valve MU-
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i V-18 failing open. Engineering initiated a QDR to document the safety concern and ensure
proper resolution of the issue. A PRG review concluded that the condition was a missed IST program surveillance; MU-V-17, the valve in series with MU-V-18, was stroked closed
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within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to address the SBLOCA concern.
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j IV. Plant Sunoort
R2 Status of P.P&C Facilities and Equipment
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R2.1 Tours of Areas j
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Inspection Scoce (71750)
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The inspectors toured various areas of the site used to process, stage, or store radioactive
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waste and radioactive materials. These areas included the radwaste handling and
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processing facility, the respirator maintenance and laundry facility, the waste storage
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facility, and the radwaste cask storage facility.
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Observations and Findinas l
The inspectors observed the general conditions and radiological housekeeping in the j
various areas used to process, store, and package radwaste. Equipment and buildings
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were kept in good condition. Radiological housekeeping was very good in most areas, and l
improvements were noted in the waste handling and packaging facility. However, some
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minor radiological concerns were identified by the inspectors to licensee representatives and immediate corrective actions were taken to resolve the concerns. A few wood pallets identified by the inspectors in the waste storage facility were determined to be an l
unintentional fire loading due to the combustion potential. Although they were not an j
immediate safety hazard, the licensee's representative stated that they would be removed and replaced with acceptable pallets.
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Conclusions
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l The areas toured by the inspectors were well maintained and radiological housekeeping
was generally very good. Some minor concerns were brought to the licensee's attention
for resolution or correction.
R4 Staff Knowledge and Performance in RP&C
R4.1 Removal of Underwater Robot from the Soent Fuel Pool
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Inspection Scope (71750)
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The inspectors observed the radiological controls during the removal of an underwater
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robot from the spent fuel pool.
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Observations and Findinas Radiological controls for the removal of the underwater robot included a roped-off area for the contaminated equipment, protective clothing and dosimetry for the workers, and radiological monitoring and dose rate measurements by a radiological controls technician.
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A decontamination technician sprayed water on the robot as it was raised out of the water to rinse potential contamination from the equipment. The radiological controls technician performed surveys for dose rates and removable contamination, including hot particle controls.
c.
Conclusions The inspector observed very good radiological controls by all workers involved in the removal of the equipment from the spent fuel pool.
R6 RP&C Organization and Administration R6.1 Chanaes to the Licensee's Radiation Protection Proaram a.
Insoection Scope (71750)
The inspectors reviewed changes to the licensee's organization through interviews with licensee staff and review of licensee documentation.
b.
Observations and Findinas A recent change to the. licensee's organization was initiated when the Radiological Controls / Occupational Safety (RC/OS) Director retired from the company. The inspectors reviewed the qualifications of the individual assigned to temporarily replace the RC/OS Director. The licensee's Technical Specifications require the management position responsible for radiological controls shall meet or exceed the qualifications of Regulatory Guide 1.8 (1977). Regulatory Guide 1.8 states that the radiation protection manager should have the qualifications described in section 4.4.4 of ANSI /ANS 3.1-1981. Section
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4.4.4 of ANSI /ANS 3.1-1981 states that the individual v/ho temporarily replaces the radiation protection group leader shall have a Bachelor Degree in a science or engineering subject and two years exp3rience, one of which shall be nuclear power plant experience and six months experience shall be onsite. The individual acting as the RC/OS Director had a Bachelor Degree in Biology and over 20 years nuclear power plant professional experience at the Three Mile Island site.
Other organizational changes occurred when the Manager, Radiological Engineering was assigned as the acting RC/OS Director. Temporary changes included combining the Radiological Health and Radiological Engineering groups under one manager. The manager reported to the RC/OS Director. In addition, the Occupational Safety group was removed from the Radiological Health group and the Occupational Safety Manager directly reported
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to the RC/OS Director.
c.
Conclusions
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The inspectors concluded that the individual acting as the RC/OS Director exceeded the minimum qualifications for the position as required by the licensee's Technical Specifications. Other organizational changes were not expected to have a negative effect on the radiological controls program.
P1 Conduct of EP Activities P1.1 Emeraency Response Drill for Centaminated iniured Individual a.
Inspection Scooe (71750)
The inspectors observed the annual emergency response drill involving a contarninated, injured worker. The licensee response team used the licensee's procedure number EPIP-TMI.16, Revision 5, titled " Contaminated injuries." The drill also involved simulated response at one local hcspital (Hershey Medical Center).
b.
Obst. Ions adFindinas s
After notification to the Control Room, the response drill was initially announced over the public address system. A first response team comprised of 2 radiological control technicians,1 radiological controls supervisor, 2 security guards, and 3 other emergency assistance technicians arrived at the location of the injured individual within a few minutes.
