IR 05000324/1998005

Insp Repts 50-324/98-05 & 50-325/98-05 on 980315-0425. Violations Noted.Major Areas Inspected:Operations, Maintenance,Engineering & Plant Support

ML20249A238
Person / Time
Site:	Brunswick
Issue date:	06/01/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20249A209	List: IR 05000324/1998005 ML20249A208 ML20249A219
References
50-324-98-05, 50-324-98-5, 50-325-98-05, 50-325-98-5, NUDOCS 9806160211
Download: ML20249A238 (36)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION II

Docket Nos: 50-325, 50-324 License Nos: DPR-71. DPR-62 Report No: 50-325/98-05. 50-324/98-05

Licensee: Carolina Power & Light (CP&L)

Facility: Brunswick Steam Electric Plant. Units 1 & 2 Location: 8470 River Road SE Southport. NC 28461 Dates: March 15 - April 25.1998

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Inspectbrs: C. Patterson. Senior Resident Inspector E.. Brown. Resident Inspector E. Guthrie.- Resident' Inspector E.. DiPaolo Resident Inspector. Browns Ferry

~ Approved by: '

L. Plisco. Director- l Division of Reactor Projects '

Enclosure 2

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Y$ W4 PDR

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EXECUTIVE SUMMARY Brunswick Steam Electric Plant. Units 1 & 2 NRC Inspection Report 50-325/98-05, 50-324/98-05 This integrated inspection included aspects of licensee operations, maintenance, engineering. and plant support. The report covers a 6-week period of resident inspection: r

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Ooerations

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The inspector identified that a brass fitting was inappropriately attached to a safety-related electrical conduit support. The operability of the support was not affected. The inspector identified a violation for erected scaffolding too close to safety-related equipment located in the Unit 1 Reactor Building (Section 01.1). S

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The inspector observed a minor High Pressure Coolant Injection flow controller setting difference which did not affect system operability (Section 03.1).

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The alignment for two systems under equipment clearances were verified 5 as correct. The clearances reviewed were consistent with the procedural requirements (Section 04.1).

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The high tem]erature annunciator for the drywell/ suppression poo located on t1e remote shutdown panel, was fot nd lit on multiple occasions. An inspector follow-up item was initiated for the review of the adequacy of the annunciator procedure, operator knowledge, and training for the remote shutdown panel (Section 04.2).

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Acceptable procedural adherence, communication, and knowledge were demonstrated by the Unit 1 Reactor Building Auxiliary Operator during routine activities (Section 04.3).

Maintenance

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The ins)ector determined that observed maintenance activities. Serformed by the rix It Now team, were conducted satisfactorily and met t1e licensee's procedural requirements (Section M1.1).

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The use of a checklist contributed to a thorough pre-job briefin The 3ersonnel performing the Residual Heat Removal and Core Spray Low Reactor Pressure Permissive Trip Unit Channel Calibration properly stopped the surveillance and informed supervision when an abnormal reading was obtained. Personnel performing the Average Power Range Monitor Flow Bias Flow Units C&D Channel Calibration positioned the wrong switch resulting in an unexpected half scram (Section M1.2).

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The inspector concluded that signing of a Quality Control (OC) procedure step by a non-qualified OC inspector was a violation of 3rocedure requirements. There were inconsistent requirements in t1e mechanical maintenance surveillance procedure for Diesel Generator (DG)

inspections. Nuclear Assessment Section had not reviewed a procedure revision that deleted OC steps (Section M3.1).

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A Fire Protection technician performed procedural steps out of sequence during the performance of the Transformer Deluge System Functional Tes PT-34.16.1.1. This was determined to be a violation of Fire Protection Procedures (Section M3.2).

. The inspector determined that good 3rocedural adherence and documentation occurred during the o]servation of sampling of a Unit 1 drywell iodine and particulate air sample. The inspector noted satisfactory work practices and communications throughout the evolution (Section M3.3).

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The inspector concluded that the licensee improperly modified twenty-eight fire doors which introduced a common mode failure. The modification degraded the 3-hour fire barriers, which could have allowed the spread of a fire from one DG to the other, one emergency bus to the other or one 4-day tank to the other. A violation was identified for the failure to maintain 3-hour fire rated barriers between redundant equipment (Section M4.1).

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The failure to report a condition outside the 10 CFR 50 Appendix R design basis was identified as a violation. This condition was not reported although questioned by the inspector (Section M4.2).

. The inspector concluded that the licensee had a long standing history of fire door hardware deficiencies. The licensee's root cause determined j that there was insufficient engineering documentation about the importance of com]onents and function of the locksets. However, the inspector noted t1at the various self-ns essments identified the hardware deficiencies, these assessments represented missed 1 opportunities to correct the adverse conditions identified (Section '

M7.1).

Enoineerino I

. The licensee had taken acceptable actions for steam leaks that developed ;

on a main steam line drain. An Inspection Follow-Up Item was opened to -

review recent changes to the inservice inspection program (Section E1.1).

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. The inspector concluded that the licensee provided effective oversight for the design and installation of the Control Building Air-Conditioning 4

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upgrade modification (Section E2.1).

. The inspector determined that the licensee had implemented a more j limiting administrative limit than what existed in the TSs due to the l

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inability of the Standby Liquid Control (SLC) system piping heat trace to maintain adequate temperatures to maintain sodium pentaborate chemicals in solution during cold temperatures (Section E3.1).

P_] ant Sgoport

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The inspector concluded that work activities on the refuel bridge grapple that resulted in skin contaminations were covered by the radiological work permit and work instructions. However, work practices to prevent the spread of contamination were not effective (Section R4.1).

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The root cause analysis for a cited violation concerning chemical control was excellent. Subsequent self-assessments were appropriately self-critical and effective in the identification of issues (Section R8.1)

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The inspector found security activities to be satisfactory during observations in the Central Alarm Station and the Secondary Alarm Station. Review of lighting inside the protected area fence was found acceptable with one exception which was corrected (Section S1.1).

. The engine-driven fire Jump flow test procedure would not demonstrate that the pump was opera)le until the procedure was revised and acceptance criteria changed. The failure to pro)erly maintain the fire pump procedure was identified as a violation. T1is issue was similar to an event in June 1997 where a dual unit shutdown was commenced as a result of erroneous data from an inadequate test procedure (Section F3.1).

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Reoort Details Summary of Plant Status Unit 1 operated continuously during this report period until April 2 , when the unit was shutdown to begin a scheduled 35 day refueling outage. The unit had been on-line continuously for 162 days. Unit 1 implemented final feedwater temperature reduction on March 28, 1998, to obtain maximum power 03eration as the unit coasted down at the end of the operating cycle. Jnit 1 has operated with four control rods inserted to suppress power around leaking fuel assemblie Unit 2 operated continuously during this report period. At the end of the report period the unit had been on-line continuously for 185 day Unit 2 developed indications of leaking fuel on March 14. 1998. Two control rods were inserted to suppress power around the leaking fuel assembly. On April 9. 1998, the unit went to 30 percent power for single loop operations to replace recirculation pump motor-generator brushes. The unit returned to full power on April 10. 199 Due to concerns about the control room dose in the event of a main steam line break, the licensee imposed an administrative limit on Iodine until a Technical Specification (TS) amendment submitted was approved. The licensee made a procedure change to Administrative Procedure OAl-8 Water Chemistry Guidelines, setting the limit at 0.1 microcurie per gram dose equivalent Iodine 131 comaared to the TS value of 0.2 microcuri per gram. Also, the licensee las been providing weekly water chemistry data to NRR and the Resident Inspector for review. None of the data i reviewed has exceeded the administrative limit. However. on April 1 l 1998, the licensee withdrew their TS amendment request and the weekly I water chemistry data update reports have been stopped. The licensee has committed to maintain the administrative limit until NRR completes their review of the licensee's analysis. The residents will not continue to )

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follow this issue until the review of the analysis is complete and inspection items are identifie Due to a reconstitution of the Environmental Qualification (EO) program and items identified. there are nine Justifications for Continued 0)eration (JCO) that remain open for both units. The following provides tie status of the open E0 JCOs and associated Engineering Service Requests (ESRs): 1 00en I i

1) ESR 97-00330. Motor Control Center (MCC) E0. closure date to be I determined (TBD)

2) ESR 97-00529. Failure of Unit 1 Drywell Motor, closure date TBD  ;

3) ESR 96-00627. Qualification Data Package for Marathon 300 Terminal Blocks, closure date TBD 1 4) ESR 97-00256. Main Steam Isolation Valve Hiller Actuator, closure date TBD 5) ESR 97-00343. Qualification of Kulka Model 600 Terminal Block closure date TBD

