IR 05000266/1999013
| ML20212K821 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 09/29/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20212K770 | List: |
| References | |
| 50-266-99-13, 50-301-99-13, NUDOCS 9910070059 | |
| Download: ML20212K821 (16) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGIONlli Docket Nos: 50-266; 50-301 License Nos: DPR-24; DPR-27 i I
Report No: 50-266/99013(DRP); 50-301/99013(DRP)
Licensee: Wisconsin Electric Power Company Facility: Point Beach Nuclear Plant, Units 1 & 2 Location: 6610 Nuclear Road Two Rivers, WI 54241 Dates: July 14 through August 30,1999 Inspectors: F. Brown, Senior Resident inspector P. Louden, Resident inspector J. Lara, Senior Resident inspector, Kewaunee l M. Kunowski, Project Engineer Approved by: R. Lanksbury, Chief Reactor Projects Branch 5 Division of Reactor Projects 9910070059 990926 PDR G ADOCK 05000266 PM
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EXECUTIVE SUMMARY Point Beach Nuclear Plant, Units 1 & 2 NRC Inspection Report 50-266/99013(DRP); 50-301/99013(DRP)
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This inspection included aspects of licensee operations, maintenance, engineering, and plant !
support. The report covers a 6-week inspection period by the resident inspector j
- Operations l
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Operators responded well to an instrument air leak effecting Unit 1. (Section 01.1)
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Operators responded well to a power transient caused by an unexpected change in the position of a Unit 2 main turbine govemor valve. Engineering and Operations Department personnel implemented a troubleshooting approach which resulted in a second unexpected load change. A subsequent licensee-imposed limitation on reactor power (while engineers formulated a plan for monitoring the turbine control system during troubleshooting) was appropriate. (Section 01.2)
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The Unit 2 control rod control selector switch was inadvertently mispositioned by a reactor operator. The oversight provided by the duty operating supervisor was not l adequate to identify the problem prior to a failed attempt to move control rod l (Section 01.3) '
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The licensee failed to recognize tht.t a leaking valve was identified as a containment isolation valve in the Fi.ial Safety Analysis Report (FSAR) until questioned by the inspectors and did not take the applicable Technical Specification actions for an inoperable containment isolation valve. The licensee sub,sequently concluded that the Technical Specification limiting condition for operation requirements had not been violated because the FSAR was in error in its classification of the subject valve. The licensee planned to correct the FSAR. Review of the licensee's position will be tracked as an Unresolved Item. (Section O2.1)
Maintenance
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The inspectors identified that air dampers for two emergency diesel generator rooms were propped open during maintenance activities. The licensee's temporary modification process had-not been followed. (Section M2.1)
Enaineering
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In general, the conclusions contained in written licensee operability determinations were technically correct. (Section E1.1)
- During the June 1998 refueling outage, the licensee installed reactor vessel head plugs on Unit 1 that did not conform with the FSAR and did not comply with Section lit of the American Society of Mechanical Engineers boiler and pressure vessel code. The licensee closed the corrective action item for this issue without requesting the required relief the NRC for code noncompliance until August 1999 and without addressing the nonconformance with the FSAR. (Section E1.2)
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Licensee engineers responsible for implementing the.American Society of Mechanical Engineers,Section XI program performed well in taking the initiative to submit a relief request for a Section 111 code noncompliance after the issue had been closed in the licensee's corrective action program. (Section E1.2)
- The licensee was meeting conunitments for testing the reliability of the G-05 station blackout _ gas turbine generator. However, the inspectors identified weaknesses in the formality and quality of the test program documentation, including an unclear designation of what support equipment was subject to reliability goals and requireme nt (Section E3.1)
