IR 05000387/1988017
| ML17156A941 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 11/14/1988 |
| From: | Blough A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17156A940 | List: |
| References | |
| 50-387-88-17, 50-388-88-20, NUDOCS 8811300338 | |
| Download: ML17156A941 (24) | |
Text
U.
S.
NUCLEAR REGULATORY COMMISSION
REGION I
Report Nos.:
50-387/88-17.
50-388/88-20 Docket Nos.:
50-387 50-388 License Nos.:
NPF14.
NPF-22 Licensee:
Penns lvania Power and Li ht Com an 2 North Ninth Street Allentown Penns 1 vania 18101 Facility Name:
Sus uehanna Steam Electric Station Inspection At:
Salem Townshi Penns lvania Inspection Conducted:
Se tember
1988 October
1988 Inspectors:
Approved By.:
A. Randy Bl
, Chief Reactor Projects Section No.
3B F. Young, Senior Resident Inspector, SSES J. Stair, Resident Inspector, SSES R. Barkley, Reactor Engineer, Reactor Projects Section No.
3B Date Ins ection Summar
Areas Ins ected:
Routine resident inspection of plant operations, physical security, plant events, surveillance, and maintenance activities.
Specifi-cally, items reviewed in more detail in the facility operations areas were a
reactor recirculation pump run back, a
low level'adioactive waste shipment, emergency service water (ESW) pipe corrosion allowance, cross-tying of ATWS-RPT divisional power supplies, a
damaged battery cable, and MSIV stroke time test-ing.
Other items reviewed included the licensee's August monthly operating report, and licensee action on previous inspection findings.
Results:
Routine review of maintenance and surveillance activities noted good control and performance.
Licensee event reports (LERs)
and monthly reports were complete and accurate.
Control of, activities associated with removal of cable tray insulation demonstrated a weakness in the licensee's work controls over'his type of evolution.
The licensee's engineering review of ESW pipe corrosion allowance and subsequent documentation of their findings adequately resolved the license condition and their resolution was found acceptable.
In general, sufficient management involvement and attention was applied to operate both units'in a safe manner.
88i1300338 88iil4 PDR ADOCK 0 000387
FDC
TABLE OF CONTENTS 1.0 Introduction and Overview..................................
~Pa e
1.1 NRC Staff Activities (Module Nos.
71707, 71709, 71881, 30703)..
1.2 Unit 1 Summary.....
1.3 Unit 2 Summary.....
1.4 Persons Contacted.
2.0 Routine Periodic Inspections 2. 1 Scope of Review.
2.2 Reactor Recirculation Pump Run Back Unit 1 (Module No. 93702)...........
2.3 Review of Low Level Radioactive Waste Shipment 88-207 (Module No. 71709)................
3.0 Surveillance and Maintenance Activities.............
3. 1 Surveillance Observations (Module No. 61726).......
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3.2 Maintenance Observations (Module No. 62703)............
4.0 Licensee Reports
5 4. 1 In-Office Review of Licensee Event Reports (Module No. 90712)...........
4.2 Review of Periodic Reports and Special Report (Module No. 90713)
5.0 Emergency Service Water Pipe Corrosion Allowance (Module No. 92701)
e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
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6.0 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Divisional Power Supplies Cross-tied Unit 2 (Module No. 93702).
7.0 Damaged Battery Cable Unit 2 (Module No. 93702)..........
~
8.0 Hain Steam Isolation Valve (MSIV) Stroke Time Testing - Unit
(Module No. 93702).....
9.0 Licensee Action on Previously Identified Items (Module No. 92701)
10.0 Exit Meeting (Module No. 30703).
14
DETAILS 1.0 Introduction and Overview NRC Staff Activities The purpose of this inspection was to assess licensee activities at Susquehanna Steam Electric Station (SSES)
during power operation as it related to reactor safety and worker radiation protection.
Within each area, the inspectors documented the specific purpose of the area under review, scope of inspection activities and findings, along with appropriate conclusions.
This assessment is based on actual observa-tion of licensee activities, interviews with licensee personnel, measurement of radiation levels, or independent calculation and selective review of applicable documents.
1.2 Unit 1 Summar Unit
operated at or near full power until September 21, when a
recirculation pump run back occurred automatically lowering power to 60 percent (see Detail 2.2).
