IR 05000395/1998005
| ML20236W750 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 07/27/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20236W740 | List: |
| References | |
| 50-395-98-05, 50-395-98-5, NUDOCS 9808060182 | |
| Download: ML20236W750 (44) | |
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U. S, NUCLEAR REGULATORY. COMMISSION
REGION II
Docket No.:
50-395 License No.:
NPF-12 Report No.:
50-395/98-05 Licensee:
South Carolina Electric & Gas (SCE&G)
Facility:
V. C. Summer Nuclear Station Location:
P. O. Box 88 Jenkinsville, SC 29065 Dates:
May 17 - June 27, 1998 Inspectors:
B. Bonser Senior Resident Inspector M. King Resident Inspector (In-Training)
C. Smith, Reactor Inspector, RII (Section El.1)
D. Jones, Reactor Inspector, RII (Sections R1.2, R1.3, R1.4.
R8.1 and R8.2)
W. Bearden. Reactor Inspector, RII (Sections M1.3 and M3.1)
R. Musser, Senior Resident Inspector-Surry (Section 07.1)
L. Garner, Project Engineer, RII (Sections 07 2, 07.3, 07.4, 07.5, and 07.6)
R. Chou, Reactor Inspector, RII (Section 07,7)
H. Whitener, Reactor Inspector. RII (Sections M8.3 and M8.4)
Approved by:
R. C. Haag, Chief, Reactor Projects Branch 5
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Division of Reactor Projects l
l-Enclosure 980006o182 980727
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.[DR ADOCK 05000395 PDR
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EXECUTIVE SUMMARY V. C. Summer Nuclear Station NRC Inspection Report No. 50-395/98-05 This integrated inspection included aspects of licensee operations, maintenance, engineering, and plant support. The report covers a six-week period of resident ins)ection: in addition it includes the results of announced inspections Jy six regional inspectors and a senior resident inspector from the Surry facility.
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Ooerations
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Licensee actions in response to a control room alarm on liquid radiation
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monitor. RM-L1, Primary Coolant Letdown Monitor, were adequate to ensure that the-failed fuel detection function of the monitor was being met.
Redundant means for performing RM-L1's function of detecting a Chemical and Volume Control System line break outside containment were available (Section 01.2).
A detailed system walkdown of the Emergency Diesel Fuel Oil Transfer
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System, the S)ent Fuel Pool Cooling System. the Control Room Ventilation System, and t1e Reactor Make-up (Emergency Boration) System found that these systems were properly aligned per system drawings and procedures.
L Material condition of components and housekeeping were good, and component labeling was adequate (Section 02.1).
A licensed operator simulator requalification exam scenario observed was
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challenging and well conducted. The examination criticue was thorough and provided a comprehensive assessment of crew and incividual competencies (Section 05.1).
An inspection follow-up item was issued to track improvements needed in
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the Condition Evaluation Re)orting process.
Licensee personnel demonstrated ownership of t1e corrective action program (Section 07.1).
Quality Assurance Audits adequately examined and evaluated the
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corrective action program (Section 07.1).
i A review of Shift Supervisor and Shift Technical Adviser logs revealed e
L that Condition Evaluation Reports (CERS) and Nonconformance Notices (NCNs) were being written as required.
Closed CERs and NCNs indicated i
items were being satisfactorily addressed (Section 07.2).
Root cause analyses were thorough and technically valid. Corrective
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actions adequately addressed equipment, procedure and program problems, as well as. human performance issues.
Corrective actions were tracked until completed (Section 07.3).
Engineering evaluations performed per the Engineering Services (ES)-508
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process to evaluate equipment problems were technically adequate and typically focused on hardware and procedure corrective actions to preclude recurrence. Human performance issues addressed by ES-508 evaluations were informally dispositioned (Section 07.3).
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Trending reports of condition evaluation r, sorts provided general
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station trends to senior site management; l.owever, the reports were of limited value to department level or lower managers to aid in identifying specific areas in which to focus management attention (Section 07.4).
Quality Assurance (0A) audit program procedure OSP-106 met applicable
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requirements in 10 CFR 50. Appendix B. Criterion XVIII.
The procedure i
provided sufficient instructions to perform audits which would independently assess plant programs and provide meaningful feedback to
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plant management.
A sampling of five OA audits in the operations and engineering areas revealed that the OA audits were performed in accordance with the procedures.
Audit firding closure documentation was readily available and justified that audit findings had been satisfactorily addressed (Section 07.5).
Self-assessments activities in operations were generally comprised of
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enhanced routine management oversight.
The self-assessments provided
meaningful feedback to Operations personnel (Section 07.6).
Management's commitment to improving operations was demonstrated by
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efforts to improve operational practices, procedures and programs j
through bench marking visits by operations personnel to other power J
plant facilities (Section 07.6).
The procedures used for the Operating Experience (OE) program were
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adequate in providing details for processing OE information.
The OE program successfully processed, evaluated and dispositioned OE information into station programs and procedures (Section 07.7).
A review of corrective action for three equipment storage violations
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concluded that there had been significant improvement in the storage and control of material in the plant (Section 08.1).
Maintenance Observed maintenance activities on a centrifugal charging pump, an
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emergency feedwater flow transmitter, the service water intake structure, and a component cooling water pump were conducted using the appropriate procedures, tools, and techniques. The maintenance technicians were knowledgeable and demonstrated good work practices (Section M1.1).
Observations of centrifugal charging pump, reactor core flux mapping,
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and solid state protection system surveillance testing revealed good communications, self-checking, and procedural adherence (Section M1.2).
Completed emergency feedwater surveillance tests and instrument loop
calibration packages for reactor protection setpoints demor.'strated acceptable test results (Section M1.3).
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The corrective maintenance following an A Diesel Generator (DG) trip on
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high crankcase pressure was performed such that the potential causes of the trip were identified and corrected.
Strong system engineering support to maintenance was observed during the repair effort.
Clear management involvement was exhibited in the maintenance review before declaring the A DG operable (Section M1.4).
The licensee's periodic Maintenance Rule assessment report provided
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sufficient detail to demonstrate that the licensee had adequately evaluated performance condition monitoring, associated goals and preventive maintenance activities for Structures. Systems, and Components within the scope of the Maintenance Rule.
The licensee's assessment met the requirements of NUMARC 93-01 and paragraph (a)(3) of 10 CFR 50.65 and is a strength that resulted in significant improvements in the Maintenance Rule Program (Section M3.1).
A non-cited violation was identified for failure to perform Technical
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Specification required channel checks on Post Accident Monitoring System neutron flux intermediate range channels N1-35 and NI-36 (Section M3.2).
A licensee maintenance self-assessment was thorough and provided a
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critical evaluation of the maintenance program.
The findings were consistent with NRC assessments of licensee performance (Section M7.1).
A non-cited violation of Technical Specification 4.0.5 was identified
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for failure to adequately implement Inservice Test (IST) requirements for the testing of pressure relief devices in the second ten year IST interval which started January 1. 1994 (Section M8.3).
Enaineerina A positive finding was identified concerning im] lamentation of the 10
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CFR 50.59 Program.
Personnel assigned responsi)ility for implementing the program were trained and procedural controls provided clear guidance for determination of an Unreviewed Safety Question.
10 CFR 50.59 screenings / evaluations performed for design changes were determined to be technically adequate (Section E).1).
An observation of the Principal Engineering Review Group concluded that
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the review served as a useful check of design input requirements (ANSI N45.2.11) that have the potential to impact the design basis of the plant (Section E1.2).
A review of service water building fan 480 volt breaker failures
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indicated that the potential for a common mode failure of safety related l
components existed. The licensee established a root cause team to determine the cause of the event (Section E1.3).
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The licensee demonstrated good work practices. controls, and vendor
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oversight during the disassembly and cut-up of a fuel assembly skeleton for shipment off site (Section E1.4).
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A review of an Unresolved Item concerning differences between the Final
Safety Analysis Report (FSAR) and plant operating practices concluded that the practice of offloading the full core during each refueling outage did not represent a change to the facility or a change to the procedure described in the FSAR and thus did not require a review
pursuant to 10 CFR 50.59 (Section E8.1).
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Plant Suocort Health physics technicians and other workers did not meet management's
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expectations concerning contamination control practices during the conduct of work in posted contaminated areas (Section R1.1).
The licensee maintained an effective program for the control of liquid
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and gaseous radioactive effluents from the plant.
There was an overall decreasing trend in the amounts of activity released from the plant in liquid and gaseous effluents and the radiation doses resulting from those releases were a small percent of regulatory limits (Section R1.2).
The r tological environmental monitoring program was effectively
implemented.
The licensee complied with sampling, analytical and reporting program requirements. and sampling equipment was being well maintained (Section R1.3).
The water chemistry control program for monitoring primary and secondary
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water quality had been implemented in accordance with the Technical Specification requirements and the Electric Power Research Institute guidelines for pressurized water reactor water chemistry (Section R1.4).
The licensee demonstrated the ability to respond effectively to an
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emergency to recognize and classify emergency action levels, and demonstrated the adequacy and proper utilization of the emergency response facilities (Section Pl.1).
Security compensatory actions during plant work activities were
sufficient to ensure that the appropriate level of security for vital areas was maintained (Section 51.1).
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Report Details Summary of Plant Status Unit 1 began this inspection period at 100 percent power.
On May 28 power was reduced to 95 percent power for feedwater bocster pump repairs.
On May 30 power was returned to 100 percent.
On June 4 power was reduced to 97 percent for a pressure switch replacement in the secondary plant.
Power was returned to 100 percent on June 5.
The unit remained at full power for the remainder of the inspection period.
I, Operations
Conduct of Operations 01.1 General Comments (71707)
The inspectors conducted frequent reviews of ongoing plant operations.
In general, the conduct of operations was professional and
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safety-conscious; specific events and noteworthy observations are
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detailed in the sections below.
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01.2 Primary Coolant letdown Monitor Alarm (RM-L1)
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Insoection Scope (71707)
The inspectors reviewed the licensee's actions in response to a I
control room annunciator alarm on the Primary Coolant Letdown Monitor, j
RM-L1.
b.
Ob.Fervations and Findinas As a result of reducing Chemical and Volume Control System (CVCS)
letdown flow to one 60 gpm orifice (see NRC inspection report 50-395/98-04) Primary Coolant Letdown Monitor, RM-L1, went in to constant alarm due to low flow through the monitor.
The primary purpose of RM-L1 is to detect the presence of failed fuel.
The accident analyses section of the Final Safety Analysis Reaort (FSAR) also states that the low flow alarm on RM-L1 will be used ]y operators to detect a CVCS line break outside containment and allow operators to isolate the break within 10 minutes.
