IR 05000387/1999004
| ML17164B069 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 06/04/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17164B068 | List: |
| References | |
| 50-387-99-04, 50-387-99-4, 50-388-99-04, 50-388-99-4, NUDOCS 9906160169 | |
| Download: ML17164B069 (40) | |
Text
U.S. NUCLEAR REGULATORYCOMMISSION
REGION I
Docket Nos:
License Nos:
50-387, 50-388 NPF-14, NPF-22 Report No.
50-387/99-04, 50-388/99-04 Licensee:
Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 19101 Facility:
Susquehanna Steam Electric Station Location:
P.O. Box 35 Berwick, PA 18603-0035 Dates:
March 16, through April26, 1999 Inspectors:
S. Hansell, Senior Resident Inspector J. Richmond, Resident Inspector A. Blarney, Resident Inspector B. Welling, Peach Bottom, Resident Inspector J. McFadden, Senior Reactor Inspector T. Burn, Reactor Inspector A. Lohmeier, Senior Reactor Inspector Approved by:
A. Randolph Blough, Director Division of Reactor Projects 9906160169 990604 PDR ADQCK 05000387
EXECUTIVESUNlMARY Susquehanna Steam Electric Station (SSES), Units 1 & 2 NRC Inspection Report Nos. 50-387/99-04, 50-388/99-04 This inspection included aspects of Pennsylvania Power and Light Company's (PPB L's)
operations, maintenance, engineering, and plant support at SSES.
The integrated report covers a six-week period of routine resident inspection activities and inspections of your residual heat removal valve failure event, radiological controls outage program, and inservice inspection (ISI)
program.
~Oerations
~
During the Unit 2 refueling outage, the control, execution, and performance of major activities were good. Management emphasized to the staff the importance of human performance, attention to detail, and personnel safety throughout the outage.
(Section 01.1 and 01.2)
~
On February 27, 1999, PPB L did not notifythe NRC, within one hour of identification, that the Unit 1 RHR loop 8 was in a condition that was outside of the design basis.
The failure to make the notification within one hour is a violation of 10 CFR 50.72. This Severity Level IVviolation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy. This violation is documented in PPB L's corrective action program as part of condition report 90981.
(Section E2.1)
ln a Licensee Event Report, PP&L identified that two main steam isolation valves did not meet a seat leakage specification.
PP&L's corrective actions, including valve seat repair and re-test activities, were good. This Severity Level IVviolation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy. This violation is documented in PPB L's corrective action program as condition report 92338.
LER 50-388/99-001 is closed.
(Section 08.1)
Maintenance During a Unit 2 under vessel maintenance activity on a local power range monitor (LPRM), technicians momentarily unseated the seal between the LPRM and reactor vessel, spraying about two quarts of contaminated water on themselves.
The additional total radiation exposure (both internal and external) to the individuals, as a result of the contamination, was non-consequential.
PP&L initiated condition reports 92527 and 92528 to review this event.
PPB L's initial response and proposed actions were appropriate.
No violations of NRC requirements were identified. (Section M1.1)
The inservice inspections had been performed acceptably and had included acceptable ASME program coverage, qualified personnel, approved procedures, proper implementation, appropriate examination documentation, and PPB L oversight. The inspections performed were thorough and of sufficient extent to determine the integrity of the components inspected.
Indications of nonconforming conditions were identified, explored, evaluated, documented and dispositioned in accordance with established requirements.
(Section M2.1)
e
~En ineerin Engineering personnel performed a comprehensive failure analysis and a thorough root cause determination of the Unit 1 and 2 low pressure coolant injection valve stem failures. Corrective actions, including the replacement of internal parts with an improved design and material, were acceptable.
(Section E1.1)
Design control deficiencies in the mid-1980's led to the use of material in the stem of the low pressure coolant injection valves that was susceptible to stress corrosion cracking.
These deficiencies constituted a violation of 10 CFR 50 Appendix B, Criterion III, "Design Control," In accordance with the NRC Enforcement Policy,Section VII.B.3, Violations Involving Old Design Issues, the NRC exercised enforcement discretion and did not cite this violation. (Section E1.1)
PP8L identified and resolved a potential common cause failure of all RHR injection valves prior to any need for the system to function. The operator's initial identification of the "1B" residual heat removal system slow pressurization reflected a good questioning attitude. The system engineer appropriately focused station priorities to complete troubleshooting and identify a failed low pressure coolant injection valve. (Section E2.1)
PP&L did not troubleshoot the unexpected slow pressurization of the residual heat removal system in a structured and methodical manner, which extended the time needed to identify the'failed "1B" low pressure coolant injection valve. While troubleshooting was in progress, and unaware that the "1B"valve was failed, PP&L removed the "A"residual heat removal system from service for maintenance for 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />.
During this 17 hour1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> period, neither Unit 1 loop of residual heat removal was available for low pressure coolant injection. (Section E2.1)
The PP&L independent safety engineering group review of the slow pressurization of the residual heat removal system event was limited and missed opportunities to reveal additional insights related to the station staff response to the failed valve. (Section E2.1)
Although PP8L identified problems with roles and responsibilities during troubleshooting, they did not identify a change management issue.
Specifically, the station staff was not fullyaware that the system engineering interface with operations and maintenance during troubleshooting had been changed.
(Section E2.1)
Access controls to radiologically controlled areas were effective, and appropriate occupational exposure monitoring devices were provided and used.
Personnel occupational exposure was maintained within applicable regulatory limits and as low as reasonably achievable (ALARA). The ALARAefforts and results for 1998 were good, including the management of radiologically significant outage work. The annual and Unit 2 refueling and inspection outage collective dose goals for 1999 were aggressive and challenging.
(Sections R1.1 and R1.3)
Radiological housekeeping conditions were noted to be good.
In particular, the equipment and personnel work activity control and coordination for the Unit 2 refuel floor, suppression pool, and drywell were excellent.
The number and type of personnel contaminations were tracked, trended, and evaluated for cause and corrective actions.
(Section R1.2)
PP8L's self-identification and corrective action processes in the area of radiation protection were effective. Nuclear Assessment Services surveillance reports, HP self-assessments, and the corrective action program continued to be effective in identifying, at a low threshold, deficiencies and improvement opportunities.
Effective corrective actions were implemented for findings. (Section R7)
TABLEOF CONTENTS EXECUTIVESUMMARY.
TABLE OF CONTENTS I. Operations
.
Conduct of Operations 01.1 Unit Operations and Operator Activities 01.2 Unit 2 Refuel Outage Activities
Operational Status of Facilities and Equipment 02.1 Operational Safety System Alignment
Miscellaneous Operations Issues...
08.1 Licensee Event Report Review
1
.
1
~.. 2 II. Maintenance..
M1 Conduct of Maintenance......
M1.1 Surveillance and Pre-Planned Maintenance ActivityReview
.
M2 Material Condition of Facilities and Equipment.
M2.1 Inservice Inspection (ISI) Review M8 Miscellaneous Maintenance Issues M8.1 Followup of Open Items..
.3
..7
III. Engineering
.
