IR 05000482/1999013

Insp Rept 50-482/99-13 on 990725-0904.Noncited Violations Noted.Major Areas Inspected:Operations,Maint,Engineering & Plant Support

ML20212G572
Person / Time
Site:	Wolf Creek
Issue date:	09/20/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20212G568	List: IR 05000482/1999013 ML20212G564
References
50-482-99-13, NUDOCS 9909300040
Download: ML20212G572 (25)

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ENCLOSURE l

l U.S. NUCLEAR REGULATORY COMMISSION j REGION IV l

Docket No.: 50-482 License No.: NPF-42 Report No.: 50-482/99-13

Licensee

Wolf Creek Nuclear Operating Corporation -

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Facility: Wolf Creek Generating Station i

Location: 1550 Oxen Lane, NE - '

Burlington, Kansas 66839 Dates: July 25 through September 4,1999 Inspectors: F. L. Brush, Senior Resident inspector R. V. Azua, Project Engineer, Branch B _ .

C. A. Clark, Reactor inspector, Engineering and Maintenance Branch l J. D. Hanna, Resident inspector, Callaway

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R. A. Kopriva, Senior Project Engineer, Branch B M. F. Runyan, Senior Reactor inspector, Engineering and Maintenance j L Branch i

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R. G. Quirk, Beckman & Associates, Inc., Contractor

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L Approved By: William D. Johnson, Chief, Project Branch B j I

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- ATTACHMENT: Supplemental Information

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l 990930004o DR 990920 i o ADOCK 05000482 l PDR

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EXECUTIVE SUMMARY Wolf Creek Generating Station NRC Inspection Report No. 50-482/99-13 I

k Operations

The licensee's response to a plant trip from 100 percent power and actions during the subsequent plant startup were very good. The shift supervisors provided very good oversight. The shift supervisors kept the control room distractions at a minimum following the trip and during the startup. The operators used three-way communications and peer checking at all times (Section 04.1).

Maintenance

_The licensee reacted promptly to a piping failure at another facility. The licensee's investigation was thorough. The wall thickness at a 45 degree elbow for a moisture separator reheater drain tank to feedwater heater pipe was less than the required amount. The licensee replaced the piping and continued to investigate additional areas of potential problems (Section M2.1).

Enaineerina

The licensee failed to verify the adequacy of designs on three occasions. Specifically, calculations did not adequately evaluate the acceptability of circuit cable lengths, did not provide an adequate analysis of the 120 Vac feeders and control circuits, and underestimated battery loads under accident conditions. The inspectors determined that the identified discrepancies had not caused a safety concern and that the licensee had acceptably corrected the affected design calculations. This was a violation of .

10 CFR Part 50, Appendix B, Criterion Ill. This Severity Level IV violation with three I examples is being treated as a noncited violation (50-482/9913-01), consistent with Appendix C of the NRC Enforcement Policy. The violation is in the licensee's corrective action program as the Performance Improvement Requests listed in Sections E8.1, E8.2, and E . The failure to correctly translate design input in plant procedures, specifically not having correct Class 1E battery service load profiles in the battery service test procedure, was identified as a violation of 10 CFR Part 50, Appendix B, Criterion V. This Severity Level IV violation is being treated as a noncited violation (50-482/9913-02), consistent with Appendix C of the NRC Enforcement Policy. The violation is in the licensee's corrective action program as Performance Improvement Request 97-4185 (Section E8.7).

. The licensee failed to accurately update the Updated Safety Analysis Report periodically to assure that the information therein contains the latest material developed. The original 11 Updated Safety Analysis Report discrepancies identified by the NRC constituted a violation of 10 CFR Part 50.71(e) for failure to properly update the Updated Safety Analysis Report. This Severity Level IV violation is being treated as a noncited violation (50-482/9913-03), consistent with Appendix C of the NRC Enforcement Polic ,

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-2-The Updated Safety Ar.alysis Report upgrade program had satisfactorily addressed the question of overall Updated Safety Analysis Report fidelity. The violation is in the ,

licensee's corrective action program as Performance improvement Requests 97-0547, 97-3823, 97-3958, 97-4018, 97-4052, 97-4126, 97-4179, 97-4190, 98-0062, and 98- {

0618 (Section E8.9).

' Plant Support

. The prejob brief and health physics coverage during a containment entry at 100 pNcont power were thorough. Personnel exhibited good ALARA practices while in the containment (Section R4.1).

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Report Details

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Summary of Plant Status The plant began the report period on July 25,1999, at 100 percent power. On August 5,1999, ,

the plant tripped when a feedwater isolation valve faileo closed. On August 6, the licensee started up the plant and paralleled the generator to the grid to end the forced outage. On August 7, the licensee returned the plant to 100 percent power. On August 27, the licensee reduced plant power to 91 percent to replace a portion of a moisture separator reheater drain 1 tank to feedwater heater pipe. On August 29, the licensee returned the plant to 100 percent j power. The plant operated at essentially 100 percent power the remainder of the report perio l 1. Operations j

01 Conduct of Operations O1.1 General Comments (71707)

The inspectors conducted frequent reviews of ongoing plant operations. In general, the conduct of operations was professional and safety conscious. Plant status, operating problems, and work plans were appropriately addressed during daily turnover and plan-of-the-day meetings. Plant testing and maintenance requiring control room coordination were properly controlled. The inspectors observed several shift turnovers and noted no problem Operators were knowledgeable of plant conditions, the status of annunciators, any Technical Specification action statements that were in effect, and evolutions in progres Control room communication was accurate as evidenced by frequent use of three-part communications and verbatim repeat-backs. The inspectors verified that contro! room log keeping was in accordance with the licensee's administrative procedure Operational Status of Facilities and Equipment O Enaineered Safety Feature System Walkdowns (71707)

The inspectors walked down accessible portions of the 'ollowing engineered safety features and vital systems:

. Chemical and Volume Control System;

. - Emergency Diesel Generators A and B; and

. Auxiliary Feedwater System Trains A, B, and Equipment operability, material condition, and housekeeping were acceptable,

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2-04 Operator Knowledge and Performance 04.1 Plant Trio From 100 Percent Power and Subsecuent Plant Startuo Inspection Scope (7'l707. 9370_2_)