The unit was returned to full power later the same day and'perated at full power until October 5,
when power was reduced to
percent in order to plug a
main condenser tube.
FulT power was restored on October 7, and was maintained for the remainder of the inspection 'eriod.
In addition, periodic scheduled power reductions were conducted for control
'rod pattern adjustments, surveillance testing, and scheduled maintenance.
1.3 Unit 2 Summar Unit 2 operated at or near full power until October 7, when conden-sate demineralizer influent conductivity increased.
and the licensee initiated a power reduction to 60 percent in order to locate and plug a
leaking tube in.the main condenser.
Full power was restored on October 10, however, during the following two weeks the licensee re'duced power two more times due to increases in condensate deminera-lizer influent conductivity and unsuccessful attempts to locate con-denser tube leaks.
Since the conductivity increases appear when cir-culating water temperatures decrease, the cooling tower bypass valves were closed to maintain higher circulating water temperatures.
In addition, periodic scheduled power reductions were conducted for con-
. trol rod pattern adjustments, surveillance testing, and scheduled maintenance.
None of the conductivity increases caused chemistry parameters to approach Technical Specification limit e 1.4 Persons Contacted During the course of the inspection, the inspector interviewed, dis-cussed issues, and received information from various licensee employees.
Listed below are the senior people and those individuals who sup-plied substantive information.
Members who attended the exit inter-view on October 31, 1988, are indicated by an asterisk.
J. Blakeslee, Assistant Superintendent of Plant, SSES F. Butler, Supervisor of Maintenance, SSES
- R.
G.
Byram, Superintendent of Plant, SSES
- T.
C. Dalpiaz, Supervisor of Technical Support, SSES J.
Doxsey, Reactor Engineering Supervisor, SSES E. Figard, Supervisor of I&C/Computer, SSES
- D. J.
Gandenberger, Mechanical Maintenance Supervisor, SSES J.
Graham, Assistant Manager, NQA T. Markowski, Day Shift Supervisor, SSES
- W.
E. Morrissey, Radiological Protection Supervisor, SSES T.= Nork, Plant Engineering Group Supervisor, SSES
- J.
E. O'ullivan, Insulation Engineering Group, SSES
~
R. J.
Prego, Supervisor of QA Operations, NQA H. Riley, Supervisor of Health Physics/Chemistry, SSES
" D. Roth, Senior Compliance Engineer, SSES
- R. L.'Stotler, Supervisor of Security, SSES
" H.
G. Stanley, Assistant Superintendent-Outages, SSES 2.0 Routine Periodic Ins ections The NRC resident inspectors periodically inspected the facility to determine the licensee's compliance with the general operating requirements of Section 6 of the Technical Specifications (TS) in the following areas:
review of selected plant parameters for abnormal trends; plant status from a maintenance/modification viewpoint, includ-ing plant housekeeping and fire protection measures; control of -ongoing and special evolutions, including control room personnel awareness of these evolutions; control of documents, including logkeeping practices; implementation of radiological controls;
implementation of the security plan, including access control, boundary integrity, and badging practices; control room operations during regular and backshift hours, including frequent observation of activities in progress, and periodic reviews of selected sections of the unit supervisor's log and control room operator'
log and other control room daily logs; followup items on activities that could affect plant safety or impact plant operations; areas outside the control room; and, selected licensee planning meetings.
The inspector conducted backshift inspection on September
and 26, 1988, and weekend coverage on September 25, 1988.
Also, the inspectors reviewed specific events in more detail as described in the sections that follow.
2.2 Reactor Coolant Recirculation Pum Run Back Unit
On September 21, 1988, the B reactor coolant recirculation pump auto-matically ran back to the 30 percent speed limiter.
The plant con-trol operator responded by manually running back the 'A'eactor coolant recirculation pump to 30 percent in order to maintain speed of the two pumps within the required
percent difference.
As a
result, power dropped to 60 percent with total core flow stabilizing at approximately 49 million pounds per hour.
During the run back the status light for the +13 inch level run back signal illuminated on the control panel.
Thus, the licensee first verified that no actual
+13 inch level signal existed and then directed their investigation toward the reactor coolant recirculation system run back logic.
