IR 05000309/1986013
| ML20215J724 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 09/19/1986 |
| From: | Lester Tripp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20215J719 | List: |
| References | |
| 50-309-86-13, NUDOCS 8610270163 | |
| Download: ML20215J724 (12) | |
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P-U.S. NUCLEAR REGULATORY COMMISSION
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REGION I
1-Docket / Report: 50-309/86-13 License:
DPR-36 Licensee:
Maine Yankee Atomic Power Inspection At: Wiscasset, Maine
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Dates:
July 29 through September 10, 1986 Inspectors:
Cornelius F. Holden, Senior Resident Inspector J frey obertson, Resident Inspector
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Approved:
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E. E. Trf$p, Chief, Reactor g s Section 3A
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Summary:
Inspection on July 29 through G id d 10, 1986 (Report No. 50-309/86-13)
i Areas Inspected:
Routine resident inspection (235 hrs.) of the control room, ac-cessible parts of plant structures, plant operations, radiation protection, physi-
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cal security, fire protection, plant operating records, maintenance and surveil-
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lance.
Results:
One plant trip during this-period was caused by an inverter failure.
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Personnel error during the recovery of the plant caused a momentary loss of one of the emergency buses.
Post-trip review revealed other equipment problems in-cluding feedwater regulating valve relay failure.
One allegation was received and closed. A problem with the release of contaminated sand into the owner controlled area was, in part, a result of inadequate followup to~an identified problem.
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8610270163 860926 PDR ADOCK 05000309 G
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DETAILS 1.
Persons Contacted Within this report period, interviews and discussions were conducted with various licensee personnel, including plant operators, maintenance technicians and the licensee's management staff.
2.
Summary of Facility Activities At the beginning of the report period, power was being maintained at 99%.
On August 1, a power reduction to 50% was initiated to facilitate the isola-tion and repair of an electro-hydraulic control oil leak on a reheat intercept valve to the No. 2 LP turbine.
Power was returned to 99% on August 3.
An automatic reactor trip occurred on August 10 due to low level in the No. 1 steam generator (details in section 9).
A startup was performed on August 11 and the plant was returned to 99.5% power on August 13.
The plant had not been taken to 100% power because of instabilities in the secondary systems.
During this time, the secondary systems were carefully monitored by the lic-ensee and, on September 4, power was increased to 100% where it remained until the end of the report period.
3.
Followup on Previous Inspection Findings a.
(Closed) Violation (50-309/85-05-01) and IFI (50-309/85-34-06).
No de-fined program for controlling access to warehouse or for performing pre-ventive maintenance on items stored in warehouse.
The following actions have been taken by the licensee to correct repeated deficiencies in access control to the warehouse:
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automatic door closer installed
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door locked to prevent operation without a key
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door modified to preclude use of shim material to retract locking bolt
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access controls discussed with each person issued a stores key, in-cluding disciplinary actions for failure to follow procedure.
Subsequent to these actions, access to the warehouse has been adequately controlled.
A computerized Inventory and Preventative Maintenance (PM) System has been developed and implemented for stored items.
The equipment in stores was researched by the Engineering and Maintenance Departments and a con-sultant to determine the appropriate PM requirements.
The PM program also ensures new items added to stores are incorporated into the program.
Further details of the PM Program are discussed in NRC Inspection Report
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50-309/86-07. A review of 17 purchase orders for safety-related equipment verified that the specific packaging and storage level requirement (i.e.,
A, B, C, or D) was specified.
Although the PM program is implemented and appears to meet the require-ments of ANSI 4.2.2-1972, the procedure to define and control the program is not consistent with the program in place.
The inconsistencies, pri-marily in program responsibilities, are a result of the newly implemented program evolving into a workable system. The licensee is aware of these inconsistencies and is presently revising the procedure.
This will be addressed incident to routine inspection.
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(Closed) Bulletin (50-309/86-8U-02).
Static "0" Ring Differential Pres-sure Switches.
Those licensees that have SOR Series 102 or 103 differ-ential pressure switches installed in systems subject to Technical Specifications were requested to take certain actions to assure that system operation is reliable.
In response to this bulletin, Maine Yankee stated in a letter dated July 24, 1986, that no SOR Model 102 or 103 dif-ferential pressure switches are installed as qualified electrical equip-ment required by the Maine Yankee Equipment Qualification Program and 10 CFR 50.49(6).
