IR 05000289/1998003
| ML20151V200 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 09/04/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20151V188 | List: |
| References | |
| 50-289-98-03, 50-289-98-3, NUDOCS 9809140151 | |
| Download: ML20151V200 (25) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
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Docket No.:
50-289
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License No.:
DPR 50 l
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Report No.:
98-03 Licensee:
GPU Nuclear, Inc. (GPUN)
Facility:
Three Mile island Station, Unit 1 Location:
P.O. Box 480 Middletown, PA 17057 Dates:
May 24 through July 25,1998 Inspectors:
Wayne L. Schmidt, Senior Resident inspector Joseph E. Carrasco, Resident inspector Michael Buckley, Resident inspector, Peach Bottom John R. McFadden, Radiation Specialist
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Approved by:
Michele G. Evans, Chief Projects Branch No. 7
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Division of Reactor Projects
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EXECUTIVE SUMMARY Three Mile Island Nuclear Power Station Report No. 50-289/98-03 i
This integrated inspection included asnects of licensee operations, engineering, maintenance, and plant support. The repo,1 covers a nine week period of resident inspection; in addition, it includes the results of an announced inspection by a regional inspector in the area of radiological controls.
Operations l
Control room operators performed routine tasks well. Operators were knowledgeable of plant conditions and the reasons for annunciated alarms. (Section 01)
General plant housekeeping was good, with some noted improvements in the emergency diesel generator areas. However, the housekeeping in the tendon access gallery was poor.
(Section 01)
GPUN had an effective program for controlling flood preparations using an emergency procedure. Plant equipment was staged and ready for use if a flooding condition occurred.
However, concerns were identified with the condition of and the design basis controls for the reactor building to tendon gallery flvod seal. (Section O2.1, URI 98-03-01)
Plant operators and management responded well to a failed once through steam generator (OTSG) feedwater start-up flow control valve controller, to indications of a possible leaking
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l OTSG tube plug, in preparation for maintenance on a letdown isolation valve, and in identification of a degraded nuclear river water system pump strainer. (Section 04.1)
The control room staff performed properly following identification of a plant process computer printer problem that resulted in hourly core power distribution limits not being printed for several hours on April 29,1998. GPUN complied with the applicable technical specifications and appropriately conducted a hand calculation of power distribution limits using the minimum incore detectors. (Section 08.1)
Maintenance GPU maintenance department personnel ~ performed well during the observed activities over the period. This included verification of several work packages and observation of good nrocedural usage. Maintenance activities on a nuclear closed cooling heat exchanger showed that it was in good condition. Work to correct a body-to-bonnet leak on the letdown isolation valve to the reactor coolant bleed tanks was very well coordinated and went very smoothly. (Section M1)
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Observed surveil!ance test activities were conducted well, including emergency safeguards, reactor protection, and undervoltage relay testing. (Section M1)
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No missed post-maintenance test (PMT) activities were identified in a detailed review of three safety-related work packages. However, severalidentified weaknesses indicated the need for continued improvement in the documentation of PMT activities and verification that job order PMT sheets and instructions, maintenance procedure instructions, and surveillance tests meet requirements. (Section M1.1)
Enaineerina Engineering support to plant operation continued to be strong, this included good support to operations in response to equipment problems. (Section E1)
The emergency feedwater (EFW) system was operable based on review of calculations and testing, assuming that the October 1997 new design basis accident calculation was correct. However, the UFSAR Chapter 14 LOFW design basis accident analysis for the EFW system had not been well maintained. The calculation process was weak in that a j
design basis calculation did not require a safety evaluation once the design verification of the information had been completed. The safety determination process, used to justify a temporary procedure change to allow a lower EFW capacity test flowrate to depressurized
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OTSGs, was inadequate. Further, recent changes to the design basis in October 1997 had l
not been translated into a PFU which would have informed other GPUN employees that there was a new design basis calculation. (Section E2.1, URI 98 03-02)
Severalissues were identified dealing with the controls in place to document, track, and assess identified reactor building tendon grease leakage, both from end caps on horizontal tendons and through concrete cracks at the lower portion of the vertical tendons.
(Section E2.2, IFl 98-03-03)
Plant Support Overall, effective performance in the area of radiological controls for radioactive materials, contamination, surveys, and monitoring was evident. (Section R1.1)
Overall, the sohd radioactive waste management program was effective based on the existence of appropriate procedures and controls, the acceptable condition of facilities and equipment, and the proper implementation of the program by knowledgeable and experienced personnel. (Section R1.2)
The program for shipping low level radioactive waste for disposal and transportation of other radioactive materials was generally effective. (Section R1.3)
The NRC and DOT training and retraining requirements for radioactive waste group personnel were satisfied. (Section RS)
Self-assessments by the radiological controls group in the area of radiological controls for radioactive materials, contamination, surveys, and monitoring were normally good and provided recommendations and appropriate corrective actions for identified deficiencies and problems. (Section R7.1)
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- Radioactive waste quality assurance activities were generally effective. The audit
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performed as required by technical sp'ecifications was extensive in scope and depth, i
(Section R7.2)
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TABLE OF CONTENTS E X E C UTIV E S U M M A RY.............................................. ii TA B LE O F C O NT E NT S............................................... v 1. Operations
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Conduct of Operations (71707).............................. 1 01.1 G e neral Comme nts.................................. 1
Operational Status of Facilities and Equipment................... 1 O2.1 ' External Flood Barrier Walkdown........................ 1
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Operator Knowledge and Performance......................... 3 04.1 General Operator Performance.......................... 3
Miscellaneous Operations issues............................. 4
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08.1 Quadrant Power Tilt and Core Power imbalance............. 4 11. M ai nt e n a n c e................................................... 5 M1 Conduct of Maintenance (61726,62707)....................... 5 M 1.1 Review of Post-Maintenance Testing Documentation.......... 5 111. En g i n e e r i n g................................................... 6 E1 Conduct of Engineering (37551)
.............................6 E2 Engineering Support of Facilities and Equipment..................6 E2.1 Emergency Feed Water Design Basis Flowrate Review......... 6
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E2.2 Inservice Tendon Surveillance.........................