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1 2 6) ESR 97-00435. MCC Fittings, closure date TBD 7) .ESR 97-00710. Lubricants for Joy / Reliance Fan Motors, closure date TBD 8) ESR 98-00093. Weed Resistance Temperature Detector Electric Termination Sealing, closure date TBD l

9) ESR 98-00125. Inadequate Conduit Seals Associated with Three Namco-Limit Switches, closure date TBD i

I. Operations 01 Conduct of Operations 01.1 Unit 1 Ricaina and Scaffolding Walkdown Insoection Scope (71707)

The inspector conducted general area tours of the Unit 1 Reactor Building and observed temporary rigging and the condition and location of scaffolding being erected for the upcoming outag Observations and Findinas On March 23. 1998, the inspector observed a brass fitting (that channeled two tygon hoses into one hose that drained to a floor drain)

strapped to a structural support for the 1B Standby Gas Treatment Unit (SBGT). using nylon tie wraps. The tygon hoses were attached to catch i drapes used to collect water from a leak in the floor directly over the 1B SBGT unit, i The inspector asked the licensee if it was acceptable to have the brass fitting attached to the su) port. The licensee was not immediately aware '

that the fitting was attacled to the support. After the licensee observed the fitting and its location. Condition Report (CR) 98-0071 Standby Gas Support Attachment. was generated to identify that the ,

Maintenance Management Manual 0MMM-022. Instructions For Placement of l Temporary Loads (e.g. . Rigging. Scaffolding. Ladders Personnel) did not define this configuration as a minor load which would have allowed

.it to remain in place. The licensee stated that the fitting was a non-flexible mass, estimated to be about five pounds, firmly attached to the !

support which should have warranted a rigging release per the 3rocedur j The licensee promptly removed the fitting from the support. T1e '

licensee determined that there were no operability concerns associated with the support while the brass fitting was attache The inspector noted several conditions related to multiple scaffolding 4 locations in the Reactor Building which had scaffolding touching or were positioned very close to safety-related equipment. The inspector ,

reviewed OMMM-022 to determine if the existing conditions of the i scaffolding met the requirements specified in the procedure. The inspector determined that the following conditions were in question:

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Scaffolding located on the 50 foot (ft) elevation near instrument rack 1-H21-P005-001 e the scaffolding was touching the safety-related instrument rac .

a vertical scaffold pole was located in between two safety-related electrical conduits only a couple of inches from each condui Scaffolding located in the North Core Soray (CS) corner room e the end of a horizontal scaffold ) ole was located less than an inch from a snubber attached to tie CS pip This information was given to the licensee by the ins)ector and the requirements specified in OMMM-022 were discussed. Tie licensee observed the locations described. The licensee agreed that the scaffolding on the 50 ft elevation and the North CS room were not in compliance with OMMM-022. The scaffolds at both locations were corrected. The licensee generated CR 98-00792. OMMM-022 Non-Compliance, which described the non-compliant condition OMMM-022. Section 6.3.1 stated " Unrestrained scaffold should not be used near safety-related equipment that does not.have adequate separatio When unrestrained scaffold is placed near safety-related equipment, it shall have a minimum 6 inch clearance from safety-related equipment or components". Additionally. Section 5.2.1 stated " Scaffolding must not

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interfere with normal operation of plant equipment including snubbers and spring cans."

10 CFR 50 Appendix-B. Criteria V. Procedures. recuires that activities affecting quality shall be accomplished in accorcance with documented instructions, procedures and drawings. The failure to erect the scaffolding in the Unit 1 Reactor Building on the 50 ft elevation and

.the Northwest CS room according to 0MMM-022 was a violation requirements. . This violation is identified as VIO .50-325/98-05-0 ~ Scaffolding Erection Non-Complianc The inspector noted that the licensee incurred a violation against OMMM-022 during a recent Unit 2 refueling outage for not using a temporary

.,. rigging release when rigging to or above safety-related equipment. This violation was identified as 50-324/97-11-06. Rigging to Safety Equipment Without a' Safety Evaluation, and was discussed in paragraph E4.1 of-Inspection Report (IR) 97-1 Subsequent to discussion with Brunswick management regarding the scaffolding. concerns and other recent events, the licensee conducted a

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" Site Stand-Down" on April 6,1998 to reemphasize the importance of following procedures. The licensee expressed concern for an indication

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of an adverse trend in personnel performance during routine activities.

t The inspector-noted scaffolding on April 13. 1998. located on the 20 ft elevation in the Northwest corner of Unit 1 Reactor Building, that had

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. multiple scaffolding poles which were touching a CS line and that a pole was: located less than six inches from a High Pressure Coolant Injection (HPCI) .line, located in the HPCI room. These conditions were deviations o

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from OMMM-022 but the licensee informed the inspector that they were identified as scaffolding erection problems during erectio These conditions were being evaluated by Engineering according to procedur However, having scaffolding erected for a period of time during which the licensee did not know whether it was acceptable or not was a concern and was discussed after the initial procedural compliance proble Subsequent to .the April 13. 1998 discussions, by the inspector, the licensee revisited the concern for having scaffolding erected which-deviated from OMMM-022 without necessary evaluations. The licensee established interim guidance, given to the scaffolding workers on April 20.1998 during a scaffolding work stand-down, to ensure that any scaffolding. which did not meet procedure requirements, got an Engineering evaluation prior to scaffolding erection. The licensee was -

considering a procedure change to 0MMM-022 at the close of this inspection period, Conclusions The inspector identified that a brass fitting was inappropriately attached to a safety-related electrical conduit support. The operability of the support was not affected. The inspector identified a violation for erected scaffolding too close to safety-related equipment

' associated with two scaffolding units located in the Unit 1 Reactor Buildin Operations Procedures and Documentation 03.1 HPCI Flow Controller Settina a. .Insoection Scooe (71707)

The inspector performed a routine tour of the Control Room observing the Unit 1 and Unit 2 control board status on April 13. 199 Observations and Findinos The inspector observed the Unit 1 and Unit 2 HPCI flow controller settings to verify whether they were set at the procedurally required setting of 4300 gallons per minute (gpm). The inspector noted that the two controllers were not set at the same valu The inspector observed the setting on the Unit 2 controller and i' determined it to be set at 4300 gom. The licensee agreed with the inspector that the Unit'2 controller was correct and that the Unit 1 controller was set lower than 4300 gpm. The Unit I controller was set at 4300 gpm and was-determined to have been set less than one increment i below 4300 gpm. The controller increments are 50 gpm increments. The

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operator estimated the setting to be between 25 and 50 gpm below 4300 gp ]

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The required design flow rate of the HPCI system is 4250 gam. The l Unit 1 HPCI system was fully operable. The disparity of tie controller l setting was due to the height difference between individuals setting the controller and those reading the controller afterward Conclusion The inspector observed a minor HPCI flow controller setting difference which did not affect system operabilit .2 Soecial UFSAR Review A recent discovery of a licensee operating the facility in a manner contrary to the Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a special focused review that compares plant 3ractices procedures. and/or parameters to the UFSAR descri)tion hile performing the inspections discussed in this report. t1e inspectors reviewed the applicable portions of the UFSAR that related to the areas inspected. The inspectors verified that the UFSAR wording was consistent with the observed plant practices procedures, and/or parameters The inspector reviewed the UFSAR. Section 6.3.1.2.1. Performance and Functional Requirements for the High pressure Coolant Injection Emergency Core Cooling System (ECCS). The inspector concluded that operation of the HPCI flow control automatic setting was consistent with UFSAR wording regarding minimum design basis flo ' Operator Knowledge and Performance 04.1 -Standbv Liouid Control (SLC) and Diesel Generator (DG) Fuel Oil Clearances Insoection Scope (71707)

The inspector reviewed safety-related tagouts for the SLC and DG. fuel !

oil system to verify proper configuration control and conformance with the procedural requirements maintained in Nuclear Generation Standard

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-0PS-NGGC-1301. Equipment Clearance. and associated administrative instructions, Observations and Findinas The inspector verified 3 roper valve / breaker alignment for clearances 2- I 98-629 and 2-98-497. T1ese clearances were for work activities on the SLC and the DG fuel oil system. The inspector noted for clearance 2-98-497 that the control switches for #2 DG auxiliary fuel oil booster pump, and the 2A and 28 fuel oil trarsfer pumps.were not tagged although the

, switches were in the correct position. These switches were located on .

l the local #2 DG control panel. Administrative Instruction 0Al-5 '

L Equipment Clearance (EC) Procedure step 6.3.5 states "...EC tags shall-be hung on the control panels such that indications and controls are not

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obstructed...EC tags shall be hung on the equi) ment identified on the EC Form in the correct sequence, as specified." )iscussions with the

