ML20207K868
| ML20207K868 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 10/04/1988 |
| From: | Linville J, Williams J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20207K856 | List: |
| References | |
| 50-352-88-19, NUDOCS 8810170065 | |
| Download: ML20207K868 (67) | |
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U.S. NUCLEAR REGULATORY C0KMSSION REGION I Report No.
8.8-19 Docket No.
50-352 License No.
NPF-39 Licensee:
Philadelphia Electric Company Correspondence Control Desk P.O. Box 7520 Philadelphia, Pa 19101 Facility Name:
Limerick Generating Station, Unit 1 Inspection Period: August 15 - September 24, 1985 Inspectors:
T. J. Kenny, Senior Resident Inspector L. L. Scholl, Resident Inspector T. F. Dragoun Senior Radiation Specialist Reviewed by:
A['y'/f/J Date
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Approved by:
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Chief. Prnjects Section 2A l
l Ro(utine laytire ( 69 hours7.986111e-4 days <br />0.0192 hours <br />1.140873e-4 weeks <br />2.62545e-5 months <br />) and backshif t/ holiday (26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br />)
Suwa ry :
inspections of Unit I by the resident inspectors consisting of (a) plant tours, (b) observations of maintenance and surseillance, (c) review of LERs and periodic reports, (d) review of operational events and (e) system walkdowns.
During this inspection period the licensee:
Operated the plant at 57 to 85*. power while monitoring the fuel cladding defect (section 2.0).
Submitted several LERs (section 6.0); the monthly operating report (section 5.0); safeguards event reports SS-505, ES-506 and 88-507 and a special report on the reteorological monitoring instrumentation operability (section 5.0).
Investigated increased of fgas activity levels (section 2.3).
Briefed NRC Region I staf f on the fuel cladding cefect (section 8.0),
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OETAILS J
r 1.0 _ Persons Contacted i
Within this report period, interviews and discussions were conducted with members of licensee management and staff as necessary to support 3
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inspection activity.
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2.0 Operational $afety_ Verification (71707, 70709 _71710 and 71881) t I
2.1 Documents Reviewed l
Selected Operators' Logs
$hift Superintendent's Log i
Temporary Circuit Alteration Log 1
Radioactive Waste Release Permits (liquid and gaseous) l
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3 elected Radiation Work Permits (RWP)
Selected Chemistry Logs I
Selected Tagouts Health Physics Log i
l 2.2 The inspector conducted routine entries into tne protected areas of I
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the plant, including the control room, reactor enclosure, fuel floor, and drywell (when access is possible).
During the j
inspection, discussions were held with operators, technicians (HP &
I&C), mechanics, security personnel, supervisors and plant management.
The inspections were conducted in accordance with NRC Inspection Procedures 71707, 71709, 71710 and 71531 and affireed the 1
licensee's cor.mitments and compliance with 10 CFR, Technical
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$pecifications License Conditions and Acministrative Procedures.
No violations were identified.
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J 2.3 Inspector Comments / Findings 1
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The inspection period began with the plant at 87*. power.
On August 17, a main contr ol roem isolation occurred when the 'O' l
channel chlorine detector morentarily spiked high.
The spike was l
apparently caused when the chlerine detector becate wetted during a t
thundurstorm.
The Heating Ventilation and Air Conditioning (HVAC) l control room e,tergency fresh air systets isolated the inlet dampers j
per design.
The NRC was inforced via the ENS and licensee event report SS-028 was submitted and is discussed in section 6.0 of this J
report.
t On August 20, reactor power was reda:ed to approairately 75*. prior to performing control red scram ti.re testing on 19 control rods.
The testing was satisfactorily completed and power was increased to l
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O 87%. On August 22, power was reduced to 81*. when of fgas l
pretreatment activity had increased to 42.800 micro curies per second.
Prior to the control red testing and associated power maneuvers offgas pretreatment activity was approximately 15,000 micro curies per second.
Technical Specification limit for offgas pretreatment activity is 330,000 micro curies per second.
On August 17, the licensee informed the NRC that an auxiliary plant operator had tested positive for illegal drugs. A confirmatory test was also positive.
The individual's protected area access has been revoked.
The testing was performed as part of a newly ieplemented drug policy which requires all personnel with unescorted access be randomly tested.
On August 22, the licensee infortred the NRC that a second auxiliary plantoperatortestedpositiveforillegaldrugs. A confirmatory test was also positive.
The individual 5 protected area access was revoked. Approximately one third of the licensee's personnel had been tested by this date.
On August 27, the licensee perforced control rod scram time testing in accordance with technical specifications.
The movements of rods aggravated the already existing fuel cladding defects and possibly caused additional cladding cefects.
The offgas pretreatment and dose equivalent gross iodine activities increased since the rod movements, hcwever reactor coolant and offgas activities remained less than 10'. of technical specification limits. At approximately 3:00 p.m. en September 1, the licensee moved control group 21 out f rcm eight ste -s to :ero in orcer to decrease reactor power and decrease offgas activity by recucing pcwer f rom S2*. to 79*..
After discussions with General Electric (GE), the licensee reduced power by decreasing recirculation flew to 75.4*. which is a power level that limits the linear heat generation rate of fuel to approxtriately ere half KW/FT below the gacolinia GE pin power curve retornendation.
i These power reductions lo ered heat generation rate, the reactor coolant activities and offgas pretreatrent activity.
The licensee continued monitoring the coolant chemistry very closely to assess i
the results of the power changes.
On August 25 at 11:17 a.m., the FBI confiscated a licensed reactor operator's vehicle frem tha cwrer control area parking lot.
The individual was alleged to be involved in the sale of illegal drugs.
The operator's unescorted access was withdra.,n.
The operator was in training at the time of the vehicle confiscatien.
This incident is still under investigation by the FBI and PECo Security.
As a result of this occurrence the drug testing schedule for operations personnel was accelerated.
All test results were negative except for the two previously discussed auxiliary operators.
Tha NRC was notified of this incident via the ENS.
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On August 30, a firewatch was found asleep while on duty.
The t
firewatch was posted on elevation 313 as a compensatory measure due to the inoperable water curtain.
The firewatch was relieved and his employment was subsequently terminated.
l On September 5, po er was reduced to 65*4 in an additional effort to j
minimite offgas pretreatment activity and the potential for i
additional fuel cladding defects, i
l On September 19, control rod 50-35 drif ted f rom the full out to fully inserted position as a result of a ha!f scram initiated on ' A' channel during surveillance testing.
Subsequently, power was reduced by 100 W'e to 57*. as directed by plant procedures.
The
.I licensee investigated and found air leaks in the solenoid valves of
'A' channel which contributed to the rod insertion.
Follewing the il repair of air leaks on the scram pilot valve; the rod was withdrawn and power was restored to 65*..