The inspectors evaluated the operators' response to an automatic plant trip from 100 percent power and actions during the subsequent startup and return to power, Observations and Findinas The inspectors observed the control room staff's actions following a plant trip from 100 percent power. The reactor tripped on Lo Lo steam generator water level when steam generator feedwater regulating Valve D closed due to a failed circuit card. All safety systems responded as required. The licensee replaced the failed circuit card and restored the feedwater regulating valve to operatio The shift supervisor provided very good supervisory oversight following the trip. Control room and plant operators used three-way communications and peer checking. The operators followed the appropriate emergency response procedures. The shift supervisor limited the number of personnelin the control room to an appropriate leve The control room operaturs were not subject to external distractions. The operators ,

immediately recognized that a main condenser steam dump valve had failed open and i isolated the valve to prevent excessive plant cool down. The operators stabilized the plant in Mode 3 without any significant complications. The shift supervisor made the appropriate offsite notifications within the required time I The inspectors also observed the control room staff during portions of the plant startup. l

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The shift supervisors provided very good supervisory oversight. The shift supervisors and supervising operators conducted detailed briefings before major evolutions. Control room and plant operators participated extensively in the briefings. The operators used three-way communications and peer checking when relaying information and manipulating components. The licensee returned the plant to full power operation with very few complication ,

) Conclusions i The inspectors concluded that the licensee's response to a plant trip from 100 percent power and actions during the subsequent plant startup were very good. The shift supervisors provided very good oversight. The shift supervisors kept the control room distractions at a minimum following the trip and during the startup. The operators used three-way communications and peer checking at all times. The inspectors did not hhve any concern E

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3-11. Maintenance M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Review of Material Condition Inspection Scope (62707)

The inspectors reviewed the licensee's actions when a portion of a moisture separator reheater drain tank to feedwater heater pipe was replace Observations and Findinos The licensee replaced a portion of a moisture separator reheater drain tank to feedwater heater pipe due to a failure of the same pipe at another facility. The inspectors I

discussed the results of nondestructive testing of the piping that was performed before and after the replacement. The licensee took the following actions on the piece of pipe that was replaced:

- Performed a radiographic examination of the piping prior to replacement. The pipe was approximately 0.100 to 0.120 inches thick at its thinnest point at a 45 degree elbow. The minimum required wall thickness was 0.109 inche . Performed an ultrasonic examination of the 45 degree elbow following replacement. This examination confirmed the radiographic examination measurement Compared the wear to an industry computer model. Wear of the piping was in close agreement with the model, except for the 45 degree elbow that had the wall thinning. Yhe model predicted 0.8 inches of wear, but the actual wear was double that amount. An industry representative was onsite coordinating the

' investigatio Tested the metallurgical content to determine the Chromium content of the pip l The amount of Chromium affects the erosion / corrosion wear of the pipe. The ]

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licensee sent the pipe to an offsite laboratory for further analysi The licensee identified an additional 11 places for radiographic examination and expected to identify a total of 15 to 18 places. The licensee planned to start additional examinations on October 11,199 I Conclusions )

I The inspectors concluded that the licensee reacted promptly to a piping failure at another facility. The licensee's investigation was thorough. The wall thickness at a i 45 degree elbow for a moisture separator reheater drain tank to feedwater heater pipe !

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was less than the required amount. The licensee replaced the piping and continued to j investigate additional areas of potential problem I M8 Miscellaneous Maintenance issues (92902) k l

M8.1 (Closed) Violation 50-482/9713-05: unauthorized work performed and not documente l The licensee identified two examples of unauthorized and undocumented welding- 1 related activities. Example A.1 of the violation noted that on March 28,1996, an unauthorized alteration (grinding) of vendor-installed welds on a safety-related essential service water system elbow on the Train A component cooling water room cooler occurred as documented in Problerr Identification Report 97-0469. Example A.2 of the violation noted that, on approximatoly August 29,1996, unauthorized base metal weld repair (plug welding) was performed on a safety-related normal charging pump (

baseplate as documented in Performance improvement Request 97-0470. Tne welds identified in the two examples of this violation were reexamined and found satisfactor l

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The inspectors reviewed Performance improvement Requests 97-0469 and -0470, Self l Assessment Report SEL 97-031, " Review of Welding Reinstatement Plan," Self )

Assessment Report SEL 98-013," Wolf Creek Nuclear Operating Corporation Welding Program," and other related documents. As part of the licensee " Weld Recovery Plan" for this violation, welders were retrained on how to control and perform grinding and welding activities at the licensee's facility. The inspectors considered appropriate the l licensee's corrective actions designed to prevent recurrence of the violatio M8.2 (Closed) Violation 50-482/9713-06: welders failed to return unused weld filler material to i the place of issue. During February and March 1997, the licensee's staff discovered 5 stainless steel welding rods and 10 E7018 covered electrodes in the weld shop are The licensee concluded that the welders had not returned unused weld filler materials to the point of issue and documented this incident in Performance Improvement Request 97-065 The inspectors reviewed Performance Improvement Request 97-0652. As part of the licensee's " Weld Recovery Plan" for this violation, welders were retrained on how to control and retum unused weld filler material. The inspectors considered the licensee's corrective actions for this issue to be appropriat ,

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M8.3 (C_losed) Inspection Followuo item 50-482/9717-01: review performance improvement requests on maintenance rule program implementation. In October 1997, inspectors ,

established that the apparent reason for the lack of maintenance rule program fctmal training for the integrated plant scheduling organization's staff was because of the changing character of the licensee's program at that time. The changing character of the program was the result of problems encountered duiing program implementation and subsequent changes made to resolve the problems encountered. For example, it ;

was noted that from January 1 to October 31,1997, the licensee's staff had written 26 performance improvement requests on programmatic problems related to implementation of the maintenance rule. The licensee's staff documented the identification of this adverse trend in Performance improvement Request 97-020 .