E1 Conduct of Engineering..
E1.1 Residual Heat Removal System Injection Valve Stem Failure Root Cause E2 E8 Determination and (Closed) LER 50-387/99-001 Engineering Support of Facilities and Equipment
.
E2.1 Station Response Related to the Failure of the Unit 1 Residual Heat Removal Injection Valve HV-151F017B..
Miscellaneous Engineering Issues E8.1 Followup of Open Items
10
14
IV. Plant Support
..........................
~..
~.. ~...... ~..........15 R1 Radiological Protection and Chemistry (RP8C) Controls................
R1.1 Radiological Controls-External and Internal Exposure.............
R1.2 Radiological Controls-Radioactive Materials, Contamination, Surveys, and Monitoring R1.3 Radiological Controls-As LowAs Reasonably Achievable (ALARA)..
17 R5 R7 F8 Miscellaneous Fire Protection Issues........
Staff Training and Qualification in RPBC Activities..
. 18 Quality Assurance in RPBC Activities.......................
F8.1 Followup of Open Items (92701)
.
V. Management Meetings
.
X1 Exit Meeting Summary
.
21
INSPECTION PROCEDURES USED ITEMS OPENED, CLOSED, AND DISCUSSED LIST OF ACRONYMS USED
23
Re ort Details Summa of Plant Status Susquehanna Steam Electric Station (SSES) Unit 1 was at 100% power at the beginning of the inspection period. On April 3, power was reduced to 75% to perform a control rod sequence exchange and control rod speed adjustment.
On April4, power was returned to 100% and remained at 100% for the remainder of the inspection period.
SSES Unit 2 shutdown on March 13, to start the ninth refueling outage.
After completion of the refuel and inspection outage, a reactor startup was performed on April24. On April25, Unit 2 was manually shutdown to repair main steam safety relief valve acoustic monitors, a traversing incore probe instrument tube, and the loose-parts monitors in the drywell. Unit 2 remained shutdown for the remainder of the inspection period.
I. 0 erations
Conduct of Operations
'1.1 UnitO erationsand0 eratorActivities (71707)
The inspectors determined that routine operator activities were planned, communicated, and conservatively performed in accordance with SSES procedures.
Effective controls were implemented for safe plant operation.
Control room logs accurately reflected plant activities. Control room operators displayed good questioning perspectives during work planning meetings and prior to releasing equipment for scheduled work activities.
01.2 Unit 2 Refuel'Outa e Activities (71707)
The control, execution, and performance of major outage activities were good.
Management emphasized to the staff the importance of human performance, attention to detail, and personnel safety throughout the outage.
The radiological control coverage for the entire outage and coordination of the refuel floor equipment and work activities were excellent.
Management oversight and reactor engineering involvement were evident during refueling activities. Management decisions were conservative throughout the outage and contributed to providing a safe and controlled work environment.
Management appropriately stopped fuel handling activities when problems were noted with the fuel handling equipment.
Refueling activities were not resumed until the equipment problems were corrected.
The control and execution of the fuel movements were performed without error by a dedicated group of refueling personnel.
'Topical headings such as Ot, M8, etc., are used in accordance with the NRC standardized reactor inspection report outline.'ndividual reports are not expected to address all outline topic PPLL evaluated plant risk when they developed the refuel outage sequence of activities.
Plant risk was re-evaluated as the sequence of activities changed during the outage.
The plant risk changes were reviewed and approved by plant management, but management did not always effectively communicate the risk changes throughout the organization.
For example, the risk increase associated with the removal of the reactor fuel supplementary decay heat removal system was not initiallyclear or well understood by control room and station personnel.
Operational Status of Facilities and Equipment 02.1 0 erational Safet S stem Ali nment (71707)
During routine plant tours, the proper alignment and operability of various safety systems, engineered safety features, and on-site power sources were verified. A partial walkdown of the following systems was performed:
Unit 2 residual heat removal (RHR) systems Unit 2 RHR shutdown cooling flowpath Supplemental decay heat removal system Unit 2 reactor core isolation cooling (RCIC)
Unit 2 suppression pool, drywell and refuel floor areas Unit 2 main steam system Unit 1 RHR low pressure injection flowpath Overall equipment operability, material condition, and housekeeping conditions were good.
Miscellaneous Operations Issues 08.1 Licensee Event Re ort Review (92700)
Closed LER 50-388/99-001-00 92700 61726 and 62707 Main Steam Isolation Valve (MSIV)Seat Leakage PP8L identified that two of eight main steam isolation valves (MSIVs) did not meet Technical Specification (TS) 3.6,1.3 local leak rate testing (LLRT) acceptance criteria due to valve seat leakage.
The excessive leakage for the "C" outboard and "D" inboard MSIVs was attributed to a low spot on the "C" valve main seat and a void on the "D" valve main poppet seat.
The leakage was detected during the Unit 2 refuel outage and corrected prior to plant startup.
The inspectors reviewed the MSIVleak rate history, TSs, and observed MSIVvalve repairs and re-test activities in the plant. The inspectors'n field review verified that the
.
~ deficiencies were corrected, engineering was evaluating the historical leakage data, and the problem was entered into the corrective action program for additional review. If required, the MSIVwould have closed and the postulated total leakage to the main condenser would have been less than the TS limit. The inspectors determined that
PP&L properly identified and reported this issue, and found PPBL's proposed and completed corrective actions to be good.
In conclusion, in a Licensee Event Report, PP8L identified that two main steam isolation valves did not meet a seat leakage specification.
PP8L's corrective actions, including valve seat repair and re-test activities, were good. This Severity Level IVviolation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy. This violation is documented in PP8L's corrective action program as condition report 92338.
LER 50-388/99-001 is closed.
(NCV 50-388/99-04-01)
II. llaintenance Conduct of Maintenance Surveillance and Pre-Planned Maintenance Activit Review
761726, 6 7 74600),
The inspectors observed and reviewed selected portions of pre-planned maintenance and surveillance activities, to determine whether the activities were conducted in accordance with NRC requirements and SSES procedures.