Fellowing the restoration of power l
the offgas pretreatment activity and iodine cose equivalent values began to increase.
These values continue to remain less than 10*o of technical specificatisn limits and are being monitored closely by the licensee and resident inspectors.
On September 19, the licensee reported, via the ENS, that a current 1
PECo employee (a former worker at Limerick) had purchased metham-j phetamines within the protected area on two occasions ( August 1936 and October 1937).
The employee had worked at Limerick at that time and hasn't worked at Limerick since 1937.
The licensee has also identified the person who sold the drugs as an employee who is
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currently undergoing rehabilitation for drug use. Neither have i
j access to the protected ared at this time. At the end of the period i
the licensee was evaluating the employment of the trdividuals and l
the involvement of law enforcement agencies, See section S.0 of this report for a synopsis of the licensee's actions regarding drug investigations.
No violations were identified.
l 2.4 Followup on Previous Violations and Unressived items j
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2.4.1 (Closed) Unresolved Item (87-16-OL) l
)I Construction Division Procedures (CDs) and modification instructions to the personnel lacked the details needed to l
perform tasks on plant equip ent.
The lack of detailed I
procedures contributed to tne irproper re-connection of i
the rotor oil cooler for the 'A' RHR pump which was later corrected.
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4 The licensee has revised procedures EROP 2.2, Construction Engineering Procedure; CD-5.3, Procedure for Installation of Electrical Equipre a and CD-5.7, Procedure for Installation of Mechanical Equipment.
The procedures now include precautions that address the pre-inspection and review cycle of the modification data package prior to work performance.
The inspector reviewed the procedure changes and concluded that the concerns have been addressed.
This item is closed.
2.4.2 (Closedl__ Unresolved Item _(83-14-01)
This item pertains to the licensee's inadequate documentation to support review and implementation of proper actions for Generic Letters (GL) 85-03, 55-13 and 85-22.
The licensee reviewed these GLs cnd provided documentation to support the completten of the required actions, as applicable.
In addition, the licensee has established a Nuclear Group Interim Administrative Procedure (dated July 1, 1988) which delineates responsibilities for prompt review of GLs and Information Notices, and assessment of significant operating experiences by the cogni: ant licensing organi:ation.
Accordingly, such activities are assigned commitment action numbers.
Upon completion of the required actions, the commitment action items are verified, documented and properly closed out.
The licensee's commitment tracking program tracks and updates the status of the action item.
Based on the review of the licensee's corrective action impleTentation, the licensee action is censidered adequate.
This item is closed.
2.4.3 (Closed) Insoector Followup Item (88-06-03)
On completion of the analyse ef water samples (spiked samples) by the licensee and Brookhaven National Laboratory, a statistical evaluation was to be made.
The analyses were completed and an evaluation was performed.
The analytical ccmparisons for the analyses were acceptable.
See Attachment 4 2.4.4 (Closed) Violation ($3-03-01)
Administrative procedure A-3, Procedure for Temporary Changes to Approved Procedures was revised effective August 22, 1953 to provide a notice in controlled copies of procecures that a Temporary Procedure Change (TPC) exists.
This notice will direct the user to a TPC book maintained in the Centrol Roem.
This book and a similar
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book in the Remote Shutdown Panel will contain copies of f
the active TPCs identified by a TPC notice. Based on this i
strengthening of temporary procedure change controls, this t
item is closed, f
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2.4.5 (Closed) Unresolved item (85-30-01) l l
The licensee further reviewed NRC concerns relative to the l
l use of a third offsite poner source in the event one of l
J the normal sources is deenergized inadvertently, The NRC l
f concerns were addressed by the licensee in correspondence i
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to the NRR Project Directorate dated October 31, 1956.
The headquarters inspector had n; further questions, t
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Previsions for the use of the alternate power source have not yet been implemented and some aspects may be different i
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than originally anticipated (e.g. use of a 66 KV source 1
instead of a 33 KV source).
However, any plant changes to l
facilitate the use of a third power source will be made in accordance with the plant modification procedures.
This 1
will ensure the change receives all the requisite reviews l
and documentation.
This item is closed.
3 2,5 Health Physics _ Events 1
2.5.1 Radiological Program _ performance Indication l
Total annual exposure of workers at Limerick station has
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remaired low since the beginning of commercial operations.
I Thus far in 1983, a total of 37 person-rem have been F
i receised.
If the current rate continues through year end, i
i the station will have achieved a record low exposure for a j
domestic C'a'R.
i Personnel contamination events, which were occurring at i
frequency well above industry averages, is now apparently 1
under control. A total of 74 events have teen reported to l
I date.
The current trend will exceed the best inriustry quartile data.
This excellent performance was achieved l
during operation with degraded fuel conditions since March i
1939.
j 2.5.2 Degraded Fuel Impact I
Due to the degraded fuel conditions in the reactor, the licensee has implemented several unique reports that 1
j inform the staf f about key radiological conditions.
Selected area dose rates, offgas activity ard calculated 1
i off site cumulative exposures are discussed in the TRIPOO j
(three day, plan of the day) meeting of managers.
The HP
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t,echnician night orders routinely give poner level, reactor coolant conductivity, and pretreatrrent offgas
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4 activity levels.
HP supervision tracks the occurrences of I
polyester clothing contaminations by short lived noble gas l
i daughters as a rapid indicator of changing fuel I
conditions.
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i The actual impact of the degraded fuel is slight. Offgas l
doses resulting from offgas activity remain negligible.
General area dose rates and airborne activity in the l
l equipment spaces are unchanged from previous levels.
The Radiological Engineering group is continuing to identify and flush "hot spots" in system piping to keep plant dose e
rates at a low level.
j The inspector concluded that the licensee's radiological l
protecticn activities relative to operations with degraded fuel are appropriate.
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2.5.3 Contaminated Resin Container 1
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Radioactive resin is dewatered on site prior to shipment I
for burial by a centrifuge process.
In anticipation of a change to a new process, the licensee removed unused resin l
shipping containers from the radwaste processing area.
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Three of the containers were placed outside the J
radiologically controlled area. One of these was l
1 subsequently found to have surface contamination in the j
j lid area.
This caused skin and clothing contamination of I
three personnel involved in quality control checks being
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performed on the containers.
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A preliminary investigat, ion by the licensee determined 1
that the root cause for this event was personnel errors by i
HP technicians controlling the movement of the containers
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and some confusion regarding the unconditional release of potentially contaminated equipment.
The licensee's l
completed enalysis and corrective action will be reviewed j
in a future inspection, i
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2.5.4 Dy pped Transfer Cart In an August 10, 1988 minor event, a transfer cart remained attached to the bottom of a full resin container i
during a crane hoist.