From an initial review of the logic system the licensee suspected that level run back actuating relay (B31-K47B)
had failed The licensee removed the run back actuating relay, placed the suspect relay on a test bench and verified its improper functioning.
The relay was replaced with a
spare relay and subse-quently surveilled in place.
Following Plant Operations Review Committee review of the event, determination of cause, and corrective
.
actions taken, the unit was returned to full power later that day.
The inspector discussed the event with control room personnel, wit-nessed the troubleshooting effort by Instrument
&
Control tech-nicians, and verified proper plant response.
The inspector found the licensee's actiops in response to this event acceptabl.3 Review of Low Level Radioactive Waste Shi ment 88-207 The inspector examined the physical condition of the shipping con-tainer and carrier vehicle for low level radioactive waste shipment 88-207.
The shipment contained dewatered powdex resin and was being shipped to Barnwell, South Carolina for disposal.
The inspector reviewed the health physics survey on the container and the vehicle as well as the radiation postings on the shipping cask.
No uaaccept-able conditions were identified.
3.0 Surveillance and Maintenance Activities On a sampling basis, the inspector selected several surveill.ance and main-tenance activities to ensure that specific programmatic elements described below were being met.
Details of this review are documented in the following sections.
3. 1 Surveillance Observations The inspector observed the performance of surveillance and special tests to determine that:
the test procedure conformed to Technical Specification requirements; administrative approvals and tagouts were obtained before initiating the test; testing was accomplished by qualified personnel in accor'dance with an approved procedure; test instrumentation was calibrated; limiting conditions for operations
~
were met; test data was accurate and compl,ete; removal and restora-tion of the affected components was properly accomplished; test results met Technical Specification and procedural requirements; deficiencies noted were reviewed and appropriately resolved; and the surveillance was completed at the required frequency.
These observations included:
SI258-204, Monthly Functional Test of The Scram Discharge Volume (SDV) High Water Level Channels LIS-C12-2N601 A,B,C,D, performed on October 7, 1988.
Ultrasonic Examination (UT),
eddy current testing and oxide measurement examinations on one spent fuel bundle to identify failed fuel rods for replacement and to measure the amount of oxidation/film buildup on the surface of the spent fuel clad-ding.
The applicable Reactor Engineering procedures governing the conduct of this work are RE081-034, Revision 0,
'Determina-tion and Inspection of Failed Fuel Rods'nd RE-081-035, Revision. 0, 'nspection and Repair of'ailed Fuel Bundles'
The testing observed was performed on October 13, 1988.
No unacceptable conditions were identifie.2 Maintenance Observations The inspector observed portions of selected maintenance activities to determine that the work was conducted in accordance with approved procedures, regulatory guides, Technical Specifications, and industry codes or standards.
The following items were considered during this review:
Limiting Conditio'ns for Operation were met while components or systems were removed from service; required administrative approvals were obtained yrior to initiating the work; activities were accomplished using approved procedures and gC hold points were estab-lished where required; functional testing was performed prior to declaring the particular component(s)
operable; activities were accomplished by qualified personnel; radiological controls were implemented; fire protection controls were implemented; and the equipment was verified to be properly returned to service.
These observations included:
Instrument calibrations on 'D'iesel generator day tank level transmitter LT 034770, and'evel indicator LI 03477D, per Work Authorization P81069, performed on October 5, 1988.
'leaning, inspection, and repair coating of reactor building
. closed cooling water heat exchanger
'A', per.Work Authorization S75131, performed on October 12, 1988.
Disassembly and repair of the motor driven fire pump, per Work Authorization S73258, performed on October 12, 1988.
No unacceptable conditions were identified.
4.0 Licensee Re o ts 4. 1 In-Office Review of Licensee Event Re orts The inspector reviewed LERs submitted to the NRC to verify that details of the event. were clearly reported, including the accuracy of description of the cause and adequacy of corrective action.
The inspector determined whether further information was required from the licensee, whether generic implications were involved, and whether the event warranted onsite followup.
The.
following LERs were reviewed:
Unit
88-018-00 One Channel of Reactor Water Cle'anup System Area High Tem-perature Isolation Instrument Found Inoperable.