Observations and Findin s The inspectors observed portions of the following work activities and surveillances:
Work Authorizations A83376 S91 069 V90153 Z81 024 H80177 Y80001 H90113 U96555 V82007 P81762 P82355 S98649 P82544 H80276 V98709 2D660 Unit 2 250 Volt Battery Replacement Unit 2 125 Volt Battery Swap From the "E" Emergency Diesel Generator Unit 2 "B" RHR Heat Exchanger Performance Test Compensatory Action for Inoperable HELB Barrier Unit 2 HCU 22-27 Valve 247115 and Valve 247126 Rework Unit 2 Snubber Test for DBA-201-H6 (RWCU Suction in Drywell)
Unit 2 "C" MSIVHV-241F028C Rebuild Unit 2 Acoustic Monitor Cable Replacement (DCP 97-9112)
Unit 2 RHRSW HV-21210A Rework Unit 2 Drywell-Suppression Pool Vacuum Breaker PMs Unit 2 RWCU Outboard Isolation Valve HV-244F004 Votes Test 1ATS229 Rework (RHR Injection Valve MCC Automatic Transfer Switch)
Local Power Range Monitor (LPRM) Detector Replacement 1&C Activities Unit 2 LPRM Replacement per ME-ORF-102 Unit 2 SRV Acoustic Monitor Repairs and Re-test
Surveillances SR-255-004 SE-024-C04 SO-256-004 SO-231-003 SO-256-006 SI-299-225 SI-299-228 SI-299-229 SE-283-202 SI-280-303 SO-251-A02 SM-059-001 SO-200-007 SE-259-023 SE-200-002 Unit 2 Control Rod Scram Time Testing After Refueling 24 Month Diesel Generator "C" Auto Start on ECCS Actuation Test Signal Rod Sequence Control System (RSCS) Diagnostic Functional Test Rod Worth MinimizerOperability Determination during Decreasing Power RSCS Rod Block Demonstration during Power Reduction Unit 2 Excess Flow Check Valve 24-month Functional Test Unit 2 Excess Flow Check Valve 24-month Functional Test Unit 2 Excess Flow Check Valve 24-month Functional Test Unit 2 SRV Accumulator Check Valve Exercise Test (241F040N)
Reactor Vessel Water Level 24-Month Calibration (LIS-B21-2N031C)
Unit 2 "A"Core Spray Quarterly Flow Verification Test Unit 2 Vacuum Relief Valve Set Pressure Test (PSV-25704xx)
Unit 2 Ops Daily Surveillance, Attachment "C", Daily Fire Door Check Unit 2 "C" MSIVLocal Leak Rate Test (HV-241F028C)
ASME Class I Boundary System Leakage/Hydrostatic Pressure Test In addition, selected portions of procedures, drawings, and vendor technical manuals, associated with the maintenance and surveillance activities, were also reviewed and determined to be acceptable.
In general, maintenance personnel were very knowledgeable of their assigned activities. The station battery maintenance and housekeeping were excellent.
The battery cells and storage racks were spotless with no sign of sulfuric acid buildup.
l&CActivities for Local Power Range Monitor (LPRM) Detector Replacement On March 21, 1999, instrument and control (l&C)technicians were under the Unit 2 reactor vessel preparing ten LPRMs for replacement.
While removing the seal nut on the fifth LPRM, the nutjammed on the lower end ofthe LPRM. In an attempt to loosen the jammed nut, a technician pushed up on the nut and inadvertently lifted the LPRM, which resulted in momentarily unseating the seal between the LPRM and reactor vessel.
About one to two quarts of contaminated water from the reactor vessel sprayed on the individuals in the immediate area.
Allindividuals exited the under vessel area, notified health physics, and the control room. The individuals were subsequently frisked and decontaminated, and then received a whole body count.
The inspectors interviewed the l&Ctechnicians, and reviewed the work packages, procedures, and training used for the LPRM replacement activity. The inspectors concluded that the procedures, work package instructions, and training provided adequate detail to accomplish the activity. The inspectors noted that the on-the-job training was conducted in a classroom using drawings, slides, and a video, but did not include any "hands-on" instruction using the under vessel LPRM mockup trainer. The inspectors noted that the technicians who performed the activity.did.not recognize the
"jammed nut" as an abnormal condition and did not recognize that the force they exerted upward on the nut could cause the LPRM to liftand break the seal
PP8 L initiated condition reports 92527 and 92528 to review this event.
Initial corrective actions included a work stoppage for all under vessel work activities, a through event review, and revisions to procedures and training. The technicians received additional training prior to the restart of under vessel work. The inspectors concluded PP8L's initial response and proposed actions were appropriate and timely. No violations of NRC requirements were identified.
A NRC health physics specialist reviewed the radiological consequences and the PP8L dose assessment for this event.
PP8L performed a detailed and complete evaluation of the dose consequences to the nine individuals who were contaminated.
The maximum individual CEDE was less than one-tenth of a percent of the respective NRC annual occupational dose limit, and the maximum individual skin dose was less than one percent of the respective NRC annual occupational dose limit. The inspector concluded that the dose assessment was detailed and that the additional total radiation exposure (both internal and external) to the individuals, as a result of the contamination, was non-consequential.
No violations of NRC requirements were identified.
Conclusions During a Unit 2 under vessel local power range monitor (LPRM) maintenance activity, technicians momentarily unseated the seal between the LPRM and reactor vessel, spraying about two quarts of contaminated water on themselves.
The additional total radiation exposure (both internal and external) to the individuals, as a result of the contamination, was non-consequential.
PP8L initiated condition reports 92527 and 92528 to review this event.
PPBL's initial response and proposed actions were appropriate.
No violations of NRC requirements were identified.
Imaterial Condition of Facilities and Equipment M2.1'nservice Ins ection ISI Review TT 5533 The inspector reviewed plans and schedules for the current ISI interval (ninth outage, second interval, second period) to verify compliance with the requirements ofAmerican Society of Mechanical Engineers (ASME)Section XI and 10 CFR 50.55a(g).
Specific areas inspected included qualifications and certifications ofthe non destructive examination (NDE) personnel, ISI NDE procedures, invessel visual inspection (IVVI),
ultrasonic examination (UT) of the core shroud and PP8L review and disposition of results.
The inspector selected one weld in each of the residual heat removal (RHR) and reactor recirculation (RR) systems for observation of nondestructive examination.
Observations and Findin s PP8L provides oversight of NDE contractors that perform ISI examinations.
The oversight, which involves review and approval of test procedures, inspector qualifications, test results and surveillance of selected non-destructive testing in
progress, was provided by the ISI and Nuclear Assurance organizations.
The inspector examined surveillance plans, logs, checklists and reports which documented the oversight of inservice inspection activities. The inspectors observed PP&L's oversight of
. contractor NDE activities during the magnetic particle test (MT) and UT of one RHR weld and UT and PT of one RR suction pipe weld.
The work performed was found to be thorough and of sufficient extent to determine the integrity of the components inspected.
The inspector reviewed the ultrasonic, liquid penetrant and magnetic particle test procedures used by NDE personnel and found them to be adequate for the NDE tasks performed.
The inspector found the inspection implementation consistent with the approved procedures. The personnel qualification records for twelve NDE inspectors were examined and found to be in compliance with the ASME code requirements.
Examination data and documentation were reviewed and found to be in accordance with ISI procedures and ASME Code requirements.
NDE personnel performing inspections had properly identified and explored indications to determine ifthey were relevant.
The tracking of ISI examination results indicated that the ISI program was in compliance with the ASME Code,Section XI for the specified period.
The inspector reviewed the results of the invessel visual examination of core spray piping, brackets, fittings, tee boxes, spargers and nozzles.
No indications were identified on these items.
The results of the visual examination of the steam dryer was also reviewed by the inspector.
This examination revealed several new cracks when compared to the previous examination performed. A condition report was initiated to document the new cracks and provide a reference for future examinations.
PPBL performed an evaluation of the new cracks and concluded that the steam dryer could continue to operate for an additional operating cycle.
The automated UT of the circumferential and vertical welds of the reactor core shroud was underway during the period. Allexaminations during this period were conducted from the outside diameter.