The FSAR also stated that there were other means of leak detection in addition to RM-L1. The inspectors reviewed the operability of RM-L1, the licensee's actions in response to the alarm, and other means of leak detection.
The licensee's review of the reduced flow through the monitor concluded that the monitor was operable. Although flow was reduced through RM-L1 the inspectors concluded that there was still adequate flow to obtain reliable readings on the high and low range radiation monitors and detect the presence of failed fuel.
The licensee also implemented a compensatory action of monitoring flow through RM-L1 every six hours to help ensure reliable radiation monitoring.
The inspectors also verified that other means of leak detection were available to detect a CVCS pipe rupture.
At the end of the report period, engineering was evaluating f
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corrective actions to restore full functionality to RM-L1.
The ins)ectors concluded that the licensee's actions in response to the locced-in trcuble alarm on RM-L1 were adequate.
c.
Conclusions Licensee actions in response to a control room alarm on liquid radiation monitor. RM-Ll. Primary Coolant Letdown Monitor, were adequate to ensure that the failed fuel detection function of the monitor was being met.
Redundant means for performing RM-L1's function of detecting a Chemical and Volume Control System line break outside containment were available.
Operational Status of Facilities and Equipment 02.1 Enaineered Safety Feature System Walkdown a.
Insoection Stone (71707)
The ins)ectors conducted a detailed system walkdown of the Emergency Diesel ruel Oil Transfer System, the S)ent Fuel Pool Cooling System, the Control Room Ventilation System, and tie Reactor Make-up (Emergency Boration) System.
b.
Observations and Findinas The inspectors conducted the system walkdowns to assess the general condition of system components including labeling, to verify that system valve positions matched system drawings and station operating procedures, and to assess plant housekeeping around system components.
There were no misaligned valves identified and component labeling was adequate.
Housekeeping was found to be good.
The inspectors concluded that the systems reviewed were properly aligned and material condition of components was good.
No concerns w :^ ;dentified.
c.
Conclusions A detailed system walkdown of the Emergency Diesel Fuel Oil Transfer System, the S)ent Fuel Pool Cooling System the Control Room Ventilation System. and t1e Reactor Make-up (Emergency Boration) System found that these systems were properly aligned per system drawings and procedures.
Material condition of components and housekeeping were good, and component labeling was adequate.
Operator Training and Qualification 05.1 Licensed Doerator Reoualification Trainina Annual Simulator Examinations a.
Insoection Scone (71707)
The inspectors observed a simulator examination and the critique of the
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examination.
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Observations and Findinas On May 27, the inspectors observed licensed operator requJ,ification simulator examinations.
The simulator scenarios were challenging and well conducted.
Operator performance was acceptable.
The examination critique was thorough and provided a comprehensive assessment of individual and crew performance.
c.
Conclusions i
A licensed operator simulator requalification exam scenario was challenging and well conducted.
The examination criticue was thorough and provided a comprehensive assessment of crew and incividual competencies.
Quality Assurance in Operations 07.1 Corrective Action Proaram a.
Insoection Scone (40500)
The inspectors performed a review of the licensee's corrective action program which included the deficiency reporting and resolution system.
Station Administrative Procedure (SAP)-1122. " Condition Evaluation.
Reporting, and Trending." Revision 2. and SAP-1141. "Nonconformance Control Program." Revision 6, were reviewed.
Additionally, a review of I
the licensee's audits of the corrective action program was performed.
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Observations and Findinal I
Corrective Action Proaram The inspectors reviewed the licensee's procedures which govern the corrective action program and process for identifying, documenting. and resolving discrepant conditions.
Procedures SAP-1122 and SAP-1141.
either separately or sometimes collectively provided a mechanism by which identified deficiencies in equipment, plant procedures and programs, and implementation of plant procedures and programs are 1)
reviewed for operability and deportability. 2) reported to management, and 3) evaluated to determine cause and corrective actions.
Condition Evaluation Reports (CERs) and Nonconformance Notices (NCNs) also provide documentation of the final deposition of the deficiencies.
The CER and NCN 3rograms define an adequate basis for a correction action program whic1 meets 10 CFR 50. Appendix B. Criterion XVI. " Corrective Action."
The licensee's primary deficiency reporting document is the CER.
With few exceptions. SAP-1122 requires that all discrepant conditions are to be documented via the CER.
Once the CERs are written, they are assigned I
different levels of significance (High or Low) based on defined criteria.
If the CER is deemed of low significance. it is routed
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directly to the dispositioning group for resolution.
If the CER is i
assigned a significance level of high, it is routed to the control room i
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l for the shift supervisors review for matters affecting operability, deportability and significant material condition problems.
All CERs are routed through a screening committee, which is comprised of personnel from the operations, maintenance, engineering licensing, and quality systems departments. The committee is chaired by the CER coordinator, whose full time position is to coordinate and administer the CER program.
The committee meets twice per week and reviews all CERs for adequacy, proper assignment for disposition, and the need for root cause evaluation.
The inspectort attended a screening committee meeting tc observe and evaluate the performance of the committee.
The inspectors determined that, based on the observations made during the meeting, the committee was appropriately reviewing CERs as required by SAP-1122.
In addition, the ins)ectors determined that the persons which made up the committee were of t1e appropriate level and position.
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The inspectors conducted intarviews with various persons at different levels at the station concernua the CER process.
All persons who were interviewed were knowledgeable of the process or demonstrated the ability to locate the applicable prwedural guidance. The inspectors also performed interviews with the CEk coordinators from plant systems
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engineering and design engineering.
These individuals demonstrated ownership'of the CER process for their individual departments necessary for proper implementation of the program.
During the initial stages of review of the corrective action program, the licensee informed the inspectors of the existence of a recently issued quality assurance (0A) deficiency report Corrective Action Request (CAR) 0A-CAR-91-1 dated May 27 and June 8, 1998.
The CAR stated that the station had a number of CERs which had exceeded the established disposition date for corrective actions. Additionally, in some cases, corrective action due dates had not been established for CERs. These matters were identified during a OA surveillance (OA-SUR-98023-0) performed in April, 1998. The licensee identified no significant safety issues which were left uncorrected as a result of these deficiencies.
Prior to the issuance of the CAR, the licensee was not focusing the required attention on details which could have lead to a significant problem with adequate corrective action. The inspectors reviewed this matter in detail and concluded that the licensee is currently focusing the proper resources to resolve this deficiency.
The current CER process is very " paper" intensive which has contributed to these problems.
The licensee stated that by June 30. 1998, the CER
process would be totally computerized (via the Primary Identification l
Program (PIP)) which should greatly reduce the " paper" portion of the
process and allow for more timely actions. Although the inspectors
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determined that licensee personnel demonstrated ownership of the corrective action program, the CER process is still in its final stages of evolution.
Based on this, an Inspection Followup Item (IFI) is being opened to track the implementation of PIP and the resolution of the CAR.
This matter is identified as 50-395/98005-01.
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Ouality Assurance (0A) Audits of Corrective Action Proaram The inspectors reviewed 0A audits OA-AUD-97022-0. 0A-AUD-97011-0, and 0A-AUD-96017-0 which evaluated the effectiveness of the corrective action program.
These QA audits represent the last three audits performed on the corrective action program.
The inspectors determined that the audits adequately evaluated the effectiveness of the corrective action program.
A number of noteworthy findings were raised by the audits.
However, the inspectors determined that based on the issues raised in the above listed audits, the OA organization should have raised the issues delineated in CAR OA-CAR-91-1 (discussed above) in a more timely manner, c.
Conclusions An inspection follow-up item was issued to track improvements needed in the Condition Evaluation Re]orting process.
Licensee personnel demonstrated ownership of t1e corrective action program.
Quality Assurance Audits adequately examined and evaluated the corrective action program.
07.2 Ouality Assurance Proaram Implementation a.
Insoection Scope (40500)
Implementation of SAP-1122 and SAP-1141 were inspected by randomly selecting ten closed CERs and any associated Nonconformance Notice (NCN)
from the period October 1997 to April 1998.
The threshold for identifying CERs and NCNs was reviewed by comparing issues listed in the Shift Supervisor (SS) logbook and the Shift Technical Adviser (STA)
logbook for the period April 14 through May 25. 1998. with the issuance of CERs and NCNs.
b.
Observations and Findinas A review of ten closed CERs and associated NCNs revealed that: 1)
operability determinations were valid and appropriate Technical Specification (TS) limiting conditions were entered when necessary, 2)
no items recuired a 10 CFR 50.72. 50.73 or Part 21 report. 3) items were correctly icentified as e: Nr high or low CER significance. 4)
Maintenance Rule applicabii.ty was evaluated. 5) causal codes were assigned. 6) corrective actions addressed the identified deficiency.
7) associated Engineering Services (ES)-508 evaluation of abnormal conditions or events was technically adequate and 8) management reviews were documented.
Operational events and equipment problems as described in the SS and STA logs for the period April 14 through May 25. 1998 were reviewed. The inspectors determined that CERs or NCNs were originated as required by SAP-1)22 and/or SAP-1141.
The inspectors did not find any issues which
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required either a CER or NCN that had not been documented as either a CER or NCN.
c.
Conclusions A review of Shift Supervisor and Shift Technical Adviser logs revealed that Condition Evaluation Reports (CERs) and Nonconformance Notices (NCNs) were being written as required.
Closed CERs and NCNs indicated items were being satisfactorily addressed.
07.3 CER and NCN Evaluation Processes a.
Insoection Scooe (40500)
The inspectors reviewed implementation of two of the processes utilized by the licensee to determine causes and corrective actions for deficiencies described in CERs or NCNs.
The inspectors reviewed Rcot Cause Analysis (RCA)-1078. ~0)eration of Component Cooling Water Pump Speed Switch Under Load" and RCA-1087 " Missed and Incorrectly Performed Surveillance" performed in accordance with SAP-900. " Root Cause Analysis." Revision 3. and earlier revisions.
Five evaluations performed in accordance with Engineering Services (ES)-508. " Evaluation of Abnormal Conditions or Events." Revision 2. and earlier revisions, were reviewed.
The following LS-508 evaluations were reviewed:
ES-508-102. " Service Water Pump Motor"
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ES-508-113. " Card Failure to Flow Transmitter"
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ES-508-119. " Diaphragm Rupture"
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ES-508-123. " Charging Pump Seal Failures"
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ES-508-139. " Sequencer Failure"
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b.
Observations and Findinas The RCA evaluations thoroughly analyzed the deficient condition and surrounding circumstances, as appropriate, to determine the probable cause(s) and to develop corrective actions.
For equipment related deficiencies vendors and other utilities were contacted, as necessary, to provide additional technical expertise and component failure history.