' licensee, indicated that.the control switches did not serve as a clearance boundary and therefore did not need an EC tag. The control switches were under clearance to prevent cycling of the indicated pumps during logic testing of the DG local control panel. The inspector found that the clearances conformed to procedural guidanc No problems were identified during the clearance verification walkdow Conclusion The alignment for two systems under equipment clearances were verified

'as correc The clearances reviewed were consistent with the procedural requirement .2 Remote Shutdown Panel (RSDP) Annunciator Procedures

' Insoection Scooe (71707)

The inspector reviewed the use and adequacy of annunciator procedures for the RSD Observations and Findinas On multiple occasions, during routine tours of the Unit 2 on the 20 ft 1 elevation, the inspector noted annunciator window 2-5 for panel 2-VA-29 in alarm. This annunciator signals a high temperature in the drywell/ suppression pool as indicated by an adjacent strip chart recorder 2-CAC-TE-778. Review of the strip chart recorder revealed normal conditions. Upon notifying the licensee. -the inspector was informed that point 1 on the CAC-TE-778 recorder had failed and was responsible for the lit annunciator. The inspector noted that, unlike other control panels located throughout the site, the annunciator panel procedure was not available at the RSD Further review by the inspector revealed that the auxiliary operators (A0s) who conducted Reactor Building rounds were not familiar with the procedure. Any problems were relayed to the Control Room for disposition. The inspector reviewed Annunciator Panel Procedure 1(2)

UA-29. Annunciator Procedure for Panel UA-29. The procedures indicated that 1(2) APP-UA-29 were only to be used in conjunction with Abnormal Operating Procedures 0A0P-32. Plant Shutdown From Outside Control Roo After review of 1(2)A PP-UA-29. the inspector .;oted that the procedures l

'had not been revised since 1991 despite changes in allowable drywell temperature (see IR 50-325(324)/97-12). In addition. review of the licensee training simulator revealed no mock-up of the RSDP. Based on questions concerning the adequacy of the annunciator procedure, operator knowledge, and training for the RSDP. this item is identified as an-Inspection Followup Item (IFI). This item is identified as IFI 50-325(324)/98-05-02. Adequacy of RSDP Procedure and Trainin ,

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c. . . Conclusion The high temperature annunciator for the drywell/ suppression poo . located on tie remote shutdown panel, was found lit on multiple occasions. An IFI was initiated for the review of the adequacy of the annunciator procedure, operator knowledge, and training for the remote shutdown pane .3 Auxiliary 00erator Rounds l Insoection Scoce (71707)

The inspector observed the Unit 1 Reactor Building Auxiliary 0)erator (RBA01) during completion of Operating Instruction 101-3.4.2. Jnit 1 Reactor Building Auxiliary Operator Daily Check Sheets and Periodic Test OPT-15.6. Standby Gas Treatment System Operabilit Observations and Findinas On April 24, 1998, the inspector accompanied the RBA01 during the performance of 101-3.4.2 and OPT-15.6. The inspector observed acceptable adherence to the applicable procedure. The RBA01 was i knowledgeable about the periodic test being performed. Satisfactory communication was observed during the performance of both activities and appropriate follow-up actions were taken for an identified deficienc Conclusion Acceptable procedural adherence. communication and knowledge were demonstrated by the Unit 1 Reactor Building Auxiliary Operator during routine activitie II. Maintenance M1 Conduct of Maintenance M1.1 Fix It Now (FIN) Team Maintenance Activities Insoection Scooe (62707)

The inspector observed maintenance activities conducted by the FIN team

- to verify that the activities were performed in accordance with the

-. licensee's procedures, Observations and Findinas The inspector observed the FIN team maintenance activities on March 31, 1998. The maintenance troubleshooting activities were being conducted on fire protection panel 2-FP-PNL-J42. located in the Unit 2 Reactor l Building. The activities were conducted under work request / job order (WR/J0) 98-AB0A1. The work request specified troubleshooting to

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' determine the cause and repair of a trouble alarm which was sealed in on l the pane '

The inspector found that the' FIN team personnel were well prepared and I knowledgeable regarding the maintenance activity. The inspector noted that the Vendor manual was present for the maintenance and was-referenced during the activity. The inspector found the work request instructions, precautions, and limitations to be thorough and accurat The FIN team coordinated their activities with operations personnel appropriatel The inspector reviewed documentation after the maintenance was complete and found the paperwork to be satisfactorily complete. Additionall WR/JO 98-ABYXI. which replaced an annunciator joy stick on the Unit 2 XU-1 panel in the Control Room, was reviewed and found ;o be i satisfactor l Discussions with FIN team personnel and observations throughout the ,

inspection period. regarding the type of maintenance performed by the l FIN team, substantiated that these activities were lower risk and were-  !

chosen for the team because they did not require significant coordination or planning. Activities performed by the FIN team were  !

generally not listed in the Plan of the Da Conclusions

The ins metor determined that observed maintenance activities, performed ~

by the rix It Now team, were conducted satisfactory and met the licensee's procedural requirement ~

M1.2 Maintenance Surveillance Test (MST) Observations Insoection Scooe (61726. 71707)

The inspector observed the performance of portions of the following i reactor instrumentation' MSTs: -!

. '0MST-APRM11W. Average Power Range Monitor (APRM) Flow Unit Gain l

' Adjustment '

. OMST-RHR260. Residual Heat Removal and Core Spray Low Reactor i Pressure Permissive Trip Unit Channel Calibration

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2MST-APRM290. APRM Flow Bias Flow Units C&D Channel Calibration These surveillance were performed in order to satisfy portions of  ;

periodic testing requirements of TS Table 4.3.1-1. The testing was  :

L performed by instrumentation and controls (I&C) technician '

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9 Observations and Findinas On March 31. 1998, the inspertar observed the performance of OMST-APRM11W. The procedure measures the instrument output voltage which was proportional to core flo Prior to testing, an acceptable range of output. voltage was calculated using actual core parameters. "2 .

performer of this calculation made a rounding error whicF s identified during review by an independent verifier. The rounding error would have resulted in more restrictive (i.e., more conservative) acceptance criteri The inspector observed that the output voltages of the instrument oscillated during the performance of the measurements. This was recognized by the procedure and was due to normal small variations in core flow. Three of the flow instruments varied in out relatively small range, however, one of the instrument'put s output over voltage a

varied over a larger range. The inspector observed that determination of an average value of output voltage for this instrument was difficult due to the digital readout of the measuring device. This was also acknowledged by the I&C technicians. The inspector observed that good self checking techniques (repeat backs and touching label plates) were used during the surveillanc On April 1.1998, the. inspector observed the performance of portions of OMST-RHR260. The pre-job brief was observed to be thorough and expectations were clearly communicated. These results were aided by the use of a checklist for pre-job briefing The inspector observed that.the I&C technicians used good self checking techniques and communications during the surveillance. During Performance of a relay contact closure confirmation step, an unusually ligh resistance was observed on the measuring device (40 ohms vice approximately zero ohms required by the procedure). The I&C technicians properly stopped the surveillance and informed the control room and their supervision. The I&C technicians explained that due to the high impedance of'the digital multimeter specified by.the procedure, unusually high measured values of resistance are sometimes encountere The I&C supervisor initiated a work instruction to remeasure the resistance with an analog resistance / voltage meter (with a lower impedance). This was performed in order to determine whether a degraded-condition of the contact assembly existed. The check with the analog resistance meter showed that no degradation of the contact assembly existed. The supervisor explained to the inspector that the digital multimeter was specified for all surveillance as a corrective action from past errors resulting from improper use of the analog resistance / voltage meter.