The inspector found the examination of the core shroud to be well planned, coordinated and being conducted with appropriate licensee involvement and oversight.
Conclusions The inspector concluded that the inservice inspections (ISI) had been performed acceptably and had included acceptable ASME program coverage, qualified personnel, approved procedures, proper implementation, appropriate examination documentation, and PP8L oversight. The inspections performed were thorough and of sufficient extent to determine the integrity of the components inspected.
Indications of nonconforming conditions were identified, explored, evaluated, documented and dispositioned in accordance with established requirement M8 M8.1 Miscellaneous Maintenance Issues Followu of0 en Items Closed VIO 50-387 388/98-03-04 (62707,92902)
Repair of the "A"and "C" Emergency Diesel Generators In February 1996, PP8L installed repair parts on the "C" EDG that had not received proper quality receipt inspections.
In September 1997, PP8L installed a defective cylinder head on the "A"EDG that had not received a proper quality receipt inspection.
The inspectors performed an in-field review of PP8L's corrective actions, which included revisions to contracts management procedures, procurement receipt inspection procedures, fulltime source inspection at the diesel vendor for quality related vendor work, and a detailed event review with maintenance personnel.
The inspectors concluded the corrective actions, performed or proposed, appeared reasonable to correct the condition and prevent recurrence.
This violation is closed.
III. En ineerin E1 Conduct of Engineering Residual Heat Removal S stem ln'ection Valve Stem Failure Root Cause Determination and Closed LER 50-387/99-001 Ins ection Sco e 37550 The inspectors reviewed licensee documentation of the failed residual heat removal (RHR) low pressure coolant injection (LPCI) valve failure analysis, stress analysis, root cause determination, corrective actions, and American Society of Mechanical Engineers (ASME) Code considerations.
The inspectors also met with PP8L engineering personnel to assess their root cause determination and corrective action implementation.
b.
Observations and Findin s On February 27, 1999, PP&L identiTied that the valve plug on Unit 1 RHR LPCI valve, HV-151F017B, had separated from the valve stem. The separation of the valve plug from the valve stem resulted in the inoperability of the Unit 1 "B" RHR LPCI system.
Subsequently PP8L identified that a similar Unit 2 RHR LPCI valve, HV-251F017A had a crack that extended through 75% ofthe stem.
The respective Unit 1 "A"valve and the Unit 2 "B"valve were examined and determined not to have any cracks.
The valve stem and plug assemblies for the affected valves were'replaced (Inspection Report 50-387,388/99-03, Section 02.2).
PP8L conducted an investigation into the event.
PP8L determined that during modifications to the valves in the mid-1980's, they did not ensure that material requirements for the valve stem were properly specified.
In addition, PP8L determined
that during valve assembly with these modifications, interferences between the valve stem and plug resulted in over stressing the stem.
These combined effects resulted in the valve being susceptible to stress corrosion cracking and resulted in the observed failure.
Root Cause Determination:
The inspectors reviewed Altran (an independent engineering consultant) and PPB L root cause analysis reports. The reports indicated that the valve stem crack began at a high stress area on the filletof the valve stem collar when the plug thread interfered with the stem collar during assembly.
This was confirmed by detailed structural analysis of the crack initiation area which indicated high local preload tensile stress beyond the material yield strength.
Chemical analysis of the stem indicated that the material was SA 479 Type 410 steel with a Rockwell hardness (Rc) of 43 to 44. Certification documents indicated the valve stems had been tempered at 900 'F versus 1100 'F, which resulted in the stems exceeding the maximum hardness of 32 Rc specified for SA 479 Type 410 steel.
The harder stem material is susceptible to stress corrosion cracking. The combined effect of a material susceptible to stress corrosion cracking, significant tensile stress, and an environment that promotes cracking resulted in failure of the Unit 1 RHR injection valve, HV-151F017B and significant cracking on the Unit 2 RHR injection valve, HV-251 F017A.
PPB L determined that the root cause of these failures, was the result of deficient control of the design process, material selection process, and the absence of adequate review and procurement procedures to assure that appropriate design and material selections were made.
In addition, the inspectors determined that PP8L did not provide the necessary oversight of stem and plug material procurement to ensure that the component composition requirements were being met. These deficiencies occurred in 1984, while the injection valves were being modified. Based on independent review of documentation and discussions with engineering personnel, the inspectors concluded that the PP8L root cause analysis was detailed and complete.
10 CFR 50 Appendix B, Criterion III, "Design Control", requires, in part, that measures be established for the selection and review for suitability of application of materials, and parts that are essential to the safety-related functions of the systems and components.
Contrary to this requirement, PP&L failed to provide adequate selection and reviews for suitability for materials used in design change packages (DCP)82-051 and 83-449 for modifications to the RHR system LPCI valves implemented in 1985 and 1986.
Consequently, inappropriate stem material, susceptible to stress corrosion cracking, was installed in these valves, resulting in the failure of HV-151F017B and inoperability of the Unit 1 "B" RHR LPCI system.
The inspectors noted that the valve failure was licensee-identified, through troubleshooting activities by station personnel.
The inspectors also determined that this issue was not likelyto be identified by surveillances or normal quality assurance activities.
In addition, the inspectors concluded that station personnel took reasonable and effective corrective actions.
In accordance with the NRC Enforcement Policy,
Section VII.B.3, 'Violations Involving Old Design Issues", the NRC exercised enforcement discretion and did not cite this violation. This violation is documented in PP&L's corrective action program as condition reports 90971 and 90981.
(NCV 50-387/388/99-04-02)
Corrective Actions:
PP&L modified the valve internal design with materials having high strength and moderate hardness levels to reduce the potential for stress corrosion cracking. Also, PP&L changed the stem and plug design to eliminate interference between the stem and plug during assembly.
The Unit 2 "A"and "B"valves were replaced with new parts from the modified design during the refueling outage in April. Both Unit 1 valves are scheduled for replacement with the modified design at the next opportunity, in accordance with the CR 90971 evaluation and action plan. As an interim corrective action, the Unit 1 "B"valve was replaced with new parts of the original design.
Engineering personnel performed an operability assessment and concluded that this condition is acceptable, based on the lack of existing cracks on the valve stems and the very slow nature of crack propagation by environmentally-assisted stress corrosion.
The inspectors determined that the PP&L assessment provided an acceptable basis for continued operation for the remainder of the cycle.
Licensee Event Re ort Review The inspectors performed an onsite review of Licensee Event Report (LER) 50-387/99-001. No new issues were identified by the LER. This LER is closed.
Conclusions Engineering personnel performed a comprehensive failure analysis and a thorough root cause determination of the Unit 1 and 2 low pressure coolant injection valve stem failures. Corrective actions, including the replacement of internal parts with improved designs and materials, were acceptable.
Design control deficiencies in the mid-1980's led to the use of material in the stem of the low pressure coolant injection valves that was susceptible to stress corrosion cracking.