IV. Plant Support
................................................12 R1 Radiological Protection and Chemistry (RP&C) Controls............ 12 R1.1 Radiological Controls-Radioactive Materials, Contamination, Surveys,
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and Monitoring
...................................12 R1.2 Implementation of the Solid Radioactive Waste Program......
R1.3 Compliance with NRC and DOT Regulations............... 14 R5 Staff Training and Qualification in RP&C Activities............... 15 R7 Quality Assurance and Self-Assessment in RP&C Activities.........
R7.1 Radiological Controls Activities........................ 15 R7.2 Radioactive Waste Activities..........................
R8 Miscellaneous RP&C issues................................
R8.1 '(Closed) Inspection Follow-up item 50-289/97-05-01........
R8.2 (Closed) Inspection Follow-up item 50-289/97-06-02........
R8.3 (Closed) EA 50-289/97-533-03014(and eel 50-289/97-09-05).. 17 V. Management Meetings..........................................18 X1 Exit Meeting Summ a ry...................................
lNSPECTION PROCEDURES USED.......................,............. 19 ITEMS OPENED, CLOSED, AND DISCUSSED..............................
LI ST O F AC RO NYM S U SED.......................................... 20
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Report Details Summarv of Plant Status The unit remained at 100% power throughout the inspection period.
l l. Operations
'01 Conduct of Operations (71707)
01.1 General Comments Control room operators performed routine tasks well. Operators were knowledgeable of
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l-plant conditions and the reasons for annunc;ated alarms.
General plant housekeeping was good, with some noted improvements in the emergency diesel generator (EDG) areas. However, the house. keeping in the reactor building tendon access gallery was poor.
Operational Status of Facilities and Equipment O2.1 External Flood Barrier Walkdown
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inspection Scoce (71707)
'The inspectors reviewed Emergency Procedure (EP) 1202-32." Flooding," and walked down the flood control barriers listed in the Updated Final Safety Analysis Report (UFSAR), Section 2.6.5 and the GPUN, December 1994, Individual Plant Examination for External Events (IPEEE), Section 5.2.3 for the control, diesel generator', auxiliary, fuel handling, and intermediate buildings. 'At TMI, external flooding contributes the majority to core damage frequency for an external event, as documented in the IPEEE.
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Observations and Findinas Plant WaW r'
wm, EP 1202-32 provided good guidance on actions needed to preclude plant damage due to a flooding condition, including direction for a plant shutdown to a cold condition and installation of flood control barriers. The UFSAR stated that the flood
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gates would protect safety related structures to a flood height of 311 feet, while plant grade is approximately 305 feet.
'The inspectors found that all barriers listed in the UFSAR and IPEEE were staged and appeared in good condition. However, in review of the flood barriers, the inspectors found some oroblems with the material condition and identification of j
other flood barriers protecting the intermediate building. The barriers in question
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protected the tendon access gallery, which would overflow into the intermediate building, from external flooding.
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in review of the history of tendon access gallery flooding, the inspectors found that during the mid 1980's, GPUN had completed changes to the tendon access gallery to allow internal flooding of the entire area following a feedwater line break in the intermediate building. These changes also necessitated that flood seals be installed between the reactor building and the west section of the tendon gallery where it would be directly exposed to the external flood at a ground level of approximately 305 feet, and at the upper and lower access from the external gallery ladder access (ladder C). Without the flood seals, flood water would fill the tendon access gallery and then the intermediate building, disabling the emergency feedwater (EFW)
system, which would affect the ability to remove decay heat if the unit was not in cold shutdown with the decay heat removal system operating.
Upon observation of water in-leakage, degradatior 'f sealing backing material, and
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visible light from outside, the inspectors questioned the ability of the reactor building to tendon access gallery seal to stop flood water. The inspectors discussed these concerns with the licensee structural engineer, who walked down that area and identified an area which required resealing. The engineer entered these concerns into the corrective action process (CAP). The plant review group (PRG)
reviewed the CAP, finding that it was not outside the design basis since the UFSAR did not mention the sealin this area.
In review of the CAP, the inspector questioned why this was not a condition outside the design basis, since this seal appeared to be needed as much as any other barrier listed in the UFSAR. The inspectors reviewed the safety evaluation completed for the installation of the seal, which stated that a watertight seal was needed to prevent in-leakage due to a flood. The safety evaluation cover sheet stated that this was not a change to the UFSAR. It appeared to the inspectcrs that it actually was, since 'the UFSAR was never changed to include the seat bccween the reactor building and the tendon access gallery as a required flood barrier.
At the end of the period, the maintenance department was conducting repairs to the seals.
The inspectors considered these issues associated with the reactor building to tendon gallery flood seal as unresolved (URI 50-289/98-03-01)pending completion of licensee corrective actions, which in:luded repair of the seal and evaluation of the affect of the degraded seal on the operability of the emergency feedwater system, and further evaluation of the weaknesses noted with the content of the UFSAR and the historical safety evaluations associated with modifications to the flood barriers.
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Cor ctusion GPUN had an effective program for controlling flood preparations using an emergency procedure. Plant equipment was staged and ready for use if a flooding condition occurred. However, concerns were identified with the condition of and the design basis controls for the reactor building to tendon gallery flood seat.
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04 Operator Knowledge and Performance 04.1 General Operator Performance a.