4.
Review of Licensee Event Reports (LER)
The inspector reviewed the following LERs to verify that the details were clearly reported, including accuracy of the description of cause and adequacy of corrective action.
The inspector had previously verified that appropriate corrective action was taken or responsibility assigned and that continued operation of the facility was conducted in accordance with Technical Specifi-cations and did not constitute an unreviewed safety question as defined in 10 CFR 50.59.
No discrepancies were identified.
LER No.
SUBJECT 86-04 Emergency Feedwater Pump inoperable due to a faulty circuit breaker.
86-05 Plant Trip on Low Steam Generator Level due to Static Inverter
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Failure
5.
Routine Periodic Inspections a.
Daily Inspection During routine facility tours, the following were checked: manning, ac-cess control, adherence to procedures and LCO's, instrumentation, recor-der traces, protective systems, control rod positions, containment tem-
t perature and pressure, control room annunciators, emergency power source operability, control room logs, shift supervisor logs, and operating i
orders.
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b.
System Alignment Inspection Operating confirmation was made of the Fuel Pool Cooling System.
Ac-cessible valve positions and status were examined.
Power supply and breaker alignment was checked.
Visual inspection of major components was performed.
Operability of instruments essential to system perform-ance was assessed. One minor discrepancy was identified in the labeling of a pressure instrument.
There were no further concerns.
c.
Biweekly Inspections During plant tours, the inspector observed shift turnovers, chemistry sample results and the use of radiation work permits and Health Physics procedures.
Area radiation and air monitor use and operational status was reviewed.
Plant housekeeping and cleanliness were evaluated.
d.
Plant Maintenance The inspector observed and reviewed maintenance and problem investigation activities to verify compliance with regulations, administrative and maintenance procedures, codes and standards, proper QA/QC involvement, safety tag use, equipment alignment, jumper use, personnel qualifications, radiological controls for worker protection, fire protection, retest re-quirements, and reportability per Technical Specifications.
e.
Surveillance Testing The inspector observed portions of tests, including the following, to assess performance in accordance with approved procedures and LCO's, test results, removal and restoration of equipment, and deficiency review and resolution:
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Diesel Fire Pump Surveillance
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Pressurizer Surge Line Sample l
CEA Exercising
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Hot Leg Injection Flow Indicator Calibration
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Weekly Battery Surveillance
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Post Accident Purge System Monthly Surveillance l
Operational Check of BD-T-32
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No unacceptable conditions were identified during the above routine periodic inspections.
6.
Observations of Physical Security
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i Checks were made to determine whether security conditions met regulatory re-quirements, the physical security plan, and approved procedures.
Those checks included security staffing, protected and vital area barriers, vehicle searches (
and personnel identification, access control, badging, and compensatory meas-j ures when required.
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7.
Radiological Controls Radiological controls were observed on a routine basis during the reporting period.
Standard industry radiological work practices, conformance to radio-logical control procedures and 10 CFR Part 20 requirements were observed.
Independent surveys of radiological boundaries and random surveys of non-radiological points throughout the facility were taken by the inspector.
No unacceptable conditions were identified.
8.
Air Ejector Radiation Monitoring System The licensee has experienced recent problems with the Radiation Monitoring System (RMS) and the main condenser air ejector (AE) detector.
Non-condens-able gases from the main condenser are removed by the air ejector.
The gases are vented to the plant stack.
The air ejector RMS consists of a detector in the vent line and control room indication and alarm.
During certain plant conditions (e.g., startup and shutdown) large amounts of steam are present in the air ejector vent line.
The moisture in the air ejector vent line has caused recent failures of the detectors.
As a result of these premature failures of the AE-RMS detector, the licensee purchased new style detectors from the manufacturer.
These detectors are more resistant to water vapor damage; however, recent plant experience has shown the detectors are difficult to calibrate.
As a result, the licensee has in-stalled one of the new type detectors, which is within calibration, at the low end of the scale but exceeds the tolerance for calibration at the high end of the scale.
Tne calibration specification is i 10 percent.
The detec-tor actually reads conservatively at +20 percent.
Even with this conservative reading the air ejector RMS channel has been declared inoperable.