These deficiencies constituted a violation of 10 CFR 50 Appendix B, Criterion III, "Design Control," In accordance with the NRC Enforcement Policy,Section VII.B.3, Violations Involving Old Design Issues, the NRC exercised enforcement discretion and did not cite this violatio Engineering Support of Facilities and Equipment Station Res onse Related to the Failure ofthe Unit 1 Residual Heat Removal In'ection Valve HV-151F017B Ins ection Sco e
71707 37550 37551 NRC inspection report 50-387,388/99-03, Section 02.2, discussed a failure of the stem of residual heat removal (RHR) low pressure coolant injection (LPCI) valve, HV-151F017B that was discovered on February 27, 1999. The inspectors performed a
'ollow-up review ofthe adequacy and timeliness of the station's response to the first indication of a potential problem, which was an abnormal pressurization of the "B" loop of RHR during a filland vent evolution on February 11, 1999. The inspectors reviewed the PP8L root cause analysis documented on Condition Report (CR) 90981, interviewed several station personnel, and reviewed station documents and procedures.
Observations and Findin s Event Summa On February 11, 1999, during restoration of the Unit 1 "B" loop of RHR following planned maintenance, operators observed slow pressurization of the system.
The PP8L staff considered that the slow pressurization was caused by a flow restriction in a portion of the condensate transfer keep-fill system and documented this on CR 90981. Operations staff concluded that the RHR system was degraded but operable because keep-fill pressure returned to its normal value and remained there.
i Maintenance, operations, and system engineering personnel performed troubleshooting under a work authorization (WA) on February 12 and 13. This troubleshooting did not find a flow restriction in the suspected portion ofthe condensate transfer system and the WAwas closed.
Engineering personnel briefly discussed the possibility of a problem with the LPCI valve, but they considered it as highly unlikely, and still believed the slow pressurization was related to a problem in the condensate transfer system.
Since there was no definitive answer to the slow pressurization, CR 90981 remained open with additional follow-up actions to be pursued at a later date.
The operations staff again concluded that the RHR system was degraded but operable.
On February 16, 1999, station personnel removed the Unit 1 "A"loop of RHR from service for about 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> for planned maintenance.
When the system was refilled and vented after the planned maintenance the station staff observed normal pressurization of the "A"loop of RHR.
On February 17, 1999, the RHR system engineer and system engineering supervision determined that the follow-up actions for CR 90981 needed to be pursued with a higher priority. They were concerned about the possibility of further degradation in the condensate transfer system.
Engineering initiated a new Priority 2 WA, developed a
troubleshooting plan, and arranged for the scheduling of additional troubleshooting for the following week.
On February 26, 1999, maintenance personnel used the troubleshooting plan prepared by the RHR system engineer to evaluate the problem. The results indicated that the RHR HV-151F017B valve was closed.
On February 27, 1999, maintenance personnel disassembled the valve and found the stem severed near the valve plug, which was in the closed position. The valve stem and plug were replaced the following day and within the seven day Technical Specifications allowed outage time.
Licensee Anal sis PP&L management assembled a multi-disciplinary team and performed a root cause analysis to review the station's response to the event. The team concluded that the investigation of the keep-fill anomaly on February 12 and 13, 1999, was not adequate to discover the problem with the RHR LPCI valve. Three causal factors were identified:
No comprehensive written investigation plan was developed and used.
No programmatic requirement existed to define the requirements of a written investigation plan and to detail the use of a written investigation plan.
Maintenance and system engineering were not fullyengaged with operations on February 13 when the results of a keep-fill system check valve inspection did not reveal a clear cause for the keep-fill anomaly, and the problem still existed.
PP&L's primary corrective action for this event was to develop and clearly communicate the roles, responsibilities and process for investigation and troubleshooting plant problems.
NRC Review Based on interviews and independent review, the inspectors determined that the root cause analysis team identified the principal causes for PP&L's inability to identify the F017B valve failure on February 12 and 13. The PP&L root cause analysis team identified that PP&L needed a structured, methodical troubleshooting process, including a written plan and documentation of troubleshooting progress.
This finding was consistent with the inspectors'indings.
The two positive actions were important to the ultimate identification of the RHR valve failure. A nuclear plant operator and a plant control operator, working together, identified, based on their experience, an anomalous RHR system pressure response following maintenance.
The RHR system engineer and engineering supervision kept the station focused on the perceived condensate transfer system problem after the first work authorization was closed and until a definitive solution was obtaine The inspectors determined that station personnel took reasonable action in response to the slow RHR pressurization condition, given the information available at the time and the lack of troubleshooting guidance.
Although the root cause analysis report addressed problems with roles and responsibilities during troubleshooting, the inspectors found there was a change management issue with a re-defined role of the system engineer.
Based on interviews, the inspectors determined that station personnel had not fully recognized that systems engineers were expected to oversee/monitor maintenance and troubleshooting, rather than be directly involved, as was done previously. This was evident during troubleshooting, because operations and maintenance personnel were expecting more direction from the system engineer.
The inspectors found that the second work authorization (WA)to investigate the perceived condensate transfer system problem, which was initiated as a Priority 2 item on February 17, was not scheduled consistent with the existing guidance in the Nuclear Department Administrative Procedures (NDAPs). Procedure NDAP-QA-502, 'Work Authorization System," states that a Priority 2 WA should be scheduled for completion within one week (7 Days) of the "identified date." The WAwas scheduled outside of this one-week window. Had the WA been scheduled consistent with NDAP-QA-502, the detailed troubleshooting that identified the RHR valve failure would have been performed a few days sooner.
The inspectors determined that this scheduling problem was a minor issue not subject to formal enforcement action.
In addition to the PP8L root cause analysis, the independent safety engineering group (ISEG) reviewed the event and documented its findings in ISEG report 2-99. While the report supported the station's findings, it provided limited critical, independent assessment.
The inspectors determined that the ISEG review missed opportunities to reveal additional insights related to the troubleshooting efforts. Further, the report discussed but did not recommend implementing a troubleshooting plan for extended troubleshooting efforts nor did the report recommend any corrective actions for the missed 10 CFR 50.72 report.
ISEG personnel stated that the review was constrained somewhat by other ISEG commitments.
Re ortabilit Determination PP8L's initial event reportability determination did not recognize that the "B" RHR injection valve failure resulted in a condition that was outside the design basis and required the NRC to be notified within one hour. On February 27, 1999, at approximately 4:30 p.m., PP8L determined that the Unit 1 RHR injection valve disk had separated from the valve stem and had been in this configuration since February 11, 1999.
During these sixteen days, the "B" loop of RHR was not capable of performing its design function of low pressure coolant injection. This condition placed the plant in a condition outside the design basis because suitable redundancy was not available for an extended period of time during plant operation.
On February 28, 1999, the NRC identified that PP8L failed to make the required notification. PP8L performed the notification on February 28, 1999 at 10:54' On February 27, 1999, PP8 L did not notify the NRC, within one hour of identification, of a condition of Unit 1RHR loop "B"that was outside the design basis.