The cart, held to the bottom ' y i
a suction, is used to roll the container out of the resin fill station. No damage occurred to the container or the 4]
crane. Minor damage occurred to the cart and tre guide rails.
i The inspector toured the area and discussed this event with the Radwaste Supervisor (operations) and the Radwaste Technical Support Engineer.
The causes of this event include the malfunction of a vacuum release solenoid and i
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l inexperience of the radwaste operator manipulating the i
crane. An analysis of the solenoid failure is underway.
j For the radwaste operators, an "operating aid" has been j
posted giving the approximate weight of a full resin a
container for comparison with the weight gauge on the l
crane.
In addition, procedure $67.8 A "Operating Resin i
Fill Stations" has been revised to specify an approximate weight.
A weight above this range would alert ti,e j
operators to the potential that tha cart did not i
disconnect.
Licensee actions on this matter are
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appropriate.
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3.0 Surveillance /$pecial Test Observations (61726) i Ouring this inspection period, the inspector reviewed in-progress j;
surveillance testing as well as completed surveillance packages.
The
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inspector verified that surveillances were performed in accordance with i
licensee approved procedures and NRC regulations.
The inspector also i
verified that instruments used were within calibration tolerances and j
that qualified technicians performed the surveillances.
l The following surveillances were reviewed:
4 ST-6-049-230-1 Reactor Core Isolation Cooling (RCIC) Pump, Valve and i
Flow Test l
ST-6-107-590-1 Daily Surveillance Log j
ST-3-048-230-1 Standby 1.iquid Control System Pump. Valve and Flow Test
$T-5-020-814-1 Emergency Diesel Generator Monthly Fuel Oil Ar.ilysis i
On August 31, the licensee identified high particulate matter contamination 1
in the fuel oil and declared the No. 14 Emergency Diesel Generator (EDG) inoperable in accordance with technical specifications.
The limiting condition for an inoperable EDG is 92 days. Within several days the i
licensee circulated the contents of the fuel oil tank associated with No.
l 14 iDG generator through s portable filtration system lowering the particulate content to an acceptable level below the t,echnical specification limit. The EDG was returned to se vice before the technical specification limiting condition for operation action statement expired.
Fuel oil particulate matter is formed naturally by the formation of carbon particles, a process that has been identified in EDG fuel oil that is stored for long periods of time.
No violations were identified.
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4.0 Maintenance Observations (62703)
The inspector reviewed the follcwing safety related maintenance l
activitier, to verify that repairs were made in accordance with approved procedures, and in compliance with NRC regulations and recognizec codes i
j and standards.
The inspector also verified that the replacement parts j
and quality control utiliced on the repairs were in ccepliance with tht lice.nsee's QA program.
8 Work Order Number Description 8805140 RCIC Inverter Replacement 8804051 Repack HV-050-1F045 (RCIC Steam Supply Valve) 8805500 Seplace Automatic Depressurization System (ADS)
Logic Relay
- 880543P, Division III Battery Charger Repair 8804551 Cnange Motor Bearing 011 in IB Core Spray Pump 4.1 During the inspection period preventive maintenance outages were conducted on the Reactor Core Isolation Cooling (RC.t), High l
Pressure Coolant Injection (HPCI) and Core Spray (CS) systems.
The work performed was routine preventive maintenance which could be l
pe-formed with the plant operating and thus reduce the refueling outage workload.
The system outages were well planned and managed effectively to ensure system ur.svailability times were minimized.
During restoration from the RCIC outage several problems were encountered.
The steam admission valve antirotation collar became loose and required rework.
The maintenance department is currently reviewing the installation to determine if a double set screw configuration should be used to secure the collar as is currently in place on other similar valves.
During the performance of the retest proper automatic flow control could not be obtained.
The turbine governor was.'ycled and flow transmitter FT-49-1N003 was revented.
No further flow control problems were encountered.
RCIC Suppression Pool Supply Isolation Valve HV-49-1F-031 required adjustments to the torque switch bypass circuit.
l These problems have been documented in the Plant Incident Tracking System which ensures that the Plant Operations Review Committee (PORC) will review and approve corrective actions taken to prevent recurrence.
No significant problems were encountered during the HPCI or CS system outages.
5.0 Review of Periodic and Special Reports (90713)
Upcn receipt, the inspector reviewed periodic and special reports. The review included the following:
inclusion of information required by the NRC; test results and/or supporting information consistent with design predictions and performance specifications; planned corrective action for
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resolution of problems, and reportability and validity of report information.
The following periedic report was reviewed:
Unit 1 Monthly Operating Reports - July and August 1988 The inspector had no questions concerning these reports.
The inspector reviewed a special report written in accordance with l
Technical Specification Sections 3.3.7.3 and 6.9.2 which require reporting whenever meteorological monitoring instrumentation is inoperable for more than seven days.
On August 17, the primary and secondary meteorological towers were struck by lightning making instrumentation on both towers inoperable.
T_he wind speed, wind direction and air temperature diff?rence sensors provide data for estimating potential radiation doses to the public in the event of a radioactive material release to the atmosphere.
The capability to perform off site dose assessment was maintained since backup meteorological data could be obtained from the National Weather Service in Philadelphia or from meteorological instrumentation at the Peach Bottom Atomic Power Station in Delta, Pennsylvania, i
l The primary tower was fully repaired on September 9 and the secondary l
tower repairs were in progress at the end of the period.
l The station personnel have requested that engineering design a modification to add lightning protection to the towers to avoid future problems.
The inspector had no further questions concerning this report.
i 6.0 Licensee Event Report Followup (90712, 92700)
The inspector reviewed the following LERs to determine that reportability i
requirements were fulfilled, that immediate corrective action was taken, and that corrective action to prevent recurrence was accomplished in accordance with technical specifications.
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88-026 I
88-027 88-028 These LERs each report actuations of the Control Room Emergency Fresh Air I
System (CREFAO when rain water wetted the chlorine detectors resulting 4
in an upscale spib on the chlorine instrumentation. As discussed in the
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reports two system modifications have been designed to eliminate the j
spurious isolations from rain water or other momentary spikes.
The first i
modification relocated the detectors to prevent rain water from wetting
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the probes and was completed during this report period. No further
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spurious actuations of the CREFAS system have occurred since this 4
installation, i
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10 A second change to the system will revise the instrumentation logic such that a single instrument spike will not initiate the CREFAS system as is presently the case.
Thus, any single upscale spike regardless of the cause should not cause a spurious CREFAS initiation.88-505 This report addresses the July 16, 1988 incident of inattentfveness by a security force member as discussed in section 2.3 of Inspector Report 50-352/88-17.
The inspector had no further questions regarding this event.88-506 88-507 The inspector reviewed the above reports, issued by the licensee, describing the results of drug testing on certain individuals.