The licensee's quality verification audit of the design engineering area was an in-depth review, conducted by knowledgeable and experienced personnel. The use on the team of a technical specialist from another utility contributed to the overall effectiveness of the audit. The corrective action documents generated by the team and the conclusions drawn about audit area weaknesses and strengths were appropriate. (Section E7,1)
Plant Sucoort
While the performance of some licensee participants during the annual emergency preparedness exercise was adequate, the overall program performance was good. The scenario was a challenging combination of security and reactor operation events, and insightful self-criticism of the exercise was provided by a group of knowledgeable and experienced controllers and quality assurance personnel. (Section P1.1)
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.s Report Details Summarv of Plant Statu_g Unit 1 began the inspection period at 100 percent power and remained at or near that level for the remainder of the period, except from July 31 to August 2,1999, when power was reduced to about 68 percent for turbine valve testing and replacement of a faulty main turbine electrohydraulic control (EHC) system speed channel car Unit 2 began the inspection period at 100 percent power. On July 25, power was reduced to 99 percent while the cause of an unexpected repositioning of a main turbine govemor valve was investigated. Full power operations were resumed on July 31. Power was reduced to approximately 60 percent on August 29 for quarterly valve test l. Operations 01 Conduct of Operations 01.1 Operator Resoonse to Instrument Air Leak (71707)
On July 21,1999, while conducting routine control room monitoring, the inspectors observed control room senior reactor operators (SROs) respond to an air leak in the Unit 1 facade area. An auxiliary operator was dispatched to the area and determined that an instrument air line associated with the Unit 1 containment purge and exhaust boot seal was leaking. The control room operators entered Abnormal Operating Procedure AOP 5B, ' Loss of Instrument Air," Revision 13. The control room operators then began to assess the air leak and methodically proceeded to determine the most appropriate isolation poin The Duty Shift Superintendent (the lead on-s lift SRO) assessed the air leak and promptly notified Maintenance Department supervision of the condition and arranged for repairs. Maintenance Department personnel repaired the air line leak in about i hou The inspectors determined that the worker who reported the air leak displayed good plant awareness. Likewise, operators proceeded in a deliberate manner to assess the situation and effectively involved Mainbnance Department personnelin a timely resolution of the proble .2 Unexpected Unit 2 Main Turbine Governor Valve Movement
, Inspection Scope (71707)
The inspectors reviewed the operator response to an unexpected movement of a Unit 2 main turbine govemor valv i
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On July 25,1999, the Unit 2 main turbine No. 4 govemor valve unexpectedly moved
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from its normal 32 percent open position to 40 percent open. This movement resulted in reactor thermal output increasing from 1512.5 to 1520.5 megawatts-thermal (the licensed power level is 1518.5 megawatts-thermal steady-state). The reactor operators (ROs) responded promptly and retumed the valve to its original position. The cause of the valve movement was not apparent to the operators and a work order was written for troubleshooting the valve. Initial reviews by Engineering and Operations Department, department personnel indicated s possible fault in the valve position limiter circuitr I Reactor power was reduced to tr9 percent (to help protect against power excursions) l while the problem was reviewed furthe !
A preliminary evaluation by licensee engineers attributed the valve movement to an isolated system perturbation. On July 27, the ROs attempted to place the turbine's EHC system in the turbine impulse pressure control mode (" imp-in"), and to retum Unit 2 to 100 percent power. Again the No. 4 govemor valve unexpectedly repositioned, from 27 percent to about 44 percent open. This resulted in the control rods moving out from i the 220-step position to 228 as the reactor control system attempted to maintain reactor coolant system average temperature within the programmed band. The operators promptly placed the turbine control in manual, repositioned the valve, and stabilized the !
