IR 05000320/1980019
| ML19346A069 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/20/1981 |
| From: | Conte R, Fasano A, Oneill B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19346A067 | List: |
| References | |
| 50-320-80-19, NUDOCS 8106040522 | |
| Download: ML19346A069 (11) | |
Text
o-U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT 50-320-80-10-27 Region I 50-320-80-11-08 50-320-81-01-13 Report No. 50-320/80-19 Docket No.
50-320 Category c
License No. OPR-73 Priority
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Licensee:
- ietropolitan Edison Company P. O. Box 480 Middletown, PA 17057 Facility Name:
Three Mile Island Nuclear Station, Unit 2 Inspection at: Middletown, Pennsylvania Inspection conducted:
December 7, 1980, - January 31, 1981 Inspectors: JNcLOW 4/ l'; [? !
R. ContF, Senior Resident Inspector date signed dual /$!
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B.O'Ne/l,pdia on Specialist date signed
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Approved by:
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A. Fasano, Chief, Three Mile Island Resident
'date' signed Section
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Inspection Summary:
Inspection on December 7,1980 - January 31,1981 (Inspection Report No. 50-320/80-19)
Areas Inspected:
Routine inspection by resident inspectors of licensee action on previous inspection findings; plant operations; licensee event reports (in-office review); plant operations review comittee membership; i
temporary nuclear sample system radiological controls; portable chemistry laboratory radiological controls; and, building cork seal contamination.
The inspection involved 87 inspector-hours by 2 NRC resident inspectors.
Results: Of the seven areas reviewed, one item of roncompliance was identified in one area (Violation - failure to follow radiological I
control procedures to support the operation of the temporary nuclear sample system, Paragraphs 7 and 8).
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DETAILS 1.
Persons Contacted General Public Utilities (GPU) Nuclear Group
- C. Adams, Operational Quality Assurance Inspector J. Barton, Director Site Operations, TMI-2 L. Beeman, Chemistry Technician
- J. Brasher, Radiological Controls Manager, TMI-2 J. Chwastyk, Plant Operations Manager, TMI-2 H. Collins,BabcockandWilcox(B&W) Chemist R. Croll, Unit 2 Health Physics (HP) Foreman
- R. Fenti, Site Audit Supervisor, Quality Assurance K. Harner, Chemistry Supervisor, TMI-2 E. Heffner, Instrument and Control (I&C) Technician, THI-2 C. Hitz, Engineer, TMI-2 K. Hofstetter, Supervisor Radiation Chemistry, TMI-2 P. Keegan, Unit 2 HP Technician
- L. King, Plant Operations Director, TMI-2
- G. Kunder, Technical Specification Compliance Supervisor, TMI-2 S. Levine, Manager Plant Maintenance, THI-2 D. Moyer, Unit 2 HP Technician P. Newkirk, Unit 2 HP Foreman
- R. Newman, Licensing Engineer, TMI-2 K. Norman, Unit 2 HP Technician W. Pitka, B&W Chemist B. Presgrove, Radwaste Engineer J. Renshaw, Manager Radiological Field Operations
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D. Tuttle, TMI Radiological Assessment O. Weaver, I&C Supervisor, TMI-2 NRC TMI Program Office
- L. Barrett, Acting Deputy Program Director, TMI Program Office R. Bellamy, Chief, Technical Support Section J. Wiebe, Systems Engineer Other members of the operations, radiological controls and administrative staffs were also interviewed.
- denotes those present at the exit interview.
2.
Licensee Action on Previous Inspection Findings (0 pen) !'oncompliance (320/79-20-06):
failure to properly control drawings. The adequacy of drawing control was further reviewed by the NRC (paragraphs 3.b and 3.c).
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Plant Operations a.
On a periodic basis the resident inspector obtained infor-mation on plant conditions, reviewed selected plant parameters for abnormal trends, ascertained plant status from maintenance /
modification viewpoint, and assessed logkeeping practices in accordance with administrative controls.
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During the review the resident inspector made random visits to the control room during regular and back shift hours, discussed operations with control room personnel, reviewed selected control room logs and records and observed selected licensee plan-of-the-day meetings.