Inspection Scope The inspectors observed operator performance during several different conditions including; replacement of a start-up feed control valve card, indications of possible increased once through steam generator (OTSG) primary to secondary tube leakage, maintenance on the waste drain valve to the reactor coolant bleed tanks, and identification of a nuclear river water system (NR) strainer that f ailed to rotate when the associated pump was started, b.
Observations and Findinas Control room operators and operations management responded well to several issues including:
During the first week of June, operators noted increases in the radiation
levels of the condenser offgas radiation monitors from approximately 40 counts per minute (cpm) to approximately 80 cpm. Operators closely monitored the offgas flowrate and the radiation detectors. Within one week, the radiation detector readings had decreased close to near their original values. The inspectors reviewed the OTSG primary to secondary leakage procedure and discussed the issue with operations management and the shift crews. It appeared based on the indications of offgas radiation monitors that a previously plugged OTSG tube may have begun to leak at a plug.
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Operators noticed that a startup feed control valve had modulated closed
during operation. This valve closure did not effect the operation of the feed pumps or cause a perturbatica in OTSG level, since the normal feed flow control valve modulated further open to control OTSG level. Operators took prudent action to place the controller in hand and instrument & controls (l&C) personnel replaced a f ailed control module. The shif t supervisor conducted a very good briefing prior to the module replacement and operators performed very well in monitoring the plant for any possible changes.
The plant crew performed wellin preparation for work on the letdown
isolation valve to the reactor coolant bleed tank. Good pre-job briefings were conducted and the work was completed smoothly.
l During pump switching operations the local auxiliary operator noted that a
NR pump discharge strainer failed to turn when the pump was started.
l Maintenance completed a repair to the shaft key.
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Conclusions l
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- Plant operators and management responded well to a f ailed OTSG feedwater start-l up flow control valve controller, to indications of a possible leaking OTSG tube plug, l
in preparation for maintenance on a letdown isolation valve, and in identification of
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a degraded NR pump strainer.
Miscellaneous Operations issues l
08.1 Quadrant Power Tilt and Core Power imbalance
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Inspection Scoce (71707)
The inspectors completed a review of an April 29,1998, issue dealing with the
temporary loss of hourly log print-outs due to an unplugged printer. The hourly log
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printout between 2:00 p.m. and 4:00 p.m. did not occur, as such the plant process computer (PPC) core power distribution outputs for quadrant tilt and axial power imbalance 'uere not available. Technical Specification (TS) 3.5.2 requires that these
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parameters be monitored and verified within acceptable limits on a minimum frequency of once every two hours.
-This review included an assessment of control room activities related to reactor f
quadrant power tilt compliance with the TS and the adequacy of hand calculations for quadrant power tilt.
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Observations end Findinas t
Operators properly recognized that the printout did not occur for the 2:00 p.m.
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readings. Upon examining the situation, the operating crew noted that the full incore system, which provided the normalinput to the PPC for power distribution limit i
j calculations and the PPC were still operable. Had the power distribution parameters exceeded a TS limit, overhead alarm G-2 6, which h fed by the PPC, would have alerted the operators. Core power distribution ~information was also available via the PPC video screens in the control room.
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L While discussions were ongoing, the operators, as'a conservative measure, l<
conducted a hand calculation of quadrant power tilt and imbalance following L
operating procedure 1203-7 using the minimum incore detector (backup) readings.
Shortly after 4:00 p.m., GPUN restored the PPC printers in the control room. A print out of power distribution limits showed that limits had not been exceeded.
The hand calculation using the backup incore readings was completed and showed
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3.5.2.4 clearly stated the full incore system as the preferred source for tilt me.asurements. The backup system is used only if the full incore system and out-i of-core systems are not operable.
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Conclusion The control room staff performed properly following identification of a PPC printer problem that resulted in hourly core power distribution limits not being printed for several hours on April 29,1998. GPUN complied with the applicable TSs and appropriately conducted a hand calculation of power distribution limits using the minimum incore detectors.
11. Maintenance M1 Conduct of Maintenance (61726,62707)
GPUN maintenance department personnel performed well during the observed activities over the period. This included verification of several work packages and observation of good procedural usage. Maintenance activities on a nuclear closed cooling heat exchanger showed that it was in good condition. Work to correct a body-to-bonnet leak on the letdown isolation valve to the reactor coolant bleed tanks was very well coordinated and wer.t smoothly.
Observed surveillance test (ST) activities were conducted well, including emergency safeguards, reactor protection, and undervoltage relay testing.
M1.1 Review of Post-Maintenance Testina Documentation a.
Insoection Scope The inspectors selected several safety related mainten'ance packages, completed during a maintenance outage in May 1998, on the reactor building spray (BS),
decay heat removal (DH) and decay heat river (DR) systems, to review tne adequacy of post-maintenance testing (PMT) documentation. This included reviews of job order (JO) packages, maintenance procedures, and completed STs.
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Observations and Findinqs Maintenance package review for work on the "A" BS pump suction valve (BS-V-3A), the "A" DH pump discharge check valve (DH-V-16A), and
"A" DR pump discharge strainer (DR-S-1 A) indicated that no specific PMT requirements had been missed. However, there were instances where the documented PMT alone, absent other ST documentation or undocumented operator checks, would not have been sufficient:
On DH-V-16A (JO 00148039),the only requirement listed in the general
check valve maintenance procedure was a body-to-bonnet leak check and a local leakrate test if required. However, this valve was in the inservice test (IST) program, and as such should have required a verification of flow following maintenance, if possible. The inspector did verify that tne JO PMT sheet stated that a body-to-bonnet leak check should be conducted during
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performance of ST 1300-3B,the routine system / pump / valve IST. The inspector verified that the ST had been completed and that in f act the valve had been stroked via the completed ST.
On DR-S-1 A (JO 00147180)the written JO PMT sheet did not agree with
the job order instruction. Specifically, the JO sheet stated that only a leak check was required, while the JO instruction stated that the strainer should operate as designed with no leakage under operating conditions. The DR normal IST procedure, ST 1300-3D, used to conduct the system surveillance following the maintenance did not verify that the strainer was operating.