plant, in response to this latest incident, the station operations manager restricted ,
Unit 2 power level to 99 percent, even though hot weather was placing a high demand i on the regional electrical distribution grid, until a more extensive evaluation was complete ,
The power level restriction was removed on August 1 after licensee engineers developed a plan for monitoring EHC system performance. The plan included the installation of a test electronic card and a data recorder at the Unit 2 turbine electrohydraulic panel. The Unit 2 EHC system was operated in " imp-out" mode for the majority of the period. When switching between " imp-in" and " imp-out," for testing minor ,
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governor valve movement was observed, but no other notable load swings occurre Conclusions Operators responded well to a power transient caused by an unexpected change in the position of a Unit 2 main turbine govemor valve. Engineering and Operations Department personnel implemented a troubleshooting apcroach which resulted in a second unexpected load change. A subsequent licensee-imposed limitation on reactor I
power (while engineers formulated a plan for monitoring the turbine control system during troubleshooting) was appropriat .3 Incorrect Manloulation of the Control Rod Selector Switch (71707)
On July 31,1999, the Unit 2 RO attempted to withdraw control bank "D" control rods to increase reactor power. The expected response of control bank "D" was not observed after the control rod withdrawal demand. The licensee reviewed this unexpected reactivity control response, and concluded that the reactor operator had mispositioned the control rod selector switch. The RO had apparently turned the selector switch from
" Auto" to * Bank A" (a clockwise to of one position) instead of from " Auto" to " Manual" (a counter :lockwise tum of one position). The bank "A" demand counter display had
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s changed from Step 228 to Step 230. The control rod bank position had not moved because the control rods were already fully withdrawn to Step 228. The procedure in effect directed a power change using control rods, but did not provide detailed instructions on the use of the control rod selector switch. An SRO, the Duty Operating Supervisor (DOS), had not been close enough to the control panel to identify the incorrect selector switch motion. The DOS was responsib% for providing oversight of reactivity changes, but licensee procedures did not specify the level of oversight expected from the DOS. Specifically, the procedures did not specify whether the DOS was expe :ted to be able to intervene if an error was made during a reactivity chang Operational Status of Facilities and Equipment O2.1 Pressurization of Residual Heat Removal (RHR) System Pioina (Unit 2) (71707) Insoection Scope The inspectors reviewed the licensee's identification and evaluation of the pressurization of the RHR system piping due to leakage past a normally closed valve. The inspectors reviewed the facility's Final Safety Analysis Report (FSAR) and Technical Specifications (T/Ss) as part of this evaluatio Observations and Findinas On July 26,1999, the inspectors found that the licensee had identified possible leakage from the chemical and volume control system to the RHR system through a normally closed valve (CV-133) located inside containment. An RO identified this condition ,
during the completion of mid-shift logs. The operator had identified that the "A" RHR I pump discharge piping pressure was 262 pounds per square inch - gauge, whereas the normal pressure was approximately 25 pounds per square inch - gauge. This condition had been identified on July 20. Subsequent licensee reviews revealed that the normally closed chemical and volume control system valve CV-133 was leaking by its seat due to a slight air leak into the valve's operator. The air was sufficient to keep the valve slightly open. The licensee performed an operability review and did not identify any operability )
concerns or impact on T/S complianc i On July 26, the inspectors reviewed this issue, and identified that Figure 5.2-8 of the ]
FSAR listed CV-133 as a containment isolation valve. As a containment isolation valve, CV-133 would be subject to the operability requirements of T/S 15.3.6. Based upon the FSAR classification, a leak through CV-133 required entry into T/S Limiting Condition for Operation (LCO) Section 15.3.6.A.1.b(2). This LCO required that the inoperable valve l be closed and de-activated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and the affected penetration flow path be verified isolated every 31 days. This issue was discussed with Operations Department personnel. The licensee subsequently concluded that the T/S LCO should have been entered and required actions completed once the condition was identified. The entry into the T/S LCO was not done due to Operations Department personnel not recognizing that T/S requirements were applicable in this system configuration. The licensee repaired the valve operator and restored the valve to operable statu While reviewing this issue for licensee event report applicability, the licensee concluded that CV-133 was not a containment isolation valve, and that T/S 15.3.6 was not applicable. This conclusion was based upon the position that valves in systems with