The failure to make the notification within one hour is a violation of 10 CFR 50.72(b)(1)(ii)(B). This severity Level IVviolation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy. This violation is documented in PP8L's corrective action program as condition report 90981. (NCV 50-387/99-04-03)
Additionally, the inspector identified on February 28, 1999, that PP8L had not evaluated the condition of Unit 1 RHR loop "A" in the same 16 day period to determined if a complete loss of the low pressure coolant injection design function had occurred.
Subsequently PP8L determined that on February 16, 1999, the "A"loop of RHR was removed from service for 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> to perform maintenance.
During this 17 hour1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> period, both loops of RHR were not capable of performing their design function of low pressure coolant injection.
PP8L included this information in the report made to the NRC on February 28, 1999.
The PP8L team also noted that reporting of the event to the NRC was not handled well.
They found that NRC prompting was necessary for the station to recognize the complete loss of the Unit 1 LPCI function of RHR due to the F017B valve failure while the "A" loop of RHR LPCI was out of service for maintenance.
The PP8L planned corrective actions for this issue were to revise reportability administrative procedures and provide focused training.
Conclusions PP8L identified and resolved a potential common cause failure of all RHR injection valves prior to any need for the system to function. The operator's initial identiTication of the "1B" residual heat removal system slow pressurization reflected a good questioning attitude. The system engineer appropriately focused station priorities to complete troubleshooting and identify a failed low pressure coolant injection valve.
PP8L did not troubleshoot the unexpected slow pressurization of the residual heat removal system in a structured and methodical manner, which extended the time needed to identify the failed "1B" low pressure coolant injection valve. While troubleshooting was in progress and unaware that the "1B" valve was failed, PP8L removed the "A"residual heat removal system from service for maintenance for 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />.
During this 17 hour1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> period, neither Unit 1 loop of residual heat removal was available for low pressure coolant injection.
Although PP&L identified problems with roles and responsibilities during troubleshooting, they did not identify a change management issue.
SpeciTically, the station staff was not fullyaware that the system engineering interface with operations and maintenance during troubleshooting had been changed.
The PP&L independent safety engineering group review of the slow pressurization of the residual heat removal system event was limited and missed opportunities to reveal additional insights related to the station staff response to the failed valv J'
On February 27, 1999, PP&L did not notify the NRC, within one hour of identification, that the Unit 1 RHR loop B was in a condition that was outside the design basis.
The failure to make the notification within one hour is a violation of 10 CFR 50.72. This Severity Level IVviolation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy. This violation is documented in PP&L's corrective action program as part of condition report 90981.
ES Miscellaneous Engineering Issues E8.1 Followu of0 en Items Closed VIO 50-387 388/98-1 2-02 (37551,92903)
"A"Emergency Diesel Generator Inoperable due to Heavy Rains In June 1998, heavy rains resulted in approximately 300 gallons of water entering the "A" emergency diesel generator (EDG) fuel oil storage tank. PP&L determined that the use of an incorrect gasket material was a root cause for the significant water intrusion into the fuel oil storage tank. The NRC determined that PP&L failed to adequately translate a 1989 system design change into appropriate specifications, drawings, and procedures, and on two separate occasions substituted gasket material without a review for suitability of materials.
The inspectors performed an in-field review of PP&L's corrective actions, which included field inspections of storage tank flange gaskets and replacement of incorrect gaskets, revisions to maintenance work instructions, design drawing changes, a review of the foreign material exclusion program, and revisions to the maintenance training and task certification program.
The inspectors concluded the corrective actions, performed or proposed, appeared reasonable to correct the condition and prevent recurrence.
This violation is closed.
Closed VIO 50-387 388/98-05-01 (92701)
Three Examples of a Failure to Translate Regulatory Requirements and Design Basis into=Test Acceptance Criteria or Accident Analysis The Core Spray system had discrepancies between the surveillance test acceptance criteria and the system design basis which resulted in non-conservative test results, and an incorrect assumption was used in an accident analysis calculation.
This Severity Level IVviolation was issued in a Notice of Violation prior to the March 11, 1999, implementation ofthe NRC's new policy for treatment of Severity Level IV violations (Appendix C of the Enforcement Policy). Because this violation would have been treated as a Non-Cited Violation, in accordance with Appendix C, this violation is being closed out in this report. This violation is in PP&L's corrective action program as Condition Reports 98-1197 (NIMS 69648) and 98-3070 (NIMS 71492). This violation is close IV. Plant Su ort Radiological Protection and Chemistry (RP&C) Controls R1.1 Radiolo ical Controls-External and Internal Ex osure Ins ection Sco e 83750-02 The inspector evaluated the effectiveness of selected aspects of the applied radiological controls program. The evaluation included a review of the adequacy and implementation of radiological controls program elements and outage activities.
The inspector observed Unit 2 refueling and inspection outage activities, toured the radiologically controlled area (RCA), interviewed stations staff, and reviewed applicable station procedures.
Observations and Findin s PP8L implemented effective access controls to the radiologically controlled areas of the station including use of RWPs, bar code readers, and computerized log-in stations.
No access control deficiencies were identified. Appropriate personnel monitoring devices for access to the RCA were supplied and used.
TLDs and personnel alarming electronic dosimeters were observed to be properly worn to measure external dose.
Access controls for high radiation areas (HRAs) were effective. Radiological postings and labels throughout the toured areas provided additional administrative controls and information to the worker. Survey maps with radiological data were kept by the health physics (HP)
technicians at the main HP control point and were used for RWP briefings.
PPBL maintained personnel occupational radiation exposures (external and internal)
within applicable regulatory limits and as low as reasonably achievable (ALARA). A review of personnel exposure data for 1998 and for 1999 (year-to-date) identified that individual exposure results for total effective dose equivalent (TEDE), lens of the eye dose equivalent (LDE), shallow-dose equivalent (SDE), and extremity dose equivalent were well below regulatory requirements.
Further, the maximum individual committed effective dose equivalent (CEDE) for any one individual was well within applicable NRC limits in Title 10, Part 20.1201 of the Code of Federal Regulations (10 CFR 20.1201).
Conclusions PP8L implemented effective applied radiological controls. Access controls to radiologically controlled areas were effective, and appropriate occupational exposure monitoring devices were provided and used.
Personnel occupational exposure was maintained within applicable regulatory limits and as low as reasonably achievable
. (ALARA). The radiation work permit program was properly implemented...
Radiolo ical Controls-Radioactive Materials Contamination Surve s and Monitorin Ins ection Sco e 83750-02 The inspector evaluated the effectiveness of PPBL's surveys, monitoring and control of radioactive materials and contamination.
The evaluation included a selective review of the adequacy and effectiveness of the following radioactive material and contamination control program elements:
surveys and monitoring of radioactive material and contamination the calibration status of survey and monitoring equipment the proper use of personal contamination monitors and friskers the tracking of personnel contamination events and goals The inspector evaluated performance in the above selected areas via observation of work activities, tours of the RCA, discussions with cognizant personnel, review of historical documentation, and review and evaluation of applicable station procedures.
Observations and Findin s PPB L implemented an effective radioactive material and contamination control program.
Continuous air monitors were in use in the RCA. Hand-held contamination monitors (friskers) and radiation survey meters exhibited current calibration stickers and were appropriately used by personnel.