These l
reports were submitted in accordance with section 2.790 of the Commission's regulations regarding the withholding of certain information from public disclosure.
Because of the licensee's ongoing investigations this information is being treated as confidential. Based on a review the inspector noted that the reports are complete and in accordance with the
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reporting criteria.
No violations were identified.
7.0 Licensee and NRC Meeting _in Region I Pertainin2_to Limerick Fuel Cladding Defect On September 12, the licensee briefed NRC Regional Management on the status of the fuel cladding defect and actions being taken to minimize the consequences of the cladding defect.
The areas discussed were:
1)
Radiological Impact 2)
Operating Strategy i
3)
Reload Strategy 4)
Copper Reduction Results 1
5)
Long Term Copper Reduction Strategy f
In general, the radiological impact of the fuel cladding defects have been minimal. Offsite dose calculations based on recent releases show that dose projections remain at less than one percent of the technical specification limits.
In plant airborne activity has not posed any significant problems.
Activity levels have been minimized in part by a concerted effort to promptly identify and correct steam leaks.
The hot spot monitoring program has been expanded in an effort to determine if the fuel leaks may be causing any long term elevated radiation levels.
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11 The licensee's operating strategy is to continue to operate at approximately 65*4 power. According to GE, operation at reduced power reduces the plant pretreatment offgas activity and coolant iodine inventory and also reduces the potential for additional fuel defects.
Control rod pattern changes and power level changes are being minimized in an effort to prevent new fuel leaks and to minimize further degradation of the existing leaks.
The core reload strategy will attempt to avoid fuel failures during cycle 3 operation by:
a) inspecting initial : ore fuel before clearing it for additional
- service, b) minimizing the fuel duty (KW/FT) on the reloaded fuel bundles and c) maximizing the use of fuel with heat treated cladding.
Recent licensee efforts to reduce copper levels in the feedwater appear to have been successful.
The copper reduction was achieved when a premixed resin was used in the condensate filter-demineralizers in lieu of an individual anion / cation / overlay regeneration process.
Long term copper level reduction methodt were discussed. The two methods under review are:
1) replacemen+ of the admiralty brass.iain ennden;er tubes with titanium c. Jenser tubes and 2) installation of deep bed condensate demineralizers.
The licensee will decide on which option will be used at a future date.
See Attachment 2 for licensee slides used during the presentation and for the list of attendees.
8.0 Drug Investigations by the Licensee Following the recent corporate management changes announced at PECo, a new random drug testing policy was introduced for all PECo employees.
As a result of the drug testing at Limerick several instances of drug i
usage have been identified.
These are delinated in section 2.0 of this report.
The licensee is continuing investigations, as a result of the identified drug usage, into the bredth and depth of employee drug involvement.
The licensee is keeping the NRC informed through discussions and reportable conditions as ' hey arise.
Due to the sensitivity of this issue and the ongoing invest ations, reports to the NRC are being submitted in accordance with 10 CFR Se;.mn 73.71 (b) (2) and 2.790 which involves a request for withholding from public disclosure.
PECo is continuing drug testing of all personnel who work at Limerick, which includes contractor personnel.
The licensee has set a deadline of November 1, 1988 beyond which anyone who has not been drug tested may not enter the protected area.
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12 No violations were identified.
9.0 Qualifications of Martin J. McCormick as Plant Manager Via a letter of December 1, 1987, the licensee informed the NRC of the intention to install Martin J. McCormick as plant manager at Limerick.
l The inspector has reviewed the letter and Mr. McCormick's qualifications l
and compared them to the "Nuclear Power Plant Experience" Section of f
ANSI /ANI 3.1 and has determined that Mr. McCormick has successfully l
completed the training outlined in the above mentioned letter and is considered qualified to be plant manager.
10.0 Assessment of plant Temperature Parameters for 1988 Summer Months Because of the unusually hot summer months the resident inspector surveyed j
the licensee to determine if any systems or components were adversely affected by the higher than normal temperatures.
In Attachment 1, a series of curves is included that delineate thermal power, river water temperatures and spray pond temperature for the months of June, July and August for 1986, 1987 and 1988.
These curves reveal higher temperatures in 1988 in the river water and spray pond.
However, the plant is cooled by cooling towers which make up from the river, and the ultimate heat sink is the spray pond which can be kept cool by recirculation of the water through spray jets.
There was more recirculation of the spray pond in 1988 but the system was not adversely affected by this additional recirculation.
The only other components within the plant affected by the warmer weather were the instrument air compressors that required additional portable blowere to keep the compressor oil temperatures from reaching the alarm l
setpoint.
' bis method of adaitional cooling was successful.
The resident determined, because of the method of cooling Limerick Station, that the plant, although somewhat affected by the hotter weather, was not hampered in maintaining all technical specification and operating parameters within the prescribed temperature limits.
No violations were identified.
11.0 Safety _ Relief Valves As stated in previous reports the Safety Relief Valves (SRV) have experienced seat leakage causing the suppression pool to heat up during operation and have experienced setpoint drift causing them to lift at other than the specified setpoint.
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The licensee intends to address both of the problems during the shutdown for refueling and the following cycle.
The first problem, seat leakage, will be addressed by Modification Number 5546, Engineering Evaluation of MSRV Main Seat Cutting.
The evaluation revealed that the main steam
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flange opening was smaller than the SRV opening allowing water to accumulate in the valve around the seat area.
The accumulation of the water probably caused thermal gradients which may have warped the seat causing the leakage.
The modification will be to bore out the main steam flange opening to the same size as the SRV opening.
This will prevent the accumulation of water by allowing any condensation to drain back to the main steam system. The modification is to be performed on all 14 main steam flange openings.
The second problem, set point drift, will be addressed by the changing pilot valve disk material.
Seven of the 14 installed SRVs will be replaced with a valve that has had the seat disk material changed from the currently installed stellite to the new material PH13-8MO.
This modification was recommended by the BWR Owners Group /MSRV Setpoint Orif t Fix Development Committee.
The remaining seven SRVs will be replaced with rebuilt valves with the original disk material.
The two sets of seven valves will be evaluated over the next operating cycle.
If the new material is successful, then during all subsequent refueling outages SRVs will be repiaced with PH13-8M0 disk material, as necessary.
12.0 Assurance of Quality During this inspection period, the resident inspectors observed the assurance of quality and thorough followup of the licensee's investigation and management eversight in the following areas.
Minimizing of safety system outage times (section 4.1)
Low personnel radiation exposure and contamination (section 2.5.1)
Fuel leak management (sections 2.5.2 and 7.0)
Drug testing program (section 8.0)
However, an area where assurance of quality was not demonstrated is presented in section 2.5.3 of this report in which a contaminated radwaste container was removed from the radiologically controlled area and resulted in a personnel contamination incident.