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Subsequently, the licensee had written other performance improvement requests on maintenance rule problems such as Performance Improvement Request 97-2305, which identified the failure to implement a structural monitoring program. In addition, the licensee wrote Performance improvement Request 97-2032 on a trend of not properly evaluating maintenance preventable functional failures. These failures to implement the maintenance rule program were licensee-identified, and the licensee had implemented corrective actions to address the problem During this inspection, the inspectors reviewed applicable maintenance rule program documents and discussed the current maintenance rule program, documents, and activities with members of the licensee's maintenance rule . staff. The inspectors reviewed NRC Inspection Report 50-482/98-05 and Procedure AP 23M-001, " Wolf Creek Generating Station Maintenance Rule Program," Revisica 3, the licensee's current corrective actions implemented per Performance improvement Requests 97-0204, -2032, and -2305 and noted the following:

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• Corrective actions were implemented by the licensee's staff to review past structure, system, or component failures to identify functional failures and any associated maintenance preventable functional failures that were not previously identifie * The following documents had been issued to implement a structural monitoring program:

Procedure Al-23M-007," Engineering Structural Monitoring Walkdowns,"

Revision 0, and Engineering Desktop Instruction (EDI) 23M-010," Determination of Structures, Systems and Components Within the Scope of The Maintenance Rule," Revision 1; EDI 23M-020," Determining the Safety Significance of Structures, Systems and Components Within the Scope of The Maintenance Rule;" Revision 1, and EDI 23M-030," Establishing Performance Criteria for Structures, Systems and Components Within the Scope of The Maintenance Rule," Revision * The initial maintenance rule data base had been revised and upgrade * Additional maintenance rule problems were still being encountered as the maintenance rule program was integrated into other site programs, and these new problems were being addressed through the licensee's corrective action progra * initial meintenance rule formal training has been provided for the integrated plant scheduling organization. Additional training for this group was scheduled for a later date, after future plan changes have been implemented into the maintenance rule progra p*

4-6-The inspectors concluded that the corrective actions irnplemented in response to the violation and Performance improvement Requests 97-0204, -2032, and -2305 were satisfactory to resolve this ite lil. Enaineerina E8 Miscellaneous Engineering issues (92903)

E (Closed) Unresolved item 50-482/97201-04: motor control center circuit length. This issue involved nonconservative assumptions used to determine maximum allowable safety-related motor control center control circuit cable lengths contained in Calculation E-B-10. " Determine Voltage Drop in MCC Control Circuits," Revision 3. For example, all auxiliary loads were erroneously assumed to be at the location of the control power transformers, in response, the licensee initiated Performance Improvement Request 97-4159. As corrective action, the licensee issued Calculation XX-E-012, " Safety-Related MCC Control Circuit Wire Lengths," Revision 0, which superseded the safety-related circuit sections in Calculation E-B-1 Performance Improvement Request 97-4159 also resulted in Design Change Package 07741, " Containment Cooler Fans A & C Control Circuit Rewiring to Reduce Voltage Drop," Revision 0, which was issued to increase the voltage drop margins. The licensee's representative stated that this change was implemented to enhance an already acceptable margin. The inspectors agreed with this characterizatio Calculation XX E-012 provided some additional justification for the nonconservative calculational assumptions, but the inspectors observed that the assumptions were nevertheless still inadequate to justify some circuits which had very little margin between calculated maximum allowable and actual cable lengths. A licensee electrical design engineer agreed with the inspectors' observation and on August 5,1999, initiated Performance Improvement Request 99-2678 to further evaluate several circuits to ensure that the end devices would operate properly. Based on additional information provided in Performance improvement Request 99-2678, the inspectors determined that the low-margin circuits were acceptable. The licensee's engineer stated that the calculation would be modified to include this additional informatio CFR Part 50, Appendix B, Criterion Ill, " Design Control," requires, in part, that design control measures shall provide for verifying or checking the adequacy of the desig The inadequacies in Calculations E-B 10 and XX-E-012 to evaluate the acceptability of circuit cable lengths reflected a failure to verify the adequacy of the dasign and were identified as a violation of 10 CFR Part 50, Appendix B, Criterion Ill. This Severity Level IV violation is being treated as a noncited violation (50-482/9913-01), consistent with Appendix C of the NRC Enforcement Policy. The violation is in the licensee's corrective action program as Performance Improvement Requests 97-4159 and 99 2673. The inspectors determined that the identified discrepancies had not L

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caused a safety concern and that the licensee had acceptably corrected or

. had plans to correct the affected design calculations (Sections E8.1, E8.2, and E8.3).

E8.2 (Closed) Insoection Followuo item 50-482/97201-05: 120 Vac short circuit and voltage drop analysis. This issue involved the lack of an analysis showing that 120 Vac feeders and control circuits were protected during faulted conditions, as well as the lack of an }

analysis demonstrating adequate terminal voltages for 120/208 V safety-related load j in response, the licensee completed the following new analyses:

Calculations RL-E-00, " Class 1E Vac Voltage Drop to Loads Powered from Panel RLO11," Revision 0;

NN-E-002, " Class 1E 120 Vac Instrument Power Distribution System (NN ,

System) Voltage Drop and Short Circuit Protection," Revision 0; and

• RP-E-003," Class 1E 120 Vac Voltage Drop and Cable Protection for Loads Powered from Panet RP332 and RP 333," Revision .

I During development of the calculation revisions, the licensee issued Performance j Improvement Request 97-4041, which identified several potential deficiencies, that were '

subsequently assigned to Performance Improvement Requests 98-1170,98-1985, 98-1986,98-2632,98-2633,98-2634,98-3445,99-0924,99-0927, and 99-0928 and Action Request 28151/2815 The inspectors reviewed the above-listed performance improvement requests, which addressed problems such as minor under- and over-voltage situations and failure to ensure short circuit protection under extreme design basis conditions of grid voltage and environmental conditions. The inspectors determined that none of the problems impacted operability, but that some of the problems did warrant further investigation and could potentiaily require plant modifications for margin improvement. The licensee's electrical design supervisor stated that their approach was to ensure that all of the issues were captured before starting further investigation under an overall 120 Vac project. The individual also stated that this project was in its early stages and that corrective actions would be scheduled in accordance with standard site work scheduling procedures. The inspectors were satisfied with the licensee's corrective action pla CFR Part 50, Appendix B, Critsrion lil, requires, in part, that design control measures provide for verifying and checking the adequacy of the design. The failure to provide an adequate analysis of the 120 Vac feeders and control circuits and to provide an analysis demonstrating adequate terminal voltages was a failure to verify the adequacy of the 4 design and was identified as a violation of 10 CFR Part 50, Appendix B, Criterion Il i

- This is a second example of the noncited violation (50-482/9913-01) identified in !