Since the air ejector RMS normally reads at the low end of the scale, the licensee con-tinues to utilize the information for trending purposes but follows the Tech-nical Specification (T.S.) Remedial Action Statement.
No alarm functions are affected.
Technical Specification 3.28 requires grrb samples be taken at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
This specification was resised on July 1, 1986 in order to conform with Standard Technical Specifications for Radiological Effluents.
Because there was some uncertainty over what the revised Technical Specifica-tions required, discussions were held on the matter.
The licensee decided to implement the following instructions for sampling the air ejector HMS dur-ing periods when the instrument was inoperable.
A 24-hour period will be designated based upon a failure of the air ejector RMS.
The air ejector sample will be taken as close to the end of the 24-hour period as possible but not beyond it.
If the sample frequency were to exceed the designated 24-hour period, the Technical Specification would not be satisfied.
The desig-nated sample period can be altered provided samples are taken more frequently to ensure the 24-hour period is met.
In all cases, a sample must be taken in each 24-hour period.
Due to variations in daily sample routines, there will be slight variations in the time of each sample.
The inspector reviewed
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the above instructions and determined the 24-hour period designation meets the intent of the Technical Specification.
Small variations in the sample time would not contradict the Technical Specification and would allow for trending of results.
However, the Technical Specification would not be satisfied if the sample is not taken within the 24-hour designated period.
This was consistent with the licensee's instruction.
Based on this review, the inspector had no further questions.
9.
Plant Trip At 4:02 a.m. on August 10, 1986, an undervoltage condition on the 125-volt A.C vital bus was detected and alarmed in the control room.
The plant tripped on low steam generator level in No. 1 generator.
Number 1 main feedwater regulating valve failed shut on loss of power causing the trip.
Several problems were caused by the loss of No. 1 A.C. vital bus ~and several problems occurred unrelated to the loss of the bus.
The No. 1 main feedwater regulating valve gets its normal power supply from No. 3 A.C. vital bus.
On a loss of power, a normally closed relay should de-energize and transfer power to an alternate source (No. 1 A.C. vital bus).
Subsequent troubleshooting revealed this relay had previously failed and the alternate power supply (No. 1 A.C. vital bus) was in use prior to the trip.
When No. 1 inverter failed and deenergized No. 1 A.C. vital bus, No. 1 feed-water regulating valve lost power and failed shut.
Level was lost in No. 1 steam generator and the plant tripped on low level in No. 1 steam generator.
Further investigation revealed that there is no status indicator or periodic testing of this relay.
The first-out annunciator is designated to indicate the cause of the trip.
In this case, it indicated high pressure was the cause of the trip.
By re-viewing computer data (and lack of any indication of actual high pressurizer pressure) the licensee determined that the actual cause of the trip was low steam generator level in No. 1 generator.
Subsequent troubleshooting revealed a bad card in the pressurizer pressure annunciator.
This card was replaced.
Three of the eight reactor trip breakers did not appear on the sequence of events printout following the trip even though they indicated open.
Further investigation revealed the printer used to display the sequence of events be-came overloaded with data and lost power during the event.
This combination resulted in the initial printout of the trip eliminating these three breakers.
Subsequent printouts contained these three reactor trip breakers and their trip time was consistent with the other breakers (approximately 33 msec.).
Following a plant trip, the station service power automatically transfers from normal to off-site power.
The power supply breakers, upon switching to off-site power, display a green flag on the breaker switch in the control room.
This is to alert the operators that the breaker has switched to the off-site power. When conditions permit, the operators rotate the breaker switch to the correct position which then displays a red flag.
The operator selected the wrong position causing power to be lost to bus No. 2.
This caused a loss
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of voltage to the safety bus No. 6 and the emergency diesel started and sequenced loads onto the bus. Bus No. 2 also powers Nos. 2 and 3 reactor
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coolant pumps (RCP), and as a result, RCP's Nos. 2 and 3 tripped.
Under this condition the operators promptly secured No. 1 RCP as required by procedure (No. 2 RCP has no. backstop device.
To prevent damage to this pump, the other RCP's are secured as a precaution when No. 2 RCP could reverse direction).
Number 1 RCP was secured and the operators verified natural circulation.