The scenario involved a worker who was performing electrical work in a contaminated area above the Unit 2 radwaste control panel. The worker's simulated injuries included a cut on the head and a burn on the leg. The worker was dressed in protective clothing that had become torn.
The response team practiced very good contamination controls and monitoring for potential contamination. Simulated radiological contamination was found on the worker's face and leg. The responders removed the individual's protective clothing and wrapped / bandaged the simulated wounds. The individual was put in a neck collar, and placed on a backboard for transport to the hospital. Two medical personnel arrived later to transport the individual
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to the hospital via the licensee's ambulance. The responders transported the individual from the radiologically controlled area in approximately 30 minutes, and arrived at the
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hospital in approximately 70 minutes after the drill announcement.
Communications between the responders was good, but some confusion regarding
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command and control was evident at the injury site. Also, the ambulance driver, a emergency responder, did not take the designated route to the hospital. Communications J
between the site responders and the hospital personnel were very good, including patient information relaycd from the ambulance while on the way to the hospital. The NRC site office was notified by drill responders within approximately 15 minutes of the incident regarding a potentially contaminated, injured worker. Another drill notification was made to the NRC office, approximately 60 minutes after the incident, regarding an event of potential public interest because a contaminated individual was transported to the hospital.
The licensee's transfer of the contaminated patient at the hospital was performed with appropriate contamination controls; however, the hospital staff was less careful than the licensee's staff regarding the potential spread of contamination. An informal critique was held after the exercise and the licensee's staff developed a list of areas for improvement for response to events in the future.
c.
Conclusions The licensee provided very good overall response during the annual medical response exercise. Appropriate contamination controls were used during preparation and transport of the potentially contaminated, injured individual to the local hospital. Some minor problems were encountered with communications, but overall communications were good.
A list of areas for program improvement were developed after the exercise.
S1 Conduct of Security and Safeguards Activities a.
Insoection Scoce (81700)
The inspector reviewed the security program during the period of October 7-10,1996.
Areas inspected included: previously identified items; effectiveness of management control; management support and audits; protected area detection equipment; alarm stations and communication; testing, maintenance and compensatory measures; and training and qualification. The purpose of this inspection was to determine whether the licensee's security program, as implemented, met the licensee's commitments and NRC regulatory requirements.
b.
Observations and Findinas Management support is ongoing as evidenced by the procurement and installation of three new metal detectors, procurement of additional training aids to add realism during firearms training, and the hiring of five additional site protection officers (SPOs). Alarm station operators were knowledgeable of their duties and responsibilities, security training was being performed in accordance with the NRC-approved training and qualification (T&O)
plan and was well documented and available for review. Management controls for identifying, resolving, and preventing programmatic problems were effectiv _.
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i Protected area (PA) detection equipment satisfied the NRC-approved physical security plan (the Plan) commitments and security equipment testing was being performed as required by the Plan. Maintenance of security equipment was being performed in a timely manner i
as evidenced by minimal compensatory posting associated with non-functioning security equipment, and two previously identifi9d items noted during two previous reactive i
inspections conducted in September 1995 and February 1996 were closed.
c.
Conclusions The inspector determined that the licensee was conducting its security and safeguards activities in a manner that effectively protected public health and safety.
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S2 Status of Security Facilities and Equipment S2.1 Protected Area Detection Aids
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Insoection Scoce j
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The inspector conducted a physical inspection of the PA intrusion detection systems (IDSs)
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to verify that the systems were functional, effective, and met licensee commitments.
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Observations and Findinas
On October 8,1996, the inspector determined by observation that the IDSs were functional and effective, and were installed and maintained as described in the Plan.
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Conclusion i
The PA intrusion detection aids met the licensee's Plan commitments.
S2.2 Alarm Stations and Communications a.
Insoection Scoce
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Determination whether the Central Alarm Station (CAS) and Secondary Alarm Station (SAS) are: (1) equipped with appropriate alarm, surveillance and communication capability,
(2) continuously manned by operators, and that (3) the systems are independent and diverse so that no single act can remove the capability of detecting a threat and calling for assistance, or otherwise responding to the threat, as required by NRC regulations.
b.
Observations and Findinas
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Observations of CAS and SAS operations verified that the alarm stations were equipped with the appropriate alarm, surveillance, and communication capabilities. Interviews with
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CAS and SAS operators found them knowledgeable of their duties and responsibilities.