RCA corrective actions and/or recommended enhancements focused not only j
on correcting the deficient condition but also on preventing future
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recurrence of the deficiency. Corrective actions were tracked until (
completed.
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l The ES-508 evaluations were typically used to address equipment failures j
in which the failure's cause was not readily known but an RCA was not J
warranted.
With the focus on equipment problems, corrective actions i
normally addressed hardware or procedure fixes.
Human performance problems that contributed to the equipment failure were noted: however.
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the associated corrective actions were treated more like suggestions.
For exam]le. ES-508-119 corrective actions stated that the " findings need to )e discussed with shop personnel." Discussions with cognizant engineering services personnel indicated that followup for this type of corrective action was handled informally.
The assigned engineer determined what followup and completion verification, if any, these corrective actions received.
c.
Conclusions Root cause analyses were thorough and technically valid.
Corrective actions adequately addressed equipment, procedure and program problems, as well as. human performance issues.
Corrective actions were tracked until completed.
Engineering evaluations performed per the Engineering Services (ES)-508 process to evaluate equipment problems were technically adequate and typically focused on hardware and procedure corrective actions to preclude recurrence.
Human performance issues addressed by ES-508 evaluations were informally dispositioned.
07.4 Trend Reports a.
Inspection Scooe (40500)
The inspectors reviewed the following trend reports:
Virgil C. Summer Nuclear Station Su]plementary Trend Evaluation o
for the period November 1995 througl May 1997.
Virgil C. Summer Nuclear Station 97-01 Trend Report for the period
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January 1, 1997, to June 30, 1997, and Virgil C. Summer Nuclear Station 97-03 Trend Report for the period
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July 1. 1997 to December 31, 1997.
The purpose of the review was to assess the usefulness of the trending program to provide insights into station performance, b.
Observations and Findinos
'The trend reports are generated by sorting and counting corrective action documents, such as CCRs.
The trend reports were successful in j
identifying significant st6; ion trends such as an increase in the number
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of human performance errors in 1997 as compared with 1996.
From discussions with the Manager of Nuclear Licensing and Experience.
l providing such a high level overview for senior station management was l
the primary ]ur)ose of the trend report.
The inspectors noted and verified wit 1 t1e Manager of Operations that the trends reports have only limited value in identifying specific areas which require management attention.
For example, a aie chart is provided for Operations which shows the number of CERs ~ generated" by each major L-___-________--___--___________
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The numbers of CERs actually " caused" by l
each group is not 'rovided and the listing of CERs in the trend report
does not provide tais information.
Thus, an increase in CERs generated i
by a subgroup could be a good finding (more problems were being identified by the subgroup) or a negative finding (the subgroup was making more errors) or a combination of the two.
This observation was discussed with the Manager of Nuclear Licensing and Experience.
c.
Conclusions l
Trending reports of condition evaluation reports provided general station trends to senior site management; however, the reports were of limited value to department level or lower managers to aid in identifying specific areas in which to focus management attention.
07.5 Ouality Assurance Audits a.
Inspection Scone (40500)
The inspectors reviewed Quality Systems Procedure (OSP)-106. " Conduct of Quality Assurance Activities." for compliance with 10 CFR 50. Appendix B. Criterion XVIII. " Audits." To ensure compliance with OSP-106, the inspectors sampled various aspects of the following five audits:
QA-AUD-97001-0:
" Station Emergency Plan" dated March 3. 1997
OA-AUD-97003-0:
" Radioactive Waste" dated March 18, 1997
OA-AUD-97009-0:
" Annual Station Fire Protection ~ dated July 31,
1997 0A-AUD-97010-0:
" Station Fire Protection Audit (Biennial)" dated
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July 31. 1997 0A-AUD-97019-0:
" Station Security / Fitness for Duty" dated
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January 12. 1998 l
b.
Observations and Findina Procedure OSP-106 clearly defined the responsibilities of the individuals who implemented the audit program, and provided an
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l appropriate framework for planning and performing an audit.
The procedure also established an acceptable process by which audit findings and ocher items such as weaknesses and recommendations were reviewed and
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Auditors are required not to be directly associated with the activities j
being evaluated.
The inspectors determined that OSP-106 met the applicable requirements of 10 CFR 50 Appendix B. Criterion XVIII.
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From the random sampling of the five audits. the inspectors verified that audit checklists were properly prepared findings and weaknesses were identified and correctly documented, and sufficient documentation l
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was readily available to justify closing audit findings.
Previous audit findings and weaknesses were incorporated, as appropriately, into audit checklists to monitor the effectiveness of the associated correction actions. The Quality Assurance Group maintained an adequate system to track the status of audit findings.
c.
Conclusions Quality Assurance (OA) audit program procedure QSP-106 met applicable requirements in 10 CFR 50. Appendix B. Criterion XVIII. The procedure provided sufficient instructions to perform audits which would independently assess plant programs and provide meaningful feedback to plant management. A sampling of five 0A audits in the operations and engineering areas revealed that the OA audits were performed in accordance with the procedures.
Audit finding closure documentation was readily available and justified that audit findings had been satisfactorily addressed.
07.6 Self-Assessment Activities in Ooerations a.
Insoection Scooe (45000)
The inspectors discussed the Operations self-assessment program with the Manager of Operations and a shift supervisor.
The Semi-annual Review Of VCSNS Operations Department Crew Observation And Feedback Program Report for the period May 19. 1997, through July 3.1997, and five bench marking trip reports to other power facilities were reviewed.
b.
Observations and Findinas The semi-annual review report assessed nine major areas involving shift operations. The areas included turnovers, operator rounds, testing activities, configuration control training, communications, and supervision. The self-assessment was performed by members of management from the Operations Department, as well as, from other departments at the station. The self-assessment identified positive observations and recommended actions to improve areas such as procedure adherence.
temporary relief practices, and control board attentiveness.
Since that self-assessment, the licensee has continued to perform periodic management observations of Operation's activities.
These subsequent observations were recorded in a special control room log, reviewed with the observed operations crew, and distributed to other Operations personnel as appropriate.
The inspectors considered that this self-assessment process provided meaningful feedback to enhance the overall performance in Operations.
The licensee sends shift crew members to other nuclear and non-nuclear power plants to observe operating practices. After a trip, a report was prepared summarizing practices in which V. C. Summer was considered superior and in which the visited facility was considered superior.
From those practices considered su)erior at another facility, at least one, if not more, were chosen by t1e shift to be implemented at the
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station.
Through the review of bench marking trip reports and discussions with operating personnel, the inspectors determined that the bench marking program was viewed as providing insights into various operating practices and was enhancing operations at the station.
The bench marking program demonstrated a strong management commitment to Operations for improving operational practices, procedures and programs.
c.
Conclusions Self-assessments activities in operations were generally comprised of enhanced routine management oversight.
The self-assessments provided meaningful feedback to Operations personnel.
Managements commitment to improving operations was demonstrated by efforts to improve operational practices, procedures and programs through bench marking visits by operations personnel to other power plant facilities.
07.7 Doeratina Exoerience (DE) Proaram a.
Insoection Scone (40500)
The inspectors reviewed the licensee's OE Program to see if it met licensee commitments, regulatory requirements, and industry standards.
b.
Observations and Findinas The licensee used the following two procedures for receiving and processing various operating experiences:
NL-102. " Distribution, Review, and Processing of Various
.
Regulatory and Industry Documents." Revision 18 NL-120. " Nuclear Network." Revision 5
.
NL-102 is used for the regulatcry and industry documents related to licensing and other significant operating experience issues.
NL-120 is used for the administration and maintenance of the Institute of Nuclear Power Operations (INPO) Nuclear Network Communications Program.
The OE Program includes reviewing or processing of the following documents:
Generic Letters (GLs) and Information Notices (ins) issued by NRC.
.
Significant Event Reports (SERs). Significant Operating Event
..
i Reports (SOERs). Operation & Maintenance Reminders (OMRs).and l
Significant Event Notifications (SENs) generated by INPO and sent l
to licensees.
0)erating Experiences (OEs) entered by the nuclear licensees into
11e Nuclear Network.
_ - _ _ _ _ _ _ _ _ _ _ _.
j l
Vendor Reports (VENs) are 10 CFR Part 21 notifications or reports
.
issued by Nuclear Steam Supply System (NSSS) vendors or users and j
are related to component or part defects.
'
The inspectors reviewed the procedures and concluded that they were adequate for processing and implementation of OEs.
Numerous exam)les of OE documents issued during the last two years were reviewed by t1e inspectors.
The inspectors reviewed the licensee's processing and corrective actions to ensure that the requirements of procedures NL-102 and NL-120 were met.
Five OEs which received the inspector's detailed review and verification are listed below:
GL 920008 Thermo-Laa 330-1 Fire Barriers:
The inspectors verified the corrective actions in the field for the response sent to the NRC for GL 920008.
The licensee removed the Thermo-Lag, installed and replaced the regular cables with a fire rated Rockbestos Firezone R Cable, modified the existing nuclear instrumentation enclosure, rerouted the cable to meet the separation criteria, and installed a 1-hour fire rated Gypsum board enclosure for cable separation requirements.
IN 970066 Failure to Provide Special Lenses for Operators Usino Resoirator or Self-Contained Breathina Aooaratus:
The inspectors went to the control room, reviewed the control room operator duty list. and verified that the operators required to wear eye glasses or contact lenses were doing so.
The inspectors also verified that the operators had prescription lenses available for respirator & Self-Contained Breathing Apparatus (SCBA) use.
IN 970068 Loss of Control of Diver in a Soent Fuel Storaae Pool: The inspectors verified that the health physics procedure HPP-413. " Diving Operations." Revision 2. was revised to maintain the location of the diver at all times during a diving operation.
IN 980005 Criminal Historv Record Information:
The inspectors verified that an individual yellow envelope was set up to contain the individual Criminal History Record Information (CHRI) obtained from the Federal Bureau of Investigation (FBI) and the envelopes were stored in locked cabinets.
The licensee authorized limited employees to have access to those files.
The inspectors noted to the licensee that the l
l locked cabinets did not have a warning label indicating that FBI files can not be released to individuals or to third parties.
Subsequently the licensee attached a warning label to the front of each cabine _-_
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VEN 970020 RHR Heat Exchanaer Partition Plate Desian:
The inspectors l
reviewed the calculations generated by the engineers to requalify the RHR heat exchangers based on the vendor part defect notification.
The inspectors found that the methods used for the design requalification were different from the methods used in the example recommended in the notification.
The inspectors inquired whether the methods used by the engineers were adequate and acceptable to the vendor.
To address this question, engineering contacted the vendor to discuss the different design conditions for Summer, presented the design methods used at Summer, and obtained verification from the vendor's engineers that the licensee's methods were acceptable.
The inspectors did not identify any deviations in the corrective actions verified above.