Number 1 A.C. vital bus also powers channel No. 1 of the Reactor Regulating System (RRS).
Loss of the bus caused RRS channel No. 1 to fail low.
The steam dump and turbine bypass system get a signal from RRS for quick opening on a plant trip to reduce steam header pressure.
Since the RRS had failed low, the steam dump and turbine bypass system did not function.
Reconstruc-tion of the event several hours later indicated that the steam safety valves (one or more) must have opened.
Personnel in the area of the safety valves were questioned and stated that the noise level was higher than normal.
The licensee concluded that the safeties had lifted.
Prior to cross-tying the No. 1 A.C. vital bus with vital bus No. 3, the lic-ensee checked vital bus No.1 for grounds or other ~ problems. Additionally, the Recirculation Actuation Signal (RAS) was defeated momentarily since the repowering of the RAS relays could result in an inadvertent actuation of RAS.
The two buses were cross-tied.
Maintenance personnel discovered a broken wire in a shorting board for the inverter.
This wire was repaired and the SCR's were replaced.
Number 1 A.C.
vital bus was returned to its normal power supply.
The inspector reviewed the licensee's actions associated with this trip.
The licensee conducted an in-depth review of the error that caused the loss of power to safety bus No. 6.
An outside consultant was utilized to perform an evaluation utilizing a Human Performance Evaluation System developed by an industry group.
This evaluation recommended two hardware and one administra-tive change as ways to prevent recurrence.
Additionally, the licensee is re-viewing the use of a Human Performance Evaluation System to help identify potential problems before they result in performance errors.
The inspector had no further questions.
A plant startup was conducted at 7:00 a.m. on August 11, 1986.
10.
Ceiling Tile Removal from the Radiological Controlled Area The inspector received an allegation concerning release of ceiling tiles from the radiological controlled area.
Specifically, the individual was concerned that sufficient measures were not established to ensure that the ceiling tiles and related hardware were properly surveyed for radiological contamination prior to release'for disposal.
The inspector reviewed the circumstances sur-rounding this inciden d
- Maine Yankee is in the process of remodeling the Control Point area of the plant.
Relocation of the warehouse and Maintenance Departments has provided additional space for the radiological control area.
In support of this acti-vity, contractors are removing the suspended ceilings, stripping offices and erecting new walls.
The ceiling tiles used in the old suspended ceiling were sampled for asbestos and were found to contain some asbestos fibers (approxi-mately 5 to 10 percent).
The licensee implemented their asbestos control procedures which specify precautions to be utilized during the removal of products containing asbestos.
A radiological survey of the ceiling tiles was performed prior to their removal.
Procedures required the ceiling tiles to be removed under controlled conditions (filtered ventilation, plastic drop cloths covering area, wipe down and air sample prior to restoration of normal traffic).
Ceiling tiles were removed and packaged in double plastic bags (5 to 10 per bag).
The tiles were then surveyed through the plastic bag using an RM-14 meter with HP-210 probe.
The bags were set on the portal monitor and released if both the frisk through the plastic bag and the portal monitor check were passed.
If either test indicated activity, the bag was processed as radiological waste.
The individual's concern was that the initial survey of the ceiling tiles was done in place.
As a result, this frisk and swipe survey was performed only on the bottom side of the tiles.
Because dust could accumulate on the top of the tiles, the individual felt that this survey may be inadequate in de-tecting contamination of the tiles.
Additionally, a survey of the tiles through the plastic bag was not in conformance with established procedures for disposal of radiological building waste. The inspector reviewed the lic-ensee's radiological survey performed on August 12, 1986, of the ceiling tiles in the areas scheduled for ceiling removal.
Those areas were: radiological control check point, hotside locker room and head, secondary lab, primary chemistry lab and chemistry counting room.
This swipe and direct frisk survey of the lower portion of the tiles showed no contamination of the tiles.
(Some of the ventilation duct openings were included in this survey and showed some internal contamination.
These were annotated on the survey.
The ventilation ductwork was not removed during.the ceiling tile removal.) There were 58 swipes taken during this survey.
The inspector also reviewed a survey per-formed on August 13, 1986.
This swipe survey was performed on the top of the tiles through areas where the tiles had been removed for preliminary work.
Two swipes were taken in each of the 7 locations for a total of 14 swiper.
This survey did not find any contamination.