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The I&C technicians continued with the surveillance until ordered to stop by the-Unit 2 operator when an unexpected half scram occurred. The half scram occurred when two other I&C technicians Jerforming 2MST-APRM290. APRM Flow Bias Flow Units C&D Channel Cali] ration, operated an

[ incorrect switch. This surveillance performs a calibration to assure

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i operability of flow unit inputs to the APRM The I&C technicians j incorrectly operated the C Channel APRM Mode Switch vice the C Channel l Flow Unit Mode Switch. This resulted in the C Channel APRM becoming !

inoperable which caused the half scra The inspector observed the labeling for the two switches and found that they were properly marked.- Discussions with control room operators indicated that since the I&C technicians were not familiar with the quarterly surveillance, the senior reactor operator directed that each step of the procedure be performed with concurrent verification. The licensee performed an investigation of the incident and initiated CR 98-0078 Conclusion The use of a checklist contributed to a thorough pre-job briefing. The personnel performing the Residual Heat Removal and CS Low Reactor i Pressure Permissive Trip Unit Channel Calibration properly stopped the surveillance and informed supervision when an abnormal reading was obtained. Personnel performing the APRM Flow Bias Flow Units C&D Channel Calibration positioned the wrong switch resulting in an unexpected half scra M3 Maintenance Procedures and Documentation M3.1 Diesel Generator 18 Month Insoection Insoection Stone (61726)

On March 31, 1998, the inspector reviewed work activities associated with the #2 DG outage, Observations and Findinas 1 i

The inspector toured the DG Building and work in progress for return of the #2 DG to service. The inspector reviewed Maintenance Surveillance Test OMST-DG500R. Emergency Diesel Generator 18 Month Inspection, at the job-site. The inspector noted that one of the persons signing off the steps in the procedure was also signing for the Quality Control (OC)

steps for verification of torque valve The inspector obtained a listing of qualified OC inspectors and peer OC inspectors from the Nuclear Assessment Section (NAS) organization. The individual in question was not on either lis This was discussed with ;

the individual and his supervisor. These steps were signed because !

another procedure. Maintenance Management Manual 0MMM-17. Maintenance ,

Methods and Guidelines for Torquing, did not require OC verification of :

torquing for equipment less than four inches in diameter. Torque verification for this equipment required a concurrent verification (CV) ;

instead of OC. The individual further stated that his signatures for QC l steps were on piping less than four inches in diamete !

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The licensee initiated CR 98-00820. Initialing of OC Signoff to address this problem. This CR addressed that there should be no need to-

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' interpret" the procedure steps. The work should have been stopped until the procedure was revised. The CR also addressed that the mechanical maintenance technicians had been routinely signing for torques on equipment that was less than four inches in diameter. The licensee considered the DG operable because it passed an o)erability run. Because the components were less than four inches, t1ey were adequately torque Also discussed with the licensee was that procedure OMST-DG500R had been revised on November 20, 1997, with revision 13 to remove QC signoffs for equipment less than four inches in diameter. Apparently, some of the OC signoffs were missed with revision 13 of the procedure. Also noted by the inspector were conflicting requirements on 3rocedure page 18. A note required OC verification for torquing of tie starting air system valves. but the signoff steps only required CV. The inspector discussed these issues with mechanical maintenance supervisio Accordingly, this failure to follow procedure was identified as a violation. TS 6.8.1 requires that procedure for surveillance and test activities of safety-related equipment be used. Nuclear Generator Group Standard Procedure NUA-NGGC-1530. Equipment. Pressure Test. Protective Coatings, and Special Process Inspection, required that QC personnel be qualified and perform the required inspections / verifications set forth in procedures. This violation is identified as VIO 50-325(324)/

98-05-03. Initialing of OC Signoff By Unqualified OC Perso Also discussed with the NAS organization was the basis for not having QC verification for pi)ing less than four inches in diamete NAS 3rovided documentation for t1e establishment of OC hold points. One of t1e factors considered was the simple torque pattern required for smalle size ecuipment. This would-typically be a flange with only four bolts insteat of many bolts -It was discussed that revision 13 to 0MST-DG500R was not reviewed by QC or NA5 although there was a reduction in O steps. The'. inspector discussed this issue with NAS and the corrective action assigned for the root cause analysis will be part of thi c '. Conclusions i The inspector concluded that signing of a OC procedure step by a non-qualified OC inspector was a violation of procedure requirements. There were inconsistent requirements in the mechanical maintenance surveillance procedure for DG inspections. NAS had not reviewed a procedure revision that deleted OC step M3.2 Transformer Deluae System Functional Test Insoection Scooe (61726. 71750)

The inspector observed the performance of Periodic Test 2PT-34.16. Transformer Deluge System Functional Test, on March 26- 199 ,

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12 Observations and Findinas The purpose of the Transformer Deluge System Functional Test was to demonstrate the operability of the deluge system for the Startu Auxiliary. Caswell Beach and Main Transformers. The procedure was written to satisfy the surveillance requirements of the National Fire

. Protection Association (NFPA) 15. Stancard-for Water Spray Fixed Systems for' Fire Protectio The inspector observed'the Caswell Beach Transformer Deluge and Main Transformer A Deluge System portions of the periodic test. The pre-job briefing was attended by the inspector. Four personnel were required to perform the test and all attended the briefing. The brief covered the actions of each individual, the intent of the procedure, and the point

.at which the test would be started. The inspector noted-participation by attendees and. determined the brief to be adequate. The inspector noted that the brief did not cover contingencies or actions in the event that something went wrong with the tes Test team personnel were located at the Control Room, the switchyard, and the Turbine Building. The inspector observed the test performance at the switchyard and the Turbine Building. During observation of the test.in the Turbine Building the ins test were performed out of sequence.pector The test noted that incorrectly performer two steps in the reset a deluge valve by not following the procedure. The 3erformer then initialed steps prior to actually Jerforming them. When t1e individual had difficulties determining why tie valve would not reset, the inspector informed the individual that two steps had been performed out of sequence. Step 7.5.44, was performed prior to step 7.5.43. ' Steps 7.5.40 and 7.5.41. which inspected the valve seat surface areas, were not performed prior to initialing the steps'in the procedur TS 6.8.1.f requires that procedures shall be established, implemented.

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and maintained for the Fire Protection Program. The Fire Protection

technician, observed by the inspector, did not adhere to the procedure L when performing steps.7.5.40 through 7.5.44 of 2PT-34.16.1.1. The i

failure to pro)erly implement 2PT-34.16.1.1 is a violation. This violation is t1e first example of VIO 50-325(324)/98-05-04. Fire J

,- Protection Procedure. A similar violation was given in IR 97-08 which L

was identified as 50-325(324)/97-08-13. Failure to Follow Fire Protection Program Procedures. This violation was discussed in Section F1.2 and F4 1 of that repor After prompting by the inspector, the test member contacted supervision to inform them of.the situation and test status. The test was resume subsequent to evaluation, with a procedure that was marked showing the point of reentry into the test and all the necessary steps which needed to be performed to continue on with the test. It was necessary to trip the deluge valve again to be able to continue with the test. so the test resumed where a fire detector was heated to trip the deluge valve. All other portions of the. test observed by the inspector, were performed adequatel _ _ ____-- _

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The licensee identified that original pages of this surveillance documenting unsatisfactory test performance due to mechanical problems with the deluge valve, which had occurred prior to the start of the inspectors observation, were discarded prior to performing the retes This problem was captured in CR 98-00757 Deluge Valve Surveillanc The inspector discussed the procedural implementation problems with the Operations Manager who agreed that the licensee was having problems with Fire Protection Technician procedural compliance. The Operations Manager discussed with the inspector that a memorandum had been sent to the Shift Superintendents to express the need for them to be involved in communicating the proper standards and expectations for procedural complianc Conclusions A Fire Protection technician performed procedural steps out of sequence during the performance of the Transformer Deluge System Functional Tes PT-34.16.1.1. This was determined to be a violation of Fire Protection Procedure M3.3 Samolina and Analysis of Drywell Monitors Insoection Scone (61726)

The inspector observed the performance of Environmental and Radiation Control Procedure OE&RC-2016. Sampling and Analysis of Dryvell Monitors, on April 16, 199 Observations and Findinas The purpose of OE&RC-2016 was to establish the method, frequency, and-limitations on obtaining the drywell Containment Atmospheric Control (CAC) iodine and particulate air samples. The procedure provided instructions for obtaining a drywell atmosphere gas sample and provided the-limits under which the drywell may be purged. Additionally, the procedures purpose was to assure compliance with the TS 4.11.2.8 and Table 4.11.2-1. Item A. The inspector verified that the procedure met the requirements of the TS The inspector observed a weekly routine sample from the 1-CAC-AQH-1262 Drywell Monitor. This component was located in the Unit 1 Reactor Building on the 20 ft elevation. Three aersonnel were used to conduct the sample. One person was located in t1e Control Room and two personnel were located in the Reactor Building. The inspector observed the sample performance in the Reactor Building. The individual located in the Control Room performed a portion of the procedure, which the inspector did not observe, however the procedure documentation was reviewed.