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accident mitigation functions (valves that allow flow into containment) were excluded from the containment isolation T/S. Using this position, the plant FSAR was in erro The license was planning to correct the FSAR. The inspectors were informed of the licensee's position on this matter at the exit meeting. Review of the licensee's position by the inspectors will be tracked as an Unresolved item ((URI) 50-301/99013-01(DRP)). Conclusions The licensee failed to recognize that a leaking valve was identified as a containment isolation valve in the FSAR until questioned by the inspectors and did not take the applicable T/S actions for an inoperable containment isolation valve. The licensee subsequently concluded that T/S LCO requirements had not been violated because the FSAR was in error in its classification of the subject valve. The licensee planned to correct the FSAR. Review of the licensee's position by the inspectors will be tracked as a UR Operations Procedures and Documentation
.- 03.1 FSAR Discreoancies (71707)
The inspectors reviewed the licensee's Emergency Operating Procedure (EOP)-1.3,
" Transfer to Containment Sump Recirculation," Revision 20, and FSAR Section 6,
" Changeover from injection Phase to Recirculation Phase," to verify that the FSAR accurately described the changeover from Si to the 'long-term recirculation mode of operation. During the review, the inspectors noted that EOP-1.3 required operators to rJign the "A" train of Si for containment sump recirculation followed by the "B" trai However, in FSAR Section 6.2, the train order was reversed, with the "B" train aligned before the "A" trai Additionally, the inspectors identified that the changeover from injection to recirculation required local operator actions in the auxiliary building whereas the FSAR implied that the changeover to recirculation was performed through manual actions in the control room. These minor discrepancies were discussed with the licensee representatives who initiated FSAR changes to reflect the actual practice .2 Operator Aides and Temoorarv information (71707) i During routine tours of the plant, the inspectors observed that placards containing information on operating plant equipment, including valves in the safety-related service l water (SW) system, were posted at various locations. Some of these placards were l marked as being controlled, but many others were not. This issue was similar to an '
observation that the inspectors had shared with the Operations Department in early 1998. . The licensee documented several of those most recent examples of inspector- ;
observed placards in CR 99-2019. The licensee found that an open corrective action item (CR 08-0824, Action Number 4) identified the need to establish a control program ,
for permanently installed informational plaques. The inspectors were concemed that j this CR had been initiated on March 2,1998, but as of August 24,1999, there was no ;
scheduled commitment to actually implement such a program (including the actual inventory of plaques in the field). However, the inspectors did not identify any instances whero the information on the plaques conflicted with information contained in approved equipment operating procedure i
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11. Maintenas,ce
. M1 - Conduct of Maintaance M1.1 Surveillance Testina (61726)
The inspectors observed non-licensed operators perform activities specified in two surveillance test procedures: Inservice Test (IT) 380, *[ Containment] Purge Valve Air
' System Check Valve," (Quarterly), Unit 1, Revision 6, and IT 15, *[ Control Room Ventilation) Chill Water Pumps and Valyss, (Quarterfy)," Revision 1 While conducting activities in accordance with P'rocedure JT 15, the operators noted a minor discrepancy between the procedure data sheets and the related data screens on the computer-based pump motor monitoring instrument. The operators stated this discrepancy would be communicated to their supervisor for correctio M2 Maintenance'and Material Condition of Facilities and Equipment M2.1 Imoronerly Controlled Temoorary MM&mtion to Venti!=tiorQamoers (62707)
During tours of the auxiliary, turbine, and facade buildings, the inspectors observed severalleaking components. The leaking components were manifested by steam leaks near bolts at the base of the Unit 2 "A" and "C" heater drain pumps and relatively large boric acid crystal accumulations on the Unit 2 letdown isc!ation valve (2GS-GW738) and the Unit 1 boric acid transfer pumps discharge motor-operated valve (1CV-350). The leaks did not affect system opert bility and work orders had been written by the licensee for repai On August 9,1999, the inspectors observed that the safety-related intake ventilation dampers for G01 and G02 emergency diesel generators were propped open (the conservative position) with pieces of wood and metal. This configuration was discussed with the ventilation system engineer who stated that he was unaware that the dampers were being blocked open but would review it further. Subsequently, the engineer stated that he observed the dampers and had the maintenance group that was working in the area return the dampers to the normally closed position. The dampers had been blocked open periodically over the past week, with the knowledge of control room operators, to a.Ilow cooler outdoor air into the diesel rooms and to allow passage of a welding cable during maintenance work in the area. The engineer wrote CR 99-1929 to document that the dampers had been blocked open without following all of the requirements of Nuclear Power Business Unit Procedure (NP) 7.3.1, " Temporary Modific'ations," Revision 10. The failure to follow NP 7.3.1, was contrary to 10 CFR Part 50, Appendix B, Criterion V, " Instructions, Procedures, and Drawings,"
which required that this procedure be adhered to, constituted a violation of minor significance, and was not subject to formal enforcement action. The inspectors were concemed that this failure to follow the temporary modification process was not an isolated occurrence. An additional example was described in CR 99-202 y ._ .