In addition, a plant tour was conducted to assess housekeeping and fire protection measures.
Specific review items follow.
b.
On January 13, 1981, at approximately 4:45 p.m. an estimated 600 gallons of concentrated sulfuric acid were inadvertently spilled onto the ground on the south side of the Unit 2 Turbine Building.
The acid was from the plant's Water Treat-ment System Acid Storage Tank, WT-T-7, located in the Coagulator 3uilding adjacent to the south side of the Turbine Building.
The spill was due to the opening of TF-WT-V3, Tank Truck Supply Isolation Valve, which established a flow path from the i
T-7 tank to a truck "take down" connection (which had no installed blank flange) outside the Coagulator Building.
As a result, acid flowM into an outside security cable chase and an Emergency Feed Water (ER4) pipe chase and into the plant storm drain system.
Licensee's immediate actions included stoppage of the spill (V-3 shut), damming of the storm drain system to preclude release (nonradioactive) to the river, and neutralization of the acid by application of lime to the affected areas and subsequent flushing with fire service water.
An event of " potential public interest" was declared and local government officials (including an NRC resident inspector) were immediately notified of the event.
Licensee's supplemental actions included an investigation into the cause of the event.
A description of the event and licensee corrective actions were documented in an internal report " Superintendent's Event Report", dated January 16, 1981.
Significant outstanding actions included: ground area turnover to assure complete neutralization of acid that soaked into the ground, and analysis of the long term effects of the acid on chase cables, pipes and concrete.
Due to the neutralization it was noted that no adverse short term effects (immediate failures) were observed on equipment serviced by the subject cables / pipes.
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-4-In summary the licensee attributed the cause of the event to operator error.
Based on further discussion with licensee representatives it was also determined that a contributing factor to the cause of this event was the fact that the subject valve and associated piping were part of a 1978 plant modification and that these components were not incorporated into the system drawings in the control room.
The modifi-cation package drawings were in a separate file in,a room adjacent to the control room.
These drawings were usually the basis upon which switching and tagging orders were generated.
For this event FT-WT-V3 was opened under a switching and tagging order for restoration of the components following maintenance.
Apparently the responsible operators did not sufficiently review all applicable drawings for the system.
Licensee disciplinary action was taken against the operators involved in the event.
The subject of adequacy of drawing control was discussed with licensee representatives as noted in Paragraph 3.c below.
NRC review of licensee outstanding actions as a result of this event, as noted above, continues and is considered unresolved (320/80-19-01).
c.
During resident inspector observations in the control room on January 29, 1981, it appeared that a Standby Pressure Control
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(SPC) System flow diagram posted near the SPC operating panel was outdated (October 1979).
The SPC System was installed as a post accident system to provide the Reactor Coolant System (RCS) with another source of borated water and another means of RCS pressure control.
It was noted that pump SPC-P-3 was mis-labelled on the drawing as 40 gpm instead of 10 gpm.
The licensee representative acknowledged the error, however,it was stated that this drawing was the drawing that was updated to provide the operator with as-built information.
During the
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review of the event noted in Paragraph 3.b above, it was determined that the licensee's present system of maintaining as-built information in the control room for the operator was to maintain a file of complete modification drawings in addition to the bank of controlled drawings.
The inspector stated that this methodology appeared to be cumbersome in that a complete system condition (" picture")
could only be achieved by multiple drawing review.
The licensee representative acknowledged the above and stated that in light of the numerous ongoing modifications to the plant since the accident an instantaneous " picture" of a system on one drawing was unrealistic.
Further it was stated that the licensee's quality assurance (QA) department was required to conduct a special audit in the drawing control area and that the posted SPC drawing was to be revised.
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-5-This area continues to be reviewed by the licensee (previous inspection finding - 320/79-20-06).
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During a review by the onsite NRC staff of a change (Temporary Change Notice (TCN) No. 2-81-06, dated January 12,1981) to Surveillance Procedure 4303-M26 A/B, Boric Acid Pump Functional Test, Revision 1, September 9,1980, it was noted that a system header pressure gage, CA-14-PI-12 was used in place of the "A" pump discharge pressure gage, CA-P-2003.