88-019-00 Valves Not Tested per ASME Code Section XI Requirement Unplanned ESF Actuation When Zone III Heating, Ventilation, and Air Conditioning.Isolated.
88-003-01 Misscheduling of Surveillance Procedure Results in Opera-tions Prohibited by Technical Specifications.
Unit 2 88-014-00 Reactor Water Cleanup System Isolation on High Room Dif-ferential Temperature Trip Signal.
The above LERs were found acceptable.
4.2 Review of Periodic and S ecial Re orts Upon receipt, periodic and special
.reports submitted by the licensee were reviewed by the inspector.
The reports were reviewed to determine that it included the required information; that test results and/or supporting information were consistent with design predictions and performance specifications; that planned corrective action was adequate for resolution of identified problems; and whether any information in the report should be classified as an abnormal occurrence.
The following reports were reviewed; Monthly Operating. Report August, 1988, dated September 15, 1988 Special Report on Meteorological Monitoring Instrumentation (ER 100450),
dated October 19, 1988
.
No unacceptable conditions were identified.
5.0 Emer enc Service Water Pi e Corrosion Allowance The operating license for Susquehanna Steam Electric Station Unit
required that complete corrective actions for a deficiency involving the corrosion allowance for emergency service water (ESW) piping to and from the RHR pump motor oil coolers be completed by January 1988.
The FSAR required a
minimum corrosion allowance of 0.25 inches for class HRC pip-ing.
However, this portion of piping installed in the ESW system was class HPC piping with a different corrosion allowance of.080 inches.
The initial prediction based on using class HRC piping showed that this cor-rosion allowance would be insufficient to insure piping integrity over the lifetime of the facility (forty year lifetime);
During the second refuel-ing for Unit 2 the licensee conducted an inspection of this pipin A representative sample of the HPC piping was removed and analyzed.
Labor-atory analysis demonstrated that a sufficient corrosion allowance existed for the piping and it would ensure that it would last for the design life of the plant.
In a letter dated March 10, 1988, to William T. Russell, Region I Administrator, USNRC, from H. Keiser, Sr. Vice-President-Nuclear, PP&L, the licensee restated the information above
'and considered their action completed on this item.
The licensee also stated that they would continue to remove and analyze piping samples as necessary to ensure that the corrosion allowance existed during the life of the plant.
The inspector reviewed and discussed this item with the licensee and determined that the licensee had taken actions appropriate and conserva-tive for this item.
A review of the documentation associated with the analysis demonstrated that the installed piping appeared to have suffici-ent corrosion allowance to last for the design life of the plant.
Based on this information the inspector concurred with the licensee's resolution of this issue and considered this item completed.
6.0 Antici ated Transient Without Scram - Recirculation Pum Tri ATWS-RPT Divisional Power Su lies Cross-Tied Unit 2 On September 19, 1988, during an investigation of electrical grounds on 125 VDC batteries 2D610 5 2D620, jumpers were discovered between terminal blocks (TB)2C005-Dl and TB2C005-D2 and between TB2C004-Dl and TB2C004-D2.
These jumpers cross-tied ATWS-RPT logic division I and II power supplies.
This occurred when a design modification was installed during the Unit 2 second refueling outage and the jumpers were inadvertently left in place.
As a result 'of the cross-tie, a detailed study of th'e loads and circuit protection provided was performed in order to determine the safety signif-icance of the event.
The licensee determined from the study that the Unit
125 VDC systems would still have been able to fulfill thei r required safety functions, although the installed condition resulted in a decrease in the independence of redundant class 1E power supplies.
Nonconformance Report (NCR) 88-0631 was written to identify the condition and determine a
disposition which included both the study performed and the removal of the jumpers.
The inspector verified the removal of the jumpers, reviewed the noncon-
~formance report and discussed the event with the licensee.
The inspector determined that the licensee'.s initial corrective actions in response to this event were acceptable, however, the safety significance of this event and the inadequacy in the design change'ackage which led to this event will be reviewed by a
NRC Region I specialist in Inspection Report No.
387/88-18; 388/88-2.0 Dama ed Batter Cable - Unit 2 On September 23, 1988, a construction worker who was removing thermolag (fire retardant insulation)
on a
safety related cable tray, cut through the 'A'50 volt battery charger cable grounding it to the cable tray.