Based on the inspectors' review of the samples and the discussions with personnel, the inspectors observed that the licensee divided the information on the operating experiences received from the industry or the NRC into higher tier and lower tier groups.
The higher-tier group included GLs. ins SOERs, SERs. OMRs, and VENs.
The remainder were included in the lower tier.
The processing used for the higher tier included formal reviews, evaluations, recommendations, corrections, and documentation.
Therefore, all information received in the higher tier was processed formally and also included information not applicable to the plant.
The processing on the lower tier was accomplished by having the program coordinator perform a peer review on the information received.
The coordinator made a decision to select which information was applicable or valuable to the plant.
During the middle of 1997, an external organization performed an audit on the licensee's OE program for receiving, processing, evaluating.
correcting. implementing, and documenting various types of information.
As a result of the audit, weaknesses were identified in the handling of the lower tier documents.
These included not formally processing these documents and having very few inputs to the Nuclear Network. Since that time, the OE coordinator logs each lower tier documents and screens them for applicable items that may require formal processing.
The OE coordinator also uses flyers, posters, newsletters and e-mail to distribute the information which is applicable or could be applicable to the plant. The licensee also inputs more information to the Nuclear Network to share their OEs with other utilities.
The inspectors reviewed the materials and information provided by the OE coordinator and found that the licensee was taking adequate actions or approaches to resolve the weakness.
The licensee had adequate procedures and performed good actions to utilize industry operating experiences for the plant.
c.
Conclusions The procedures used for the Operating Experience (OE) program were adequate in providing details for processing DE information. The OE
,
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l program successfully processed, evaluated and dispositioned OE l
information into station programs and procedures.
Miscellaneous Operations Issues (92901)
l 08.1 (Closed) Violation (VIO) 50-395/96008-01: failure to follow station housekeeping program procedure:
(Closed) VIO 50-395/96014-01: failure to follow station housekeeping
,
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program procedure:
(Closed) VIO 50-395/97005-02: failure to im)lement effective corrective action for equipment storage violations.
T1e three violations involved issues associated with the adequacy of the station housekeeping program and the implementation of designated storage area change requests. The licensee's corrective action for the first two violations. VIO 50-395/96008-01, and 50-395/96014-01 failed to correct the equipment storage deficiencies identified in the violations. As a result a third violation was identified and characterized as an ineffective corrective action issue.
The licensee's corrective action for the third violation involved comprehensive programmatic changes, training, designation of a single point of authority for overall designated storage responsibility, and a comprehensive overview by quality services during and after implementation of the new program.
The inspectors reviewed the revised administrative procedure. SAP-142. " Station Housekeeping Program."
Revision 12: reviewed training attendance records for training on SAP-142: observed several designated storage areas in the plant: and reviewed the results of the quality assurance verification of corrective action.
The ins)ectors identified no concerns and observed significant improvement in t1e storage and control of material in the plant.
The inspectors concluded that the housekeeping pro, gram was functioning in accordance with the revised program.
II.
Maintenance M1 Conduct of Maintenance M1.1 Observation of Work Activities a.
Insoection SCoDe (62707)
l The inspectors observed selected maintenance activities.
,
b.
Observations and Findinos On May 19 the inspectors observed portions of a Work Request (WR) on the C Centrifugal Charging Pump (CCP). WR 9800108. " Rework The Welds On XPP0043C Per CER-971178." The subject undersized welds on the pump equalizing line were enlarged by weld build-up.
The inspectors observed the work in progress, fire protection and radiological control practices, reviewed the weld travelers. and verified that the other CCPs' operability were not affected.
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On June 3 the ins)ectors observed technicians perform (PMTS 9805832)
ICP-195.003. "Turaine Driven Emergency FW Pump Discharge Flow IFT03525,"
Revision 7.
The technicians identified an arror in their procedure data sheet and corrected it before proceeding. An out of tolerance reading was also identifiec' and corrected, j
On June 9 the inspectors observed diving activities for the inspection of the Service Water (SW) System intake structure (STTS 9721303).
l On June 22 the inspectors observed portions of WR 9806825. " Repair Or j'
Replace Inboard Seal." on the C Component Cooling Water (CCW) pump.
!
The observed maintenance activities were conducted using the appropriate
)rocedures, tools, and technicues.
The maintenance technicians were (knowledgeable and demonstrated good work practices.
No concerns were identi fied.
c.
Conclusions Observed maintenance activities on a centrifugal charging pump, an emergency feedwater flow transmitter, the service water intake structure, and a component cooling water pump were conducted using the appropriate procedures, tools, and techniques.
The maintenance technicians were knowledgeable and demonstrated good work practices.
M1.2 Surveillance Observation
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Insoection Scooe (61726)
a.
The inspectors observed or reviewed surveillance testing activities.
l b.
Observations and Findinos On May 20 the inspectors observed performance of Surveillance Test Procedure (STP)-205.003, " Charging / Safety Injection Pump and Valve Test." Revision 5, following maintenance on the pump.
The technician followed the procedure and was familiar with how to operate the equi ament used for vibration measurement. The inspectors reviewed the cali) ration records for the vibration monitor and the probe used to obtain the vibration data.
The inspectors identified no concerns.
On June 2 the inspectors observed performance of surveillance test STP-212.001. " Reactor Core Flux Mapping." Revision 6.
The inspectors observed good communications and procedural adherence.
The inspectors identified one deficiency in the procedure. The procedure indicated a flux thimble was blocked: however, the thimble was actually usable. The surveillance was performed and the test acceptance criteria were met without using this thimble.
The licensee indicated that a procedural revision was in progress that would correct this minor deficiency in the procedure.
.
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On June 12 the inspectors observed STP-345.074 " Solid State Protection System Actuation Logic and Master Relay Test (Train B)." Revision 9.
The inspectors observed good communications and self-checking practices by the technicians performing the test, c.
Conclusions Observations of centrifugal charging pump. reactor core flux mapping, and solid state protection system surveillance testing revealed good communications, self-checking, and procedural adherence.
M1.3 Review of Completed Surveillance Test Packaaes a.
Insoection Scone (61726)
The inspectors reviewed a sample of completed surveillance test packages to verify that the documentation satisfied the referenced TS Surveillance Requirements (SRs).
b.
Observations and Findinas The inspectors reviewed STPs and completed surveillance test packages for the following tests:
STP-345.001
" Delta T - TAVE Protection Loop 1 Calibration L
Procedure." Revision 9 STP-345.002. " Delta T - TAVE Protection Loop 2 Calibration
.
Procedure." Revision 9 STP-345.003. " Delta T - TAVE Protection Loop 3 Calibration
.
,
Procedure." Revision 10 STP-120.005. " Emergency Feedwater Actuation Test." Revision 6
.
STP-220.001A. " Motor Driven Emergency Feedwater Pump and Valve
.
Test. Revision 5 STP-220.002. " Turbine Driven Emergency Feedwater Pump and Valve
.
Test." Revision 2 For these completed test packages, each SR referenced by the licensee's Surveillance Test Task Sheets had been satisfied.
Completed emergency feedwater surveillance tests and instrument loop calibration packages for reactor protection setpoints demonstrated acceptable test results.
No 3roblems were identified with the STPs or the completed surveillance pac cage _ _ _ - - _ _ _ _ _ _ - _
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c.
Conclusions Completed emergency feedwater surveillance tests and instrument loop calibration packages for reactor protection setpoints demonstrated acceptable test results.
M1.4 Diesel Generator (DG) Failure Durino Surveillance Test a.
Insoection Scooe (62707)
The inspectors reviewed the licensee's efforts to restore the A DG t
following a trip on high crankcase pressure during a routine l
surveillance test.
b.
Observations and Findinas l
At 9:58 p.m on June 15. the A DG had been running fully loaded for
'
about 15 minutes for surveillance testing when the A DG tripped on high
crankcase pressure.
According to the operator, crankcase pressure had L
been checked a few minutes before and was normal.
Following the A DG l
trip operators ran the A DG crankcase vacuum pump in manual to check the l
pump's operability.
The vacuum pump drew the expected vacuum.
Several i
minutes later operations ran the vacuum pump again. When the vacuum pump switch was returned to the auto position several abnormal indications were observed on the local and remote A DG control panels.
Operators documented their observations for review by engineering.
l An engineering review of the event concluded that an erroneous signal l-from the crankcase pressure switch had probably caused the A DG trip.
l The abnormal-indications on the A DG panels described by operators appeared to have been caused by a time delay relay (T3 relay) which
l enables the A DG trips after the A DG starts.
All the indications l
observed by operators could not be logically explained by engineering l
when the circuit diagrams were reviewed.
l A troubleshooting plan was developed by maintenance and engineering.
The )lan included replacing the crankcase pressure switch, monitoring
!
cranccase pressure during A DG operation and replacing the T3 relay and verifying it operated properly.
Following successful completion of the maintenance and a surveillance test the A DG was declared operable on June 17 at 4:10 p.m.
The licensee was unable to identify any problems
.
with the pressure switch or the T3 relay, and no equipment was found inoperable.
'
The inspectors observed portions of the A DG work and reviewed completed work documentation. The inspectors also observed management meetings following the completion of the A DG repairs and surveillance testing.
The meetings were held to review the maintenance performed and to ensure all issues with the A DG were resolved before declaring the A DG operable.
The management review included a review of maintenance that was performed earlier in the week to verify that it was not the cause of the A DG trip.
During the troubleshooting efforts the inspectors
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observed strong system engineering support to maintenance.
System engineering was involved in all aspects of the A DG repair effort by maintenance.
The inspectors noted that valuable information may have been lost for the troubleshooting 3rocess when operators ran the crankcase vacuum pump after the A E trip.
c.
Conclusions The corrective maintenance following an A Diesel Generator (DG) trip on high crankcase pressure was performed such that the potential causes of the trip were identified and corrected.
Strong system engineering support to maintenance was observed during the repair effort.
Clear management involvement was exhibited in the maintenance review before declaring the A DG operable.
M3 Maintenance Procedures and Documentation M3.1 Maintenance Rule Periodic Evaluation a.
Insoection Scoce (62706)
Paragraph (a)(3) of the Maintenance Rule. 10 CFR 50.65, requires that performance and condition monitoring activities and associated goals and preventive maintenance activities be evaluated taking into account.
where practical, industry-wide operating experience.
This evaluation is required to be performed at least one time during each refueling cycle, not to exceed 24 months between evaluations.
The inspectors reviewed the licensee's completed periodic assessment to determine if it met the requirements of 10 CFR 50.65, paragraph (a)(3).
b.
Observations and Findinas At the time of the Maintenance Rule baseline inspection. during May 1997, the licensee had not completed their first periodic evaluation.