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14 1 The inspector noted good procedural adherence by the individuals aerforming the filter change out. Satisfactory communications occurred

)etween the Jerson in the Control Room and the Reactor Building throughout tie performance of the sample. Dual concurrence procedural steps .were conducted properly at the drywell monitor. The inspector reviewed the completed procedure after completion of the sampling and determined that the paperwork had no error The inspector verified ~ subsequent to performance of the sampling that the correct TS Limiting Condition For Operation (LCO) was entered and exited to perform the sample. Entering the LCO was necessary since the drywell monitor was required to be secure Conclusions The inspector determined that good arocedural adherence and documentation occurred during the caservation of sampling of a Unit 1 drywell iodine and particulate air sample. The inspector noted satisfactory work practices and communications throughout the evolutio M4 Maintenance Staff Knowledge and Performance M4.1 Modifications Defeat Fire Door Ratina Insoection Scooe (62707. 71750)

The inspector reviewed the circumstances surrounding the modification of the doors and the actions taken once the degradation of the fire doors was' discovered on January 13. 199 Observations and Findinas In April 1996 the licensee began upgrading the fire door hardware as a result of various fire door hardware issues identified previously. In July 1996 the licensee determined that differential pressure problems with the diesel generator (DG) Building ventilation . caused slamming and jamming of the DG Building doors. The jammed doors were difficult to

.open and the door hardware was in a bad condition. The maintenance-organization determined that if certain components from the doors'

lockset were removed, the doors would not jam. The licensee added to the ongoing hardware upgrade efforts the action to perform the lockset modifications. On January 13, 1998, a maintenance supervisor determined that the modifications previously made to the fire doors, had degraded

.the 3-hour rating for twenty-eight doors in the OG Building. The inspector reviewed Limiting Condition for Operation (LCO) A-2-98-F01 .

This LCO was initiated on January 13. 1998 and stated " Doors in the Diesel Generator Building are INOP [ sic] as fire barriers due to the

. removal of a device necessary for the three hour fire rating of the doors.". The licensee initiated a fire impairment for the DG Building and a fire watch was initiated. This event was captured by the licensee in CR 98-74 Fire Doors Lockset. The licensee contacted the vendor and received instructions for the proper modification which would not

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degrade or eliminate the 3-hour fire door rating. Work tickets were initiated for all twenty-eight doors and by January 18. 1998. the locksets were reconfigure to maintain the 3-hour rating. ESR 98-2 Evaluation of Yale Mortise Modifications, was completed to approve 'he reconfigure lockset Hardware Modification Timeline September 8. 1995 Underwriter's Laboratory (UL) at the request of Procurement Engineering, performed an assessment and made recommendations for correction of identified fire door hardware deficiencie March 1996 Hardware upgrades began to fire door June 1996 NAS Assessment (B-FP-96-01) identified . . . fire door hardware conditions issues were a result of DG Building ventilation differential pressure problem August 1996 Licensee began modifying fire door locksets to prevent door jamming due to high DG Building differential pressur November 1997 Maintenance personnel attended sendor training on fire ;

doors and hardwar December 1997 Fire door modifications ar.d upgrades complete Licensee received new " passage only lockset December 12. 1997 NRC inspector identified lack of documentation of fire l

door design basis (IR 50-325(324)/97-13).

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January 13. 1998 Maintenance supervisor returned from vendor training i and compared new locksets with current installed locksets.

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Maintenance supervisor identified past modifications l degraded 3-hour. rating (CR 98-74). Doors declared inoperabl January 18, 1998 All. doors replaced with new " passage only" locksets:

3-hour rating restore The inspector walked-down the DG Building observed the doors in ,

questica, reviewed the WR/Jos which modified the doors. CR 98-74 and ESR {

98-23. WR/JO 97-ADHS1 provided instructions for the modification to fire door 111. .This door provides passage between #1 and #2 DGs. These DGs supply different 4160 volt Number 1 DG supplies the division I components and # power divisions.2 DG supplies the division II This ensures proper separation and redundancy of the power supplies for

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those components needed to mitigate the consequences of an accident.

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Other' doors affected provided separation for the four 4160 volt emergency buses located on the 50 ft elevation, and on the -6 ft elevation.two doors provided separation for three' rooms which each contained two DG fuel oil transfer pumps and a 4-day DG fuel oil tan The 3-hour rating for the fire doors and locksets have various UL labels identifying them as UL fire component Licensee Condition 2.B.(6) requires that the licensee shall implement and maintain in effect all provisions of the approved fire protection 3rogram r

as described in the Final Safety Analysis Report for the acility and as ap22, November as 1997, proved April supplemented in the Safety 1979,~ Evaluation June Report dated 11,1980, and December 30, 198 Plant.0perating Manual 0PLP-01.1, Fire Protection Commitment Document, contains commitments for design, operation, and administrative controls that implement. the fire protection 3rogram. Commitment number FB-03 requires labeled 3-hour fire doors )e provided in all fire areas in the DG Building. The licensee failed to properly maintain labeled three-hour fire doors as designed in all fire areas in the DG Buildin Between September 1996 and January 1998, the licensee performed unreviewed

. modifications to 28 labeled 3-hour fire doors that provided separation between the Diesel Generators Emergency Buses, and 4-Day Fuel Oil Tanks. The door lockset design was modified by removing o)erating levers such that, during a fire, the bimetallic strip in t7e lockset could not lock the door to prevent opening if the panic bar were struc This violation is identified as VIO 50-325(324)/98-05-05. Twenty-eight Fire Doors Outside Desig . Conclusions The inspector concluded that the licensee improperly modified twenty-eight fire doors which introduced a common mode failure. The modification degraded the 3-hout fire barrier, which could have allowed i the spread of a fire from one DG to the other, one emergency bus to the other or one 4 day tank to the other. ' A violation was identified for the failure to maintain a 3-hour fire rated barrier between redundant equipmen M4.2 Failure to Reoort Condition Outside Accendix R Desian Basis Insoection Scooe (62707'. 71707)

The inspector reviewed the operability assessment for the January 1 , discovery that 28 fire doors were modified degrading the 3-hour fire ratin ~ Observations and Findinas l' On February 12, 1998, the inspector questioned the licensee concerning 10 CFR 50.72 and/or 10 CFR 50.73 applicability. The inspector reviewed CR 98-74, and noted that the licensee determined that this event was not )

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immediately reportable. .The inspector determined that although this event was not immediately reportable, this event met the 1-hour non-emergency criteria due to this condition resulting in the nuclear power plant being outside the 10 CFR 50 Appendix R design basis of the plant as a result of the fire door modification CFR 50 Appendix R(III)(G)(2) requires that a means of ensuring that one of'the redundant trains is free of fire damage shall be provide One means of providing adequate separation is the separation of cables and equipment and associated non-safety circuits of redundant trains by a fire barrier having a 3-hour rating. The modifications, made to the 28 DG doors, degraded the 3-hour fire rating for those doors connecting the DG rooms, emergency buses. and DG 4-day tank The inspector

, reviewed NUREG-1022, Revision 1.1998 which provides clarification of

' ' the requirements of 10 CFR 50.72 and 10 CFR 50.73. Section 3.2.4 states that a " violation of fire protection commitments regarding safe shutdown capability may indicate that the plant is outside of its design basi For example, if fire barriers are found to be missing, such that the

. required degree of separation for redundant safe shutdown trains is lacking, the plant would be outside of the design basis with respect to Appendix R to 10 CFR Part 50." The NUREG also states that when applying engineering judgement, and there is doubt regarding whether to re) ort the condition or not, the Commission's policy is that licensees siould

.make the repor CFR 50.72(b)(1)(ii) requires that the licensee, within one hour, report any event or condition during operation that results in the condition, of the nuclear power plant, including its principal safety barriers, being seriously. degraded; or results in the nuclear power plant being in.a condition that is outside the design basis of the plant. Similarly.10 CFR 50.73(a)(2)(ii) requires that the licensee

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make a re) ort, within 30 days of discovery. The failure to report this event witlin one hour and within 30 days in accordance with 10 CFR 50.72 and 10 CFR 50.73 is a violation. This failure is iaentified as VIO 50-325(324)/98-05-06, Failure to Report Unanalyzed Fire Door Conditio Conclusions i

The failure to report a condition outside the 10 CFR 50 Appendix R

, design basis was -identified as a violation. This condition was not reported although questioned by the inspecto M7 Quality Assurance in Maintenance Activities M7.1 History of Fire Door Hardware Problems l a. :Insoection Scooe (62707.'71750)

I The inspector reviewed the adequacies of 3ast fire door issues as recorded in various Condition Reports, ES1s. NAS assessments, and Maintenance self-assessments.