The worker was in the process of removing the thermolag =in order to repair cracks in the cable tray's insulation.
Because the thermolag forms a hard concrete type structure, the worker was using a chisel to breakup and remove damaged sections of the thermolag which needed repairs.
In the process the chisel being used penetrated the cable's protective covering and cut into the charger cable.
Grounding of the cable caused an elec-trical arcing. to occur which burned a'ortion of the battery charger cable.
No injuries occurred as a result of the grounding.
8.0 The licensee entered the Technical Specification L.C.O. action statement for an inoperable battery charger and verified that the affected battery bank met applicable technical specification'
surveillance criteria.
The licensee debriefed the workers and increased surveillance testing of the battery bank until repairs to the cable were made.
The inspector discussed the event with the licensee's staff and reviewed the significant operating occurrence report.
The inspector found the licensee's immediate corrective actions acceptable.
The inspector, how-ever, noted that failure to properly remove cable tray insulation
'demon-
. strated a weakness in the licensee's work practices over this type evolu-.
tion.. The licensee acknowledged the inspector's concern and stated that the applicable controls would be factored into their program.
The inspec-tor discussed the additional controls and concluded that the actions adequately address the concerns.
Main Steam Isolation Valve MSIY Stroke Time Testin Unit I The unit's Technical Specifications (T.S.) requires that the MSIVs be full stroke tested every 91 days.
Actual stroke time must be with the range of 3-5 seconds.
The licensee uses the associated limit switches and a stop-watch to determine this time.
The closed limit switch is the switch nor-mally used to indicate a
closed position for the valve.
However, when these.limit switch indicates close the valve is actually about 90 percent closed.
Because of this fact, the licensee had been adjusting the str'oke time limit to 2.7 to 4.5 seconds (90 percent of 3-5 seconds).
In preparing to perform the MSIV stroke time test, the system engineer questioned the actual setting of the closed limit switch.
A review of the maintenance records indicated that a
range of 3 to 10 percent open indi-cation was allowable for setting the limit switch by maintenance.
A review of current limit switch settings indicated that MSIV 1F022B had been set at 3 percent open.
Thus, based on the actual previously recorded stroke time of 2.72 seconds compared to the recomputed allowable range of 2.91 to 4.85 seconds, (97 percent of 3-5 seconds)
the licensee determined that valve had failed the surveillance and declared the valve inoperabl On September 21, 1988, the licensee reduced power to
percent and retested the valve.
Retesting determined that the valve was operable based on the stroke time of 3.0 seconds.
(Allowable range of 2.91 to 4.85 seconds)
Review of other MSIV's records determined this to be the only case where a valve had failed to meet its intended stroke time.
The inspector reviewed the licensee's corrective actions and considers this action appropriate and timely.
The inspector did express a concern that this item could occur again if appropriate programmatic changes were not implemented.
The licensee acknowledged the inspector's concerns and stated the applicable procedure changes to address the concerns were being formulated.
Discussions with the system engineer indicated that he would be reviewing the data with respect to the problem to assure that a failure of the same type would be identified earlier.
Discussions with the licen-see also indicated that the licensee was not satisfied with this metho-dology of obtaining the data (i.e.
~
The present method requires an operator'o start and stop a
stopwatch based on seeing a
change in valve status lights.
The licensee is independently evaluating other ways of obtaining the information more easily and precisely.
Based on the licensee immediate and proposed long term corrective actions, the inspector considers this problem resolved.
9.0 Licensee Action on Previousl Identified.Items 9. 1 Closed Unresolved. Item 387/87-16-01:
Licensee to investi ate the cause of and conduct re airs to the leakin service water su
lines to the
'B and C
Core S ra room coolers.
On September 8,
1988, the licensee identified a significant leak from a flexible supply line from the Emergency Service Water System to the
'B'ore Spray (CS)
room cooler.
At that time, the 'B'S loop was declared inoperable and the flexible supply line was replaced.
Later that day, another leak on the 'C'S room cooler was noted.
An inspection of all of the Emergency Core Cooling System (ECCS)
room coolers was subsequently conducted.
Fourteen leaking flex line connections were identified as a result of that inspection.