The inspectors reviewed the licensee's completed Maintenance Rule Periodic Assessment dated February 3,1998.
This first periodic assessment covered the period from July 10. 1996, to November 7. 1997.
The Jeriodic assessment report consisted of a high level summary report whic1 summarized individual system engineer monthly system assessment reports rather than a single comprehensive evaluation report.
This method was an option allowed by NUMARC 93-01. " Industry Guidelines for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants."
Revision 2.
The periodic assessment report also documented the balance
!
between unavailability and reliability for risk-significant Structures.
l Systems and Components (SSCs) as required by paragraph (a)(3) of the Maintenance Rule.
The periodic assessment was conducted in accordance with Section 6.9,
" Periodic Assessment," of SAP-1252. " Maintenance Rule Program." The
!
inspectors determined that the licensee's assessment satisfied the l
requirements of 10 CFR 50.65 and NUMARC 93-01. Revision 2.
This
)
assessment resulted in significant program improvements including the i
addition of one new SSC to the scope of the Maintenance Rule, t
reclassification of one SSC. function as risk-significant, and revision of reliability and unavailability performance criteria for several risk-significant SSCs due to historical performance or industry experience.
Additionally, monitoring for a portion of one system was changed to require the use of SSC specific performance criteria rather than plant level performance criteria.
The use of industry operating experience
,
was verified as being well integrated with system engineering evaluations, scoping, reviews of functional failures, and cause
. determinations.
The next periodic assessment is scheduled for performance within the next 18 months (normal fuel cycle) to meet Maintenance Rule requirements for conducting an assessment every refueling cycle, not to exceed 24 months between assessments.
c.
Conclusions The licensee's periodic Maintenance Rule assessment report provided sufficient detail to demonstrate that the licensee had adequately evaluated performance, condition monitoring, associated goals and preventive maintenance activities for SSCs within the scope of the Maintenance Rule.
The licensee's assessment met the requirements of NUMARC 93-01 and paragraph (a)(3) of 10 CFR 50,65 and is a strength that resulted in significant. improvements in the Maintenance Rule Program.
M3.2 Missed Surveillance Requirement On Accident Monitorina System Neutron Flux Instrumentation a.
Insoection Scooe (61726)
The inspectors reviewed a missed surveillance requirement on post accident monitoring neutron flux intermediate range instrumentation NI-35 and NI-36.
b. -Observations and Findinas On June 11 the licensee identified that channel checks for the Post Accident Monitoring System (PAMS) neutron flux channels NI-35 and NI-36
'
were not performed in'accordance with the TS 4.3.3.6, " Accident Monitoring Instrumentation," surveillance requirements.
The licensee documented the details and the cause of the missed surveillance in Licensee Event Report (LER) 50-395/98006-00. The cause of the event was a procedural inadequacy.
Procedures did not adequately identify the
instruments to be checked and the requirements for the channel checks.
I As a result the tests were not performed.
The inspectors reviewed the licensee's corrective action.
Corrective action consisted of performing a 3 roper channel check, verifying other PAMS instrumentation channel chects were performed, and correcting the procedure. This non-repetitive, licensee identified and corrected violation is being treated as a Non-Cited Violation (NCV) consistent
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._ __________- __ _ _ ____ _ ___ _ ___
with Section VII.B.1 of the NRC Enforcement Policy.
This is identified as NCV 50-395/98005-02.
Based upon this inspection LER 50-395/98006-00 is considered closed.
c.
Conclusions A non-cited violation was identified for failure to perform Technical Specification required channel checks on Post Accident Monitoring System neutron flux intermediate range channels NI 35 and NI-36.
M7 Quality Assurance in Maintenance Activities M7.1 Maintenance Self-Assessment Exit Meetina a.
Inspection ScoDe (40500)
The inspectors attended a maintenance self-assessment exit meeting.
b.
Observations and Findinas On May 20. the inspectors attended an exit meeting for a maintenance self-assessment inspection conducted by the licensee's Quality Assurance (0A) group. The self-assessment, conducted the week of May ll, identified several strengths and areas for improvement.
The inspectors also discussed the findings with the OA supervisor. The inspectors concluded from their review of the assessment findings that the assessment had provided a thorough and critical evaluation of the Summer maintenance program.
The issues identified during the self-assessment were consistent with recent NRC assessments of licensee performance.
c.
Conclusions A licensee maintenance self-assessment was thorough and provided a critical evaluation of the maintenance program.
The findings were consistent with NRC assessments of licensee performance.
M8 Miscellaneous Maintenance Issues (73756, 92902)
M8.1 (Closed) VIO 50-395/97011-02:
failure to maintain delta T - TAVE protection loop calibration procedure.
(Closed) LER 50-395/97003-00:
entry into TS 3.0.3.
The licensee identified that STPs listed an incorrect error tolerance for the Over Temperature Delta Temperature (0 TDT) and Over Power Delta Temperature l
(OPDT) reactor trip setpoint time constants that were set during l
protection loop calibrations. As a result of this error. the time l
constant values used were less conservative than the values specified in the TS.
The inspectors reviewed the licensee's response to the violation dated November 3. 1997. along with LER 50-395/97003-00 which reported this event as required by 10 CFR 50.73 (a)(2)(i). The licensee determined
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]
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that the root cause of the problem had been the failure to correctly implement the instrument calibration revisions required by Technical Specification Amendment 75 resulting in the failure to revise the acceptance criteria for the time constants for OTDT and OPDT.
During the review of completed STPs documented in Section M1.3 of this report, the inspectors veriTied that the affected procedures had been revised to include the correct time constant values for the OTDT and OPDT reactor trip setpoints.
Based on this inspection. the inspectors determined that the licensee's corrective actions were adequate.
M8.2 (Closed) VIO 50-395/96015-03: numerous examples of missed technical specification required surveillance tests associated with the turbine and motor driven errwgency feedwater pumps.
(Closed) LER 50-395/96010-00: missed surveillance on slave relay (K634).
The licensee identified that several quarterly surveillance tests associated with steam generator low-low actuation slave relays K633 and K634 for the turbine driven and motor driven emergency feedwater pumps had not been performed within the interval specified by TS SR 4.3.2.1.
Functional Unit 6.b. " Emergency Feedwater." of Table 4.3-2 "Fngineered Safety Feature Actuation System Instrumentation."
requires that a slave relay test for Automatic Actuation Logic and Actuation Relays be performed quarterly.
The inspectors reviewed the licensee's response to the violation dated April 4, 1997 along with LER 50-395/97010-00 which reported this event as required by 10 CFR 50.73 (a)(2)(I).
The licensee attributed the cause of the problem to personnel oversight in not uniquely identifying individual surveillance requirements referenced in STPs.
The licensee also reviewed other plant procedures and determined that the problem was j
limited to the STPs for actuation logic circuits for the emergency feedwater pumps.
Surveillance testing performed subsequent to each of the missed surveillance tests verified that the affected equipment had been capable of performing its safety function hing the interval between tests.
During the review of completed STPs documented in Section M1.3 of this report. the inspectors verified that SR 4.3.2.1 had been satisfied for the period between October 1. 1997. and March 31. 1998.
Specific STPs reviewed included emergency feedwater pump surveillance tests STP-120.005. STP-220.001A. and STP-220.002.
Licensee )ersonnel made effective use of attached Surveillance Test Task Sleets to ensure each required SR had been performed.
Cased on this inspection. the inspectors determined that the licensee's corrective actions were adequate.
M8.3 (Closed) URI 50-395/98001-02:
Inconsistencies between Code requirements and implementing procedures for testing pressure relief valves.
This item was issued to document and track the licensee identified condition that there were apparent differences in the Inservice Testing (IST)
Program document. General Test Procedure (GTP)-302 " Inservice Testing of Valves. Second Ten Year Interval." Revision 9. and the implementing test
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procedures. STP-401.003. " Code Relief Valves ASME XI Test." Revision 4 and Mechanical Maintenance Procedure (MMP)-445.005 " Maintenance Repair
.
and Testing of Relief Valves." Revision 13. The inspectors reviewed the l.
licensee ~s IST program for pressure relief valves as described in GTP-l 302, and the test procedures.
The licensee's IST program second ten year interval began on January 1.
l 1994.
The American Society of Mechanical Engineers (ASME) Boiler and l
Pressure Vessel (B&PV) Code.Section XI requirement for this interval is the 1989 Code edition In 10 CFR 50.55a the 1989 Code edition is
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l identified as the ASME/ ANSI Operations and Maintenance Standards (0M).
l 1987 edition, with the 1988 addenda as the requirements for pump and valve inservice testing.
Part 1 of the OM-1987 edition addresses
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" Requirements for Inservice Performance Testing of Nuclear Power Plant Pressure Relief Devices." GTP-302 was updated, almost verbatim. to the new code (OM-1987. Part 1) requirements: however, the implementing
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procedures (STP-401.003 and MMP-445.005) were not upgraded to the new requirements. A CER 98-0075 was issued on January 22. 1998, based on the observation of a Quality Control (OC) engineer who identified this problem during a review of STP-401.003.
CER 98-0075 documented the deficiencies and established certain activities to correct the problem. Tne plan of action included immediate suspension of all relief valve testing until program changes were completed.
Maintenance was to provide System / Component Engineering
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(SCE) with data for all ASME Class 2 and Class 3 relief valve test data l
from January 1. 1994 to the suspension of testing in January 1998. SCE was to evaluate these data against technical and administrative requirements of OM-1987. Part 1 (OM-1) for operational impact. The test i
Unit /0C engineer were to review OM-1 ASME Section XI Code requirements and industry practices to develop guidance for procedure changes.
The OC Engineer was assigned to disposition CER 98-0075 as the single point of contact.
Doerational Imoact Review The inspectors discussed the corrective actions with various licensee personnel and reviewed the documentation of these activities as follows:
CER-98-0075 Operability, deportability evaluation form dated
.
1/22/98.
Disposition, cause and corrective action form dated 1/28/98.
.
Status report of actions dated 2/17/98.
- Action completion form dated 3/3/98.
(Retrieval of previous test
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data for analysis).
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Action completion form dated 6/2/98.
(completion of engineering
.
evaluation of test data.
Evaluation concluded no operability impact on relief valves but some failed tests were identified).
Engineers Technical Work Records 15150-M36 dated 4/30/98 and
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15150-36M1 dated 5/18/98, documenting the evaluation of relief valve tests the basis for conclusions and recommendations to improve weak areas.
CER 98-0470 dated 5/20/98.
This deficiency report was issued to
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generate an LER when the evaluation report identified that additional valve tests had not been performed for failed as found tests.
LER 50-395/98005-00 dated 6/1/98. " Inadequate ASME Code Relief
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Valve Surveillance Testing."