The cali-bration/ operability of PI-12 was questioned during the review of this procedure change.
The resident inspector reviewed this area and noted the following:
The last calibration for CA-14-PI-12 was conducted in
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March 1977 while the last calibration for CA-PI-2003 and 2004 was conducted in June 1980.
However, it did not appear that these gages were incorporated into the licensee's program to calibrate instruments used to comply with Technical Specifications (TS) although not specifically addressed (implied) by TS.
Per TS 4.1.1.1(e)l-3, either the "A" or "B" Boric Acid
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pump was to be operable.
Based on a review of completed data for Procedure 4304-M26 A/B for 'the period June 1980 to November 1980 the "B" Boric Acid Pump was satisfactorily
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tested with a calibrated gage CA-PI-2004.
This TS was subsequently deleted November 14, 1980, by NRC Order amending the TS.
The header gage CA-14-PI-12 was used for troubleshooting
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the problem with gage CA-PI-2003 on the "A" Boric Pump.
This troubleshooting led to the determination of a clogged sensing line.
Repair was subsequently hampered due to relatively high area radiation levels.
As a part of this review the resident inspector questioned inordinate time to calibrate CA-14-PI-12 and further questioned why these gages were apparently not put into a formal calibration program. At the close of the inspection period this area was under review by the licensee and the licensee's representative stated that the TS would again be reviewed to assure no gages were missed from the established program for calibration of instruments used to comply with existing TS.
The resident inspector stated that this area is unresolved pending completion of licensee action as stated above and subsequent NRC:I review (320/80-19-02).
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4.
In-Office Review of Licensee Event Report; (LER's)
The resident inspector reviewed LER's submitted to NRC: Region I to verify that the details of the event were clearly reported, including the accuracy of the description of cas w and adequacy of corrective action.
The inspector determined wherner further information was required from the licensee, whether generic implications were indicated, and whether the event warranted onsite followup.
The following LER's were reviewed.
80-49/0ll-0, Unauthorized Use of an ASME (American Society of
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Mechanical Cngineers) Code On relief Valve for the Long Term
"B" Cooling System by the Valve Supplier 80-50/01L-0, Incore Thermocouple, H-9, Exhibited Erratic
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Behavior and was Therefore Declared Inoperable No items of noncompliance were identified.
5.
Emergency Drills During this period the resident inspectors observed licensee conducted emergency drills.
On December 15, 1980, a site emergency level drill was conducted to comply with NRC: Region I Imediate Action Letter No. IAL 80-31, dated September 17, 1980.
The resident inspectors assisted the Region I Drill Team in this effort.
Details are documented in Office of Inspection and Enforcement Inspection Report No. 50-320/80-20.
On Monday, January 5,1981, a small scale emergency drill (mini-exercise) was conducted by the licensee and monitored by the resident inspector.
The exercise simulated two injured, contaminated individuals within a TMI-2 restricted area.
Although some minor comunications problems existed, overall response was excellent, and within fifteen minutes all necessary actions had been successfully completed by licensee personnel.
The resident inspector has no further coments in this area.
6.
Plant Operations Review Committee (PORC)
During this period a licensee quality assurance audit identified a discrepancy between actual PORC membership and the required PORC membership per Technical Specification (TS) 6.5.1.2.
The TS references Section 4.4 of ANSI 18.1-1971; however, this section addresses technical personnel only in the areas of Reactor Engineering /
Physics, Instrument and Controls, Radiochemistry and Radiation Protection.
The actual membership (seven minimum) does not comprise only those areas, that is, all seven members do not only fit those categories of technical expertise.
Other areas such as Mechanical, Electrical and Radwaste Engineering are also represented.
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-7-T.a licensee indicated that the present membership met the intent of the TS, that is, to provide the PORC with a broad base of engineers from various disciplines as noted above.
The licensee also ind,cated that a TS change would be submitted to clarify this area.
The resident inspector, in consultation with the licensing project manager, acknowledged the above and stated that this area was unresolved pending completion of stated licensee action and sub-sequent NRC:I review (320/80-19-03).
7.