The inspector also reviewed the licensee's procedures for surveying and re-lease of material from the radiologically controlled area.
Procedure 9.1.1, Plant Radiological Surveys, revision 11, discusses release of material from the controlled area in section 7.3.
The two criteria used by this procedure to release an item specify no loose contamination greater than 1000 disinte-grations per minute (dpm) per 100 square centimeters (cm sq.) and no fixed contamination greater than.1 millirem per hour (mrem /hr).
The procedure also states that small or difficult to sample items can be carefully surveyed and released utilizing an RM-14 detector and HP-210 probe if they indicate less than 100 counts per minute (cpm) above backgroun.
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Procedure 9.1.22, Sorting of Primary Side Debris, revision 1, was also re-viewed.
This procedure requires trash with low radiation levels to be sorted in order to eliminate unnecessary clean trash from being included in radio-logical waste.
After segregating, the clean trash is disposed of as normal plant trash after it has passed through the portal monitor at the radiological controlled check point.
The portal monitor is a walk in, wait, then walk out survey instrument which utilizes large scintillation detectors to detect the presence of contamination.
As with all portal monitors, it is dependent on the geometry of the individual and the contamination.
The worst case geometry is approximately three feet from the top of the detector on the center line of a person's body.
During the two weeks of ceiling removal, approximately 60 to 70 bags of tile and debris were removed.
Of these, a total of 12 bags failed to be released from the area (i.e., were classified as contaminated).
These bags wcre seg-regated for disposal as low activity waste.
All other bags of tiles were double-bagged and segregated for disposal as asbestos waste.
During a routine tour of the plant, the inspector requested that three of the bags of tiles that were labelled as contaminated be resurveyed.
All three bags passed the through-the-bag frisk test and two of the bags passed the portal 'onitor test.
Since one more room was scheduled for tile removal the m
licensee decided to take advantage of the asbestos controls established for that room (the chemistry counting room) and conduct a detailed frisk survey of the contents of those bags.
Additionally, three bags which had previously been released were randomly selected and included in the survey.
All bags passed this test and were released from the radiologically controlled area.
The inspector discussed with the licensee the possible causes for the bags not passing the portal monitor test on one occasion but passing a second try and a more stringent frisk test.
The licensee replied that, since the portal monitor is geometry sensitive, it can be detecting contamination below the release level of 1000 dpm.
Because the bags were placed on the foot location, they were closer to the lower detectors. Additionally, portal monitors are more sensitive to dispersed radiation than a normal frisk since the area they look at is larger.
Based on the above, the licensee believes that the portal monitor was detecting radiation levels below the releasable limit.
The lic-ensee was classifying the ceiling tiles as contaminated based on this conser-vative measurement.
Although the tiles were processed differently than normal trash, the measures used appeared adequate to prevent unrestricted release of contaminated material.
The inspector discussed the results of this inspection with the alleger, who was not aware of the additional survey performed on top of the tiles prior to the start of the ripout.
The results of the licensee's frisk of the six bags of tiles on August 25, 1986, was also discussed.
The individual agreed that the results of this investigation showed no uncontrolled release of radioactive material had occurre.
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The inspector discussed the results of this inspection with the licensee.
Although no violations of requirements were found, there were areas where the inspector believed performance could be improved.
Those were a better under-standing of the portal monitor's ability to detect contamination and the lack of a questioning posture when tiles which were believe'l to be free of contami-nation were detected as contaminated at the portal monitor.
The licensee is reviewing these comments and the inspector will revisit this area in a future inspection.
Allegation RI-86-A-0102 is closed.
11.
Outdated Temporary Procedure Change in Control Room Procedure File On August 1, 1986, a copy of the operating procedure for the Excess Flow Check Valve Air Compressors was found at the compressors.
The procedure contained a Temporary Procedure Change Report (PCR) that used set points that were no longer applicable.
This temporary PCR was also still in the procedure file located in the control room.
This temporary change was only to remain in effect until August 1, 1986; how-ever, EFCV air compressor set points were returned to normal during a shutdown in July to replace the rupture discs.
The temporary PCR should have been terminated at that time.
The following points were discussed with the licensee:
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Method and responsibilities to ensure temporary PCR's are current
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Procedure left by air compressors was either not recognized as incorrect or not followed, in that-the final conditions step requires verification that the system is controlling air pressure within the stated set points.