However, an operator verified that the strainer rotated as part of normal rounds conducted during the ST. Subsequently, GPUN also revised the ST and the normal operating procedure to ensure that the strainer is rotating and
that differential pressure is less than 6 psid on a pump start.
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Conclusion
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No missed PMT activities were identified in a cetailed revie'.v of three safety-related work packages. However, several identified weaknesses ir dicated the need for continued improvement in the documentation of PMT activi':ies and verification that
JO PMT sheets and instructions, maintenance procedure instructions, and surveillance tests meet requirements.
111. Enaineerina E1 Conduct of Engineering (37551)
Engineering support to plant operation continued to be strong, this included good support to operations in response to equipment problems, t
E2 Engineering Support of Fad.ses and Equipment E2.1 Emeraency Feed Water Desian Basis Flowrate Review a.
Inspection Scoce The inspectors completed a design basis review of the emergency feedwater (EFW)
system to evaluate the suitability of lowering the system flowrate requirement, based on an October 1997 design basis accident calculation for a loss of feedwater accident (LOFW).
The inspectors reviewed a temporary change notice (TCN) to the EFW capacity test conducted during the 1997 refueling outage. GPUN measured EFW flow to depressurized OTSGs and compared it to acceptance criteria to ensure that the EFW system could deliver at least the design basis flowrate with the OTSGs at 1050 psig. The acceptance criteria were developed using a new lower EFW design basis flowrate and calculating the flowrate that would be achieved if the OTSGs were depressurized, using the pump head curves and system resistance relationships.
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Observations and Findinas The new design basis calculation completed in October 1997, appeared to support reduction in EFW flowrate to the OTSGs, and the capability of the system as verified by outage testing. The new design basis minimum flowrate was calculated as 480 gpm (240 to each OTSG) for a LOFW without a loss of offsite power (LOOP)
(reactor coolant pumps (RCPs) running.) This calculation along with other system resistance calculations were used as a basis to determine new lower minimum flows to the depressurized OTSGs during the outage. The method used appeared appropriate.
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The original EFW design basis flow requirement was not clear based on a review of documented restart requirements including NUREG-0680 and NUREG 0737, the present TSs, and UFSAR Update 14, dated April 1998, as discussed below.
With respect to NUREG-0680,TMI-1 Restart:
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Section C. SHORT TERM ACTION, item la. EFW TIMELINESS AND
RELIABILITY was reviewed. Item 4 of this section discusses the TS that
"eded to be put in place and the minimum EFW flow requirements.
Specifically stated was, "The licensee proposes in the future (not required for restart) to incorporate cavitation venturies in the auxiliary feedwater lines to limit auxiliary feedwater flows so that a minimum of 500 gpm will be available to each steam generator, with a total of 1,000 gpm to be available j
at the 100% flow condition for both steam generators." The NUREG goes i
on to discuss accidents where EFW would be required, stating that loss of feedwater with four reactor coolant pumps would be the most limiting accident. It is clear that the staff reviewed a Babcock and Wilcox (B&W)
report on LOFW accidents stating that 370 gpm was the minimum for a i
LOFW without RCPs running and that approximately 130 gpm would be needed to remove RCP heat. The report stated, "The effect of pump heat would raise the requirement to 500 gpm." In conclusion, the report stated that, " Based on the above we conclude that the minimum required auxiliary flowrate of 500 gpm to each steam generator is acceptable and is in compliance with this part of the Order."
With respect to NUREG-0737, TMI Action Plan, item II.E.1.1, Auxiliary Feedwater System Evaluation:
In review of GPUN letter dated April 29,1985, the inspector found
documentation of changes made to the system prior to restart, including installation of the heat sink protection system and EFW system modifications, including maintaining the pump recirculation values open and installation of cavitating orifices in the line to limit flow to a depressurized OTSG. These changes included making the EFW system single failure proof.
Table 1 of this letter provided a summary of the maximum deliverable EFW flowrates for different pump combinations, with the recirculation valves open, as follows: one motor-driven pump 440 gpm (220 to each OTSG), two
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I motor-driven pumps 730 gpm (365 to each OTSG), the turbine driven pump 660 gpm (330 to each OTSG), one motor driven pump and the turbine driven pump 800 gpm (400 to each OTSG), and all three pumps 860 gpm j
(430 to each OTSG).
With respect to TS:
The licensee submitted a TS amendment request dated May 1981 and the
NRC approved the request issuing TS Amendment #78 on October 29, j
L 1982, which incorporated requirements for EFW operability and testing into i
the TS. The bases for EFW TS 3.4 Decay Heat Removal-Turbine Cycle stated, in part:
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"Both motor driven pumps are required initially to remove decay heat
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with one eventually sufficing, and
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. The requirements for TS 3.4.1 assure that before the reactor is heated to above 250*F, adequate auxiliary feedwater capability is available. One turbine driven pump full capacity (920 gpm) and the l
two half-capacity motor-driven pumps (460 gpm, each) are specified.
l However, only one-half capacity motor driven pump is necessary to l
~ of a small break loss-of-coolant accident."
supply auxiliary feedwater flow to the steam generators in the onset
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In review of the present TS 3.4.1 bases, there was a minor change to the
first statement above, in that either both motor-driven or the turbine driven pump are required initially to remove decay heat, with one EFW pump eventually sufficing. There has been no change to the other statement in the '
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bases since it was originally issued in Amendment #78.