The inspector also verified through observation and interviews that the CAS and SAS operators were not required to engage in activities that would interfere with the
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assessment and response functions, and that the licensee had exercised communications methods with the local law enforcement agencies as committed to in the Plan.
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Conclusion The alarm stations and communications met the licensee's Plan commitments and NRC requirements.
S2.3 Testina. Maintenance and Compensatorv Measures a.
Inspection Scoce Determination whether programs were implemented that will ensure the reliability of j
security related equipment, including proper installation, testing and maintenance to replace
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defective or marginally effective equipment. Additionally, determination whether security related equipment failed, the compensatory measures put in place was comparable to the effectiveness of the security system that existed prior to the failure.
b.
Observations and Findinas j
Review of testing and maintenance records for security-related equipment confirmed that
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the records were on file, and that the licensee was testing and maintaining systems and equipment as committed to in the Plan. A priority status was assigned to each work request and repairs were normally being completed in a timely manner from the time a work request, necessitating compensatory measures, was generated.
c.
Conclusions Security equipment repairs were being completed in a timely manner. The use of compensatory measures was found to be appropriate and minimal.
S5 Security and Safeguards Staff Training and Qualification a.
Inspection Scone Determination whether members of the security organization were trained and qualified to perform each assigned security related job task or duty in accordance with the NRC-approved T&O plan.
b.
Observations and Findinas The inspector selected at random and reviewed the training, physical, and firearms qualification /requalification records of nine SPOs.
On October 9,1996, the inspector met with the security training staff and discussed the training department enhancements and program initiatives implemented since the previous program inspection conducted in January 1996. The enhancements included the procurement of additional scenarios for the firearms training system (FATS), which is conducted during security requalification training, and the development of additional local law enforcement agency (LLEA) interface activities. The LLEA interface activities included providing a plant tour, classroom training, which consisted of an overview on nuclear
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power, and FATS training. Additionally, the inspector interviewed a number of SPOs to determine if they possessed the requisite knowledge and ability to carry out their assigned
duties.
c.
Conclusions The inspector determined that training had been conducted in accordance with the T&Q
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plan, and that it was properly documented. Interface activities with LLEA was appropriate.
Based on the SPOs responses to the inspector's questions, the training provided by the security training staff was effective.
i S6 Security Organization and Administratior;
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insoection Scope A review of the level of management support for the licensee's physical security program
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was conducted.
b.
Observations and Findinas The inspector reviewed various program enhancements made since the last program inspection, which was conducted in January 1996, with security management. These enhancements included the installation of three new metal detectors, procurement of additional training aids to add realism during firearms training, and the hiring of five j
additional site protection officers (SPOs).
c.
Conclusions i
Management support for the physical security program was determined to be excellent.
S7 Quality Assurance in Security and Safeguards Activities S7.1 Effectiveness of Manaaement Controls a.
Insoection Scope A review was conducted to determine if the licensee had controls for identifying, resolving and preventing programmatic problems, b.
Observations and Findinas
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i The inspector determined that controls were in place. They included the performance of a semi-annual assessment by corporate security, the NRC-required annual quality assurance (QA) audit, and a continual shift observation program by supervision. Additionally, the licensee has implemented a formalized self-assessment program which requires each site
protection shift supervisor to conduct two assessments per year, in areas determined by
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the Security Manager. The licensee also utilizes industry data, such as violations of regulatory requirements identified by the NRC at other facilities, as a criterion for self-assessmen.
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c.
Conclusions A review of documentation applicable to the licensee controls, including results, indicated that security performance errors were being minimized and were effectively implemented to identify and resolve potential weaknesses.
S7.2 Audits a.
Insoection Scope The inspector reviewed the licensee's NRC-required audit of the security program to determine if the licensee's commitments as contained in the NRC Plan were being satisfied.
b.
Observations and Findinas The inspector reviewed the 1996 QA audit of the security program conducted February 1 -
March 7,1996, (Audit No. S-TMI-96-02). The audit was found to have been conducted in accordance with the Plan. The audit identified no adverse findings and three minor deficiencies, which were not indicative of programmatic weaknesses but, if corrected, would enhance program effectiveness. Audit results had been dissiminated to the appropriate levels of management. The inspector determined, based on discussions with i
security management and a review of the responses to the deficiencies, that the corrective actions were effective and were completed prior to the issuance of the audit report.
c.
Conclusions The review concluded that the audit was comprehensive in scope and depth, that the findings were appropriately distributed and that the audit program was being properly administered.
S8 Miscellaneous Security and Safeguards issues (92904)
(Closed) Violation 50-289/95-15-01. Enforcement Action 95-217: The licensee failed to provide security compensatory measures during maintenance activities, which resulted in the existence of three, and the potential for a fourth, unmonitored and unprotected pathways, with cross sectional areas significantly greater than 96 square inches, from the owner controlled area into the protected area.