Status Report of CER-98-0075 actions dated 6/18/98. This report
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summed up the items to be completed for CER 98-0075 actions as completion of procedure revisions for relief valve testing including, incorporation of valve grouping and scheduled frequency of tests, incorporation of additional valve tests for as found test failures, and incorporation of the proper definition of as-found test setpoint.
Engineering was to provide the technical basis for any temperature or media correlations.
The inspectors reviewed test procedures STP-401.003 and MMP-445.005 against the requirements of OM-1.
The following deficiencies were noted in the procedures:
A minimum 10 minute hold time between valve tests was not
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speci fied.
Valves were not grouped and monitored for 20% tested within any 48
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months.
j Additional valves were not tested when as-found setpoints exceeded
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the stamped setpoint by greater than 3%.
Visual inspection was not documented and procedural guidance for
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inspection was not specified.
l Leak tightness acceptance limits were not specified in the
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procedure acceptance limits.
Verification of integrity of balancing devices not required by
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procedure.
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As-found setpoint incorrectly defined by procedure.
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Correlation of test condition to operating condition for j
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temperature or test media not provided.
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Except for the 10 minute hold time. these same deficiencies had been identified and documented by the licensee in CER 98-0075. The licensee performed a thorough review of all Class 2 and Class 3 relief valve tests aerformed since January 1. 1994.
Analysis of this data indicated that t1ere was not a relief valve operability concern although there were several ASME/ ANSI OM-1987. Part 1 Code requirements that were not properly performed.
The inspectors reviewed the licensee's evaluation and determined that the evaluation of relief valve operability was reasonable and adequate.
The inspectors were informed that the procedure revision process was just beginning and that the above procedural deficiencies will be corrected in the revised procedures.
Programmatic Asoects (
The licensee conducted a RCA to determine why these procedures were not
'
revised to meet the ASME/ ANSI OM-1987 Part 1 requirements.
Results of the RCA indicated that a series of mis-communications and mis-understandings had occurred among managers, procedure writers and contractors.
,
The root cause team indicated that this issue should have been prevented or, at least, identified many times.
The RCA characterized the contributors as:
Lack of authority of the single point of contact for IST.
- Inadequate written documents.
- Lack of supervision.
- Lack of personnel accountability.
.
Lack of a questioning attitude.
- l.act of change management.
.
Inaccurate verbal communications.
.
The RCA contained five recommendations, accepted by site management. to prevent future breakdowns in communications.
One has been completed and the remainder are scheduled for completion no later than August 1, 1998.
The inspectors concluded that t!ese corrective actions were sufficient to prevent recurrence.
Technical Specification 4.0.5 requires in part that inservice testing of ASME Code Class 1.2 and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a.
10 CFR 50.55a identifies the Code for ASME/ ANSI Part 10 shall be the OM-1987 Edition and OMa-1988 Addenda to OM-1987 Edition.
Part 10 references ASME/ ANSI l
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-OM-1987 Part 1 as the applicable Code for testing Class 1, 2 and 3 Pressure Relief Devices. The licensee's failure to implement OM-1987 Part 1 for testing of pressure relief devices in the second IST ten year interval starting January 1,1994 was ideritified as a violation. This non-repetitive. licensee identified and corrected violation is being treated as a NCV consistent with Section VII.B.1 of the NRC Enforcement Policy. This is identified as NCV 50-395/98005-03.
M8.4 (Closed) LER 50-395/98005-00:
Inadequate ASME Code Relief Valve Testing. This LER addressed the same issues as discussed in URI 50-395/98001-02.
The LER corrective actions were reviewed as part of the review of this URI item.
Closure of the URI item also closes this LER.
III.
Enaineerina El Conduct of Engineering El.1 10 CFR 50.59 Safety Evaluation Proaram a.
Insoection Scooe (37001)
The inspectors performed a programmatic review of the licensee's 10 CFR 50.59 3rocedures and training requirements in order to verify compliance with t1e regulations and regulatory commitments.
An evaluation of the i
licensee's performance in implementing the requirements of Section 50.59 was also performed to assess the licensee's resolution of safety issues related to change, tests, and experiments.
b.
Observations and Findinas Procedure SAP-107 "10 CFR 50.59 Unreviewed Safety Question Review Process." Revision 1. established requirements for review and evaluation of changes performed in accordance with the requirements of 10 CFR 50.59. The scope of the procedure includes changes to the facility im)1emented by plant modifications, commercial changes or by plant enhancements.
Temporary changes to the facility resulting from NCNs.
procedures, industry operating experience evaluations and temporary modifications were also included within the scope of the procedure.
Procedure revisions which require manipulation of plant equipment: are l
used for working on plant equipment: or are used to test plant equipment or describe the administrative controls associated with plant operations require a 10 CFR 50.59 screening / evaluation in accordance with the requirements of SAP-107.
Detailed procedural guidance was provided for
. performing a 10 CFR 50.59 screening and safety evaluation in order to i
determine if an unreviewed safety question was created by the change.
Responsibilities for implementation of the 10 CFR 50.59 program were assigned to specific individuals as delineated in Attachment 1. "10 CFR 50.59 Support Matrix" of SAP-107.
The licensee has established training requirements for the following personnel who implements the requirements of the 10 CFR 50.59 program:
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--___-____--_-_
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l 10 CFR 50.59 Independent Reviewer - an individual who performs
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independent reviews per SAP-107 and is certified by plant management as having satisfied the training requirements delineated on,\\ttachment VI. " Qualified Reviewer / Cross-Disciplinary Reviewer /10 CFR 50.59 Certification" of SAP-107.
l 10 CFR 50.59 Preparer - an individual who performs and documents
.
I.
required. reviews per SAP-107 and is certified by plant management as having completed the-requirements of Attachment VI.
Cross-Disciplinary Reviewer - an individual designated by their
manager to act as a procedure reviewer for procedures outside their discipline and who meets the requirements of ANSI N18.1 (1971). Section 4.0.
The inspectors selected 25 personnel at random in order to verify that
,
l-training requirements delineated on Attachment VI had been completed for these individuals.
Based on review of the certifications for the
'
selected personnel the inspectors concluded that the training requirements of the 10 CFR 50.59 program were being adequately implemented.
!
Implementation of the 10 CFR 50.59 program was evaluated based on review of the following Modification Change Notices (MCN)/ Engineering Change l:
Requests (ECR):
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MCN No. 21745H. Close torque switch by-pass
MCN No. 21745K. Change control scheme for valve XVG0312A-SW
MCN No. 20951. Install new fire detection control system e
!
ECR No. 50025, Add test connection for emergency feedwater flow
control valve leak rate testing l
ECR No. 50004. Power Lockout feature charging /HHSI cross connect
valves MCN No. 22553. Turbine Uprate.
!
The 10 CFR 50.59 screenings / evaluations aerformed for the above design I
changes were reviewed and determined to ]e technically adequate and in compliance with the requirements of 10 CFR 50.59.
No deficiencies were identified during this review.
!
c.
Conclusions A positive finding was identified concerning im) lamentation of the 10 CFR 50.59 Program.
Personnel assigned responsi]ility for implementing the program were trained and procedural controls provided clear guidance
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for determination of an Unreviewed Safety Question.
10 CFR 50.59 screenings / evaluations performed for design changes were determined to
,
be technically adequate, E1.2 Observation of Princioal Enaineer Review Group (PERG) Meetina t
a.
Insoection Scooe (37551)
The inspectors observed the conduct of a PERG screening meeting.
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Observations and Findinas On June 2, the inspectors observed a PERG screening meeting.
The PERG consists of discipline principal engineers from the design engineering organization that as a grou) screen proposed engineering changes to determine the appropriate clange control program to be used.
The inspectors observed the PERG review Engineering Change Requests (ECR)
involving battery testing and )lant operation with two CVCS letdown ori fices.
The PERG reviewed t1e Design Input Considerations Checklist
,
arepared for.each ECR.
The checklist covers the considerations that lave the potential to impact the design basis of the plant. The
inspectors concluded that this was a useful review in the design change process.
c.
Conclusions An observation of a Principal Engineering Review Group Screening Meeting concluded that the review served as a useful check of design input requirements (ANSI N45.2.11) that have the potential to impact the design basis of the plant.
E1.3 Trio Of Service Water (SW) Buildino Fans l
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a.
Insoection Scooe (37551)
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The inspectors reviewed the trip of both SW building fan motor breakers.
b.
Observations and Findinas j
On June 10 at 2:40 a.m., the control room received a SW building high temperature warning alarm on the plant computer system.
Operators determined that the B SW building fan motor XFN80B had tripped.
The XFN80B breaker would not reset and fan XFN80A was started. After
l approximately 10 minutes fan XFN80A tripped.
The building operator found the A fan breaker tripped and noticed a smell of burnt insulation.
Both of these fans motors are powered from 480 volt vital busses. A subsequent review of SW building temperature information on the plant computer indicated that the B train SW fan motor had tripped at about 2:30 p.m. the previous afternoon.
The licensee promptly recognized that measures to control temperature in the SW building were necessary to prevent exceeding the TS 3.7.9 limit u
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and set-up temporary cooling to maintain building temperature.
The inspectors viewed both the A and B train fan motor breakers after they were removed from the switchgear.
On each breaker a line side conductor on one of the three phases was severely burned where it connected to the breaker.
The inspectors concluded based on the similarity of the breaker failures and the time at which the breakers failed was close together that the potential existed for a common mode failure.
Both breakers were 480 f
volt Square D breakers.
The licensee assembled a root cause team to evaluate the cause of the event.
c.
Conclusions A review of service water building fan 480 volt breaker failures indicated that the potential for a common mode failure of safety related components existed.
The licensee established a root cause team to determine the cause of the event.
El.4 Fuel Assembly Skeleton Recoverv a.
Insoection Scooe (37551)
The inspectors observed Jortions of a fuel assembly skeleton recovery operation conducted in tie Spent Fuel Pool (SFP).
b.
Observations and Findinas The inspectors observed work practices and controls in the SFP during the disassembly of a fuel bundle and cut-up of the fuel assembly i
i skeleton.
The inspectors verified the licensee's compliance with the fuel movement TS requirements, observed fuel movement, radiological controls, foreign material exclusion, and vendor oversight.
The technicians performing the activities were knowledgeable and
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demonstrated good work practices.
The licensee demonstrated good
>
oversight of contractor personnel.
c.
Conclusions The licensee demonstrated good work practices controls, and vendor oversight during the disassembly and cut-up of a fuel assembly skeleton for shipment off site.
E8 Miscellaneous Engineering Issues (92903)
E8.1 (Closed) URI 50-395/96005-03:
variances between the FSAR and plant operating practices for the SFP.