As a result of this finding, the licensee replaced the leaking ser-vice water connections on the 'B'nd 'C'ore Spray room coolers and sent the failed lines to their laboratory for nondestructive examina-tion.
Material analysis of the fa'iled flexible lines determined the primary mode of failure to be a form of galvanic corrosion occurring at the junction of the carbon steel and stainless steel portions of the tubing:
As a result of this finding, the licensee initiated plant modifications (DCP No.
87-9209 and 87-9210)
to replace a'll of the Unit 1 and Unit 2 ECCS room cooler flexible hoses.
The replace-ment hoses use an all stainless steel construction which serves to eliminate the problem of galvanic corrosion experienced by the exist-ing line i
The inspector reviewed the Significant Operating Occurrence Report (SOOR 1-87-239 and 87-240) covering the failures of the flexible tub-ing.
He also reviewed portions of DCP No.
87-9209 and the reporta-bility evaluation surrounding these fail'ures (the failures were determined to be nonreportable).
The inspector noted that DCPs 87-9209 and 87-9210 have been completed, and no problems have been noted with the replacement lines.
The inspector found the licensee's actions acceptable and considers this item closed.
9.2 Closed Violation 387/87-12-01:
Fire Door No.
420 was ino erable and no com ensator fire watch was established.
On 'August 13, 1987, an inspector noted that fire. door No.
420 was inoperable and no compensatory fire watch was established per Tech-nical Specification 3.7.7.
In addition, operations surveillance procedure SO-200-007 governing the checks on fire door operability did not list the fire door as inoperable.
A severity level Y viola-tion was issued as a result of this finding.
As a result of this violation, the licensee stated in their response to the violation, dated October 8,
1987, that the fire watch was promptly dispatched to the, door, the fire door was repaired and the Daily Surveillance Operating Logs SO-100-007 and SO-200-007 were changed to clearly alert operations personnel to the potentially inoperable condition of a fire door.
The inspector reviewed the changes to SO-100-007 and SO-200-007.
The inspector found the licensee's actions in response to this violation acceptable and considers this item closed..
9.3 Closed Ins ector Followu Item 387/84-26-01 u dated in Ins ection Re ort 387/85-31 388/85-26:
Licensee to investi ate the cause of the turbochar er bearin failure on diesel enerator D
.
As documented in combined inspection report 387/85-31; 388/85-26, the source of the turbocharger bearing failure was determined to be the lack of bearing prelubrication upon engine starting.
To pr event a
recurrence of this problem, the licensee incorporated a turbocharger bearing prelubrication sequence prior to starting the diesel engine during scheduled surveillance testing.
Operating Procedure OP-024-001, Diesel Generators, was revised in February 1985 to allow for manual initiation of turbocharger lubrication prior to surveillance testing.
Since this change in operating procedure, no additional turbocharger failures have occurre Since the failure of the
'D'urbocharger, the licensee contracted with the turbocharger vendor to disassemble and inspect the turbo-chargers on the 'A'nd 'B'iesel generators (the 'C'iesel genera-tor turbocharger was replaced in February 1984 after experiencing a
turbocharger thrust bearing failure).
The inspection of those turbo-chargers revealed only minor wear, although the 'B'iesel generator did require replacement of three (3) of the turbocharger blades due to material damage.
9.4 The inspector reviewed the Procurement -Source Inspection Reports SSIP 87"16 and 87-45 governing the gC inspection of the vendor's inspec-tion and rework of the
'A'
'B'urbochargers.
The inspector found the licensee's actions acceptable and considers this item closed.
Closed Unresolved Item 387/87-23-01
Licensee to develo a new adminisrative test rocedure to control all h sics related tests durin be innin -of-c cle lant startu The inspector noted, during inspection 387/87-23 that the test proced-ures used during startup physics testing were administratively con-trolled under the Surveillance Test Program (AD-gA-422).
While this program provided general administrative guidance on all plant sur-veillancee tests, it did not specifically address the type, test con-dition and test sequence of 'physics testing which is required to'. be performed to demonstrate the adequacy of the core. characteristics as a result of new fuel cycle operations.
At the exit meeting for inspection 387/87-23, the licensee agreed to generate a
new adminis-trative test procedure to control all physics related tests during beginning-of-cycle plant startup.