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' With respect to the UFSAR:
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. Section 10.6.1, Design Basis, stated, "The EFW system has a supply
l capacity of 460 gpm to each OTSG,...."
h Section 10.6.2, System Description and Operation, stated, " The EFW
system is required (per Chapter 14) during the following accidents: loss of coolant flow, loss of offsite power (including loss of all AC, main steamline break, small break LOCA (certain sizes), and loss of feedwater. The system i
is capable of safe shutdown with one motor driven pump (delivering 400 gpm) during a loss of feedwater or small break LOCA following a seismic
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Section 14.2.2.7 LOFW accident was very confusing as to the actual design
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basis accident and the resulting minimum required flowrate. Values of 500 gpm with and then without a LOOP are discussed. A statement that flow rates as low as 370 gpm were found to provide satisfactory performance, i
based on a B&W analysis was made in a section discussing a LOFW with a
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LOOP. Further, Table 14.2.-33 was referenced as an associated table of key parameters which listed 740 gpm for TMI-1.
The inspector identified three problems with the October 1997 calculation and its implementation into ST 1303-11.428,EFW Capacity Test, with TCN 1-97-0143, dated October 14,1997.
The calculation cover sheet noted that it was a design basis calculation;
however, there was no safety evaluation completed for this instance. The calculation process did not ensure that a change to a design basis calculation received a safety evaluation.
The safety determination for the TCN was inadequate, since it did not
identify that a safety evaluation was needed. The determination stated that since the new value of 480 gpm was not lower than the 370 gpm stated in UFSAR section 14.2.2.7,the implementation of the 480 gpm was not a change to the facility as described in the UFSAR. However, the inspector in review of NUREG 0680, as stated above, found that the 370 gpm was for a LOFW with a LOOP condition (RCPs not running), while the 480 gpm in the new calculation assumed a LOFW without a LOOP (RCPs running) as the worst case condition. As stated above, NUREG -0680 and at least once in UFSAR section 14.2.2.7, a LOFW with RCPs running appeared to be the worst case LOFW accident, with 500 gpm being the minimum analyzed EFW flowrate.
As such, the use of the 480 gpm as the new minimum flowrate for a LOFW without a LOOP would have to be reviewed to ensure that an unreviewed
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safety question did not exist, since it was less than the 500 gpm as stated in the UFSAR for the same accident scenario.
Because there was no safety evaluation on the calculation or the procedure
change there was no pending UFSAR update (PFU) generated following the design basis change to the flowrates.
The inspector discussed these issues as they developed with the licensee engineering staff and management who stated that GPUN was in the process of:
Completing a review of the Chapter 14 LOFW ana!ysis. During this review
the licensee generated a CAP on several aspects of the Chapter 14 analysis documented in the April 1998 UFSAR update, Completing a safety evaluation for the implementation of the October 1997
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calculation values as a design basis change to the plant as described in the UFSAR to determine that the change did not involve an unreviewed safety question (USO). GPUN stated that this safety evaluation would include a historical look at how the design basis changed since the restar _. _ _ _... _ _. _. _
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Reviewing the calculation process for changes to ensure that design basis calculations were properly captured on PFUs.
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The inspectors considered the unclear EFW system de(' sign basis flowrate
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documentation in the UFSAR and the actions taken by GPUN to use the new design
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basis calculation without completing a safety' evaluation, to be an unresolved item, pending additional review of GPUN's subsequent safety evaluation. (URI 50-289/98-03-02)
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Conclusion The EFW system was operable based on review of calculations and testing, assuming that the October 1997 new design basis accident calculation was correct.
' However, the UFSAR Chapter 14 LOFW design basis accident analysis for the EFW system had not been well maintained. The calculation process was weak in that a
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design basis calculation did 'not require a safety evaluation once the design verification of the information had been completed. The safety determination process, used to justify a temporary procedure change to allow a lower EFW capacity test flowrate to depressurized OTSGs, was inadequate. Further, recent changes to the design basis in October 1997 had not been translated into a PFU which would have informed other GPUN employees that there was a new design
- basis calculation.
E2.2 Inse vice Tendon Surveillance a.
Insoection Scooe (37551)
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The scope of this inspection included an assessment of the general structural condition of the containment tendon gallery. The inspector completed this review
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as a result of a tour of the tendon access gallery, during the flooding prevention inspection discussed in Section 02.1 of this report.
The reactor building is a reinforced structure composed of a cylindrical wall with a
- flat foundation mat and a shallow dome roof. The cylindrical walls are prestressed with a post-tensioning system in the vertical and horizontal directions. Each tendon
consists of 1691/4 inch diameter wires encased in galvanized steel conduits which
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are filled with grease to limit corrosion of the wires. The hoop direction tendons are
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anchored on vertical buttresses. The exterior of these buttresses can be accessed from the outside of the reactor building through the tendon gallery. The vertical tendons are anchored at the base and top of the reactor building, the lower anchor point and lower portion of the tendon area are accessible in the tendon gallery, b.
Observations and Findinas The inspectors identified grease leakage from at least one end cap on a horizontal tendon. The inspectors notified the system engineer who confirmed that the end cap for tendon H1312 was leaking. Work request (WR) 792292 was initiated for
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the tendon end cap which including contacting the specialty contractor to investigate and perform repairs. In addition, the system engineer stated that GPUN would verify that adjacent tendons were not leaking, and they would perform the pertinent repairs or regreasing as needed. The leakage observed by the inspectors and confirmed by the system engineer was approximately one gallon.
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Further, there was indication of grease /oilleakage from numerous. vertical tendons l
through cracks in the concrete at the upper tendon access gallery. The inspectors
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reviewed the most recent tendon ST conducted in 1995 and the associated report
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to the NRC. In accordance with the ST, any grease leakage was required to be identified and evaluated by engineering. The ST did not identify any grease leakage l
from the vertical tendon, while the report stated that there was some indication of grease leakage and that it would not affect containment integrity. The inspector j
questioned why the ST, conducted by a contractor, had not noted the leakage but i
the report did, since the report should have been generated based on the ST results.
l GPUN replied that the engineer completing the report had toured that area and noted the condit:on. However, this was not reported on a surveillance deficiency report or specifically evaluated.
