IR 05000456/1987035

Safety Insp Repts 50-456/87-35 & 50-457/87-33 on 870913-1024.Violation Noted.Major Areas Inspected:Action on Previous Insp Findings,Ler Review,Allegation,Unit 2 Fuel Receipt,Followup on TMI Action Items & Physical Security

ML20236V834
Person / Time
Site:	Braidwood
Issue date:	11/20/1987
From:	Hinds J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20236V813	List: IR 05000456/1987035 ML20236V810 ML20236V819
References
TASK-1.A.2.1, TASK-1.B.1.2, TASK-1.C.1, TASK-1.C.2, TASK-1.C.3, TASK-1.C.4, TASK-2.D.3, TASK-2.E.4.1, TASK-2.E.4.2, TASK-2.F.1, TASK-2.F.2, TASK-2.G.1, TASK-2.K.1, TASK-TM 50-456-87-35, 50-457-87-33, IEB-84-03, IEB-84-3, NUDOCS 8712070125
Download: ML20236V834 (18)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Reports No. 50-456/87035(DRP); 50-457/87033(DRP)

Docket Nos. 50-456; 50-457 Licenses No. NPF-72; CPPR-133 Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 l

Facility Name: Braidwood Station, Units 1 and 2 Inspection At: Brridwood Site, Braidwood, Illinois Inspection Conducted: September 13 through October 24, 1987 Inspectors: NRC TT~M. Tongue W. J. Kropp T. E. Taylor 1 E.G.&G. Idaho, In E. Anderson R. Larson K. Spencer

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Approved B - J. M. Hinds,

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11. 2 0 8 7 eactor Projects Section 1A Date Inspection Summary Inspection on September 13 through October 24, 1987 (Reports No. 50-456/87035(0RP); 50-457/87033(DRP))

Areas Inspected: Routine, unannounced .c=fety inspection of activities with regard to regional request; action on pretious inspection findings; licensee event report (LER) review; an allegation; Unit 2 fuel receipt, inspection, and storage; comparison of as-built plant to FSAR description; followup on TMI action items; operational safety verification; radiological protection; engineered safety feature (ESF) systems; physical security; monthly maintenance observation and modification installations; monthly surveillance observation; Unit 2 plant tour; training effectiveness; report review; and meetings and other activitie Results: Of the 17 areas inspected, no violations were identified in 16. In the remaining area one violation was found regarding inadequate short term corrective actions (Paragraph 4).

8712070125 B71201 DR ADOCK05000gg6

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1. Persons Contacted T. J. Maiman, Vice President M. J. Wallace, Manager of Projects D. L. Shamblin, Project Manager l *E. E. Fitzpatrick, Station Manager W. E. Vahle, Construction Superintendent

• C. W. Schroeder, Station Services Superintendent (Dresden)

• K. Kofron, Production Superintendent

• L. E. Davis, Assistant Superintendent, Technical Services B.-Byers, Assistant Construction Superintendent

• M. E. Lohman, Project Startup Superintendent P. Cretens, Station Startup Assistant Superintendent F. Willaford, Security Administrator

• D. Paquette, Maintenance Assistant Superintendent

• D. E. O'Brien, Services Superintendent

• E. L. Martin, Quality Assurance Superintendent l R. Benn, Assistant Security Administrator

• G. Masters, Assistant Operations Superintendent G. E. Groth, Project Field Engineering Manager

• P. L. Barnes, Regulatory Assurance Supervisor

• M. Takaki, Regulatory Assurance

• J. Gosnell, Quality Control Supervisor

• R. E. Aker, Radiation / Chemistry Supervisor J. Jasnoz, Tech Staff AR/PR Coordinator  ;

R. Lemke, Technical Staff Supervisor l G. W. Nelson, Assistant Technical Staff Supervisor E. R. Netzel, Quality Assurance Supervisor

• F. Rescek, Health Physics ,

• S. Inserra, PSG TRB Supervisor '

• J. Vonk, Health Physics Supervisor G. M. Orlov, Staff Assistant to Project Manager P. G. Holland, Regulatory Assura .ce T. W. Simpkin, Regulatory Assurance Operating Experience Group