Forty-four PCRs were reviewed to determine if the changes were still applic-able and the information current.
In all cases, the PCR was either current or terminated in accordance with the instructions on the cover sheet.
PORC reviews were completed within the required 14 days.
Because the set points were changed back to those specified in the body of the procedure (10 psig above those specified by the PCR) and no other discre-pancies could be identified, this item was determined to be unique and not a programmatic problem.
The inspector will continue to monitor this area.
12.
Linear Heat Generation Rate Limits On 8/28/86, new administrative linear heat generation rate limits were imple-mented and input into the plant computer alarms.
This is a result of concerns related to the axial power distribution assumed in the LOCA analysis.
The new administrative limits are more conservative than the existing Technical Specification limit _
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a The axial power shape, previously assumed by Yankee Nuclear Services Division (YNSD) to result in the most severe consequences in the LOCA analysis for Maine Yankee, was one with a peak near the top of the core.
As a result of model studies done for Yankee Rowe by YNSD, a flat axial power shape with the peak extended over the length of the core has been determined to result in the highest postulated peak clad temperature.
The revised administrative limits are to ensure that LOCA criteria are not exceeded by unanalyzed axial power distributions.
A review of the incore monitoring system data by the licensee indicates that they have not exceeded the new administrative limits during previous operation.
If the administrative limit is exceeded (and not the T.S. limit) YNSD will perform an evaluation to determine if an unanalyzed condition existed.
No unacceptable condition were identified.
13.
Contaminated Sand Pile Outside the Protected Area On Friday, August 29, 1986, the licensee moved a pile of sand and asphalt from outside the protected area into the radiologically controlled area.
The pile contained localized areas of contamination that had not been discovered when it was released approximately one year ago when the sand and asphalt were excavated from an alley way in order to inspect underground piping.
The highest reading particle was approximately 50 mrad /hr beta and 10 mrem /hr gamma.
The activity has been determined to be from Cs-137 and Co-60.
During the 1985 refueling outage ending in mid-October, the alley way between containment and the service building was dug up to investigate leakage from the Primary Component Cooling System.
The alley is in the radiologically controlled area and some surface contamination of the asphalt was expected.
A survey was performed of the asphalt prior to the excavation and the contami-nation found was segregated.
After releasing the asphalt and underlying sand, it was piled outside the protected area but within the owner controlled area.
The pile was surveyed on the surface after it had been dumped outside the protected area and a small amount of material was found to be "a few counts above background".
All material found above background was returned to the controlled area for disposal.
This pile was brought to management attention in April 1986 by an employee who thought it may contain contaminated material.
The pile was spot checked with a low range detector and no activity was found.
A more thorough survey was performed on August 26 to determine if the pile should be brought inside the radiologically controlled area and segregated using a method that was demonstrated to be more effective in detecting acti-vity.
This survey indicated localized areas where counts were above background.
No dose rate could be attributed to the pile.
On August 29, the pile was moved inside the Radiologically Controlled Area.
The Resident Inspector be-came aware of the situation at this time.
After removing the pile, an extensive grid survey was performed on the ground.
Residual contamination was removed and the area was release.
While the pile was being moved, the resident inspectors discussed the radio-logical controls in use with plant management.
This resulted in additional controls being implemented.
As previously discussed, the method now being used to segregate and release this type of material is adequate to ensure that a recurrence of this does not take place.
This method is discussed in an NRC routine radiological safety inspection (50-309/86-09) performed on June 2-6, 1986.
The lack of timely corrective actions, when contamination was first discovered, was dis-cussed with the plant manager.
Aggressive actions had been taken by the lic-ensee regarding this concern to ensure problems such as this one are elevated to upper management in order to receive appropriate attention.
The bases for the type of surveys performed and lack of prompt action is the subject of continuing investigation by plant management and will be reviewed further by NRC in subsequent inspections.
14.
Exit Interview Meetings were periodically held with senior facility management to discuss the inspection scope and findings.
A summary of findings was presented to the licensee at the end of the inspection.
Preliminary inspection findings were discussed with licensee management periodically during the inspection.
A summary of findings for the report period was also discussed at the conclu-sion of the inspection.
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