Section E8.1. This issue was entered into the licensee's corrective action program as Performance improvement Requests 98-1170,98-1985,98-1986,98-2632,98-2633, 98-2634,98-3445,99-0924,99-0927, and 99-0928 and Action Request 28151/28152.

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E8.3 (Closed) Inspection Followuo Item 50-482/97201-07: battery load profile. This is-sue involved nonconservative errors in Calculation NK-E-002, " Class 1E Battery Sizing,"

Revision 3. The errors resulted in underestimating the battery loads under accident conditions. The licensee corrected these items in Revision 4 of the calculation, issued on April 4,1999. The changes were minor and were detailed in Performance improvement Request 97-3988 concerning emergency diesel generator startup loads, Performance improvement Request 97-4063 for the load from Valve FCHV0312, and Performance improvement Request 97-4190 for the total value of continuous load The revised battery load profile did not result in an operability concern.

l Revision 4 of the calculation was much more detailed than Revision 3 and was reorganized in response to a finding in Inspection Followup Item 50-482/9812-03, as addressed in Section E8.12 of this repor CFR Part 50, Appendix B, Criterion Ill, requires, in part, that design control measures l provide for verifying and checking the adequacy of the design. The nonconservative l assumptions in Calculation NK-E-002, Revision 3, were indicative of a failure to verify the adequacy of the design and were identified as a violation of 10 CFR Part 50, Appendix B, Criterion Ill. This is a third example of the noncited violation (50-482/9913-01), identified in Section E8.1. This issue was entered into the licensee's corrective action program as Performance Improvement Requests 97-3988, -4063, and-419 E (Closed) Insoection Followuo item 50-482/97201-08: Technical Specification change for ,

batteries. This issue involved a failure of Design Change Package 05846, " Class 1E I Battery Replacement," and the associated 10 CFR Part 50.59 safety evaluation to properly identify battery capacity replacement and battery degradation criteria in Technical Specification 4.8.2.1.e and 4.8.2.1.f, as impacted by the recent battery replacement. As noted in the staff's letter to the licensee dated March 30,1999, the staff agreed with the licensee that Technical Specification 4.8.2.1.e battery capacity acceptance criteria should be changed from 80 to 85 percent of the manufacturer's rating, but that Technical Specification 4.8.2.1.f did not require revision. The inspectois agreed and had no further concern E8.5 (Closed) Inspection Followuo item 50-482/97201-09: battery sizing. The issue involved I a failure to require that Calculation NK-E-002, " Class 1E Battery Sizing," Revision 3, used to support the replacement of the four Class 1E batteries in 1996, used the correct design margin. The staff's position stated in NUREG-0881, " Safety Evaluation Report Related to the Operation of Wolf Creek Generating Station, Unit No.1," required batteries to be sized 50 percent greater than the base load to account for temperature, voltage, specific gravity fluctuations, and aging effects. In Revision 3 of Calculation NK-E-002, the licensee used a margin of 25 percent. The licensee issued Calculation NK E-002, Revision 4, which ensured there was at least a 52 percent margin included in the battery sizing to account for the listed effects. All four installed Class 1E batteries had adequate reserve margin to comply with the staff's safety evaluation report. The inspectors noted that the incorrect factor used in the calculation to size the replacement batteries did not result in an operability concern.

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l 10 CFR Part 50, Appendix B, Criterion ill, requires, in part, that design control measur l

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- provide for verifying and checking the adequacy of the design. The error in Calculation NK-E-002, Revision 3, was indicative of a failure to verify the adequacy of the design and was a violation of 10 CFR Part 50, Appendix B, Criterion Ill, requirements. This failure constitutes a violation of minor significance and is not subject to formal enforcement actio E (Closed) Unresolved item 50-482/97201-10: dc voltage drop calculation. This issue involved an assumption in Calculation NK-E-001, " Class 1E Voltage Drop," Revision 1, concerning minimum operating voltage for devices where the value had not been identified by the manufacturer. This was addressed in Performance improvement Request 97-4043, which concluded that this assumption was valid because it was consistent.with the licensee's standard procurement specification for 125 Vdc equipment. A iicensee design electrical engineer stated that the standard specification, which requires a 100 Vdc minimum operating voltage, was used for the procurement of all Class 1E electrical equipment. Based on this information, the inspectors concluded that the original assumption was valid. This information was captured in Calculation NK-E-001, Revision E (Closed) Unresolved item 50-482/97201-11: minimum battery voltage. This issue

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involved Calculation NK-E-002, " Class 1E Battery Sizing," Revision 3, which used the circuit from Battery NN11 to inverter NK11 as the limiting case for determining the minimum required battery voltage during a station blackout. The resolution to  ;

associated Performance improvement Request 97-4185, dated December 23,1997, l determined that Inverter NK11 was not the worst case load. Calculation NK-E-002, Revision 4, resulted in new Class 1E battery service load profiles and a reassessment of the worst case load, which were incorporated in surveillance test Procedure STS MT-021, " Service Test for 125Vdc Class 1E Batteries," Revision 12, A licensee design engineer stated that the tests with the new profile were successfully completed during the recent refueling outage. Therefore, the error in Calculation NK-E-002, Revision 3, did not result in an operability concer CFR Part 50, Appendix B, Criterion V, states, in part, that activities affecting quality shall be prescribed by procedures of a type appropriate to the circumstances. The failure to correctly translate design input in plant procedures, specifically not having correct Class 1E battery service load profiles in the battery service test procedure, was identified as a violation of 10 CFR Part 50, Appendix B, Criterion V. This Severity Level IV violation is being treated as a noncited violation (50-482/9913-02), consistent with Appendix C of the NRC Enforcement Policy. The violation is in the licensee's corrective action program as Performance Improvement Request 97-418 E8.8 (Closed) Unresolved item 50-482/97201 12: load growth control. This issue concerned a failure to ensure minor load changes associated with Design Change Package 05248,

" Swing Battery Charger Installation," Revision 0, were incorporated into applicable DC calculations, as required by Procedure Al 05-06, " Electrical Load Growth," Revision The licensee determined that the cause of this failure was personnel error associated with the extended period between the start of the design change effort and its final completion. Additionally, in Performance improvement Request 98-1951, the licensee i