• R. C. Bedford, Regulatory Assurance R. D. Kyrouac, Quality Assurance Supervisor L. Kline, Regulatory Assurance Industry Group L. W. Raney, Nuclear Safety R. J. Ungeran, Operating Engineer Unit 1 R. Yungk, Operating Engineer

• R. J. Legner, Senior Operating Engineer i R. Mertogul, Tech Staff i T. O'Brien, Tech Staff S. Hedden, Master Instrument Maintenance R. Hoffman, Master Mechanical Maintenance J. Smith, Master Electrical Maintenance W. McGee, Training Supervisor A. Iturrieta, P0AD Field Supervisor B. Tanouye, Project Construc+ ion Department

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A. J. D' Antonio, Quality Control D. H. Schavey, Training

• E. W. Carroll, Regulatory Assurance l

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• P. A. Boyle, Regulatory Assurance

• L. Johnson-Hester, Quality Assurance

• Denotes those attending the formal exit interviews conducted on September 30 and October 21, 1987, and at other times throughout the inspection perio The inspectors also talked with and interviewed several other licensee employees, including members of the technical and engineering staffs, startup engineers, reactor and auxiliary operators, shift engineers and foremen, electrical, mechanical and instrumentation personnel, contract security personnel, and construction personne . Action on Previous Inspection Findings Open Items (Closed) 457/83016-04: Inspection of as-built fuel oil piping against appropriate pipe drawings revealed that nameplate information was not consistent between the as-built conditio, and the drawing A Field Change Order was written to vibro-etch the correct cartridge number on the strainer housing data plate. The inspector verified that the correct cartridge number was vibro-etched on the data plate. The correct cartridge is controlled in stores and listed in the licensee's inventor (Closed) 457/86044-02: Reactor Coolant System (RCS) Hydrostatic Test - Unit A portion of the RCS boundary between Power Operated Relief Valve (PORV) Block Valve 2RY8000B and PORV 2RY456, and between PORV Block Valve 2RY8000A and PORV 2RY455A had not been hydrostatically tested. The inspector reviewed Hydro Test Report No. 60104. The required hydrostatic test was satisfactorily completed on January 9, 198 Unresolved Item (Closed) 457/85016-02: The nuts on the threaded rod that hold the shims for the steam generator, inner frame support corner connection, were not tight. The inspector reinspected the nuts and found that they were tigh Safety Evaluation Reports (SERs)

(Closed) 457/86000-03: Containment Isolation Valves Added to the Redundant Hydrogen Recombiner Supply and Return Lines Outside Containment. The inspector reviewed Sargent & Lundy Drawings M-150,

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Sheet 2, Revision V, 20E-2-4030G11 and 20E-2-4030G012, and field verified that the design and the installation meet the commitments in the SER. The commitments verified included: (1) isolation on

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i Phase A isolation. signal; (2) positive position indication; (3)

remote manual control switches in the main control room; and (4)

valves powered from the same Class IE emergency power source as the inside isolation valve on the same line.

l (Closed) 457/86000-04: Containment Isolation Valves Added to the Chilled Water Return Lines Inside Containment. The inspector reviewed Sargent & Lundy Drawing M118, Sheet 7, Revision W, and verified the design included the commitments. The inspector verified Valves 2W0056A and 2WO56B were installed. The commitments verified included: (1) isolation on Phase A isolation signal; (2)

positive position indication; (3) remote-manual control switches in the main control room; and (4) the valves powered from a different Class 1E power source than the associated outside containment isolation valv (Closed) 457/86000-19: Verify that Steam Generator Power Operated Relief Valves (PORV) have hydraulic operators installed. The inspector reviewed Sargent & Lundy Drawing M-120, Sheets 1, 2A, and 28, and determined that these drawings required the PORVs to have hydraulic operators. The inspector also reviewed Borg Warner Drawing 85440. The inspector verified that electro-hydraulic operators were installed for PORVs 2MS018A, 2MS018B, 2MS018C, and 2MS018 CFR 50.55(e)

(Closed) 457/82002-EE: Reinspection of Containment Spray Support A total of 2872 safEy-related hangers installed in Braidwood Units 1 and 2 prior to June 1981 were reinspected in accordar.ce with QC '