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111. Engineering
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E1 .. Conduct of Engineering'
'E1.'1' Operabildy Determinations (ODs)
. a.- Inswinn Scooe (3755L71707)
The inspectors reviewed plant equipment status, licensee CRs, and licensee ODs to
. determine whether plant staff accurately identified and tracked nonconformances and dagraded equipment, b Observations and Findinas -
With the exception described in Section E1.2 of this report, the inspectors found the -
licensee's ODs to be technically correct. However, the inspectors identified one case l (CR 99-1742) in which a condition outside of the analyzed system configuration (Sl piping in a different configuration than was originally analyzed for American Society of Mechanical Engineers (ASME) Code allowable stresses) did not receive a prompt written OD until plant management and the inspectors questioned the engineering staff about the potential for a nonconformance. Based upon subsequent written analysis, the .
effected system piping was found to be fully operable and conforming. In two other examples (CR 98-0564 and CR 99-1972), the written ODs did not appear to address the concems raised in the associated CRs. The licensee reviewed each of these ODs, and
. provided the inspectors additional information on the plant specific licensing basis, and how the licensing basb was satisfie j Conclusions In general, the conclusions contained in written licensee ODs were technically correc E1.2 Head Closure Plua Code Nonconformance Insoection Scope (37751. 92903)
The inspectors reviewed the circumstances associated with a licensee request for relief from ASME, Section lil, code requirement . Observ.ations and Findinas .
On August 16,1999, the licensee requested relief from ASME, Section lil, Article N-310, requirements for the material composition of threaded plugs installed in the reactor
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vessel heads of Units 1 and 2. The threaded plug material had been identified by the licensee'as different than that specified on the design drawings. The docketed relief request (reference No. NPL 99-0426) described the material discrepancy adequately, and the inspectors did not identify an operability concem with the material issue.