The licensee's practice of performing full core offloads during every refueling outage was identified as a variance from the FSAR description of the spent fuel pool cooling system design basi _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _
Section 9.1.3.1, " Design Bases." of the Summer FSAR (Amendment 2. August 1986) states that the spent fuel pool cooling system is designed to perform the following functions:
'
With one heat exchanger operable. maintain the spent fuel pool water temperature less than 140 F with a heat load based upon decay heat generation from 76 fuel assemblies that have been irradiated for 24.000 Effective Full Power Hours (EFPH) and cooled for six days plus 1,200 fuel assemblies that have been irradiated for 24.000 EFPH and [ cooled] one year. or more.
With two heat exchangers operable, maintain s]ent fuel pool water
!
temperature less than 140 F with a heat load ]ased upon decay heat generation from 157 fuel assemblies that have been irradiated for 2,400 [ sic] EFPH and cooled for 6 days and 1.119 fuel assemblies l
that have been irradiated for 24.000 EFPH and cooled for one year.
I l
or more.
Similarly, Section 9.1.3.3 " Safety Evaluation," of the FSAR states:
The normal design basis cooling situation is 72 fuel assemblies that have been irradiated for 24,000 EFPH and cooled for 6 days plus 1214 fuel assemblies that have been irradiated for 24.000 EFPH and cooled for one year, or more.
Thipcombinationresultsindesignbasisheatloadof16.4 X10 BTU /hr.
The calculated operating temperature of the pool, considering only heat removal through one heat exchanger, is equal to the design criterion of 140 F.
No credit is taken for the evaporative cooling effect.
The off-normal design basis cooling situation is 157 assemblies that have been irradiated for 24.000 EFPH and
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cooled for 6 days plus 1.119 assemblies that have been i
irradiated for 24.000 EFPH and cooled for one year, or more.
The design basis heat load for this case is 31.3 X 10 BTU /hr.
The calculated operating temperature of the pool, considering only heat removal through two heat exchangers, is no greater than the design criterion of 140 F.
No credit is taken for evaporative cooling effect.
The inspectors reviewed records from past refueling outages and confirmed that:
'
(1)
The licensee has offloaded the full core during each refueling since plant startup.
l (2)
During all past full core offloads, the licensee did not offload
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the full core prior to the six days assumed in the FSAR.
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(3)
The inspectors did not review records of spent fuel pool cooling system operability records but.did determine that the maximum temperature reached during a full core offload was 105.3 F.
This temperature was well below the design criteria of 140 F described in the FSAR.
The inspectors noted that Section 9.1.3.1 of the FSAR does not in any.
way characterize the frequency with which full core offloads are or might be performed at Summer.
In addition, none of the design considerations described in Section 9.1.3.1 are sensitive to the frequency with which full core offloads are performed.
The inspectors also observed that Section 9.1.3.3. by introducing the terms normal and off-normal. recharacterizes each of the design basis cases in a way that could imply that full core offloads are to occur on a less frequent basis than partial core offloads.
The inspectors were unable to conclude, however. that Section 9.1.3.3 represents a specific commitment to limit the frequency with which full core offloads are conducted. As noted above, nothing in the FSAR description of the spent fuel pool cooling design basis is sensitive to the frequency with which full core offloads are conducted. Therefore, the inspectors concluded that the practice of offloading the full core during each refueling outage did not represent a change to the facility or a change to the procedure described in the FSAR and thus did not require a review pursuant to '10 CFR 50.59. On this basis URI 50-335/96005-03 is closed.
IV.
Plant Sucoort R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 Observation of Health Physics (HP) Practices a.
Insoection Scope (71750)
The inspectors reviewed contamination control practices during the conduct of work in potentially contaminated areas, b.
Observations and Findinas On May 25 inspectors observed replacement of reactor coolant filter.
XFL-9.
The inspectors observed the removal of the contaminated filter and its placement into a lead cask.
The procedure involved the use of a long extension tool and socket for loosening bolts. The radiological controls observed by the inspectors were adequate. Contamination I
control practices, however, were not applied consistently. The extension tool used in the potentially contaminated area was placed such that it extended outside the contaminated roped area prior to swiping or surveying the end of the tool. Also, health physics technicians were
]
f observed reaching across the contaminated boundary into the contaminated I
area without gloves to assist with the lead cask bagging and taping.
]
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I The inspectors questioned HP supervision on these observed contamination
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control practices.
The licensee agreed that these practices were not consistent with their expectations.
On June 1, the inspectors observed portions of carbon replacement in the reactor building purge exhaust plenum XAA0012.
Preparations for carbon removal from charcoal filter cells in the plenum included covering cell openings with clear plastic to prevent the s) read of carbon dust and potential contamination.
The space inside t1e plenum was posted as a contaminated area with a step-off pad at the door to the plenum. The inspectors observed two technicians taping the plastic up inside the plenum.
Both technicians were wearing protective clothing, however, one technician was not wearing gloves while performing the taping. The l
technician had a cotton liner on one hand and nothing on the other hand.
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The ins 3ectors questioned the HP technician at the RCA control point and found t1at HP verbal requirements established for the job were lab coats, booties and gloves.
The HP control point technician responded immediately to the inspectors concerns to ensure the workers were aware of HP dress requirements.
The inspectors concluded that consistent and clear contamination control practices were not being demonstrated by both HP technicians and other I
workers.
The workers appeared to be unsure of and inconsistently applying contamination control work practices while working in and around posted contaminated areas.
c.
Conclusions l
Health physics technicians and other workers did not meet management's expectations concerning contamination control practices during the conduct of prk in posted contaminated areas.
R1.2 Radioactive Effluent Control Proaram a.
Insoection Scooe (84750)
The inspectors reviewed the overall results of the radioactive effluent control program as documented in the Annual Effluent and Waste Disposal Report for 1997 The amounts of radioactivity released and the resulting radiation doses for the years 1994 through 1997 were also tabulated from the annual reports to evaluate long term performance of the effluent control program relative to the design objectives in i
10 CFR 50. Appendix I for radiation doses from plant effluents.
I b.
Observations and Findinas The data presented in Table 1 below were compiled from the licensee's effluent release reports for the years 1994 through 1997.
The inspectors reviewed the report for the year 1997 and discussed it's content and the data presented in Table 1 with the licensee.
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TABLE 1 SUMMER RADI0 ACTIVE EFFLUENT RELEASES
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LIOUID EFFLUENTS Curies Released Dose (mrem)
Year F&AP
"H D.&lG T B.
Oraan
&
[3 mrem]
[10 mrem]
1994 0.35 752 1.19E-1 1.6E-2 (0.53%)
5.1E-2 (0.51%)
1995 0.11 306 2.19E-3 5.2E-3 (0.17%)
1.5E-2 (0.15%)
1996 0.14 579 1.25E-2 1.0E-2 (0.33%)
3.8E-2 (0.38%)
1997 0.04 922 1.97E-2 9.8E-3 (0.33%)
1.0E-2 (0.10%)
GASE0US EFFLUENTS Curies Released Dose (mrem)
Year F&AP Iodines Part.
"H Air Oraan
[y 10 mrad]
[15 mrem]
[# 20 mrad]
1994 135 2.12E-4 3.74E-4
y 3.8E-2 (0.38%)
2.1E-2 (0.14%)
- 4.3E-2 (0.21%)
1995 2.8 3.03E-7 6.47E-7
y 2.1E-6 (<0.01%) 9.9E-4 (0.01%)
- 1.3E-3 (<0.01%)
1996 0.6 1.73E-6 6.83E-6
y 9.0E-5 (<0.01%)
7.4E-3 (0.05%)
- 2.3E-4 (<0.01%)
1997 0.3 7.20E-7 5.00E-5
y 2.5E-4 (<0.01%) 3.0E-3 (0.02%)
- 1.1E-4 (<0.01%)
F&AP Fission and Activation Products
'H Tritium D&EG Dissolved and Entrained Gases T.B.
Total Body
[]
Limits / Unit ()
% of Limits / Unit Part Particulate y
Gamma B
Beta As indicated in the table there was an overall decreasing trend in the amounts of activity released from the plant in liquid and gaseous effluents and the radiation doses resulting from those releases were a small percent of regulatory limits.
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The inspectors also observed the collection of samples from Waste Monitor Tanks A and B, and noted that the samples were collected in accordance with'the instructions in procedure HPP-710. " Sampling and Release of Radioactive Liquid Effluents." Based on the analytical results from those samples. Liquid Waste Release Permits WM-98-79 and WM-98-90 were issued by the licensee.
Those permits indicated that the liquid waste was well within the radionuclides concentration and dose limits specified in the Offsite Dose Calculation Manual (0DCM). The permits also specified the alarm setpoints for liquid waste effluent monitors RM-L5 and RM-L9.
The inspectors visited the Control Room during the release of Waste Monitor Tank A and verified that the alarm setpoints for those two monitors were as specified by permit WM-98-79.
The inspectors noted that the setpoints were more conservative than recuired by the ODCM and that the monitor's strip chart recorders incicated that source checks had been performed as required by the ODCM.
c.
Conclusions The licensee maintained an ef fective program for the control of liquid and gaseous radioactive effluents from the plant.
There was an overali decreasing trend in the amounts of activity released from the plant in liquid and gaseous effluents and the radiation doses resulting from those releases were a small percent of regulatory limits.
R1.3 Radiological Environmental Monitorina Proaram a.
Insoection Scooe (84750)
The inspectors reviewed the overall results of the radiological environmental monitoring program as documented in the Annual Radiological Environmental Monitoring Report for 1997.
Those results were compared to the program requirements delineated in the ODCM.
b.
Observations and Findinas The inspectors noted that, in accoraance with the TS and ODCM. the report included a description of the program, a summary and discussion of the results for each exposure pathway, analysis of trends during the operational years as compared to the preoperational years, and an assessment of the impact on the environment based on program results.
The report also included a tabulation of the summarized analytical results for the samples collected during 1997.
From a review of those data the inspectors determined for selected exposure pathways that the i
sampling and analysis frequencies specified in the ODCM had been met.
I As indicated in the report conclusions, the analytical results were as
expected for normal environmental samples.
Very low concentrations of l
man-made isotopes were occasionally detected in the samples but were of i
no dose consequence.
It was further concluded, in the report. that the 1997 program results substantiated continued conformance with the design objectives in 10 CFR 50. Appendix I for maintaining radioactive effluents from the plant so doses would be As Low As Is Reasonably Achievable (ALARA).
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The inspectors also observed the licensee's collection of samples from four air sampling stations.
The inspectors noted that the sampling equipment was operable and in good working order the sampling stations were located as indicated in the ODCM, and good sampling techniques were employed by licensee personnel in collection of the samples.
c.