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l j Observations and Findinas The inspector reviewed CR 98-74 and the associated root cause. The root l

cause stated that there was insufficient documentation about the furiction of the locksets and the maintenance supervisor assumed the

" upgrades" and/cr adjustments would not affect the 3-hour fire rating of the doors. The lack of documentation of the fire door basis had been identified previously to the licensee by the NRC in Section F2.3 of IR 50-325(324)/97-13. Unlike previously identified events, review of the January 1998 event revealed that the Equipment Data Base System (EDBS)

properly listed the locksets as Fire Protection Quality (FPO) item Through discussion with the licensee the inspector determined that the maintenance planner did question the need for an ESR. However, the planner was informed that no ESR was needed. The root cause indicated that the maintenance supervisor assumed that the modifications would not affect the fire rating and therefore did not obtain an engineering analysis before performing modifications to the lockset Previous licensee assessments such as NAS Assessment (B-FP-95-01) had identified that non-FP0 parts were being used on fire rated doors. A maintenance assessment M-96-13, conducted in Nuvember 1996 identified '

several instances where non-FP0 parts where used on fire rated doors: CR 96-3928 Configuration of Fire Doors, was initiated to address this concern. Corrective actions included updating the EDBS. determining the proper requirements for the fire doors, and review of the specifications to ensure compliance with the National Fire Protection Association cod J The modification or repair of fire doors with non-FP0 parts without an evaluation was identified previously in a fire protection assessment in July 1995 as captured in CR 95-196 The root cause in 1995 indicated that the fire doors had a history. as identified in a January 1989 maintenance self-assessment, of being modified with non-Q parts without an engineering evaluation. The 1995 assessment identified this as a programmatic problem. The corrective actions included proper updates of the EDBS'to ensure all parts' associated with the fire doors had the 3 roper quality listing. Maintenance was "to ensure the Maintenance

)lanners understand that only Quality Class B3 parts are to be installed on Quality 83 (Fire Protection) Doors."  ;

c. Conclusions The inspector concluded that the licensee had a long standing history of fire door hardware deficiencies. The licensee *s root cause determined !

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that there was insufficient engineering documentation about the importance of com)onents and function of the locksets. However, the inspector noted tlat the various self-assessments identified the l hardware deficiencies, these assessments represented missed l opportunities to correct the adverse conditions identifie ;

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M8 Miscellaneous Maintenance Issues (92902)

M (Closed) Insoection Followuo Item IFI 50-325(324)/97-15-03: ECCS i Response Time Testing '

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During the observation of a surveillance test related to ECCS Response i Time a question arose regarding the adequacy of conducting the test in 1 pieces at different time The TS definition for ECCS Response Time !

requires that times include DG starting and sequence loading delays where applicable. In addition, a question arose if an integrated test of DG start loading was ever performe I The inspector reviewed 2MST-DG13R. DG-3 Loading Test. This test i provides a loss of the emergency bus E3 in conjunction with a start  ;

signal for Unit 2 Division I ECCS loads. The inspector concluded that l this was an integrated tes !

Additionally. the inspector reviewed the test method for ECCS Response Time with the licensee and discussed the same with NRR. The new test method is consistent with the test methods approved for Improved Standard Technical Specifications (ITS). The definition of ECCS Response Time in ITS was changed to specify that testing is allowed by l any means of any series of sequential overlapping or total steps such that the entire system is tested. This IFI is closed M8.2 (Closed) Violation VIO 50-325(324)/97-08-03: Safety Relay Setting Change Made as Pen and Ink Changes to Procedure The licensee replied to this violation on September 2. 1997. A supplemental reply was asked for and received on October 21. 1997. This violation occurred due to not following procedure. The procedure allowed the use of validated information to be obtained for the relay setting. The licensee provided requirements on how to obtain this validated information for relay settings. Procedures were revised to allow the system engineer to use validated information as a source of relay settings. The supplemental relay clarified that the system engineer must use only validated information (design verified). The inspector reviewed the licensee's corrective actions and this item is close M8.3 (Closed)-Violation VIO 50-325(324)/97-08-02: Failure to Verify / Check E Bus Relay Operability The licensee replied to this item on September 2, 1997. This event was that the E-2 electrical bus tripped when it was energized following maintenance activities, The cause was related to the failure of a relay. It was determined that the licensee had not incorporated vendor

! recommended gap setting for the relay contacts. The licensee revised

! 3rocedure OPIC-RLY 026. Relay Calibration Using Pulsemaster Software and l

Julsar Relay Tester, to include vendor gap settings. The inspector f asked for the change and the licensee demonstrated in the electrical

shop that the change had been incorporated into the software. The

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inspector observed this demonstration in the shop and noted that the

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procedure revision number was' cross referenced to the computer software which required the technician to obtain a computer printout which gives t

the correct setting. The inspector reviewed the licensee's root cause

' evaluation and considered this item close M8.4- (Closed) Violation VIO 50-325(324)/96-18-03: Required Nonsafety SSCs Excluded from Maintenance Rule During inspection of the licensee's compliance with 10 CFR 50.65 (the Maintenance Rule), the inspector determined that the licensee had failed

. to include the ambient chlorine detectors, radiation monitors for the turbine ventilation, service water effluent, maia steam and reactor ,

building, and emergency AC and DC lighting. These nonsafety components

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were used in the performance of emergency activities. The failure to include those nonsafety components used in emergency operating procedures (EOPs) resulted in the issuance of a violation. The licensee responded to the.-violation in a letter dated March 17, 199 The licensee performed an extensive review for the entry conditions into the E0 In addition to the items in the violation the licensee identified several other components which needed to be reviewed for inclusion into the Maintenance Rule. The licensee changed the guidance

, in Nuclear Generation Group Standard Procedure ADM-NGGC-0101.

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Maintenance Rule Program, to provide better guidance for review of structures, systems, and components for inclusion into the Maintenance Rule. The inspector verified that the committed corrective actions were

- satisfactorily completed. Based on satisfactory completion of those corrective actions and no issues related to scoping identified -in the Maintenance Rule Program Inspections as discussed in IR 50-325(324)/98- ,

02 this item is close j M8.5 (Closed) Unresolved Item URI 50-325(324)/97-15-05: Inoperable Fire Doors Based on the discussions detailed in Sections M4.1 M4.2, and M7.1 of ;

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this IR and the associated violations' issued. this item is close I

. III. Enaineerina .

El Conduct of Engineering l

El.1 Main Steam Line Drain Steam Leak Jmpection Scooe (37551)

The inspector reviewed the licensee's actions for steam leaks on the Unit 1 main steam line drain pipin l

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21 Observations and Findinos The inspector reviewed the licensee's actions for two steam leaks that developed on the Unit 1 main steam line drain piaing leading to the main condenser. One steam leak had developed in the teactor Building main i steam line pit. The other steam leak was closer to the main condenser in the Turbine Building. Due to the radiation levels in the main steam line pit, visual verification of the steam leak was not permissibl Troubleshooting efforts on December 18, 1997. determined that the leak was downstream of the main steam valve 1-MS-F02 This was based on the fact that when 1-MS-F020 was throttled shut, noise emitting from the main steam line pit disa]peared. Simultaneously, offgas flow increased indicating that air was Jeing pulled into the line and main condense For these reasons, the licensee maintained valve 1-MS-F020 in a throttled condition to minimize the steam leak downstream of the valve and also minimize the infiltration of air to the main condense The steam line was previously scheduled to be replaced in the u) coming refueling outage. The licensee attributed the cause of the leacs to flow accelerated erosion / corrosion. The inspector reviewed the American Society of Mechanical Engineers (ASME) Classification of the piping and found it to be non-code class. ESR 97-00639, inservice Inspection Technical Report, on December 17, 1997, revised the classification of this pipin Piping downstream of 1-MS-F020 to valve 1-MS-F021 was previously classified as ASME Code Class 2 pipin The licensee revised the classification of this piping based on the fact that these components were constructed to non-ASME Section III standards and are non-safety-related. Per 10 CFR 50.55a(g)(1), these components are not subject to the requirements of ASME Section XI. Due to the com31exity and number of changes made by ESR 97-00639, further review of t1e changes is warranted. This is identified as Inspection Follow-Up Item 50-325(324)/98-05-07 Inservice Inspection Technical Repor Conclusion The licensee had taken acceptable actions for steam leaks that developed on a main steam line drain line. An Inspection Follow-Up Item was opened to review recent changes to the inservice inspection progra E2 Engineering Support of Facilities and Equipment E2.1 Control Buildina Air-Conditioning Modification In section Scooe (37551)

The inspector reviewed the modification activities associated with installation of the Control Building Air-Conditioning Units Modificatio _ _ _ - _ .

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22 Observations and Findinas As previously reviewed in IR 50-325(324)/98-03, the licensee installed 0-list Control Building Air-Conditioning units to replace the non-0 units. This activity was completed on April 11, 1998, when the new units became operational. A one-time TS amendment granting a nine week LC0 for replacement of the-units was exited at this time.

, The inspector periodically observed the installation of the modification from the Control Building Ventilation Roof and inside the Control

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Building. Activities associated with this modification were closely and effectively monitored by licensee supervision. A supervisor was routinely seen at the job-site.