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In addition, the inspector found that GPUN did not have a specific process for monitoring and tracking the amount of grease leakage between the five year tendon surveillances. ST 1301-9.1, Section 9.4.3, Acceptance Criteria, stated that four l
gallons or more of the grease leakage requires an evaluation of the condition for possible degradation. TMI operations surveillance procedure No. S193, entitled
" General TMI-1 Inspection of Remote Areas by Operations Engineers," prescribed quarterly surveillance of the tendon gallery for housekeeping. This procedure would require maintenance to be conducted to cleanup any grease, but there was no specific treriding of the amount of the leakage.
The lack of continuity between the most recent (1995) tendon ST and the report to l
the NRC with respect to vertical tendon grease leakage, as well as the lack of a process for monitoring and tracking the amount of leakage allowed from a tendon
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l between the five year required ST were considered an inspector follow item (IFl 50-289/98-03-03),pending review of GPUN's corrective actions.
Several minor discrepancies were also identified concerning the condition of the tendon access gallery and the description of " Corrosion Protection" in section 5.2.2.5 of the UFSAR. These discrepancies included minor standing water and pealing epoxy coating at the lower level of the gallery. The system engineer j
acknowledged the discrepancies and documented them in a C/ P for correction.
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Conclusion
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Severalissues were identified dealing with the controls in place to document, track,
and assess r,eactor building tendon grease leakage, both from end caps on horizontal tendons and through concrete cracks at the lower portion of the vertical
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tendons. These tendon issues were considered an inspector follow item pending additional review of GPUN corrective actions. In addition, several minor
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discrepancies were identified between the actual tendon gallery and the associated UFSAR description and GPUN documented these discrepancies on a CAP for correction.
IV. Plant Support
R1 Radiological Prutection and Chemistry (RP&C) Controls (71750,83726,86750)
R1.1 Radiolooical Controls-Radioactive Materials. Contamination. Survevs. and Monitorino
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a.
Inspection Scooe A selective review of the licensee's control of radioactive materials, contamination,
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surveys, and monitoring, was performed, including adequacy of posted surveys, the establishment and maintenance of contamination area boundaries, calibration status of survey and monitoring equipment in use in the' plant, the proper u'se of personal contamination monitors and friskers, evaluations of personal contaminations,
' tracking of personal contamination events, the process for release of potentially
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- contaminated material to unrestricted areas, and recordkeeping for decommissioning _
planning. Information was gatnered through observation of activities, tours of the
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radiologically controlled area (RCA), discussions with cognizant personnel, and review and evaluation of procedures and documents, b.
Observations and Findinos
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Overall, effective performance in the area of radiological controls for radioactive materials, contamination, surveys, and monitoring was evident. The licensee used routine surveys, posteil at the RCA access control point and inside the RCA, as one
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of several methods to provide information to radiation workers on potential radiological hazards. These routine surveys were performed and updated on a periodic basis. Pre-job briefings were held as required by radiation work permits (RWPs). Electronic personal dosimeters provided real-time dose and dose-rate
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information,' and radiological barriers and postings reflected current radiological j
hazard information. Contamination areas were clearly outlined by barriers and
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postings, and no examples of materials or equipment lying across the boundary plane were found. Survey and monitoring equipment had appropriate calibration
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and performance check dates. In almost all cases, personnel and small article -
i monitors and friskers were being used properly by radiation workers. On one observed occasion, radiological control personnel at the RCA access control point j
detected and assisted several workers who were frisking articles improperly, l
Radiological control technicians effectively corrected this improper frisking which was of minor radiological safety significance. Personnel contamination events were j
being evaluated and tracked and trended. A process flow for contaminated
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materials and potentially contaminated materials from contaminated and uncontaminated areas of the RCA to the hot tool room, to radioactive waste management areas, or to release to unrestricted areas was in place. A file, containing records of information important to the safe and effective decommissioning of the facility, was maintained by the facility's licensing group.
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This file was examined and contained records appropriately responsive to the requirements in 10 CFR 50.75(g).
c.
Conclusions
Overall, effective performance in the area of radiological controls for radioactive materials, contamination, surveys, and monitoring was evident.
R1.2 Implementation of the Solid Radioactive Waste Proaram a.
insoection Scope A selective review of the availability of documentation of regulatory requirements, the licensee's verification of the license status of organizations to which they ship, operating procedures, processes and vendors used for solid waste management, radioactive shipments since the last inspection, use of scalin'g factors, radiological housekeeping, and the process control program (PCP) was performed. Information was gathered through observation of activities, tours of the RCA, discussions with cognizant personnel, and review and evaluation of procedures and documents.
b.
Observations and Findings individuals in charge of the solid radioactive waste program were knowledgeable and experienced. The licensee maintained up-to-date copies of the applicable NRC and Department of Transportation (DOT) regulations and licenses of f acilities to which radioactive waste or materials were shipped. Detailed procedures were available which addressed solid waste processing and disposalincluding the use of
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evaporators for processing liquid' waste, handling of spent cartridge filters and spent '
resin, cement solidification, resin dewatering, and waste storage. Since the last NRC inspection in this area, no shipments of dewatered spent primary resin were made, and all other waste shipments were transferred to radioactive waste processors for compaction, dewatering, or incineration. The licensee maintained appropriate qualification records for individuals authorized to certify low level radioactive waste (LLRW) shipments. Updated and audited procedures were used when scaling f actors were employed to quantify the concentration of hard-to-measure radionuclides in materials or for classification of wastes.