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l-10-identified an inconsistency between Procedure Al 05-06 and Form APF 05-002-01,

" Engineering Screening," Revision 3, which restricted design engineering evaluations to changes of 10 kW or larger. The performance improvement request was still open, but the inspector agreed with the proposed resolution. A licensee design electrical engineer stated that a review of other design change packages associated with 125 Vdc components resulted in the conclusion that this was an isolated case. The overall effect of the error was negligibl l Criterion lit of Appendix B to 10 CFR Part 50 requires, in part, that design control l i

measures provide for verifying and checking the adequacy of the design. The errors in j

Design Change Package 05248 and the failure to account for loads less than 10 kW l were indicative of a failure to ensure the adequacy of the design and were identified as a violation of 10 CFR Part 50, Appendix B, Criterion Ill. This failure constitutes a i I

violation of minor significance and is not subject to formal enforcement actio '

E8.9 (Closed) Unresolved item 50-482/97201-21: Updated Safety Analysis Report (USAR)

discrepancies. The NRC identified 11 areas in the USAR that contained one or more errors, in NRC Inspection Report 50-482/98-12, the inspectors verified that performance improvement requests had been written to investigate and cormet each of the errors. Although the performance improvement requests adequately addressed the identified problems, a broader concern existed with the overall fidelity of the USA Consequently, this item was left open pending further NRC review of the licensee's USAR upgrade program, which is documented belo The inspectors reviewed Self Assessment Plan SEL 97-044, "WCGS Updated Safety Analysis Report Fidelity Review," Revision 1, dated May 15,1998, and Self-Assessment Plan SEL 97-044, "WCGS Updated Safety Analysis Report Fidelity Review," dated October 16,1998. .The licensee's review identified 2,354 potential USAR discrepancies and generated 650 performance improvement requests to review and correct the error Approximately one-third of the errors were editorialin nature, and none impacted the operability of a safety system or component. While acknowledging that many minor discrepancies existed, the licensee concluded that the USAR was generally accurat The more significant errors were mostly the result of a failure to update or correctly update the USAR as a result of a procedure or design chang The USAR update program included a 100 percent review of the USAR text (excepting descriptions of several nonsafety-related systems), including a representative sample of tables and figures. As a result of discovering several complex discrepancies in tables and figures, the licensee initiated performance improvement requests and identified their resolution as " Phase 2" of the project. Several of these performance improvement requests were open at the time of this inspectio The inspectors noted that the USAR upgrade program was comprehensive in scope and had identified a large number of discrepancies similar in nature to the problems identified in NRC Inspection Report 50-482/97-201. The inspectors considered that the USAR upgrade program had satisfactorily addressed the question of overall USAR fidelity. However, the inspectors were also concerned with the licensee's ability to maintain USAR fidelity in the future. In discussions with licensee representatives, the

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inspectors leamed that the licensee had reviewed the processes needed to maintain the

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accuracy of the USAR and had concluded that they were effective and not in need of i revision. Accordingly, the licensee's emphasis was on individual responsibility for

following these processes and maintaining the USAR as a precise, living documen l l This message had been stressed during company meetings and in newsletters. The I licensee's representative stated that future performance improvement requests would ;

be reviewed to detect any adverse trend l l 10 CFR Part 50.71(e) states, in part, that licensees must update the safety analysis )

l report periodically to assure that the information therein contains the latest material developed. The original 11 USAR discrepancies identified by the NRC constituted a violation of 10 CFR 50.71(e) for failure to properly update the USAR. This Severity Level IV violation is being treated as a noncited violation (50-482/9913-03), consistent with Appendix C of the NRC Enforcement Policy. The violation is in the licensee's corrective action program as Performance Improvement Requests 97-0547,97-3823, 97-3958,97-4018,97-4052,97-4126,97-4179,97-4190,98-0618, and 98-006 E8.10 (Closed) Violation 50-482/9812-01: failure to receive NRC approval of a change that created an unreviewed safety question. The licensee made changes to motor-operated valve stroke times and Procedure ES-12, " Transfer to Cold Leg Recirculation," (various revisions) that involved an unreviewed safety question, without prior Commission approval and without performing safety evaluations. This was identified as a violation of 10 CFR 50.59. The licensee had revised Procedure ES-12 on several occasions to add operator action steps and had increased the stroke times of certain emergency core cooling system motor-operated valves without considering the impact on assumed i

, operator response times in the USAR. As a result, following a loss-of-coolant accident,

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the operators may not have had sufficient time to perform all actions necessary to transfer emergency core cooling pump suctions from the reactor water storage tank to the containment sump, along with other concurrent required actions, before the reactor water storage tank inventory was deplete The licensee found that the plant was operable based on Procedure ES-12 actions to secure emergency core cooling spiem pumps prior to the reactor water storage tank becoming depleted, the f act that the residual heat removal pumps automatically transferred suction, and the fact that one residual heat removal pump could, at the applicable point into the accident, supply sufficient cooling water. The NRC agreed with the licensee's operability assessmen Licensee corrective actions related to this violation were incorporated into Performance Improvement Requests 97-4018,97-4026,98-0118,98-1008, and 98-3003. The inspectors reviewed each of these performance improvement request To address the response-time problems associated with the emergency core cooling system pump suction switchover, the licensee developed Design Change Package 07849, " Automation of CCW to RHR Heat Exchanger Alignment," Revision 0, to automate component cooling water alignment during switch-over from emergency l

core cooling system injection to the recirculation mode. This modification, installed l during Refueling Outage 10 (Spring 1999), provided additional but not sufficient time for

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d f -12-operators to accomplish all other necessary actions within the assumed USAR time frame. Other time gains were obtained through calculational revisions and resequencing of operator actions within Procedure ES-12, which was revised to delete the previously-required manual operator actions needed to realign component cooling water and to implement other changes to streamline the evolutio Regarding the failures that led to this violation, the licensee found that they would not

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likely recur because the procedures and the environment for performing 10 CFR 50.59 l screenings and evaluations are mJch more rigorous than existed at the time of the {

violation examples (late 1980's and early 1990's). The licensee documented the root causes of this event in Performance Improvement Request 98-1008 as:

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(1) A lack of clear ownership and accountability to put a process in place to ensure that operator response times assumed in the USAR were actually achievable, (2) A mindset that small changes to Procedure ES-12 could not significantly affect USAR timeline assumptions, and

(3) A failure on the part of support engineering to fully regard the significance of

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increasing the stroke times of several motor-operated valves that were involved in the post loss-of-coolant accident mode transition from injection to recirculatio The licensee's representative stated that no specific training or required reading pertaining to this violation had been conducted for 10 CFR 50.59 reviewers or

, motor-operated valve personnel, but that this topic would be considered for

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, The licensee researched all safety-related and time-critical operator actions assumed in l the USAR and preliminarily determined that no other examples existed (excepting two

! that were identified prior to the violation) where an actual performance time would l

exceed the assumed response time. The licensee entered this information into a database and established plans to reverify the USAR action time assumptions every 2 year The inspectors reviewed Calculation BN-M-013, "RWST Volume Requirements for injection, ECCS, and Containment Spray Pumps Transfer and Time Available for Operator Actions," Revision 1, which provided a summary of assumed response times and water inventories, taking into account Design Change Package 07849, the changes to Procedure ES-12, and other changes to calculations and assumptions. This calculation showed that sufficient time was available for both a case with no active failures and a case with the single most limiting active failure occurring during the sequence of events. These times were updated in the USA In the simulator, the licensee validated that operating crews could meet the revised USAR loss-of-coolant accident time frames for transition from the injection to the recirculation mode. This evolution had been added to the regular curriculum for periodic operator training. Four of the six operating crews had demonstrated proficiency in l

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-13-completing all steps with a considerable margin to the USAR time assumptions. The other two crews were scheduled to complete this task during their next regularly scheduled training sessio E8.11 (Closed) Violation 50-482/9812-02: failure to follow design control procedures and failure to have adequate design control. Two examples (B.1 and B.2) were identified for this violation. The inspectors observed that the September 29,1998, cover letter for NRC inspection Report 50-482/98-12 noted that the licensee's corrective action

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response to violation example B.2 was complete and that no further response or action was required for this example. The licensee provided additionalinformation to the NRC in a letter dated November 13,1998, for violation example B.1. Subsequently, )

example B.1 of the violation was withdrawn in an NRC letter dated December 18,199 . 1 E8.12 (Closed) Inspection Followuo item 50-482/9812-03: revised battery sizing calculatio This issue involved difficulties the inspector encountered when reviewing Calculation NK-E-002, " Class 1E Battery Sizing", Revision 3. In particular, the inspector had to question the calculation preparers in order to understand the calculation. These problems were resolved in Revision 4 of the calculation on April 4,1999. All necessary j information was included in the calculation body and attachments, and information was i rearranged to permit an independent review and verificatio E8.13 (Closed) Insoection Followuo item 50-482/9812-04: perform a review of revised electrical cabinet configuration seismic qualification. The NRC noted that Electrical Cabinets NK-25, -75, and -77; NK-01 and -71; and NK-04 and -74 were connected to )

each other by 4-inch diameter, short, rigid conduit. Electrical Cabinets NK-71, -74, -75, and -77 were installed as a result of Design Change Package 05248, "NK System, Swing Battery Charger Installation," in November 1997.

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Cabinets NK-01 and -25 weighed approximately 1800 lbs. and were more flexible horizontally than Cabinets NK-71, -74, -75, and -77, which weigh approximately 500 lb each and were considered to be horizontally rigid. The NRC noted that the differences in height, plan dimensions, and mass distribution could result in different dynamic responses during a eeismic even Upon questioning the licensee's representative regarding the interaction between the cabinets, the inspectors (during the previous inspection) were informed that each ;

cabinet was independently seismically tested and qualified; however, no formal calculations existed to address interactions between the cabinets. The licensee's representative stated that the conduit had been installed in accordance with Drawing 1RG900, " Raceway Notes, Symbols, and Details," paragraph 3.35 of Revision 15, dated January 2,1997, which stated, "Only flexible conduit shall be fastened to any equipment which requires seismic qualification except when the installation is such that the equipment and the last support for the conduit are attached to the same surface (plane) . . . " The licensee's representative stated that the basis for this specification was that, during a seismic event, equipment connected via a rigid conduit and mounted on the same plane is expected to initially react in the same direction. The inspectors agreed but, given the differences in mass and dimensions, i

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-14-questioned the interaction between the cabinets during the entire seismic event. The licensee initiated Performance improvement Request 98-0986 to perform a calculation explicitly addressing the inspectors' concern The inspectors reviewed the latest issue of Perforrr ,ce Improvement Request 98-0986 and Bechtel Calculation 16577-757-C001, " Evaluation of NK Panel Interconnections with Short Rigid Conduit, Area 1, Elevations 2000 ft. and 2016 ft.," Revision 0, dated May 1, 1998. The inspectors observed that Calculation 16577-757-C001 showed that the subject cabinet installations were satisfactory during the design seismic event. The licensee had not completed its review and processing of Calculation 16557 757-C001, which was scheduled to be complete by December 3,1999. The inspectors concluded that the licensee's original design assumption that no adverse seismic interactions would affect the cabinets was correc E8.14 (Closed) Violation 50-482/9812-05: failure to translate design basis into specifications and failure to verify and check the adequacy of design control. This violation identified ,

five examples of failure to adequately implement the licensee's design change progra '

The examples involved failures to assure that design calculations were revised when applicable design changes were made, that design calculation results were consistent l l with the USAR, that all variables were addressed in the instrument uncertainty j calculations, and that correct parameters were used in the design basis calculations. As '

documented in the cover letter for NRC Inspection Report 50-482/98-12 and the associated Notice of Violation, the NRC concluded that the initial responses to each of I the violation examples were complete and that no further licersee response was require E8.15 (Closed) Inspection Followuo item 50-482/9812-07: discrepancy between the Technical i l Specification safety limits and design temperature of the reactor coolant system. This i

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issue involved a licensee-identified discrepancy between the safety limits, as defined in the Technical Specifications, and the design temperature of the reactor coolant system, as defined in the USAR. The USAR stated that the design temperature of the reactor j coolant system was 650 F, whereas the Technical Specifications (safety limit Technical l Specification, Figure 2.1-1) appeared to allow operating conditions that would l' ave