13.0 Exit Interview The NRC resident inspectors discussed the issues in this report throughout the inspection period, and summarized the findings at an exit meeting held with the Plant Manager, Limerick Generating Station, on September 22, 1988.
No written inspection material was provided to licensee representatives during the inspection period.
ATTACHMENT 1 IR # 88-19
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THERMAL POWER JUNE 1986 MWt 3600 3000 2600 2000 1600 1000 600 0
12 3 4 8 0 2 8 9 to %121314M16175192021222324262627232930 DAY RIVER WATER TEMPERATURES JUNE 1986 Ot0REES P 100 to 80 70 40 60 12 3 4 6 6 7 8 9 0111213141516f714192021222324262627262930 DAY
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ATTACHMENT 2 IR # 88-19 AGENDA LIMERICK UNIT 1 FUEL PROBLEM MEETING i
Monday, September 12, 1988 l
t RADIOLOGICAL IMPACT G. W. Murphy Limerick Senior Health Physicist j
i t
OPERATING STRATEGY M. P. Gallagher j
Limerick Reactor Engineer RELOAD STRATEGY L. F.' Rubino f
Supt. - Fuel Management i
COPPER REDUCTION RESULTS G. Barley Limerick Plant Chemist l
LONG TERM COPPER REDUCTION STRATEGY R. J. Scholz i
Supv. Engineer - Nuclear Engineering l
PANEL DISCUSSION - QUESTIONS AND ANSWERS 1
I
}
)
)
I RADIOLG CAL luPACT 3R ESE NTAT O N TO T-E \\RC SEPTEMBER 12,1988 GARY W. MURPHY SR. HEALTH PHYSICIST i
i RADIOLOGICAL IMPACT PRESENTATION 1
I CURRENT STATUS OF THE IN-PLANT RADIOLOGICAL ENVIRONMENT t
II ROUTINE SURVEILLANCE PROGRAM ENHANCEMENTS 6
A.
FUEL LEAK EFFECTS MONITORING PROGRAM HOTSk)OT/DOSERATEREDUCTIONPROGRAM B.
III FUEL FAILURE MONITORING ACTION PLAN t
IV PERSONNEL EXPOSURE - LGS, 1988 V
FUEL LEAK EFFECTS ON EFFLUENT RELEASES I
VI DETERMINING LIMITATIONS ON EFFLUENT RELEASES t
i VII CONCLUSIONS 1
i I
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o EXPANDED ROUTINE SURVEILLANCE PROGRAM 10 CONSTANT AIR MONITORS (CAMS)
CAMS IN SERVICE AT BOUNDARIES TO UNIT 2 AND SELECTED AREAS IN UNIT 1 CAMS INSPECTED EACH SHIFT TO DETECT AND MONITOR TRENDS IN AIRBORNE RADIOACTIVITY 2.
ADDITIONAL AIR SAMPLING LOCATIONS SELECTED IN UNIT 1.
3.
ROUTINE SURVEILLANCE IN UNIT 2 7 DAILY AIR SAMPLES OBTAINED AND ANALYZED FOR PARTICULATE AND IODINE WEEKLY CONTAMINATION AND RADIATION SURVEYS IN UNIT 2 TURBINE ENCLOSURE, ELEVATIONS 217, 239, AND 269
UNIT 2 SURVEILLANCE 1.
Air flow direction is documented shiftly by Unit 2 operaticns at the 23 line (U-2/O-1 Boundary) on the Refuel Floor and on the turbine deck.
If the air flow is from Unit 1 to Unit 2, the Control Room is to be notified and actions to correct the air flow direction shall be taken.
2.
Air samplers are located at the following locations in Unit 2:
North & South stairwells on the Refuel Floor Inside the Unit 2 Reactor Cavity Top e.evation of the Unit 2 Drywell Unit 2 Turbine Building at the 269' elevation Exhaust Ducts, and near the open hatchway.
3.'
PINGS are located at the Unit 1/ Unit 2 boundaries at the following locations:
i Reactor Bldg Refuel Floor Turbine Bldg at elevation 269', 239', & 217' The PING alert and alarm set points are at levels which would annunciate below 10CFR20 APP B Table II, Column I limits for unrestricted areas.
J
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FUEL LEAK EFFECTS MONITORING PROGRAM FOLLOWING PARAMETERS MONITORED, TRACKED AND TRENDED ON A DAILY BASIS:
LOCATION PARAMETER UNITS NA Off Gas Activity uCi/Sec Tb-217 el, at Reactor Airborne Activity Feed Pumos (part. & I)
MPC Fraction Tb-239 el. at Airborne Activity FW Htr. Level Inst.
(part & I)
MPC Fraction Grating Above Off Gas Pipe Tunnel Radiation Levels mR/hr MSIV Room Door Radiation Levels mR/hr Rx-177' el. ec Floor Airborne Iodine and Equip. Drain Activity Iodine MPC Fraction Sumps 4
+ - - - - _ -,,
I TB-200 OFFGAS PIPE TUNNEL i
GRATING DOSE RATE 4
mR/hr 50 N
" ~ ~
40 d
i
/
30 J
20 l
1 1
- ~ ' ' - ~ - - - - - - - - ~
i 10 l
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i 8/5 8/10 8/15 8/20 8/26 8/30 9/5 9/10
}
1988
~
Series 1
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l RX-253 OBMSIV RM DOOR DOSE RATE i
mR/hr 25
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20 777 j - -
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0 8/6 8/10 8/15 8/2C 8/26 8/30 9/6 9/10 i
j 1988 Serias 1
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TB-239 HALLWAY AIRBORNE RADIOACTIVITY MPC FRACTION 0.5 I
i, 0.4 1
0.3 1
0.2 i
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Series 1 i
I l
I RX-177 SUMP AREA l
AIRBORNE IODINE ACTIVITY MPC FRACTION O.01 r
0.008 0.006 0.004 I
0.002 f
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1988 Series 1 1
4 O
e 9
(
DOSE RATE REDUCTl_Qis._di;.OG. AM R
A PROGRAM BY WHICH F.LiAT"..
9 :"RCES CONTRIBUTING TO INCffEa tid '082. RATES IN THE PLANT ARE IDENTii-itEU, "OSTED, TRACKED, AND REDUCED.
i I
l-
i, DOSE RATE REDUCTION COMMITTEE A MULTlDISCIPLINARY COMMITTEE COMPRISED OF REPRESENTATIVES FROM HP, OPERATIONS, TEST ENGINEERING, CHEMISTRY, REACTOR ENGINEERING, AND RADWASTE TASKED WITH MAKING COST EFFECTIVE RECOMMENDATIONS FOR THE REDUCTION AND/OR ELIMINATION OF SOURCES OF INCREASED DOSE RATES.