Personnel were properly frisking at the RCA exit using whole body contamination monitors.
HP technicians were observed in the field providing job-coverage radiological surveys and instruction in a capable and effective manner.
Survey records contained appropriate radiological information. Radiological housekeeping conditions in the Unit 2 reactor building and the turbine buildings were good.
In particular, the equipment and personnel work activity control and coordination for the Unit 2 refuel floor, suppression pool, and drywell were excellent.
Radioactive material and radioactive waste were clearly labeled, segregated, and stored in an orderly manner.
Receptacles used for anti-contamination clothing, radiologically contaminated trash, and radiologically clean trash were available in the RCA and were clearly labeled.
Goals to assist in monitoring and tracking personnel and area contamination rates and percent recoverable contaminated area continued to be maintained and used to gauge the overall effectiveness of the station's programs in this area.
For 1998, the personnel contamination report (PCR) rate goal was 2.3 PCRs/10,000 RWP-hours, and the actual was 1.08; the area contamination report (ACR) rate goal was 0.8 ACRs/10,000 RWP-hours, and the actual was 0.71; and the recoverable contaminated area goal was 4.1%
and the actual also was 4.1%. Atthe start of 1998, the actual recoverable contaminated area was 6%. For 1999, the personnel contamination report (PCR) rate goal was 1.5 PCRs/10,000 RWP-hours; the area contamination report (ACR) rate goal was 0.75 ACRs/10,000 RWP-hours; and the recoverable contaminated area goal was 2.1%.
Thus, these goals for 1999 were more challenging than the ones for 199 c.
Conclusions Radiological housekeeping conditions were noted to be good.
In particular, the equipment and personnel work activity control and coordination for the Unit 2 refuel floor, suppression pool, and drywell were excellent.
The number and type of personnel contaminations were tracked, trended, and evaluated for cause and corrective actions.
I R1.3 Radiolo ical Controls-As Low As Reasonabl Achievable ALARA a.
Ins ection Sco e 83750-02 The inspector evaluated the effectiveness of PPB L's program to maintain occupational radiation exposure as low as is reasonably achievable.
The evaluation included a selective review of the adequacy and effectiveness of the following ALARAprogram elements/documents:
ALARAPre-Job Reviews:
RWP No. 19992115 Diving activities: suppression pool clean out/inspection and strainer installation and removal, RWP No.
19992120 DCP 97-3026:
Replace HV24107A/B FW check valve seats, RWP No. 19992353 CRD exchange:
under vessel work, and RWP No. 19992362 LPRM replacement and under vessel support February HP Monthly Report to HP Department 1998 End of Year Summary of Susquehanna Steam Electric Station (SSES)
Radiological Performance Monthly Health Physics Report to G.J. Kuczynski, March 12, 1999 1999 Person-Rem Goal Breakdown Unit 2 (U2) ninth Refueling and Inspection Outage (9RIO) Dose Tracking Data 1999 Annual Dose Tracking Data The inspector evaluated performance in the above selected areas via observation of work activities, tours of the RCA, discussions with cognizant personnel, review of historical documentation, and review and evaluation of applicable station procedures.
b.
Observations and Findin s Management involvement in ALARAwas evidenced by the Monthly HP Report to management, the Site ALARACommittee's involvement with an Exposure Reduction Plan, and senior management sponsorship of an ALARAEnhancement Plan designed to regain first quartile BWR collective dose performance.
The selected ALARApre-job reviews, for outage activities, which were examined, were detailed and comprehensive and incorporated lessons learned from previous evolutions.
The person-rem goal for 1998 for the site was 440 and the. actual result was approximately 333. The year of 1998 included a refueling and inspection outage for Unit 1 (U1 10RIO). There was also a goal of 2 rem for the maximum individual dose, and the actual maximum individual dose was 1.49 rem. The goal and the actual achieved
person-rem for the outage were 242 and 205, respectively.
This data was based on electronic dosimetry.
Annual and outage goals for person-rem in 1999 were based on achieving improved performance in the three-year-rolling average person-rem and were 280 and 140, respectively.
Through February of 1999, the actual collective dose was 43.5 versus 24.2 projected person-rem.
Approximately, 20 person-rem was received due to unplanned corrective maintenance (fuel pool cooling heat exchanger maintenance, 1B reactor water clean-up (RWCU) pump and inlet valve repair, and rework of RWCU non-regenerative heat exchanger) vs. a yearly goal of 6.0 person-rem for corrective maintenance.
Through April 1, 1999, the actual collective dose for the Unit 2 9RIO which started March 13, 1999 was 122.732 versus 102.2 projected person-rem.
The collective dose goals for 1999 were aggressive and challenging.
Conclusions PP8L implemented an effective program to maintain occupational radiation exposure as low as is reasonably achievable (ALARA),and the ALARAefforts and results for 1998 were good, including the management of radiologically significant outage work. The annual and Unit 2 refueling and inspection outage collective dose goals for 1999 were aggressive and challenging.
R5 Staff Training and Qualification in RP8C Activities Ins ection Sco e 83750-02 A selective review of PP&L's selection, training, and qualification program for the contracted radiological control technicians hired for the current outage was performed.
Information was gathered through discussions with cognizant personnel and review and evaluation of procedures and documents.
b.
Observations and Findin s The selection and qualification of contracted radiological control technicians was proceduralized, conducted, administered, and documented in a detailed and thorough manner.
The resumes of technicians performing radiologically risk significant work were reviewed.
Personnel selection was conducted in accordance with the technical specification requirement for two years appropriate experience and with PP8 L's procedure which provided instructions for evaluating experience.
PP8L was conservative in their evaluation of past experience to meet the technical specification experience requirement and their internal requirement of three years of experience.
Junior and senior contracted radiological control technicians were required to be qualified in a set of licensee practices and procedures, and senior technicians were
- required to be qualified in an additional set of procedures..ln addition to using a detailed procedure for selection and training, PP8L administered and documented these activities in a detailed and thorough manne i c.
Conclusions
The program for selection, training, and qualification of contracted radiological control technicians was conducted effectively. The program was in accordance with technical specifications and was proceduralized, administered, and documented in a detailed and thorough manner.
R7 Quality Assurance in RP8C Activities a.
Ins ection Sco e 83750-02 The inspector evaluated the effectiveness of PP&L's self-identification and corrective action processes.
The evaluation include a selective review of the adequacy and effectiveness of the following program elements and documents:
Plant Walkthrough-September 1998 QSR ¹98-107 Third Quarter 1998 Egress Point Monitoring QSR ¹98-122 Fourth Quarter 1998 Egress Point Monitoring QSR ¹98-125 Unit 1 Outboard Main Steam Isolation Valve Investigation (HV141F028B)
QSR ¹99-002 New Fuel Inspection NAS Weekly Outage Surveys of radiation workers's knowledge of radiological practices and conditions Fourth Quarter 1998 HP Self-Assessment Report The inspector evaluated the performance in the above area via observation ofwork activities, tours of the RCA, discussions with cognizant personnel, review of applicable documentation, and review and evaluation of applicable station procedures.