Temporary Nuclear Sample System Radiological Controls The Temporary Nuclear Sample System was constructed as a system to provide sampling capabilities (primarily RCS water) within Unit 2 rather than in Unit 1 to reduce the radiological hazard in Unit 1.
During this report period, an inspection of the Unit 2 temporary nuclear sample sink / radiation safety procedures was conducted.
The inspector observed procedures during resin column testing on December 18, 1980, and RCS sampling on January 12, 1981.
The in:pector reviewed applicable portions of the following procedures; 2104-4.62, " Temporary Nuclear Sampling System" (Tempcrary
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Change Notice (TCN) No. 2-80-419, dated December 19,1980);
and,
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2104-4.10. " Resin Column Testings" (TCN No. 2-80 314, dated
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September 23,1980).
While inspecting the sample sink area on December 18,1980, the inspector noted that two radiation monitors in the sample sink room, SNS-RE-10 (Sample Hood Area Radiation Monitor) and SNS-RE-30 (Hood Ventilation Mo.itor), were inoperable.
The inspector also noted that the last :alibration of the monitors by the Instrument and Control (I&C) Department as marked on the monitors, was performed on November 1,1979.
Licensee personnel stated that the monitors had not worked for more than six months, and that the monitors were not necessarf since work in the hood area required a radiation work permit (RWP), Health Physics (HP) escort, and special HP monitoring.
The inspector determined that although the two radiation monitors had been inoperable for an extended period of time prior to December 18, 1980, and RCS samples had continued, no proper notification, procedural change, or repair request had been initiated prior to December 18, 1980.
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-8-Procedure 2104-4.62, requires in part that when either radiation monitor SNS-RE-10 or SNS-RE-30 is inoperable, the sample sink is to be imediately shut-down and Unit 2 HP be notified.
The inspector stated that failure to follow Station Operating Procedure 2104-4.62, paragraph 4.1.4, was an apparent item of noncompliance (Paragraph 8)(320/80-19-04).
The inspector interviewed licensee personnel and reviewed records.
It was noted that:
On December 19, 1980, the lice see initiated TCN No. 2-80-419
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to Procedure 2104-4.62 to address the inoperable monitors; and, On December 23, 1980, the licensee submitted a work request to
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I&C for repair of the two monitors.
8.
Chemistry Trailer Radiological Controls The Babcock and Wilcox (B&W) chemistry trailer located outside of the Unit 2 Turbine Building was installed as a support facility for EPICOR-I, a low level pcrtable liquid radioactive waste processing system.
Later the trailer was used for testing in support of the activities performed by EPIC 0R-II.
The EPICOR-II analysis included high activity samples.
In December 1980 the B&W trailer was used as the testing laboratory for Unit 2 RCS samples.
The inspector examined the RCS samp aandling and analysis at the B&W chemistry trailer and reviewed tne applicable radiation safety procedures.
The inspector observed procedures and activities on January 6, 9,12, and 16,1981.
Numerous industrial and radio-logical inadequacies were noted during these reviews:
Fire and industrial hazards - excessive plastics and trash in
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and storage of volatile /flamable materials; improper electrical
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wiring; unsecured acetylene tank; inadequate fire protection training for personnel; and, blocked emergency exit.
i Radiological inadequacies - inadequate training; excessive
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storage of highly contaminated waste materials, and previous
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l RCS samples; and, inadequate monitoring.
The inspector interviewed licensee personnel and reviewed pertinent records.
The following findings were determined:
Three RCS samples were tested prior to January 5,1981, under
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standing RWP's rather than a specific RWP for the analyses in the B&W trailer.
A specific radiological condition for this j
evolution was a sample contact reading of 4 R/hr y.
This specific radiation level precluded the use of a standing RWP
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dated January 2,1981, (effective for one week). Since a specific RWP was not issued for these evolutions, this represents apparent noncompliance with Procedure 1691, Revision 2, dated April 14, 1980, B&W Cap-Gun-Chemistry Trailec, Paragraph 2.2, which required in part that an RWP be issued to cover the sample analysis area including anticipated monitoring requirements dependent on levels of activity.
During December 1980 the HP daily surveys of the trailer were
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not performed as required.