9 e-
DOSE RATE REDUCTION EFFECTIVENESS DECREASE IN DOSE RATE ON SELECTED LINES GENERAL AREA DOSE RATE (mR/hr) 140 10 80 I
60 0
9 10 11 12 1
2 3
4 5
6 7
8 9
l 87 l
88 l
MONTil HilR PIPING
-'- CORE SPRAY PIPING
-F RWCU F/D PIPING
-* FLOOR DRN PlPING S
y
e FUEL FAILURE MONITORING ACTION PLAN 1.
DAILY O?FGAS COMPOSITION REPORTS ARE BEING REVIEWED AND TRENDED.
3.
DAILY REACTOR COOLANT SAMPLES ARE BEING DECAYED TO IDENTIFY BUILDUP OF LONGER LIVED FISSION PRODUCTS THAT MAY BE MASKED BY NOBLE GASES.
3.
REACTOR COOLANT AND RWCU RESIN SAMPLES HAVE BEEN SENT OFFSITE FOR ANALYSIS TO IDENTIFY ALPHA AND BETA COMPONENTS AllD LONG i
LIVED FISSION PRODUCTS WHICH, IF PRESENT IN SUFFICIENT QUANTITIES, COULD RESULT IN AN INCREASE IN GENERAL AREA DOSE RATES OVECTIME.
4.
THE PORTABLE INTRINSIC GERMANIUM GAMMA SPECTROSCOPY SYSTEM IS BEING USED TO PERFORM ISOTOPIC TRENDS OF CRUD TRAPS IN ORDER TO PROVIDE EARLY INDICATION OF THE PRESENCE OF LONG LIVED FISSION PRODUCTS.
1 5.
OTHER PLANTS THAT HAVE EXPERIENCED SIMILAR PROBLEMS RAVE BEEN l
CONTACTED TO CAPTUhE LESSONS LEARNED.
l 6.
OFFSITE DOSE PROJECTIONS ARE BEING PERFORMED AS DATA BECOME AVAILABLE SO THAT EARLY INDICATIONS OF STEAM LEAKS CAN BE IDENTIFIED AND REPAIRED.
f.
ALL DATA ON SITE IS BEING Cui COLIDATED AND REVIEWED WEEKLY ~ IN"~~ ~~ ~ ~ ~
ORDER TO PROVIDE MANAGEMENT WITH INFORMATION TC SUPPORT SHORT AND LONG TERM STRATEGIC PLANNING.
I G
4
e 1988 LGS EX?0SURE MAN-REM 100.00 PLANT GOAL 00.00 SRD ACTUAL TLD ACTUAL i
80,00 t
/
70.00 I
60.00 1}
50.00 l
40.00 l
i
/
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/
30.00 1
i "0.00 i
10.00 f
l l
0.00 1
Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec MONTH i
u.
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GASEOUS RELEASE TECHNICAL SPECIFICATION LIMITS AT LGS Tnch. Spec.
Limit Applicability 3.11.2.1(a) 500 ar/yr Any "Instant" - 10CFR20 (Noble Gases) 5.11.2.1(b) 1500 ar/yr Any "Instant" - 10CFR20 (Iodines) 3.11.2.2(a) 5 mRads Calendar Quarter - 10CFR50 (Gamma) 3.11.2.2 10 mRads Calendar Quarter - 10CFR50 (Beta) 3.11.2.2(b) 10 mRads Calendar Year - 10CFRSO l
(Gamma) 3.11.2.2.(b) 20 mRads Calendar Year - 10CFR50 (Beta) 3.11.2.3(a) 7.5 ERem Calendar Quarter - 10CFRSO (Iodines) 3.11.2.3(b) 15 mrem Calendar Year - 10CFR50 (Iodines)
OTHER RELATED TECH SPECS T ch. Spec.
Limit Applicability 3.11.2.4 0.3 mPen Projected over (Iodines) 31 days 3.11.2.6 330,000 uCi/sec*
Any "Instant" o
Sum of the offgas activity of the noble gases Kr-85m, Kr-87, Kr-88, Ie-133, Se-135, Xe-138 Prior to treatment l
v l
CURRENT PLANT STATUS RELATIVE TO
~
TECHNICAL SPECIFICATION LIMITS Tech. Scec.
Limit LGS Status 3.11.2.l(a) 500 mR/yr Release point effluent (Noble Gas) monitor Hi-Hi setpoint will alarm if these limits are met or exceeded 3.ll.2.l(b) 1500 mR/yr (Iodine)
Cumulative Excosure to Date 3.ll.2.2(a) 5 mrad (Gamma)
<1E-3 mrad 10 mrad (Beta)
<1E-3 mrad in a calendar quarter 3.ll.2.2(b) 10 mrad (Gamma) 4E-2 mrad 20 mrad (Beta) 2E-2 mrad in a calendar year 3.11.2.3(a) 7.5 mrem (Iodine) 5.56E-3 mrem in a calendar quarter 3.ll.2.3(b) 15 mrem (Iodine) 6.70E-2 mrem in a calendar year 3.11.2.4 0.3 mrem (Iodine) 4.2E-2 mrem projected over 31 days 3.11.2.6 330,000 uCi/sec 21,900 uCi/sec offgas release rate 1
ACCUMULATED DOSE (YTD 1988)
DUE TO IODINE RELEASES LCQ mAEM 1.00e*002 Tecn Spec 3 11.2.3(ts)I 15mAEM/Ur 1 00e+001 Tech Spec 3.11.2.3(alt 7.5 MREM /qte 1.00e*000 1.00e=001 l
i AC3UMULATED CCSgse.067mAEM
}
1.00e=002 1
1.00e=003
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1.00e=004 f1AA31 AAA30 MAY31,W3 0 JY31 AU2 AUS AU18 AU23 AU30 SEP6 CEPS NCNTH l
[
DETERMINATION OF MOST LIMITING RELEASE RATE GOALS LOsarenu
- 1. MAINTAIN LESS THAN 10% TECHNICAL SPECIFICATION LIMITS DURING 1988.
SHORT TERM
- 2. TO TOLERATE 10% BYPASS OF THE OFFG AS SYSTEM AND MAINTAIN THE ' INSTANTANEOUS
- 10CFR20 LIMIT AT THE SITE BOUNDARY (LARGEST BYPASS TO DATE WAS DUE TO CROSS AROUND PIPING BELLOWS RUPTURE APPROXIMATELY 2% BYPASS) i l
ASSUMPTIONS
- 1. SIGNIFICANT STEAM LEAKS DO NOT DEVELOP IN THE TURBINE ENCLOSURE DURING 1988.
- 2. TEECE IS NOT BYPASSED.