I Engineering also closely monitored the design of the modification. The air-conditioning units were purchased as stand-alone units. Howeve various supports were designed and installed. The licensee identified significant technical errors in a contractor supplied civil design portion of the modification. The errors were identified in the calculation and/or installation drawings. The licensee documented these findings in CR 98-00633. CB A/C Mod Civil Design. The licensee stopped the fabrication and installation of the supports. The licensee was able l

to correct all of these errors in time to not impact completion of the modification

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The inspector concluded that the licensee provided effective oversight for the design and installation of the Control Building Air-Conditioning  !

upgrade modificatio '

E3 Engineering Support of Facilities and Equipment E3.1 Slc System Modification Insoection Scone (37551)

The inspector reviewed SLC system operability concerns and an approved

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modification to SLC piping insulatio , Observations end Findinas

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On March 26, 1998. the inspector noted an Operator Log Report entry that

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stated that the SLC tank concentration was above an administrative i

limit, specified in Operating Procedure 20P-05. Standby Liquid Control System, of.18 percent. The log entry additionally stated that an engineering evaluation was necessary to determine operability of the SLC system. Later in the shift SLC tank concentration was lowered below the administrative limit by adding water to dilute the concentration. The operator generated CR 98-00737. SLC Concentration >18 percent, which captured the necessity for an engineering evaluation (

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! The inspector reviewed 20P-05 and Environmental and Radiation Control l- Procedure OE&RC-1130. Chemical Addition and Determination of Sodium l Pentaborate Solution in Standby Liquid Control Tank, and found that the basis for.the 18 percent upper limit was due to SLC piping heat trace l design limitations of 80 degrees Fahrenheit (F). The licensee explained to the inspector that the limit was established because during the coldest winter months the heat trace was not capable of maintaining the pipe temperature above 82' degrees F. based on logged data. The inability to maintain temperatures higher affected the saturation temperatures for maintaining concentrations in the water above 18 percent, thus an administrative limit was establishe An engineering evaluation was necessary to determine system operability because if the sodium pertaborate came out of solution it would form a solid salt compound which would restrict flow. The inspector noted that the evaluation reviewed temperatures and tank concentrations back several months ensuring that concentrations and daily pipe temperatures were adequate to maintain chemicals in solution. The ins)ector found that the licensee determined the SLC system was operable )ased on the dat The inspector noted that TS 3.1.5. SLC System, allowed the SLC tank concentration to be much higher than 18 percent based on tank volume and temperatur The inspector reviewed CR-98-00780. Upgrade of SLC insulation for Improved Standard Technical Specifications. This CR had been generated one day prior, on March 24, 1998..to the SLC concentration exceeding the administrative limit on March 25, 1998. The inspector determined from the CR and in discussions with the licensee that- a modification had been approved to remove the existing piping insulation and replace it with a fiberglass blanket insulation whic according to the licensee, had better thermal properties. Brunswick engineering believed that this modification would solve the temperature limitations on the SLC system heat traced pipe. This modification was proposed to occur in the summer of 199 '

c. . Conclusions The inspector determined that the licensee had implemented a more limiting administrative limit than what existed in the TSs due to the inability of the SLC system piping heat trace to maintain adequate temperatures to maintain sodium pentaborate chemicals in solution during cold-temperature l

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IV. Plant Sucoort R1 Radiological Protection and Chemistry Controls L R1'.1 Butterfly Valve Contamination Controls

' Insoection Scooe (71750)

The inspector toured the Radiological Maintenance Control Service l Building (RMCSB) on A)ril 3.1998 to observe the status radiologically l

. controlled items in tie building, Observations and Findinos The inspector noted a butterfly valve located in the RMCSB which had a Radioactive Material label-on it. dated April 2,1997, which stated that internal contamination was present on the valve. The valve was not wrapped nor did it have any postings around it specifying it as a controlled surface contamination area. The valve internals were exposed. The inspector questioned the licensee if the correct controls were being used for the valve. The licensee expressed that the correct controls were present siace the valve would need to be taken apart to get to the contaminated area On April 15. 1998, the inspector noted that this valve was wrapped on one side with a statement that internal contamination was present, of greater than 1000 decades per minute, under the wrap. Upon questioning the licensee the inspector.was informed that contamination was found in an almost inaccessible area of the valve.- A Health Physics Technician believed the posting to be adequate but conservatively _placed a wrap over the are Conclusions The inspector concluded. based on survey results which showed that the exposed areas of the valve would not have resulted in the s) read of contamination had the valve been brushed up against, that t1e valve was controlled properl i R4 . Staff Knowledge and Performance in RP&C

.R4.1 Personnel Contaminations c Insoection Scoce (71750)

The inspector reviewed the cause for two personnel contaminations that

. occurred on the Unit 1Lrefueling floor on March 10, 199 ;

Observations and Findinos l i l

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l After reading CR 98-00573. Personnel Contaminations, the inspector

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inquired about the cause of the contaminations. The CR indicated that

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contamination was found on all areas of the body. The personnel contamination event records were reviewed for both workers. One worker had contamination on his lower front. face, forearms, and up

.The other worker had contamination on his chin and forearms.per leg The inspector reviewed the special radiological work permit (RWP). RWP B98-1026. for work on the refueling bridge. The preventive' maintenance 3rocedure OPM-CRN001. Preventive Maintenance of Refuel Platform, and WR/JO 97-AIR 02 were also reviewed. This task involved replacing the refuel bridge grapple cylinders and light The inspector interviewed the two workers invc1ved and the health physics technician assigned to the refueling floor. This task had been performed before by the technicians, but the light assembly was normally replaced as~ an entire assembly instead of bulb replacement. This was a parts cost savings. The technicians followed the RWP and procedure requirements for the task. However, disassembly of the light assembly required more time than anticipated. .This resulted in perspiration and contamination leaching through the protective clothing. In addition, working the light assembly and grapple cylinders.in parallel to. minimize time in the area led to more area of the mast not covered by damp rag This was also a contributor to the contamination spread. Additionally, more frequent change-out of gloves may have helped to limit the spread of contaminatio i Each of the individuals was successfully decontaminated. No internal contamination was found. Only a minor do!.e of about two mrem was assigned to the workers. The licensee's corrective action was to revise ,

the work instructions to provide additional guidance on grapple lamp i replacement and cover lessons learned with other workers Conclusions The inspector concluded that work activities on the refuel bridge- i

. grapple that resulted in skin contaminations were covered by the radiological work permit and work instructions. However, work practices to prevent the spread of contamination was not effectiv R8 Miscellaneous RP&C Issues (92904)-  !

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R (Closed) Violation VIO 50'-325(324)/97-G-10: SLC Chemical Addition i

. Labeling Problem I During observation of activities associated with addition of sodium p'entaborate (boron) to the SLC storage tanks, the inspector determined .

that the boron containers were not labeled in accordance with procedure requirements. Subsequent walkdowns revealed other deficiencies. The'

failure'to properly label material to prevent the use of incorrect or defective material was identified as a violation. The licensee '

responded to this. violation in a letter dated May 5. 199 I

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The licensee performed an extensive investigation with an excellent root cause analysis. The licensee revised deficient procedures assigned individuals within each affected organization with responsibility to monitor chemical control program adherence. Routine review of the storage warehouse and improvements in site training were made. The inspector conducted walkdowns of the site and found no chemical labeling discrepancies. The licensee performed several self-assessments. The assessments performed were determined to be self-critical and very effective in identifying nonconformances. Several CRs were generated as a result. Based on satisfactory completion of the associated corrective actions, this item is close P8 Miscellaneous EP Issues (92904:

P (Closed) Insoection Followuo I;em IFI 50-325(324)/96-07-01: Failure to Promptly Recognize and Classif / a Site Area Emergency During the licensee's full par .icipation exercise on January 29, 1998, the Site Emergency Coordinator and his staff in the Technical Sup] ort

} Center were effective in aromptly recognizing and classifying botl the Site Area Emergency and t7e General Emergency. This issue is therefore close S1 Conduct of Security and Safeguards Activities S1.1 Physical Security Observations Insoection Scone (71750)

The inspector observed the conduct of security activitie Observations and Findinos On April 3,1998, the inspector toured the protected area in the evening to inspect the adequacy of lighting. The inspector toured all areas inside the protective area fence and observed cameras inside the central alarm station (CAS) and secondary alarm station (SAS). All areas were clearly visible with one exception. There was a truck trailer located between two storage bins. The area under the trailer was dark. This was discussed with the security shift supervisor. This area had been added to the security rounds and a permanent securit" was located nearb Later the inspector discussed this problem with the site security manage The area was tested and found to have below the minimum required lighting. Although the compensatory measures had been taken. a light was placed under the truck traile On April 14, 1998, the inspector observed security activities in the SAS. The inspector observed the SAS operator perform their duties and via the monitoring system observed activities across the Brunswick Sit On April 29, 1998, the inspector observed security activities in the

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CAS. The same observations were conducted in the CAS as they were in l the SA The inspector found that the Alarm Station Operators were attentive and professional,- The inspector did not observe any discrepancies i associated with security activities during the observations in the CAS and SAS.