Daily survey was discontinued on November 19, 1930, when a program of biweekly survey was instituted, contrary to Procedure 1691, Paragraph 2.9.
The biweekly surveys usually occurred on Wednesdays and Saturdays, during periods when no RCS samples were processed (RCS testing occurred on Mondays and Tuesdays).
The inspector determined that Procedure 1691 was not properly implemented in that during December 1980, a special RWP was not obtained for work with RCS samples in the B&W trailer, and in that, HP's did not perform a daily survey of the facility.
These findings represent another example of an apparent item of noncompliance in the area of procedure implementation (Paragraph 2)
(320/89-19-04).
Due to the potential radiological hazards involved, the NRC onsite staff discussed with licensee representatives necessary and immediate corrective actions to assure the safety of the radiological operations at the onsite B&W chemistry laboratory.
The licensee suspended all operations at the onsite B&W laboratory pending completion / implementation of required modifications and operating procedures necessary for upgrading the safety of radiological operations at the laboratory.
The licensee committed to conduct periodic evaluation of the radiological and industrial conditions in the trailer.
9.
Cork Seal Contamination On November 26, 1980, during a routine radiological survey, the cork seam found along the auxiliary building to the control and service building interface was found to be contaminated.
This area is inside the licensee's radiological control area (RCA). The radiation and contamination levels when measured very close to t.1e seam (within 6 inches) were above the general prevailing levels in the area.
The cork was found to be essentially saturated with water.
The licensee undertook a program to determine the levels and distribution of the contamination and to identify the source.
Seams which were contiguous to the first seam were sampled and analyzed.
Three possible sources of contamination were examined:
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-10-Residual contamination from flooding of this area early in the
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March 1979 accident; Leakage from the seal injection cubicle; and,
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Leakage from the reactor building sump,
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Since significant leakage from the reactor building sump could c
result in radiological problems, much of the initial investigation was oriented in this direction.
The tendon gallery was surveyed and the tendon gallery sump was sampled.
The results indicated no leakage from the reactor building sump.
The cork seams have water stops (rubber) on the sides where they terminate (contact with the ground).
Samples taken outside the water stops yielded no detectable activity.
Results of cork samples taken showed a general decrease in activity with increasing distance from the seal injection cubicle.
The 90Sr/137 s ratio varied from 1:140 to 4:1 with a predominance in C
the range of 1:20 to 1:8.
The variability is probably due to a variety of factors including differential solubility of Sr and Cs and ion exchange between Sr and the Cs in the concrete bounding the seams.
The data essentially eliminates residual contamination from March 1979 flooding of the Auxiliary Building as a cause, in that the water 90 r.
initially flooding the building cgSr/g only traces of tai S
The containment sump water has a Cs ratio of approximately 1:60 while primary coolant seal injection leakage is on the order of 2:3.
Smears taken in contaminated areas of the plant generally have 137 s ratios than primary coolant due to a possibly low lower 90Sr/
C mobility of the 90 r.
Samples taken very near the source of a leak S
sometimes show 90 r/137 S
Cs ratios greater than primary coolant.
Thus the sampling data and observed behavior of nuclides indicate that the source of the activity could be the seal injection cubicle.
As the licensee began a systematic sampling program of the cork seam, the water saturating the seam receeded.
The licensee plans to obtain additional water samples if the seams become saturated again.
The licensee's environmental groundwater wells have shown no increases in activity which could be attributed to the cork seams.
The licensee's planned corrective actions include:
Applying a strippable coating to the seams;
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-11-Repairing flashing on the roof between the reactor building
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and adjacent buildings to prevent rainwater from spreading contamination; and, Periodically monitoring cork seams for contamination levels
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and to detect any spreading of contamination.
The area is unresolved pending(completion of action by the licensee and further review by the llRC 320/80-19-05).
10.
Unresolved Items Unresolved items are findings about which more information is needed to ascertain whether it is an item of noncompliance, a deviation, or acceptable.
Unresolved items disclosed during this inspection are discussed in Paragraphs 3.b, 3.d, 6 and 9.
11.
Exit Interview On February 5,1981, the resident inspectors met with licensee representatives (denoted in Paragraph 1) to discuss the inspection
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scope and findings.
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