! resulted in a peak reactor coolant system temperature of 655 F. The licensee l performed a preliminary review and concluded that the excessive reactor coolant j temperatures shoLid be precluded by the reactor protection system, meaning that a

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reactor trip or other protective feature would actuate before the high temperature condition could occur. The formal resolution of the remaining aspects of the issue was evaluated under Performance improvement Request 98-016 The inspectors reviewed Performance improvement Request 98-0169, which was completed on January 22,1998. The performance improvement request concluded that this was not an actual technical problem. According to the performance improvement request, the Technical Specification safety limits do not and are not required to demonstrate that a reactor coolant sptem design temperature limit is not violate Rather, they represnnt limiting conditions that should not be exceeded during certain design basis transients (as described in USAR, Chapter 15). The li.:ensee's position

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I was that there were no requirements that the design temperature (which corresponds to I normal operating conditions) is not exceeded during accident transient l I

The inspectors reviewed Technical Specification 2.1.1 and noted that Figure 2.1-1 provided a graphical display of parameters defining " acceptable operating conditions."

This terminology was misleading. First the OT-delta-T and OP-delta T reactor protection system trip setpoints would prevent the plant from operating at the more limiting points on the graph. Second, Technical Specification 3.2.5,"DNB Parameters,"

restricts operation above an average reactor coolant system temperature of 590.5*F, which correlates to a full-power maximum reactor coolant system temperature of approximately 620'F. Therefore, the plant could not be operated at a temperature that exceeded the USAR design temperatur The inspectors agreed with th9 licensee's position tnat a technical conflict did not exis l E8.16 (Closed) Inspection Followup item 50-482/9812-08: review the dose consequences for 1 a fuel handling operation. This item involved a licensee-identified discrepancy (described in Performance improvement Request 98-0179) concerning assumptions in the USAR fuel handling dose calculation. The discrepancies were: (1) an assumed 4500 F maximum center-line operating fuel temperature versus and actual 4700 F I maximum center-line operating fuel temperature, and (2) an assumed average burnup for the peak assembly of 25,000 Megawatt-Days per ton of fuel (MWD / ton) and a peak local burnup of 45,000 MWD / ton versus actual operating cycle limits of 34,430 MWD / ton core average burnup and 60,000 MWD / ton peak local burnup. The concern was that, as a result of these discrepancies, the USAR offsite and control room doses for a fuel handling accident were nonconservativ ,

The inspectors reviewed Calculation AN-99-013," Radiological Consequences of a Fuel Handling Accident for Spent Fuel Pool Rerack," Revision 0. dated July 29,1999. This calculation included a revised source term and corrections for the two discrepancies discussed above, as well as a change in the assumed gap activity contribution from lodine-131 (12 percent versus 10 percent). The revised doses were slightly increased from previous USAR values. For instance, site boundary thyroid increased from 7.90 rem to 9.34 rem. All changes were of this order and were well within regulatory limits. The licensee intended to evaluate the change under 10 CFR Part 50.59 at the time that the results of Calculation AN-99-013 are incorporated into the USAR (October 1999).

E8.17 (Closed) Inspect.,;n Followuo item 50-482/9812-09: review analysis for extended core operation. This item involved the licensee's discovery that the projected core average cumulative burnup for Operating Cycle 10 exceeded earlier projections. The revised estimate was 34,430 MWD / ton versus a previous calculation of 33,500 MWD / ton. The principal concern was that the extended burnup could change the radiological isotopic distribution following a design basis loss of coolant accident, such that the previously calculated u MR source term and dose consequences for this accident would be nonconservative. The licensee estimated that the previous burnup projection would not 2 be exceeded until March 1999 and therefore proceeded into Operating Cycle 10

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I l The licensee referred this issue to Westinghouse, the nuclear steam supply system l vendor. Westinghouse analyzed the dose consequences based on a bounding core 1 average cumulative burnup of 38,400 MWD / ton, updated the source term, and .

concluded that the extended burnup would not significantly change the USAR offsite i dose consequences. Westinghouse advised the licensee that the plant could continue power operation to the scheduled end of Operating Cycle 10 without invalidating the conclusions of dose consequences analyses described in the USA The inspectors reviewed Calculation AN-98-056, "LOCA Radiological Consequences Analysis to Support Performance improvement Requests 98-1920,97-2783, and 96-2558," Revision 0. This calculation incorporated the revised source term. Also, the calculation was updated to reflect a radiological release contribution from the minipurge system. The inspectors noted that the source term had decreased from the previous value, despite the higher assumed fuel burnup, and requested an explanation for this anomaly. The licensee's representative explained that a new computer program had been used to estimate the isotopic distribution and that the new program had predicted concentrations generally about 3 percent less than the former program. This change was ascribed to the improved accuracy of the new program. All doses remained well within regulatory limits. The licensee intended to evaluate the change under 10 CFR 50.59 at the time that the results of Calculation AN-98-056 are incorporated into the USAR (October 1999).

E8.18 (Open) Licensee Event Report (LER) 50-482/98-001-00: pressurizer code safety valves testing outside of Technical Specification allowances. This LER stated that five I;ft-test f ailures of pressurizer code safety valves had occurred over the past four refueling outages. During this time, only 12 tests had been conducted; therefore, the failure rate was almost 50 percent. The plant has three pressurizer code safety valves, and the licensee performed a lift test of each of these valves during every refueling outag Technical Specification 3.4.2.2 states "All pressurizer code safety valves (AB-V) shall be operable with a lift setting of 2485 psig +/- 1%."