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- The inspectors were concemed, however, that four of the head plugs covered by the i relief request had been installed in Unit 1 during the most recent refueling outage
' (June 1998). These four plugs installed in the reactor vessel head did not conform with
- the material specifications listed in Table 4.2-1 of the FSAR and did not comply with the
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referenced sec ion of ASME Section Ill. This material discrepancy had been entered into the corrective action program (CR 98-1259), but had been closed based upon a conclusion that the material was technically acceptable for use. .However, the conformance issues with the FSAR and ASME Section til had not been addressed within the corrective action program. The August 16,1999, relief request had been submitted because of excellent personalinitiative on the part of the licensee engineers -
responsibb for implementing the ASME Section XI program. Following the submittal of the relief request itself, CR 98-1259 was reopened to track receipt of the NRC response; however, the nonconformance with the FSAR and ASME Section lll was still not documente The inspectors discussed the above concern with the licensee. The licensee initiated CR 99-2021 to track the nonconformace and to determine why the need to address ASME Section lli and FSAR compliance was not recognized in June 1998 and to develop corrective actions to prevent recorrence. Reactor vessel components were required to conform tv ASME Section til by 10 CFR 50.55a(a)(2) un!ess NRC-approved relief was obtained in accordance with 10 CFR 50.55a(a)(3). The licensee installed reactor vessel head plegs that did not conform to the ASME Section lli material requirements. The fai!ure to reqqcst NRC-approved relief of the identified ASME Section 111 noncompliance wes a violation of 10 CFR 50.55a(a)(2). Tt.s Severity Level IV violation is being treated as a Non-Cited Violation ((NCV) 50-266/99013-02(DRP)), consistent with Appendix C of the NRC Enforcement Polic Conclusions During the June 1998 refueling outage, the licensee installed reactor vessel head plugs on Unit 1 that did not conform with the FSAR and did not comply with ASME Section Il The licensee closed the corrective action item for this issue without requesting the required relief from the NRC for code noncompliance until August 1999 and without addressing the nonconformance with the FSA Licensee engineers rusponsible for implementing the ASME Section XI program performed well in taking the initiative to submit a relief request for a Section 111 code noncompliance after the issue had been closed in the licensee's corrective action progra E1.3 Auxiliary Feedwater (AFW) Pump Qualifications (37551)
The inspectors reviewed the licensee's program for ensuring that safety-related SW loads were not negatively affected by the accumulation of zebra mussels or sitt. The results of this review were discussed in Section E2.2 of Inspection Report 50-266/99008(DRP); 50-301/99008(DRP). One issue not resolved during that program review was whether the licensee's program for flushing the safety-related SW supply lines to the AFW pumps was adequate to ensure AFW pump operabilit The licensea engineering staff had been reviewing the inspectors' concerns and had gathered data that indicated the sitt build-up .n the SW suction lines between flushes was within the analyzed acceptable limit. However, the licensee had not yet been able to provide assurance that the silt in the SW system would not effect AFW pump operability. The licensee system engineer had contacted the pump vendor, and the q vendor had requested additionalinformation on sitt particle size distribution. The 1
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inspectors were concemed that the licensee had not addressed the impact of sitt in the safety-related SW supply to the AFW pumps (until prompted by inspector questions)
when establishing a flushing program for the SW supply lines. The review of the "
licensee's evaluation of SW system sitt on AFW pump operability will be tracked as an Inspection Follow-up Item ((IFI) 50-266/99013-03(DRP); 50-301/99013-03(DRP)).
E3 Engineering Procedures and Documentation E Reliah@tv Data for Gas Turbine Generator (G-05) ; inspection Scooe (71707. 37551. 62707)
.The inspectors reviewed the program for establishing G-05 reliability data to determine whether a Duty Shift Superintendent's plan to take credit for operating G-05 as a peaking unit in lieu of performing a periodic test was appropriat I Observations and Findinos The licensee takes credit for G-05 as an alternate source of power in the station blackout analysis required by 10 CFR 50.63. In order to take credit for G-05, the licensee is required to maintain a 95 percent reliability goal for this nonsafety-related power source. During a station blackout, G-05 will only start if its auxiliary support buses are powered by a small diesel generator (G-501). When operated as a peaking unit, G-05's auxiliary support buses are powered from offsite power source The inspectors discussed the program for gathering reliability data on G-05 with Operations Department personnel and the system engineer. Based upon these discussions, the inspectors made the following observations:
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The program document for G-05 reliability, Site Engineering Manual (SEM) 5.6, did not specify what equipment was included in the target reliability goal. For example, SEM 5.6 did not specify whether G-05 needed to be started using G-501 (as would be the case in a station blackout) for purposes of establishing !