Radioactive waste facilities and equipment were well-maintained. Radioactive material and waste were properly labeled and stored. Areas and equipment were generally clean and uncluttered, and walkways were clear. To provide for greater efficiency and effectiveness, the licensee obtained a vendor-supplied computer software program for integrating the management of radionuclide scaling factors, classification of radioactive waste, determination of the proper shipment type, and generation of databases and shipping papers.
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c.
Conclusions Overall, the solid radioactive waste management program was effective based on the existence of appropriate procedures and controls, the acceptable condition of facilities and equipment, and the proper implementation of the program by knowledgeable and experienced personnel.
R1.3 Comoliance with NRC and DOT Reaulations for Shionina of Low Level Radioactive Waste for Disposal and Transportation of Other Radioactive Materials a.
Inspection Scooe A selective review was performed of partially loaded shipments and of completed shipping documentation packages. The shipping documentation for five completed radioactive shipments was reviewed. The shipments included f ailed incore detectors for evaluation, a reactor coolant sample for analysis, a waste evaporator concentrctes sample for analysis, protective clothing to a laundry vendor, and waste evaporator concentrates for offsite processing. The documentation reviewed included radiation and contamination surveys, the licensee's determination of DOT shipment subtype, packaging, marking, labeling, and placarding requirements, shipping paper requirements, driver's instructions, and emergency response information. Information was gathered through observation of activities, tours of the RCA, discussions with cognizant personnel, and review and evaluation of procedures and documents.
b.
Observations and Findinas
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Partially loaded shipments of laundry and of dry active waste were observed, and no noncompliances or discrepancies were noted. The radioactivity contained in each completed shipment was described in terms of the appropriate international system (SI) units as required, and the shipments were made in compliance with NRC and DOT regulations except for the lack of the emergency response telephone number being on the shipping paper as required by 49 CFR 172.201(d). The emergency response telephone number was included on the shipping paper for shipments of radioactive waste but was missing on the shipping paper for shipments of radioactive material. However, the emergency response telephone number was included on the driver's instructions and the bill of lading, when used, which accompanied t ' shipping paper for shipments of radioactive material. The licensee stated that a procedure change would be made to correct this issue. This failure to include the emergency response telephone number on the shipping paper for radioactive material shiprnents as required by 49 CFR 172.201(d) constitutes a violation of minor significance and is not subject to formal enforcement action.
c.
Conclusions The prograrn for shipping of low level radioactive waste for disposal and transportation of other radioactive materials was generally effective. The failure to include the emergency response telephone number on the shipping paper for
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radioactive material shipments as required by 49 CFR 172.201(d) constituted a violation of minor significance and was not subject to formal enforcement action.
l R5 Staff Training and Qualification in RP&C Activities (86750)
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a.
Insoection Scooe
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A selective review was performed of training in response to IE Bulletin No. 79-19, l
" Packaging of Low-Level Radioactive Waste for Transport and Burial" and to Subpart H, " Training," of 49 CFR 172. Information was gathered through
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discussions with cognizant personnel and review and evaluation of procedures and documents, b.
Observations and Findinas in response to IE Bulletin No. 79-19, the licensee provided applicable training and periodic retraining for radioactive waste group supervisors in the NRC and DOT regulatory requirements and waste buriallicense requirements. Separato training l
and periodic retraining in the NRC and DOT regulatory requirements for shipping, pertaining to radiological considerations, was provided to radiological control technichns. The training materials were appropriate. Current individual training
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l recoras were maintained. Training, as specified by 49 CFR 172 Subpart H, was l
provided to all radwaste personnel through a combination of General Employee l
Training and of Environmental Awareness Hazardous Chemical Worker Training
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which were both conducted on an annual basis.
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Conclusions
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The NRC and DOT training and retraining requirements for radioactive waste group
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personnel were satisfied.
l R7 Quality Assurance and Self Assessment in RP&C Activities (83726,86750)
R7.1 Radioloaical Controls Activities a.
Insoection Scope A selective review was performed of three self-assessments by the radiological controls group. These self-assessments included topics in the area of radiological l
controls for radioactive materials, contamination, surveys, and monitoring.
l Information was gathered through observation of activities, tours of the P.CA, discussions with cognizant personnel, and review and evaluation of procedures and
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r documents, b.
Observations and Findinas i
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Self-assessments by the radiological controls group in the area of radiological j
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controls for radioactive materials, contamination, surveys, and monitoring were
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normally good and provided recommendations and appropriate corrective actir,ns for identified deficiencies and problems. However,in some areas of minor radiological safety significance, identification of issues may not occur until after a potential problem becomes apparent. A selective review was performed of the following three self-assessments by the Radiological Controls Group:
Internal / External Dose Assessment, January 13,1998,
Control of Radioactive Material, March 31,1998,and 1997 Contamination Events, April 30,1998.
- The assessment of internal and external dosimetry, in part, addressed skin and internal dose evaluations, focused on pre-planned attributes, and resulted in two detailed recommendations accompanied by five radiological control action items.
The assessment of the control of radioactive material addresued transfers, storage, and waste processing and shipping, was focused on pre-planned objectives, and resulted in four rec' nmendations accompanied by three radiological control action items. The asses'
.nt of 1997 contamination events was thorough and detailed and resulted in thL recommendations, each with an associated action item.
c.
Conclusions Self-assessments by the radiological controls group in the area of radiological controls for radioactive materials, contamination, surveys, and monnering were normally good and provided recommendations and appropriate corrective actions for identified def,iciencies and problems.
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R7.2 Radioactive Waste Activities a.
Inspection Scope A selective review was performed of quality assurance end self-assessment activities conducted since the last NRC inspection, including; Audit Report S-TMI-97-02, Monitoring Reports 97-17 and 97-30, and selected CAPS. Information was gathered through discussions with cognizant personnel and review and evaluation of procedures and documents, b.