(Closed) Violation 50-289/96-03-01. Enforcement Action 96-057: A violation occurred when the licensee again f ailed to provide security compensatory measures during maintenance activities. Specifically, an opening exceeding 96 square inches existed between the protected area and the owner controlled area via a storm drain.
With respect to the above violations, the inspector determined that the corrective actions described in the licensee's December 20,1995 and April 24,1996 letters, in response to the NRC's Noti::es of Violations, were reasonable, complete, and appeared to be effectiv.
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V.
Manaaement Meetinas X1 Exit Meeting Summary At the conclusion of the reporting period, the resident inspector staff conducted an exit meeting with TMI management on November 21,1996, summarizing Unit 1 inspection activities and findings for this report period. TMI staff comments concerning the issues in this report were documented in the applicable report section. No proprietary information was identified as being included in the report.
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- TMI Systematic Assessment of Licensee (SALP) Management Meeting On September 30,1996, a public meeting was held between the NRC and GPU Nuclear Corporation at the TMl Training Center in Middletown, Pennsylvania. The purpose of the meeting was to discuss the TMl performance during the previous 18 months and provide an opportunity for TMI to provide their perspective on the SALP report. The NRC Staff provided a brief overview of each SALP functional area followed by an informal discussion a'oout the related program strengths and areas for improvemen,. _.
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PARTIAL LIST OF PERSONS CONTACTED Licensee G. Broughton, President GPUN D. Etheridge, Acting Radiological Controls / Occupational Safety Director R. Goodrich, Site Security Manager D. Hosking, NSA Manager
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'J. Knubel, Vice President and Director L. Noll, Plant Operations Director M. Ross, Director, Operations and Maintenance J. Schork, Regulatory Affairs J. Wetmore, Manager, Regulatory Affairs C. Incorvati, QV Manager l
- senior licensee site manager present at exit meeting on November 21,1996.
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NRC J. Norris, TMl Project Manager, NRR
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INSPECTION PROCEDURES USED i
IP 37551:
Onsite Engineering IP 40500:
Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems i~
IP 62707:
Maintenance Observation
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IP 61726:
Maintenance Surveillance Observation l
IP 71707:
Plant Operations
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IP 71750:
Plant Support Activities
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IP 81700:
Physical Security Program IP 92700:
Onsite Follow-up of Written Reports of Non-routine Events at Power Reactor Facilities IP 92904:
Followup-Plant Support ITEMS OPENED, CLOSED, AND DISCUSSED Opened i
50-289/96-07-01
" Failure to Follow Scaffold Procedure for Safety Related Equipment." (VIO)
Closed 50-289/95-15-01
" Failure to provide security compensatory measures during (EA No.95-217)
maintenance activities." (VIO)
50-289/96-03-01
" Failure to provide security compensatory measures during (EA No.96-057)
maintenance activities." (VIO)
Updated 50-289/96-08-01
" Potential Condition Adverse to Quality that was not Entered into the Formalized Corrective Action Program."
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LIST OF ACRONYMS USED AO Auxiliary Operator AP Administrative Procedure AS Auxiliary Steam BWST Borated Water Storage Tank CAS Central Alarm Station CRO Control Room Operator CFR Code of Federal Regulations DBD Design Basis Documents ECCS Emergency Core Cooling System EFW Emergency Feedwater ENMCF Event or Near Miss Capture Form ESF Engineered Safety Feature FATS Firearms Training System HPl High Pressure injection HRA High Radiation Area IDS Intrusion Detection Systems IPE Individual Plant Evaluation IST Inservice Testing Program LCO Limiting Condition of Operation LER Licensee Event Report LLEA Local Law Enforcement Agency MNCR Material Nonconformance Report MOV Motor Operated Valve MU Make-Up NCV Non-Cited Violation NRC Nuclear Regulatory Commission NSA Nuclear Safety Assessment
PA Protected Area
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PCi/L Pico Curies per Liter PDR Public Document Room PPM Part per Million PRG Plant Review Group QDR Quality Deficiency Report RC/OS Radiological Controls / Occupational Safety RCS Reactor Coolant System RP Radiation Protection SALP Systematic Assessment of Licensce Performance SAS Secondary Alarm Station SBLOCA Small Break Loss of Coolant Ac ident SPO Site Protection Officer
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SRO Senior Reactor Operator
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SS Shift Supervisor T&O Training and Qualification TS Technical Specification UFSAR Updated Final Safety Analysis Report VIO Violation
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