- 4. NO DRYWELL PURGES DURING REMAINDER OF FUEL CYCLE.
I
}
i 4
I 1.
i I
2 e
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i ASSESSMENT OF FUEL LEAK IMPACT ON LONG TERM GOAL I
t J
t 1
CORRELATION BETWEEN PLANT RELEASES AND OFFGAS ACTIVITY?
4 REVIEW OF OFFGAS DATA, DOSE EQUIVALENT IODINE DATA, AND IODINE RELEASE RATE DATA FROM JUNE 1 TO THE PRESENT SHOWS i
F VIRTUALLY NO CORRELATION.
THEREFORE l
i I
e l
i i
i PROJECTION RATHER THAN PREDICTION WILL BE THE METHOD THAT THE j
{
PLANT STAFF WILL USE TO ASSESS THE LONG TERM IMPACT OF FUEL t
i LEAKS ON OFFSITE RELEASES t
i P
t I
i i
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1 ACCUMULATED DOSE (YTD 1988) l DUE TO 7,0 DINE RELEASES l
mAEM 0.18 (9/9/88)
Tecn Spec 3.11.2.4 - 8.31 MREM /31 daye 0.89 Tech Spec 3.11.2.3(a) - 7.5 mAEM per quarter Tech Spec 3.11.2.3(m) - 15 MREM annual limit 8.88 i
8.87 (8.887 MREM) 8.88 i
4.45 8 e4 0.83 l
4.02 l
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.731 AU2 AUS AU16 AU23 AU38 SEP6 SEPS McNTH i
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ASSESSMENT OF FUEL LEA < IMPACT ON SHORT TERM GOAL LIMITING RELEASES TO STAY LESS THAN OR EOUAL TO 10CFR20 AT ANY INSTANT DUE TO A 10% BYPASS OF THE OFFGAS SYSTEM.
USING RECENT OFFGAS ISOTOPIC RATIOS :
8,380 uCi/sec IF RELEASED UNTREATED WILL RESULT IN ATTAINING THE 10CFR20 LIMIT AT THE STACK.
1 CONCLUSION : 83,800 uCi/sec WOULD BE THE LIMITING OFFGAS RELEASE RATE BASED ON THE ABOVE SYSTEM FAILURE.
l l
CONCLUSIONS 1.
HEALTH PHYSICS HAS HAD A COMPREHENSIVE SURVEILLANCE PROGRAM AND "HOT SPOT" REDUCTION PROGRAM IN PLACE SINCE THE FIRST INDICATION OF A FUEL PROBLEM IN MARCH.
THE RESULTS OF THIS PROGRAM CONTINUE TO INDICATE AND SUPPORT THE CONCLUSION THAT NO FAILURE MECRANISM OTHER TRAN CILC EXISTS.
2.
IN SOME AREAS, THE SURVEILLANCE PROGRAM RAS BEEN INCREASED IN SCOPE AND DEPTH TO SUPPORT CONTINUED.LONG TERM RADIOLOGICAL EFFECTS ASSESSMENT.
3.
A COMPREHENSIVE FUEL FAILURE MONITORING ACTION PLAN RAS BEEN ESTABLISHED TO PROVIDE AN FARLY INDICATION OF POTENTIAL LONG i
LIVED FISSION PRODUCT BUILDUP.
THIS DATA WILL BE PROVIDED TO MANAGEMENT TO SUPPORT SHORT AND LONG TERM STRATEGIC PLANNING.
4.
OFFSITE RELEASES ARE CURRENTLY A VERY SMALL FRACTION OF l
TECHNICAL SPECIFICATION LIMITS.
l A.
AN UPPER LIMIT FOR OFFGAS OF 80,000 uCi/SEC RAS BEEN ESTABLISHED TO ZNSURE TEAT A 10% BYPASS OF THIS SYSTEM UILL NOT RESULT IN EXCEEDING INSTANTANEOUS TECHNICAL SPECIFICATIONS.
B DOSE PROJECTIONS ARE BEING CONDUCTED AS DATA BECOME AVAILABLE TO ENSURE THAT APPROPRIATE STEPS ARE TAKEN IF THE TREND INDICATES THE LIKELIHOOD OF EXCEEDING 10% OF 10CFR50 APPENDIX I LIMITS.
5.
OFFSITE CONSEQUENCES RAVE BEEN FACTORED PROMINENTLY IN THE DECISION MAKING STEPS OF UPPER MANAGEMENT.
I 1
LIVERICK LhlT 1
NRC FUEL PRESENTATION MONDAY SEPTEMBER 12, '988 l
i CYCLE 2 OPERATING S RA EGY 4
PRESENTED BY: M. P. GALLAGHER
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CILC FAILURE: MOST LIKELY CAUSE REASONS:
= LIMERICK IS CILC SUSCEPTIBLE
= FEEDWATER COPPER HAS BEEN > 0.5 ppb l
= Cs-137 to Cs-134 RATIO INDICATED BUNDLE EXPOSURE > 15000 mwd /MT
= INITIAL CORE BUNDLES CONTAIN SOME NON HEAT TREATED CLADDING
= INCREASING OFFGAS ACTIVITY
= INCREASING DIFFUSION COMPONENT
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1/89 REFUELING INSPECTION PROGRAM e SIPPING TO BE PERFORMED DURING CORE OFFLOAD
- IRRADIATED BUNDLES RETURNING TO SERVICE TO BE SIPPED
- FUEL INSPECTIONS TO CLEAR RETURNING IC BUNDLES
- INSPECTION BASED ON 10% SAMPLING OF MAJOR Roll'S IN IC-248 FUEL
-INSPECT 75 GAD RODS
-lNSPECT 458 UO: RODS
- FUEL MUST BE VISUAL STANDARD 4 OR BETTER FOR RETURN TO SERVICE l
3
LIMERICK 1 CYCLE 3 CILC FUEL FAILURE RISK REDUCTION FACTORS
- MAXIMlZE USE OF HEAT TREATED (HT) CLADDING O RELOAD 2 224 BUNDLES,10096 HT O RELOAD 1 268 BUNDLES,10096 HT O INITIAL CORE 272 BUNDLES (INTERIOR LOCATIONS)
(8696 ALL RODS HT)
(6796 Gd RODS HT)
- MINIMlZE INITIAL CORE FUEL DUTY O 92 INITIAL CORE BUNDLES TO BE LOADED ON THE PERIPHERY, Gd ROD DUTY <4 kW/f t O 64 INITIAL CORE BUNDLES TO BE LOADED ONE ROW IN FROM THE EDGE, Gd ROD DUTY <4 kW/f t O REMAINING 116 INITIAL CORE BUNDLES ARE LOADED IN THE INTERIOR, Gd ROD DUTY PROJECTED TO REMAIN BELOW DUTY CURVE (28 BUNDLES - 0.5 kW/f t ABOVE CURVE, DURING EARLY PORTION OF CYCLE)
- INSPECTION PROGRAM FOR CLEARING INITIAL CORE FUEL FOR ADDITIONAL SERVICE O BEST VISUAL STANDARD MATERIAL USED IN HIGH POWER REGIONS O BOTH HT AND VISUAL STANDARD CRITERIA WILL BE l
USED IN IDENTIFYING FUEL FOR HIGH DUTY LOCATIONS 1