Observations and Findinas Audit Report S-TMl-97-02 covered radwaste management and the PCP. This audit resulted in the identification of three findings and nine minor deficiencies. Selective review of corrective actions on the findings showed timely resolution. The audit was programmatic and extensive in scope and highly detailed. Two surveillances in the radwaste and transportation area were conducted since the last NRC inspection.
Monitoring Report No. 97-17 involved a Type A quantity of laundry shinped as low specific activity - II, and no negative findings were documented. Mointonng Report No. 97-30 was a surveillance of a Type B quantity shipment of radioactive material
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involving a reportable quantity, fissile material, and a highway-route-controlled quantity. This surveillance did not result in any identified discrepancies. A selective review of licensee-identified issues for the past year revealed that corrective actions were completed in a timely manner, c.
Conclusions Quality assurance activities were generally effective. The audit performed as required by technical specifications was extensive in scope and depth.
RS Miscellaneous RP&C lssues R8.1 (Closed) Insoection Follow-uo item 50-289/97-05-01 Preparation to maintain constant telephonic communication with emergency response personnel until complete transfer of information regarding radioactive material shipments in transit. Additional clarification and guidance was placed in Administrative Procedure 1072, Hazardous Material Transportation incidents, to ensure that control room personnel maintain emergency response communications as recommended in NRC Information Notice 92-62. These actions were acceptable and this item is closed.
R8.2 (Closed) Insoection Follow-uo item 50-289/97-06-02 Disposition of responsibilities and authorities of the eliminated position of Corporate Radiological Health / Safety Director relative to review and maintenance of the corporate Radiation Protection Plan and to the required annual review of Radiation Program content and imp'lementation for each site. The lead responsibility and ownership for the various elements of a common corporate radirttion protection program, by mutual agreement of each site radiation protection organization, have been assigned to site personnel and documented in " process scope and boundaries" and " center of excellence scope and boundaries" documents. A team of evaluators, including two health physicists from an outside organization, conducted the annual 10 CFR Part 20 assessment between January 28 to February 6,1998. The health physicists from outside organizations provided a degree of independent review and assessment.
The inspector concluded that the licensee's actions were acceptable and this item is closed.
R8.3 (Closed) EA 50-289/97-533-03014(and eel 50-289/97-09-05)
Failure to make surveys (to verify elimination for hot particles) to assure compliance with 10 CFR 20.1201(a)(2)(ii),which limits radiation exposure to the skin. Hot particle control requirements were contained in the radiological controls procedure for radiation work permits. This procedure was revised to require that, in an emergent hot particle situation, work was to be stopped, and a radiological control supervisor was to be notified. Additionally, several existing controls were changed
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l from procedural recommendations to requirements.
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V. Manaaement Meetinas
- X1 Exit Meeting Summary i
The preliminary results of this inspection were discussed at the site with licensee
- management on July 31,1998. The licensee acknowledged the findings.
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INSPECTION PROCEDURES tlSED IP 37551 Onsite Engineering IP 61726 Surveillance Observations IP 62707 Maintenance Observation IP 71707 Plant Operations IP 71750 Plant Support Activities l
IP 83726 Control of Radioactive Materials and Contamination, Surveys, and Monitoring IP 86750 Solid Radioactive Waste Menagement and Transportation of Radioactive Materials ITEMS OPENED, CLOSED, AND DISCUSSED poened:
50 289/98-03-01 URI Adequacy of and Controls over Reactor Building to Tendon Access Gallery Flood Seal (Section O2.1)
50-289/93-08-02 URi Unclear Design Basis and Failure to Conduct a Safety Evaluation for a Change to the Emergency Feedwater System Design Basis Calculation (Section E2.1)
50-289/98-03-03 IFl Review Corrective Actions for Horizontal and Vertical Tendon Grease Leakage (Section E2.2)
Closed:
50-289/97-05-01 IFl Preparation to maintain constant telephonic communication with' emergency response personnel until complete tranafer of information regarding radioactive material shipments in transit (Section R8.1)
50 289/97-06-02 IFl Disposition of respansibilities and authorities of the eliminated position of Corporate Radiologiczl Health / Safety Director relative to review and maintenance of the corporate Radiation Protection Plan and to the required annual review of Radiation Program content and implementation for each site.
(Section R8.2)
50 289/97 533 EA Failure to make surveys (to verify elimination for hot-03014 particles) to assure compliance with 10 CFR (eel 50-289/97-09-05)
20.1201(a)(2)(ii), which limits radiation exposure to the skin. (Section R8.3)
Discussed:
None
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LIST OF ACRONYMS USED BS Reactor Building Spray (ECCS)
' B&W -
Babcock and Wilcox-
' CAP Corrective Action Process CFR Code of Federal Regulations CPM Counts Per Minute DH Decay Heat Removal (ECCS)
- DOT'
Department of Transportation DR Decay Heat River (ECCS)
EDG Emergency Diesel Generator eel Escalated Enforcement item EFW Emergency Feedwater EP Emergency Procedure GPUN GPU Nuclear, Inc.
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l&C -
Instrumentation and Controls IFl Inspection Followup Item
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IPEEE Individual Plant Evaluation External Events IR
Inspection Report
Inservice Test
JO
Job Order
Low Level Radioactive Waste
Loss of Feedwater Accident
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NR
Nuclear River Water System
NRC
Nuclear Regulatory Commission
Once Through Steam Generator
P'rocess Control Program
- PDR
Public Document Room
PFU-
Pending UFSAR Update
- PMT
Post-Maintenance Test
Plant Process Computer
Plant Review Group
RCA.
Radiological Control Area
Reactor Coolant Pump
- RP&C
Radiation Protection and Chemistry Controls
Radiation Work Permit
. international System
ST-
Surveillance Test
. TCN'
Temporary Change Notice
TMl
Three Mile Island-Unit 1
TS
Technical Specification
Updated Final Safety Analysis Report
~ URI
Unresolved item
Work Request
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