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. Conclusions The inspector found security activities to be satisfactory during observations in the Central Alarm Station and the Secondary Alarm Station. Review of lighting inside the protected area fence was found

acceptable with one exception which was correcte F3 Fire Protection Procedures and Documentation F3.1 Fire Pumo Functional Test Failure l Insoection Scope (71750. 61726)

The inspector reviewed the circumstances surrounding the unsatisfactory Performance r of Periodic Test OPT-34.5.5.0, Engine and Electric Fire Pump unctional Test on March 8, 1998. The failure of the diesel driven fire Jump to achieve rated capacity was recorded in CR 98-550 Diesel Fire 3 um Observations and Findinas On March 8,.1998, the-licensee performed OPT-34.5.5.0 to demonstrate the operability and flow capacity for both the engine-driven and diesel-driven fire pump. This test was to satisfy the fire protection surveillance requirements established in Plant Program Procedure OPLP-1.2, Fire Protection System O Requirements. Section 6.1.3. At perabilit :00 Action the operator and Surveillance logs indicated

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that the diesel-driven pump did not pump to rated capacity. Fire impairments Al-98-F098 and A2-98-F097 were issue The acceptance criteria required that the engine-driven fire pump start between 80 and 100 pounds:per square inch gauge ()sig). with a cranking time of 30 seconds or less, and demonstrate tie ability to maintain rated flow and pressure (2000 gpm and 125 psig). The methodology used in the test

. performed on March 8 was consistent with fire pump performance acceptance testing. The fire pump was started, and the flow, suction and discharge pressures at five different loads were recorded. A performance curve plotting the net discharge pressure and total flow was developed from the results. ~The results were then corrected since the-test was performed at less than rated speed. The corrected points were plotted and both curves were compared to the 2000 gpm at 125 psig limit.

L One of the corrected points did not meet the criteria. and therefore the-test was considered unsatisfactory. The licensee initiated a work ticket' to address the identified pump performance deviation.

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The licensee performed a review of the test results in ESR 98-12 Evaluation of test results for engine drive fire pump. This ESR determined that the test methodology used to determine acceptable fire pump capacity was based on verifying the net (suction - discharge)

instead of simply the pump discharge pressure. The previous test used i calculated pump performance curves to satisfy the 2000 gpm at 125 psig acceptance criteria. In addition. it was determined that the table used to convert pressure to gpm in the procedure was not for the flow element used in the test. therefore the flow values calculated were not correc The procedure was revised and the licensee declared the previously unsatisfactory results as satisfactory based on the flow and pressure results calculated using the new methodolog The inspector reviewed ESR 98-121. OPT-34.5.5.0. CR 98-550. OPLP- and NFPA 20-1974. The inspector verified the calculation of the original and corrected flows, and reviewed the test results under the revised methodology as specified in the revised OPT 34.5.5.0. The ins]ector verified that the initial test results using the correct methodology met the acceptance criteria although the pump performance curves did indicate some deviation in pump performance. The inspector concluded that using OPT-34.5.5.0 as written did not provide accurate results upon which to make a pump operability determination. This was the second time in the of procedural problems.past year that a flow test did not pass because In IR 50 325(324)/97-08 Section F1.1 a non-cited violation (NCV) was issued for not adequately maintaining Periodic Test OPT-34.7.1.0 Fire Sup3ression Water System Flow Test. As a result of the inadequate test methodology, the licensee declared the fire suppression water system inoperable and began a dual unit shutdown on June 20, 1997. It was determined, for the June 1997 event, that using the procedure as written did not provide accurate results. In addition, as discussed in IR 50-325(324)/98-03, a NAS audit of fire protection was conducted. NAS concluded that the program was ineffective based on a number of program elements being below acceptable standards. As evidenced by this issue with the fire pump, problems with the procedures continue. TS 6.8. requires that written procedures shall be established, implemented, and maintained for the Fire Protection Program. The failure to properly maintain OPT-34.5.5.0 to verify the surveillance requirements for the fire pumps in OPLP-1.2 is identified as a violation. The inspector determined that this issue was similar to violation NCV 50-325(324)/97-08-11. Inadecuate Fire Protection Flow Test Procedure. This violation is the seconc example of violation VIO 50-325(324)/98-05-04. Fire Protection Procedur c. Conclusion The engue-driven fire aump flow test procedure would not demonstrate that the pump was opera)le until the procedure was revised and acceptance criteria changed. The failure to pro)erly maintain the fire pump procedure was identified as a violation. T11s issue was similar to

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an event in June 1997, where a dual unit shutdown was commenced as a result of erroneous data from an inadequate test procedur Manaaement Meetinas XI Exit Meetina Summary The inspector presented the inspection results to members of licensee management at the conclusion of the inspection on April 29, 1998. The licensee acknowledged the findings presente The Site Licensing Manager stated that the Site Vice President did not agree with the two violations concerning the Diesel Generator Building fire doors and associated deportability of this issue. A meeting was i requested with the NRC prior to issuance of the inspection repor I l

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30 1 1 i PARTIAL LIST OF PERSONS CONTACTED

Licensee .

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- A. Brittain, Manager Security l . Christinziano, Manager Environmental and Radiation Control W. Dorman~ Supervisor Licensing and Regulatory Programs

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N. Gannon, Manager Maintenance l

' J. Gawron. Manager Nuclear Assessment Section

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15. Hinnant, Vice President. Brunswick Steam Electric Plant :

K. ; Jury..; Manager Regulatory Affairs I B. Lindgren, Manager Site Support. Services  ;

J. Lyash,-Plant General Manager '

G. Miller, Manager Brunswick Engineering Support Section

. i R..Mullis, Manager Operations

' Other licensee employees or contractors included office, operatio )

maintenance, chemistry, radiation and corporate personnel !

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E. Brown E. DiPaolo

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E. Guthrie C. Patterson M.- Shymlock j l

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l 31 I INSPECTION PROCEDURES USED IP 37551: Onsite Engineering IP 61726: Surveillance Obser.vations  !

l IP 62707: Maintenance Observations

! IP 71707: Plant Operations IP 71750: Plant Support Activities IP 92902: Followup - Maintenance IP 92904: Followup - Plant Support )

ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-325/98-05-01 VIO Scaffolding Erection Non-Compliance (paragraph l 01.1)

50-325(324)/98-05-02 IFI Adequacy of RSDP Procedure and Training l (paragraph 04.2) '

50-325(324)/98-05-03 VIO Initialing of OC Signoff By Unqualified OC Person (paragraph M3.1)

50-325(324)/98-05-04 VIO Fire Protection Procedure (paragraph M3.2. F3.1)

l 50-325(324)/98-05-05 VIO Twenty-eight Fire Doors Outside Design i (paragraph M4.1)

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50-325(324)/98-05-06 VIO Failure to Report Unanalyzed Fire Door Condition (paragraph M4.2)

i 50-325(324)/98-05-07 IFI Inservice Inspection Technical Report (paragraph El.1)

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50-325(324)/97-15-03 IFI ECCS Response Time Testing (paragraph M8.1)

50-325(324)/97-08-03 VIO Safety Relay Setting Change Made as Pen and Ink Changes to Procedure (paragraph M8.2)

50-325(324)/97-08-02 VIO Failure to Verify / Check E Bus Relay Operability (paragraph M8.3)

50-325(324)/96-18-03 VIO Required Nonsafety SSCs Excluded from Maintenance Rule (paragraph M8.4)

50-325(324)/97-15-05 .URI Inoperable Fire Doors (paragraph M8.5)

50-325(324)/97-02-10 VIO SLC Chemical Addition Labeling Problem (paragraph R8.1)

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32 l 50-325(324)/96-07-01 IFI Failure to Promptly Recognize and Classify a  !

Site Area Emergency (paragraph P8.1) l

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"s^-324/97 '11-06 VIO Rigging to Safety-Related Equipment Without a Safety Evaluation (paragraph E4.1)  :

50-325(324)/97-08-11 NCV Inadequate Fire Protection Flow Test Procedure (paragraph F3.1)

50-325(324)/97-08-13 VIO Failure to Follow Fire Protection Program Procedures (paragraph M3.2)

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