The test failures were as follows:

Refueling Valve As-Found Percent Outage Deviation from Nominal Setpoint 6 BB8010A -2.13 6 BB8010B +1.40 6 BB8010C -2.70 8 BB8010B +1.97 9 BB8010A -1.81 l

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-17-Following issuance of the LER, the valves were again tested during Refueling Outage 10. The results of this testing were one additional failure of Valve BB8010B at-1.89 percent. The other two valves were found to be in toleranc The licensee analyzed each of the as-found deviations and determined that none of these conditions would have: (a) prevented fulfillment of a safety function, (b) represented an unanalyzed condition, or (c) placed the plant outside of the safety l analysis design basi Because the safety valves exhibited no physical indications that would explain the observed setpoint drift, no root cause could be found. The licensee noted that some improvement had occurred following a change of testing medium from nittogen to steam. However, the licensee concluded that the valves were inherently not capable of consistently meeting a tolerance of +/- 1 percent of valve test pressure. Therefore, the licensee decided to pursue a change to the Technical Specifications to lower the nominal pressurizer safety setpoint by 1 percent and to change the tolerance to +/- 2 percent. At the time of this inspection, the change request was in draft and being prepared for submittal to the NR This item was left open pending attainment of either a mechanical or a Technical Specification solution to the problem. This item is in the licensee's corrective action program as Performance improvement Request 99-234 IV. Plant Support R1 Radiological Protection and Chemistry Controls R1.1 Gqneral Comments (71750)

The inspectors observed health physics personnel, including supervisors, routinely touring the radiologically controlled areas. Licensee personnel working in radiologically controlled areas exhibited good radiation worker practice Contaminated areas and high radiation areas were properly posted. Area surveys posted outside rooms in the auxiliary building were current. The inspectors checked a sample of doors, required to be locked for the purpose of radiation protection,.and found no problem F14 Staff Knowledge and Performance R4.1 Containment Entry at 100 Percent Power inspection Scope (71750)

l The inspectors accompanied licensee personnel during a containment entry at i 100 percent powe !

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-18-I Observations and Findinas i The inspectors attended the prejob brief and observed health physics coverage during the containment entry. The licensee entered the containment to search for a leak in the instrument air system. The prejob brief was thorough. In addition to health physics concerns, personnel safety issues were discussed since the temperature and humidity in the containment were hig The health physics technician provided excellent coverage while in the containmen Personnel were continually advised of dose rates in various areas. The technician was proachve in ensuring personnel did not enter high dose areas. Licensee personnel used good ALARA practices when checking for air leak l I Conclusions The inspectors concluded that the prejob brief and health physics coverage during a containment entry ai 100 percent power were thorough. Personnel exhibited good ALARA practices while in the containmen S1 Conduct of Security and Safeguards Activities l St .1 Security Comouter System Uoarade Insoection Scope (71750)

The inspectors reviewed the licensee's efforts during the security system upgrad Observations and Findinas The inspectors observed the licensee's compensatory measures during portions of the l

- upgrade of the plant security system. The licensee provided appropriate compensatory l coverage when surveillance equipment was removed from service. The licensee also ensured that operations department personnel had the means to access required area The plant trip, referenced in Section 04.1, occurred during the security upgrade. The operators were not hindered during their response to the tri Conclusions

The inspectors concluded that the licensee took appropriate compensatory measures J during the security system upgrad r

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-19-V. Manaaement Meetinas X1' Exit Meetina Summarv The exit meeting was conducted on September 3,1999. The licensee cd not express a position on any of the findings in the repor The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identifie i

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ATTACHMENT SUPPLEMENTAL INFORMATION PARTIAL LIST OF PERSONS CONTACTED

Licensee  !

M. J. Angus, Manager, Licensing and Corrective Action G. D. Boyer, Chief Administrative Officer  !

J. W. Johnson, Manager, Resource Protection l O. L. Maynard, President and Chief Executive Officer B. T. McKinney, Plant Manager R. Muench, Vice President Engineering l S. R. Koenig, Manager, Performance improvement and Assessment l

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C. C. Warren, Chief Operating Officer INSPECTION PROCEDURES USED IP 37551 Onsite Engineering IP 61726 Surveillance Observations IP 62707 Maintenance Observations IP 71707 Plant Operations l

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IP 71750 Plant Support Activities IP 92700 Onsite Review of Licensee Reports IP 92902 Followup - Maintenance IP 92903 Followup - Engineering IP 93702 Prompt Onsite Response to Events at Operating Power Reactors ITEMS OPENED. CLOSED. AND DISCUSSED Ooened 50-482/9913-01 NCV- Multiple design control errors (Sections E8.1, E8.2, and E8.3)

50-482/9913-02 NCV Minimum battery voltage calculation (Section E8.7)

50-482/9913-03 NCV Failure to update USAR (Section E8.9)

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Closed 50-482/9713-05 VIO Unauthorized work performed and not documented (Section M8.1)

50-482/9713-06 VIO Welders failed to return unused weld filler material to issue (Section M8.2)

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-2-l 50-482/9717-01 IFl Review performance improvement requests on maintenance rule program implementation (Section M8.3)

50-482/97201 04 URI Motor control center circuit length (Section E8.1) 1 50-482/97201-05 IFl 120 Vac short circuit and voltage drop analysis (Section E8.2) )

50-482/97201-07 IFl Battery load profile (Section E8.3)

50-482/97201-08 IFl Technical Specification change for batteries (Section E8.4) '

50-482/97201-09 IFl Battery sizing (Section E8.5)

l 50-482/97201-10 URI de voltage drop calculation (Section E8.6)

50-482/97201-11 URI Minimum battery voltage (Section E8.7)

50-482/97201-12 URI Load growth control (Section E8.8)

50-482/97201-21 UT;; USAR discrepancies (Section E8.9)

50-482/98f2-01 VIO Failure to receive NRC approvat of a change that created an unreviewed safety question as an unreviewed safety question (Section E8.10)

50-482/9812 02 VIO Failure to follow design control procedures and failure to have adequate design control (Section E8.11)

50-482/9812-03 IFl Revised battery sizing calculation (Section E8.12)

50-482/9812-04 IFl Review revised electrical cabinet seismic qualification (Section E8.13)

50-482/9812-05 VIO Failure to translate design basis into specifications and failure to verify and check the adequacy of design (Section E8.14)

50-482/9812-07 IFl Discrepancy between Technical Specification safety limits and c'esign temperature (Section E8.15)

50-482/9812-08 IFl Review dose consequences of fuel handling accident (Section E8.16)

50-482/9812-09 IFl Review analysis for extended core operation (Section E8.17)

50-482/9913-01 NCV Multiple design control errors (Section E8.1)

50-482/9913-02 NCV Minimum battery voltage calculation (Section E8.7)

50-482/9913-03 NCV Failure to update USAR (Section E8.9) w

-3-Discussed 50-482/98-001-00 LER Pressurizer code safety valves testing outside of Technical Specification allowances (Section E8.18)

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