-< Procedure B23. Closure of NCRs 776, 2249, and 6075 indicated satisfactory completion of the reinspection for Unit A total of 92 containment spray supports remained to be reinspected for Unit This Unit 2 effort was initiated under NCR 2258. NCR 6699 later superseded NCR 2258 and NCR 2258 was close NCR 6699 was closed September 3, 1987, indicating the completion of the Unit 2 reinspection effor The inspector reviewed all NCRs and a sample of the Component Support Drawings associated with the 92 Unit 2 containment spray hangers and concludes the effort was satisfactorily complete No violations or deviations were identifie Bulletins (IEB)

(Closed) IEB (456/84-03-BB; 457/84-03-8B): Refueling Cavity Water Seal. The Braidwood Nuclear Power Station was reviewed with respect to concerns raised by the Haddam Neck reactor cavity drain down event of August 21, 198 The following is a summary of the items considered during the review:

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(1) A Presray manufactured seal is used at Braidwood to seal the I reactor cavity floor, but the dimensions and rubber hardness  ;

are significantly different from the seal that failed at Haddam l Neck:

Braidwood Haddam Neck

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Seal Flange Width 4 inches 3.5 inches Rubber Hardness (Durometer) 60 40 1 (2) The cavity seal gap is fixed, nominally two inches in widt The gap dimensions have been verified, including side to side vertical offset, for both units. The Haddam Neck seal design was considerable more complex and more subject to failur (3) Testing was performed by Impell on a sample of the Braidwood seal to test its strength, particularly that of the overlap joint formed when both end of the seal are connected during installation. Testing showed that there was ample margin in the seal and seal joint design, but that the joint was subject to failure at high bladder pressures. The licensee has adopted measures, including the use of a safety valve, to assure that overpressurization will not occur. The test report indicates that even if overpressurization did cause failure of the joint,

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a propagating failure would not be expected to occur. (This

'was based on engineering judgement, however, and not fully tested.) 4 Previous testing at the same seal, but without the lap joint, showed excellent results for the Zion Nuclear Power Station as reported in Inspection Reports No. 50-295/85-010; No. 50-304/85-01 (4) The licensee evaluated the fuel pool physical design and determined that there was no mechanism, other than pit '

structural failure, that would cause a loss of pool inventory below the top of the fuel assemblies. This is reported in FSAR 1 Section 9.1.3.3 which also indicates the backup sources of water to be used in the event of a loss of pool inventor This was also confirmed by inspection of the pool by the inspecto (5) Abnormal procedures are approved, addressing a loss of reactor cavity level or fuel level during refueling. Level indications and alarms are available locally. The licensee also plans to install level alarms in the Main Control Room for alerting the plant operators.

l (6) Inspection of the fuel transfer tube manual isolation valve indicated that no positive indication of its position was provided near the handwheel nor was an arrow provided for the opening or close directions. The licensee agreed to provide

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' se e some provision for readily determining valve position and rotation requirement and that this would be done prior to the next plant refueling outage. It is tracked by their Action h Item Number 456-401-85-0010 (7) A question was raised whether the Presray Company was removed from the approved bidders list. The licensee subsequently indicated that Presray was currently.on the approved lis Overall, the licensee appeared to have done a comprehensive evaluation of the potential seal failure problem as well as the potential loss of water by other means. Appropriate actions were taken in response to concerns identified and to assure that appropriate emergency measures were addressed. This was done both in response to IE Bulletin 84-03 and INP0 SOER 85- j 3. Regional Request-By telephone request on October 2, 1987, Region III requested information related to the use of silicon rubber insulated cable at Braidwood. This request resulted from a 10 CFR 21 Notification from the Tennessee Valley Authority (TVA), which identified failures in wire insulatio Other information was provided to the licensee to assist in its assessment of the installed cablin The licensee conducted a record search and found that the wire insulation installed at Braidwood was not as described in the 10 CFR 21 Notification and that it is a chemical cross-linked polyethylene substanc . Licensee Event Report (LER) Review '

Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine u that deportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been or would be accomplished in accordance with technical specifications: ,

(0 pen) 456/87044-LL: Closure of RH Discharge Isolation Valve Due to Inadequate Operating Procedure Review by Shift Personnel. This event was reviewed and addressed in Section 4 of Inspection Report 456/87029(DRP);

457/87027(DRP). A violation was issued at that time (456/87029-01(DRP)).