reliabilit The program outlined in SEM 5 ~ did not specify how often to test G-05, or what procedures to use in testing G-06 for the purpose of gathering reliability dat Based upon the information provided by the licensee to the inspectors, the Operations Department was conservatively performing periodic tests of G-05 and G-501 more often than the quarterly frequency specified in correspondence from Wisconsin Electric to the NRC. However, the inspectors were concerned that the test frequency and methodology was based on verbal direction from the Engineering Department to the Operations Departmen Under the licensee's current test program, poor reliability of G-501 could be masked by successful starts of G-05 using offsite power. This would not be consistent with the intent of the required station blackout analysis. This concern was mitigated by the inspectors' observations that the licensee typically responded in an aggressive manner to any problem with G-05 or its auxiliary support systems, i
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q The inspectors discussed their concerns with the licensee. The licensee initiated CR 99-2056 to track engineering review of these concern c. Conclusions The licensee appeared to be meeting commitments for testing the reliability of the G-05 station blackout gas turbine generator. However, the inspectors identified weaknesses in the formality and quality of the test program documentation, including an unclear designation of what support equipment was subject to reliability goals and requirement E7 Quality Assurance in Engineering Activities E7.1 Audit of Desian Enaineerina Inspection Scone (37551)
The inspectors reviewed the report of a licensee quality verification (quality assurance)
audit of design engineering. In addition, the inspectors interviewed the audit team leader and reviewed the audit checklist and draft result Observations and Findinas The audit (Number A-P-99-08) was conducted from June 14 to July 1,1999, to determine if the modification, calculation, and design specification processes at Point Beach were effectively implemented and were in accordance with regulatory j requirements and site procedures. The audil also included an assessment of the effectiveness of corrective actions for previous aedit findings in this area and a review of the recently revised temporary modification procedure (NP 7.3.1," Temporary Modification") to determine if deficiencies previously identified by quality verification department personnel had been addresse The audit was conducted by appropriately trained and experienced personnel, including a technical specialist from the North Anna plant in Virginia. As part of the audit, numerous modification packages, calculations, corrective action documents, and design specifications were reviewed. Independent verifications of selected calculations were also conducted. The use of the specialist from another utility by the licensee added to the overall effectiveness of the audi The audit team concluded that the modification, calculation, and design specification processes at Point Beach were adequate and had improved since the last quality verification audit of this area, performed in mid-1998. Notwithstanding this improvement, the audit team identified four weaknesses which required continued management attention to resolve. These weaknesses were in the areas of:
1) qualifications of engineering software used for calculations,2) licensee review of vendor calculations, 3) translation of design information into modifications developed on an " emergency" basis, and 4) control of engineering specifications. Appropriate corrective action documents (CRs) were generated by the audit team members for these weaknesses and other less significant prcblems identified during the audit. The auditors also identified two strengths in the design engineering area: 1) the engineering assurance function was providing critical reviews to engineering department management of engineering department products and 2) the supervisor responsible for
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the calculation process demonstrated owr ership of the process and was practice in resolving deficiencies in the process. Regarding the revision to the temporary q modification procedure, the audit team concluded that it corrected the previously l identified deficiencies and, if effectively implemented, should improve the temporary l
modification process at the statio I Conclusions l
The licensee's quality verification audit of the design engineering area was an in-depth l review, conducted by knowledgeable and experienced personnel. The use on the team of a technical specialist from another utility contributed to the overall effectiveness of the audit. The corrective action documents generated by the team and the conclusions
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drawn about audit area weaknesses and strengths appeared appropriat IV. Plant Support R1 Radiological Protection and Chemistry Controls R1.1 Radioloalcal Work Practices (71750)