The licensee has since implemented Administrative Directive AD-gA-138, Revision 0,
'Control of Core Reactivity Changes at SSES,'o control all core reactivity changes at Susquehanna.
The inspector reviewed the procedure.
The inspector found the licensee's actions acceptable and considers this item closed.
9.5 Closed Unresolved Item 388/86-26-02
Review corrective actions in res onse to under-tor ued containment hatches.
On October 27, 1986, at approximately 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br />, it was discovered that several.nuts used to securely bolt the Unit 2 primary contain-ment suppression pool hatch covers were loose, i.e.
less than hand tight.
The control room personnel were notified and a Limiting Con-dition for Operation (LCO) was entered on primary containment integ-rity at 1720 hours0.0199 days <br />0.478 hours <br />0.00284 weeks <br />6.5446e-4 months <br />.
After further review of the event by management, the licensee decided to inspect the suppression pool and Control Rod Drive (CRD) hatches on both units.
The CRD removal hatches on both units were also found to have loose bolting arrangements.
LCO's were entered for those hatches and the appropriate Emergency Notification System (ENS)
notifications were made.
Local leak rate tests were conducted on the hatches in their 'as found'ondition with satis-factory results achieve Subsequent analysis of the as-found bolting conditions on all of the hatches revealed that all were bolted adequately enough to ensure
'hey would remain in place in the event of an accident.
To prevent a recurrence of this event, the licensee issued procedure MT-059-004, Revision 2,
'Preferred Method for Containment Suppression Pool Hatch (X-200A and 'X-200B)
Removal and Installation to ensure that the hatches are properly secured when removed.
In addition, procedures MT-059-005, -006, -007, and -008 were issued to cover the removal and installation of the CRD, drywell head access manhole, containment access, and containment equipment hatches, 9.6 The inspector reviewed procedure MT-059-004, Revision
as well as the licensee's investigation into the cause of this event and their corrective actions, as documented in the Significant Operating Occurrence Report (SOOR)
closeout package, The inspector found the licensee's actions acceptable and considers this item closed.
Closed Violation 387/86-06-05
Scram Dischar e Volume SDV level transmitters were isolated from the SDV.
On April 10, 1987, with Unit
in a
refueling outage, SDV level transmitters LT-C12-1N016C and LT-C12-)N016D were found isolated from the SDY.
The level transmitters provide scram signal's to the Reactor Protection System in the event the SDV water inventory 'is high, ensuring that sufficient volume exists for the water discharged dur-ing a reactor scram.
The discovery was made when a conflicting level indication for the SDV was observed during hydostatic pressure test-ing of the reactor vessel.
An enforcement conference with PP&L was held on this issue in Region I on May 30, 1986.
At that meeting, the licensee stated that the cause of the SDY level transmitter isolation was the inadequate completion of a step in the modification closeout process (the valves were locked closed during the modification process).
The licensee promptly corrected the situation and conducted an in depth investiga-tion into the cause of the error and into recommended corrective actions to prevent a recurrence of this problem.
The licensee's correc'tive actions in response to this violation (other than the immediate corrective actions to unisolate the trans-mitters)
included a review of almost
.400 modification packages, as well as a walkdown of electrical panels and instrumentation racks, to attempt to identify any other similar errors in the lineup or comple-tion status of recently implemented modifications.
No problems were discovered.
Long term corrective actions to prevent a recurrence of this problem include:
( 1) allotting more lead time for the review of modifications, (2) providing definitive guidance on the conduct of closeout reviews to the operations engineer, (3) promoting interac-tion between the systems engineer and the operations engineer, (4) adding a
second review to the modification procedure closeout process, and (5) incorporating post-modification in-service checks of equipment performanc ~
The inspector reviewed the Significant Operating Occurrence Report (SOOR)
1-86-113 on the incident, Licensee Event Report 86-014-01, the material presented by the licensee during the enforcement conference in Region I and Procedure Change Approval Form (PCAF)
1-86-530.
The inspector found the licensee's actions for this violation acceptable and considers this item closed.