Several surveillance reports, conducted by the onsite assessment organization since the previous radiation protection inspection, were reviewed.
b.
Observations and Findin s A memorandum, documenting a corporate surveillance of a low level radioactive waste filterreplacement evolution, resulted in five recommendations.
Surveillance reports exhibited a good level of detail. Two findings (one involved a deficiency in the procedure for personal item self-monitoring and one involving housekeeping conditions at the RCA egress points) were identified by these reports.
One of these deficiencies met PPB L's threshold for entry as a Condition Report and was properly entered into the corrective action program. The onsite assessment organization was continuing to perform weekly outage surveys of the radiation workers's knowledge of radiological practices and conditions. The percentages of acceptable responses to questions in seven topical areas for the current outage (U2 9RIO) were compared to those from the previous two outages (U1 10RIO and U2 8RIO), and the comparisons were documented.
The quarterly self-assessment by the radiation protection organization concentrated on radiological controls in the field (in the areas of RCA entry and exit controls, radiation
postings and access controls, routine HP activities, contamination controls, and high radiation area controls). Assistant HP foremen completed 112 proceduralized observation forms, and these were effective in identifying deficiencies.
The identified deficiencies were evaluated and used to determine the areas requiring increased assessment focus in the future.
HP-related Condition Reports were reviewed and found to include low threshold items.
The Monthly HP Report dated March 12, 1999, included tracking and trending for eighteen HP-related 1999 Condition Reports.
These Condition Reports were separated into 15 categories and were evaluated for cause within each category.
This review demonstrated that HP issues were being identified, elevated to an adequate management level, adequately evaluated, and responded to with timely and reasonable corrective actions which were effective.
Conclusions PP&L's self-.identification and corrective action processes in the area of radiation protection were effective. Nuclear Assessment Services surveillance reports, HP self-assessments, and the corrective action program continued to be effective in identifying deficiencies and improvement opportunities.
Effective corrective actions were implemented for findings.
FS F8.1 Miscellaneous Fire Protection Issues Followu of 0 en Items (92701)
Closed EA 98-999-01014
[reference:
EEI 50-387,388/98-09-04]
Failure to Meet the Safe Shutdown Performance Goal of Maintaining Reactor Vessel Water Level Above the Top ofthe Active Fuel The SSES safe shutdown methodology failed to meet the performance goals specified in 10 CFR 50, Appendix R.
This Severity Level IVviolation was issued in a Notice of Violation prior to the March 11, 1999, implementation ofthe NRC's new policy for treatment of Severity Level IV violations (Appendix C of the Enforcement Policy). Because this violation would have been treated as a Non-Cited Violation, in accordance with Appendix C, this violation is being closed out in this report. This violation is documented in PPBL's corrective action program as Condition Report 97-3542 (NIMS 75633), and Regulatory Action Items PIMS-R06421, PIMS-R06422, and PIMS-R06423. This violation is close V. Mana ement Meetin s X1 Exit Meeting Summary The inspectors presented the inspection results to members of PP8L management at the conclusion of the inspection period, on May 14, 1999.
PP8 L acknowledged the findings presented.
The inspectors asked PPBL whether any materials examined during the inspection should be considered proprietary.
No proprietary information was identifie INSPECTION PROCEDURES USED IP 37551 IP 40500 IP 61726 IP 62707 IP 71707 IP 71750 IP 83750 IP 92700 IP 92901 IP 92902 IP 92903 IP 92904 IP 93702 Onsite Engineering Observations Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems Surveillance Observations Maintenance Observations Plant Operations Plant Support Activities Occupational Radiation Exposure On Site Foliowup of Reports Followup Plant Operations Followup Maintenance Followup Engineering Followup Plant Support Prompt Onsite Response to Events at Operating Power Reactors
~Oened None.
ITEMS OPENED, CLOSED, AND DISCUSSED 0 ened/Closed 50-388/99-04-01 50-387,388/99-04-02 50-387/99-04-03
~udeted None.
NCV Main Steam Isolation Valve (MSIV) Seat Leakage (section 08.1)
NCV Residual Heat Removal System Injection Control Valve Stem Failure - Old Design Issue (section E1.1)
NCV Failure to make a one hour notification for the Unit 1 Residual Heat Removal Injection Control Valve failure (section E1.1)
Closed 50-388/99-001-00 50-387,388/98-03-04 50-387/99-001-00 50-387/99-001-01 50-387,388/98-12-02 50-387,388/98-05-01 LER Main Steam Isolation Valve Seat Leakage (section 08.1)
VIO Repair of the "A"and "C" Emergency Diesel Generators (section M8.1)
LER Residual Heat Removal Low Pressure Coolant Injection Valve Failure (section E1.1)
LER Residual Heat Removal Low Pressure Coolant Injection Valve Failure (section E1.1)
"A"Emergency Diesel Generator Inoperable due to Heavy Rains (section E8.1)
VIO Three Examples of a Failure to Translate Regulatory Requirements and Design Basis into Test Acceptance Criteria or Accident Analysis (section E8.1)
EA 98-999-010'I4 VIO Failure to Meet the Safe Shutdown Performance Goal of (EEI 50-387,388/98-09-04)
Maintaining Reactor Vessel Water Level Above the Top of the Active Fuel (section F8.1)
LIST OF ACRONYMS USED ASME CFR CR DBD DCP EASCC ECCS FSAR HELB HPCI l8C IR LCO LER LLRT LPCI NCV NDAP NRC OD R,
" RCIC RHR RHRSW RWCU SSES TRM TS UHS URI UT ACR ALARA CEDE CFR CRD HP HRA LDE LPRM NVLAP PCR RCA RIO RPSC American Society of Mechanical Engineers Code of Federal Regulations Condition Report Design Basis Document Design Change Packages Environmentally Assisted Stress Corrosion Cracking Emergency Core Cooling System Final Safety Analysis Report High Energy Line Break High Pressure Coolant Injection Instrument and Controls
[NRC] Inspection Report Limiting Condition for Operation Licensee Event Report Local Leak Rate Testing Low Pressure Coolant injection Non-Cited Violation Nuclear Department Administrative Procedure Nuclear Regulatory Commission Operability Determination Rockwell "C" Reactor Core Isolation Cooling Residual Heat Removal Residual Heat Removal Service Water Reactor Water Clean Up Susquehanna Steam Electric Station Technical Requirements Manual Technical Specification Ultimate Heat Sink
[NRC] Unresolved Item Ultrasonic Test Area Contamination Report As Low As is Reasonably Achievable Committed Effective Dose Equivalent Code of Federal Regulations Control Rod Drive Health Physics High Radiation Area Lens of the eye Dose Equivalent Low Power Range Monitor National Voluntary Laboratory Accreditation Program Personnel Contamination Report Radiologically Controlled Area Refueling and Inspection Outage Radiological Protection and Chemistry
WBC Reactor Water Clean-Up Radiation Work Permit Shallow Dose Equivalent Total Effective Dose Equivalent Thermoluminescent Dosimeter Unit Whole Body Count