Conclusions
Based on the above reviews and observations. the inspectors concluded that the licensee had complied with the sampling, analytical and reporting program requirements. the sampling equipment was being well maintained, and that the radiological environmental monitoring program
)
was effectively implemented.
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R1.4 Primary and Secondary Water Chemistry a.
Insoection Scooe (84750)
The' inspectors reviewed implementation of selected elements of the
,
licensee's water chemistry control program for monitoring primary and secondary water quality. The review included examination of program guidance and implementing procedures, and analytical results for selected chemistry parameters.
Those procedures and data were com)ared to specific and programmatic requirements. i.e., the TS required t1e licensee to monitor primary coolant for specific chemistry Jarameters and to implement a program for monitoring secondary water clemistry to inhibit steam generator tube degradation.
b.
Observations and Findinas
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The inspectors reviewed three SAPS (SAPS 400. 401, and 402) and three chemistry procedures (cps 613. 614. and 615) which implemented the licensee's programs for controlling the chemical environment of the primary and secondary plant systems.
The SAPS provided descriptions of l
the chemistry control programs and guidance for conducting chemistry o>erations. The cps provided instructions for implementing the L
clemistry program as described in the SAPS. The cps included provisions for sampling and analyzing reactor coolant for the TS required parameters at the s)ecified frequencies and for implementing, with a few
minor exceptions, t1e Electric Power Research Institute (EPRI)
l guidelines for Pressurized Water Reactor (PWR) primary and secondary l
water chemistry.
The exceptions were made in accordance with guidelines established by the fuel supplier for the plant specific chemistry regimes.
Guidance was also provided for actions to be taken if
,
analytical results exceeded prescribed action limits.
'
The inspectors also reviewed trend plots of analytical results for selected parameters generated during the period January through mid-May 1998. The parameters selected included dissolved oxygen. chloride, fluoride, and dose equivalent iodine-131 in reactor coolant and dissolved oxygen, copper, sodium, and conductivity in secondary systems.
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Those parameters were maintained well within the relevant TS limits and within the EPRI guidelines for power operations.
The licensee discussed with the inspectors the status of their investigation of the elevated release of insoluble particulate radioactive material into the Reactor Coolant System during the cooldown for the October 1997 refueling outage.
The investigation was focusing on the following areas: lithium, hydrogen, and oxygen concentrations and pH of the coolant, oxygen concentration in make-up water, boron
!
injection rate, letdown flow rate to the Chemical and Volume Control System, run time for the reactor coolant pumps, and chemistry controls during power operations.
The licensee indicated that no specific cause i
for the formation of the insoluble particulate chemical compound (s) had been determined but the investigation was continuing with support from vendors and industry research organizations.
c.
Conclusions Based on the above reviews, the inspectors concluded that the licensee's water chemistry control program for monitoring primary and secondary water quality had been implemented in accordance with the TS requirements and the EPRI guidelines for PWR water chemistry.
R8 Miscellaneous RC&P Issues (92904)
R8.1 (Closed) VIO 50-395/96007-09:
failure to follow procedures for contamination control.
During the 1996 refueling outage, inspectors
.
performed independent surveys and detected loose surface contamination greater than procedural limits at the step off pad for the hot machine shop.
The licensee's reply to the Notice of Violation indicated that
,
their corrective actions included decontamination and resurvey of the l
area to ensure that there was no further spread of contamination.
During this inspection the inspectors performed independent surveys for loose surface contamination around five step off pads which were then currently in use.
No loose surface contamination was detected by those surveys.
R8.2 (Closed) VIO S0-395/97013-06:
failure to follow posted radiation protection instructions. On October 19, 1997, the inspectors were conducting a routine plant tour and observed an individual in the Radiation Control Area (RCA) using a chewing tobacco product. The licensee's reply to the Notice of Violation indicated that this event appeared to have been an isolated occurrence by a contractor employee I
and resulted from a lack of familiarity with the restrictions on eating, smoking, and drinking within the RCA.
The licensee's reply also indicated that by March 1.1998, the Station Orientation Training (SOT)
l materials would be amended to better reflect the site policy regarding
!
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this issue.
During this inspection. the inspectors reviewed the Student l
Hand Out for Phase 2 of the SOT which provided training to qualify for I
unescorted access to the RCA.
The inspectors noted that revision number seven of the hand out, dated February 3.1998. specifically states that
"To reduce the possibility of internal exposures from radioactive i
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material, personnel are not permitted to eat, drink, chew, smoke or apply cosmetics while in the RCA."
P1 Conduct of EP Activities Pl.1 Observation Of Emeroency Drill a.
Inspection Scone (71750)
The inspectors observed an after hours emergency drill.
b.
Observations and Findinas On the evening of June 17. the inspectors observed an after hours drill.
The simulated emergency required mobilization and response of onsite and off-site personnel.
The drill involved activation of the emergency response organization and the Emergency Response Facilities (ERFs).
The purpose of the drill was to evaluate the capability of the emergency response organization to respond to an emergency after normal working hours.
The inspectors observed portions of the drill from the simulator control room and the Technical Support Center (TSC).
The inspectors observed response to the simulated plant conditions from the simulator; accident assessment and event classification in the simulator and.TSC: managerial direction and coordination of the emergency response organization:
activation of the TSC: TSC activities: and access control to the site and the protected area.
The inspectors concluded that the licensee demonstrated the ability to respond effectively to an emergency. to recognize and classify emergency action levels, and demonstrated the adequacy and proper utilization of the ERFs.
The insaectors also observed the ability of the staff to formulate and mace Protective Action Recommendations based on plant parameters, surveys and field monitoring information. The inspectors also observed the drill critique in the TSC immediately following termination of the drill.
The critique identified minor deficiencies for resolution.
c.
Conclusions The licensee demonstrated the ability to respond effectively to an emergency, to recognize and classify emergency action levels, and demonstrated the adequacy and proper utilization of the emergency response facilities.
S1 Conduct of Security and Safeguards Activities S1.1 Observation of Security Compensatory Actions l
a.
Inspection Scooe (71750)
The inspectors observed security compensatory actions during the observation of in-plant work activities.
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i b.
Observations and Findinas During observation of CCP work activities. SW intake structure diving I
activities, and actions to provide temporary cooling to the SW building
)
the' inspectors observed security compensatory measures to ensure the
'
appropriate levels of plant security were maintained for specific plant vital areas.
~c.
Conclusions Security compensatory actions during plant work activities were sufficient to ensure that the appropriate level of security for vital-areas was maintained.
V.
Manaaement Meetinas X1 Exit Meeting Summary The inspectors ) resented the inspection results to members of licensee management at t1e conclusion of the inspection on July 8.1998.
The licensee acknowledged the findings presented.
The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary.
No proprietary information was identified.
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PARTIAL LIST OF PERSONS CONTACTED
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Licensee F. Bacon Manager. Chemistry Services L. Blue. Manager. Health Physics l
S. Byrne. General Manager. Nuclear Plant Operations
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R. Clary, Manager. Quality Systems M.- Fowlkes. Manager. Operations L
S. Furstenberg. Manager. Maintenance Services l
D. Lavigne. General Manager. Nuclear Support Services l
G. Moffatt. Manager. Design Engineering
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K. Nettles. General Manager. Strategic Planning and Development L. Hipp. Manager. Nuclear Protection Services A. Rice. Manager. Nuclear Licensing and Operating Experience-
.G. Taylor. Vice President. Nuclear Operations
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R. Waselus. Manager. Systems and Component Engineering R. White. Nuclear Coordinator. South Carolina Public Service Authority B. Williams. General Manager. Engineering Services
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G. Williams. Associate Manager. Operations
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INSPECTION PROCEDURES USED i
IP 37001:
10 CFR 50.59 Safety Evaluation Program L
'IP 37551: Onsite Engineering-IP 40500:
Effectiveness of Licensee Controls in Identifying. Resolving, and Preventing Problems IP 61726:
Surveillance Observations IP 62706:
Maintenance Rule Inspection Procedure
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IP 62707: Maintenance Observations IP 71707:
Plant Operations
IP 71750:
Plant Support Activities L
IP 73756:
Inservice Testing of Pumps and Valves
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IP 84750:
Radioactive Waste Treatment, and Effluent and Environmental l
Monitoring i
IP 92901:
Followup - Plant Operations IP 92902:
Followup - Maintenance IP 92903:
Followup - Engineering IP 92904:
Followup - Plant Support l
ITEMS OPENED AND CLOSED Ooened 50-395/98005-01 IFI review implementation of the Primary Identification Program and resolution of 0A-CAR-91-1 (Section 07.1)
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50-395/98005-02 NCV missed surveillance requirement on accident monitoring system neutron flux instrumentation (Section M3.2)
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50-395/98005-03 NCV Failure to implement ASME Section XI Code requirements for testing Class 1. 2 and 3 Pressure relief devices as required by TS 4.0.5 (Section M8.3)
l Closed l
50-395/96008-01 VIO failuru to follow station housekeeping program procelure (Section 08.1)
50-395/96014-01 VIO failare to follow station housekeeping program procedure (Section 08.1)
l-l 50-395/97005-02 VIO failure to implement effective corrective action for
equipment storage violations (Section 08.1)
50-395/98005-02 NCV missed surveillance requirement on accident l
monitoring system neutron flux instrumentation l
(Section M3.2)
50-395/98006-00 LER missed surveillance on Post Accident Monitoring System neutron flux instrumentation channel check (Section M3.2)
50-395/97011-02 VIO failure to maintain Delta T - TAVE Protection Loop Calibration Procedure (Section M8.1)
l 50-395/97003-00 LER entry into TS 3.0.3 (Section M8.1)
50-395/96015-03 VIO numerous examples of missed technical specification required surveillance tests associated with the turbine and motor driven emergency feedwater pumps
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(Section M8.2)
l 50-395/96010-00 LER missed surveillance on slave relay (K634)
(Section M8.2)
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l 50-395/98001-02 URI Inconsistencies between Code requirements and i
implementing procedures for testing ASME' pressure relief valves (Section M8.3)
50-395/98005-03 NCV Failure to implement ASME Section XI Code requirements for testing Class 1. 2 and 3 Pressure
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relief devices as required by TS 4.0.5 (Section M8.3)
50-395/98W5-00 LER Inadequate ASME Relief Valve Testing concerned the L
failure to upgrade implementing procedures to meet the requirements of OM-1987 Part 1 for testing of Code Class 1. 2. and 3 prersure relief valves (Section M8.4)
50-395/96005-03 URI variances between the FSAR and plant operating practices for the SFP (Section E8.1)
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50-395/96007-09 V10 failure to follow procedures for contamination control (Section R8,1)
50-395/97013-06 VIO failure to follow posted radiation protection instructions (Section R8.2)
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