While observing maintenance and modification activities associated with the fuel transfer cart, the inspectors observed a lead health physics technician leaning into a posted contaminated work area. The technician was not wearing anti-contamination clothing. The technician's hand was observed to be in close proximity to a worker wearing full anti-contamination clothing in the work area. The inspectors reviewed Health Physics Manual Procedure HP 3.2, " Radiological Labeling, Posting and Barricading Requirements," Revision 31, and concluded that such a practice was not specifically prohibited as long as an individual contacted only air within the posted are Notwithstanding the licensee's procedural guidance, the inspectors considered the observed performance to be a poor radiological work practice, and communicated the concern to the licensee. The observation was acknowledged by the hcense :
P1 Conduct of Emergency Planning Activities P1.1 Annual Emeraency Plannina Exercise Inspection Scope (71750) i The inspectors observed, from the simulator control room and the technical support and operations support centers, the licensee's conduct of its annual emergency preparedness exercise on August 5,1999, and also observed the critique in the simulator control room immediately after the exercise and two subsequent meetings of the controllers to formulate the written critiqu Observations and Findinas The inspectors determined that while the performance of some of the exercise participants was adequate, the overall performance of the station was goo Performance problems observed during the exercise included simulator control room and technical support center personnel being slow to realize that fire alarms in the
7 auxiliary building were caused by an intersystem loss of coolant accident, responsibility for personnel assembly and accountability prior to technical support center activation was unclear, and emergency operations facility personnel were slow to recognize a shift in wind direction that warranted protective action recommendation On the positive side, the scenario was a challenging combination of security and reactor
, operation events, and insightful self-criticism of the drill was provided by a cadre of knowledgeable and experienced controllers and quality assurance personnel. The inspectors observed that the operations support center re-entry team coordinator, an experienced SRO, provided valuable information to technical support center personnel during periodic reassessments of station priorities for dealing with the simulated emergency. In addition, the lead controller in the simulator control room, who was an SRO and the assistant manager of operations, provided valuable perspectives to {'
sithulator control room personnel on their performance and the possible problems that personnel in the other emergency response facilities must contend with. Condition I reports were written for the major problems identified during the exercis ! Conclusions While the performance of some licensee participants during the annual emergency preparedness exercise was adequate, the overall program performance was good. The scenario was a challenging combination of security and reactor operation events, and insightful self-criticism of the exercise was provided by a group of knowledgeable and experienced controllers and quality assurance personne V. Manaaement Meetinas
' X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on August 30,1999. 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 identifie .
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PARTIAL LIST OF PERSONS CONTACTED Licensee A. J. Cayla, Regulatory Services and Licensing Managsr R. P. Farrell, Radiation Protection Manger !
V. M. Kaminiskas, Maintenance Manager j R. G. Mende, Plant Manager l B. J. O'Grady, Operations Manger l C. R. Peterson, Director of Engineering I M. E. Reddemann, Site Vice President 1 J. G. Schweitzer, System Engineering Manager f
INSPECTION PROCEDURES USED IP 37551: Onsite Engineering IP 61726: Surveillance Observations IP 62707: Maintenance Observations IP 71707: Plant Operations IP 71750: Plant Support Activities IP 92903: Followup - Engineering
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ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-301/99013-01(DRP) URI Containment isolation valve classification 50-266/99013-02(DRP) NCV Code nonconformance of head closure plug 50-266/99013-03(DRP) IFl Licensee's evaluation of SW system silt on AFW 50-301/99013-03(DRP) Pump operability l Closed !
50-266/99013-02(DRP) NCV Code nonconformance of head closure plug
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Discussed
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None
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LIST OF ACRONYMS USED AFW Auxiliary Feedwater ASME American Society of Mechanical Engineers CFR Code of Federal Regulations CR Condition Report DOS Duty Operating Supervisor DRP Division of Reactor Projects EHC Electrohydraulic Control FSAR Final Safety Analysis Report IFl Inspection Follow-up Item IP inspection Procedure IT Inservice Test LCO Limiting Condition for Operation NCV Non-Cited Violation NP~ Nuclear Power Business Unit ,
NRC Nuclear Regulatory Commission j OD Operability Determination i RHR Residual Heat Removal I RO Reactor Operator SEM Site Engineering Manual ,
SI Safety injection
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SRO Senior Reactor Operator SW Service Water T/S Technical Specification URI Unresolved item i
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