9.7 Closed Violation 387/85-12-02
- Ins ection of Fire Dam ers During inspection 387/85-12,
.the inspector
,identified
. that the safety-related fire dampers
'FPD-3-27-8-1SC,
-2SC,
-3SC and FPD-3-30-8-1SC had not been visually inspected as required
'in - the Technical Specifications.
The affected fire dampers are Ruskin 1BD23a self-actuating fire damper assemblies in the ducting between the standby gas treatment system trains and the recirculation plenum in the Reactor Building.
The surveillance procedures SM-113-009 and SM-213-009 are used to implement TS 4.7.7. 1 which requires an
month visual inspection of all safety related fire dampers.
It was deter-
. mined that fire dampers FPD-3-27-8-1SC,
-2SC,
-3SC and FPD-3-30-8-1SC were inadvertently omitted from the surveillance procedures.
This item was reviewed in NRC inspection 387/88-12 and 388/88-15.
At that time, the licensee's corrective actions were reviewed and found acceptable.
However, at that time, the licensee did'ot pro-vide the results of a review of all safety-related fire dampers as committed to in their response to the Notice of Violation.
The licensee has since provided the results of a review of the venti la-tion drawings for the facility as well as the results of an engineer-ing review conducted by PP&L's corporate office staff (EWR-MIS-85-0584).
As a result of these reviews, the licensee identified twelve (12)
additional fire dampers which were required to be surveilled under their approved fire zone analysis.
Those dampers were subse-quently incorporated into surveillance procedure SM-013-009 (which superceeded procedure SM-113-009 and SM-213-009).
The inspector reviewed EWR-MIS-85-0584 and SM-013-009, Rev.
1.
The inspector found the licensee's actions in response to this violation acceptable and considers this item closed.
9.8 Closed Unresolved Item 388/84-34-09
- Licensee to Provide an Ex lanation of The Partial Full Core Dis la Indications Which Occurred After The Loss of AC Power Event on Jul 26 1984.
On July 26, 1984, Unit 2 experienced a
Loss of All AC Power during the conduct of Startup Test ST 31. 1,
'Loss of Turbine Generator and Offsite Power'.
Following the ensuing reactor scram, the control room full core display showed a
"checkerboard" pattern with only about one half to two-thirds of the rods indicating "full in" or properly inserted into the core (the ful'1 core and four rod displays on the Standby Information Panel (SIP)
and Rod Position Indication
System (RPIS),
which are powered from Uninterruptible Power Supplies (UPSs),
provided alternate control rod position indication during the course of the event).
The licensee's action plan to identify and correct the problems which led to the loss of all AC power event committed to determining the cause of these display indications and to make appropriate recommendations.
The licensee determined that the cause of, the core display inadequa-cies was the unavailability of AC power to the Unit 2 non-class IE Instrument AC Panel 2Y218, which feeds selected control room instru-mentation.
To prevent a
recurrence of this problem, the licensee issued Design Change Packages 86-3002C and 3002D to install a
KVA Uninterruptible Power Supply (UPS)
to feed Instrument Panel 2Y218.
The inspector reviewed the safety evaluation for DCP-3002 as well as the initial startup and electrical test of the UPS.
The inspector noted that this inspection item was also reviewed during inspection 388/85-10 and left open pending the installation of this modifica-tion.
The inspector found the licensee's actions in response to this event acceptable and considers this item closed.
9.9 Closed Ins ector Followu Item 388/84-15-01 Licensee To Install and Calibrate Five Unit 2 Area Radiation Monitors.
At the time NRC inspection 388/84-15 was conducted, the licensee was awaiting the receipt of 5 area radiation monitors (ARMs) that they were experiencing difficulty in procuring.
.Since that time, "the licensee has installed those ARMs and placed them in service.
The inspector reviewed Design Change Package (OCP)
87-9132, which installed a higher range ARM for the traversing Incore Probe (TIP) room and examined the ARM in the field.
The inspector found the licensee's actions acceptable and considers this item closed.
lD. ~EfI On October 31, 1988, the inspector discussed the findings of this inspec-tion with station management.
Based on NRC Region I review of this report and discussions held with licensee representatives, it was determined that this report does not contain information subject to
CFR 2.790 restric-tions.
At the conclusion, the licensee acknowledged the NRC findings and did not disagree with the findings or their characterization.