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1/89 REFUELING CONTINGENCIES LARGER NUMBER OF FAILURES MAY INDICATE SUFFICIENT VISUAL STANDARD 4 2.48 e BUNDLES MAY NOT BE FOUND FOR RE-USE LOWER GAD ROD DUTY CURVE MAY BE REOUIRED CONTINGENCIES UNDER REVIEW:
O RECONSTITUTE 28 - 2,48 e INITIAL CORE BUNDLES PROJECTED TO EXCEED GAD ROD DUTY CURVE -
REPLACE 37 GAD RODS WITH FRESH RODS i
O RECONSTITUTE ALL 116 - 2,48 e INTERIOR INITIAL 1'
CORE BUNDLES - REPLACE 150 GAD RODS WITH l
FRESH RODS
- 100*b OF INTERIOR 'e 48 e Gd RODS HEAT TREATED i
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- 88% OF ALL INTERIOR RODS IN 2,48 e BUNDLES ARE HEAT TREATED
- 97% OF ALL INTERIOR RODS ARE HEAT l
TREATED i
e O IF BOTH GAD AND UO: RODS ARE INVOLVED IN CILC FAILURES, CONSIDER USE OF LGS 2 FRESH BUNDLES
- l AS REPLACEMENTS FOR BUNDLES IMPLICATED e
l FUEL PLANNING CONTINGENCY IN EVENT OF PREMATURE REFUELING
- LOWER 2XPOSURE LICENSING WINDOW REQUIRED FOR EARLIER REFUELING
- USE 0.94 e LGS 2 FRESH FUEL IN PLACE OF 2.48 e INITIAL CORE BUNDLES IN THE INTERIOR
- EARLIER REFUELINGS POSSIBLE WITH INCREASING USE OF LGS 2 FUEL O 76 LGS 2 BUNDLES - 11/88 0 32 LGS 2 BUNDLES - 12/88
- LGS 2 FUEL PROVIDES ADDITIONAL BACKUP IF i
SUFFICIENT INITIAL CORE 2.48 e BUNDLES ARE NOT CLEARED BY INSPECTION LICENSING ISSUES
- PERFORM DUAL RELOAD LICENSING ANALYSIS FOR BOTH BASE CASE AND USE OF LGS 2 FUEL i
j
- SUBMIT BOUNDING CASE TO NRC FOR REFUELING j
- PERFORM 10 CFR 50.59 FOR ACTUAL LOADING ACHIEVED 1
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l PROGRAM PRESENTATION TO THE NRC SEPTEMBER 12,1988 t
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GREG K. BARLEY f
PLANT SUPERVISORY CHEMIST P
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l LONG TERM COPPER REDUCTION i
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SCHOLZ SEPT 12, 1988 i
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LONG TERM COPPER REDUCTION i
l DECISION MADE TO DEFER UNIT 2 CONDENSER REPLACEMENT l
SCHEDULE IMPACT HEAT TREATED FUEL IMPROVED CONDENSER LAYUP i
COPPER REDUCTION OPTIONS l
NEW CONDENSERS l
DEEP BED DEMINERALIZERS UNIT 1
RETROFIT PLANNED FOR 3RD OUTAGE (SEPTEMBER, 1990) i e
UNIT 2 RETROFIT PLANNED FOR isT OUTAGE (MARCH, 1991) j-
=w
18 ATTACHMENT 3 LIMERICK FUEL PERFORMANCE MEETING ATTENDEES E
J. C. Linville, Chief, Reactor Projects Section Chief 2A F. J. Witt, NRR, Chemical Engineer T. T. Martin, Director, Division of Reactor Safety (DRS)
W. V. Johnston, Deputy Director, DRS R. L. Fuhrmeister, Resident Inspector, Unit 2 L. L. Scholl, Resident Inspector, Unit 1 J. H. Williams, Project Enginee-J. Gadzala, Reactor Engineer P. D. Swetland, Acting Branch Chief, Projects Branch No. 2 T. F. Dragoun, Senior Health Physicist, DRSS M. M. Shanbaky, Chief, Facilities Radiation Protection Section, DRSS W. F. Kane, Director, Division of Reactor Projects J. R. Strosnider, Chief, Materials and Processes Section S. L. Wu, Reactor Engineer PECo L. F. Rubino, Superintendent G. M. Leitch, Vice President - Limerick M. J. McCormick, Plant Manager C. A. McNeill, Executive Vice President - Nuclear G. W. Murphy, Health Physicist I
G. C. Storey, Engineer-Supervisor, Fuel Management M. P. Gallagher, Reactor Engineer W. M. Alden, Director, Licensing R. J. Scholz, Manager-PPSS G. K. Barley, Plant Supervisory Chemist
ATTACHMENT 4 P
MEASUREMENT CONTROL EVALUATION NONRADIOLOGICAL CHEMISTRY (Closed)50-352/88-06-03(lice)nsee and Brookhaven National Laboratory, aOn completion of IFI :
(spiked samples) by the statistical evaluation was to be made.
The analyses were completed and an evaluation was performed. The anlaytical comparisons for the analytes were acceptable.
Analytical Results of Spiked Solit Samo'es unut= ppb Analyte Matrix Sample ID Limerick Brookhaven Fluoride Aux Boiler Steam A
12.2 1.8 12.92 0.5 Water (ABSW)
B 25.3 3.8 25.0 0 Chloride ABSW A
20.0 3.0 12.7 0 B
32.4 4.9 20.42 0.3 Sulfate ABSW A
26.7 24.0 12.2 1.5 8
35.6 5.3 20.7 20.6 Copper ABSW A
66.4 e6.6 61.0 1.0 B
162.0 16.2 170,0 11.0 Nickel ABSW A
65.?t 6.5 59.0 1.0 B
167. e16.7 160.0 8.0 Iron ABSW A
50.4 5.4
(
B 158.0 15.8 97 (.
Chromium ABSW A
62.526.3 2.0 B
164.02 16.4 146.02 2.0 Boron BIT 116.0 11.6 (2)
(1) Brookhaven's atomic absorption was contaminated with iron, therefore, measurements were not made.
(2) Comparison was not made due to the analytical sensitivity.
Limerick - used direct current plasma technique.
Brookhaven = used manitol titration technique.
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