The Licensee has committed to a subsequent LER revision by December 31, 1987. This LER will remain open until that revision is reviewe (Closed) 456/87045-LL: Axial Flux Difference Surveillance Not Initiated when Process Computer was Rebooted due to Misinterpretatio Upon identification, the licensee performed the appropriate AFD surveillance with acceptable results. This is a repeat of the occurrence identified in LER 456/87041 and is discussed further at the end of this sectio _ _ __ _ . _ . _ _ _ _ _ _ _ _ _ - - - _ _ - - _ - - - - _ - - _ .

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(Closed) 456/87046-LL: Increase in RCS Cooldown Rate Causing Lolo Steam j Generator Level and Auto-start of 1B Diesel Driven AFW Pump. This ESF j actuation was determined to have been caused by insufficient instruction i for the special test in progress, RC-35, " Shutdown From Outside the j Control Room." The licensee has generated a test deficiency against the ^

equivalent test on Unit (Closed) 456/87048-00-LL and 456/87048-01-LL: Loss of Offsite Power Due To Inadvertent Deluge System Actuation Resulting From a Disposition Valv The cause was found to be a mispositioned auxiliary drain valve. The licensee is pursuing several corrective actions, such as procedure revisions and possibly removing that trip from the transformer circui Revision 01 to the LER was issued to add the titl (Closed) 456/87049-LL: Exceeded Analysis Frequency on Waste Gas Oxygen Analysis. Upon identification, the licensee developed formal written color coded packages for LC0AR and surveillance identification, a formal review process, and training meetings with associated personnel. This event is similar to the events described in LER 456/87001 and 456/87043, and is discussed further at the end of this sectio (Closed) 456/87050-LL: Reactor Trip Due to Inadvertent Transformer Deluge Actuation During Maintenance Activity. The licensee contends that this was a unique trip caused by maintenance work to correct a previous event (reference LER 456/87048). The plant was stabilized and the maintenance work was completed. The licensee is performing an engineering evaluation to remove that trip from the transforme (Closed) 456/87051-LL: Control Room Ventilation Switchover Due to Spurious Noise on Radiation Monitor Channel ORE-PR0338. The licensee has been unable to determine the cause of this spike. This is being evaluated with other previous events and if a cause is identified, a revised LER will be submitte (Closed) 456/87052-LL: Reactor Trip Due to Main Power Transformer Overexcitation Relay Actuation for Unknown Reason. The licensee evaluated the event and if a cause is identified, a supplemental report will be submitted. The inspectors reviewed the licensee's trip analysis and return to power, and found no discrepancie (Closed) 456/87053-LL: Containment Isolation From Loss of Power to Radiation Monitor 1RT-AR011. The cause is attributed to possible loose wiring connections and painting activities in the vicinity of the monitor. The electrical connections were cleaned, replaced, and steps were taken to control activities in the area of the monito Regarding LERs 456/87045 and 456/87049, both are repeats of other events and are the result of incomplete corrective actions. In each previous I case, the licensee identified what appeared to be acceptable corrective l actions (operator aid and a sample matrix) that were to be implemented I

i after some development. In both cases, recurrence resulted from the

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failure of short term corrective actions to be effective until permanent corrective actions are implemente It should be noted that LER 87045 is related to the Operations Department I and LER 87049 is related to the Radiation Chemistry Department, indicating that the other departments at Braidwood may be vulnerable to similar problem The failure to take adequate short term corrective action is considered to be in violation of 10 CFR 50, Appendix B, Criterion XVI (456/87035-01(DRP)).

In addition to the foregoing, the inspector reviewed the licensee's Deviation Reports (DVRs) generated during the inspection period. This was done in an effort to monitor the conditions related to plant or personnel performance, potential trends, etc. DVRs were also reviewed for assurance that they were generated appropriately and dispositioned in a manner consistent with the applicable procedures and the QA manua No violations or deviations were identifie . Allegation

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(Closed) RIII-87-A-0028: On March 12, 1987, an individual visited the Region III office with a concern that the site mechanical contractor's Site Quality Manager was not qualified for his position. The alleger stated that the Site Quality Manager's background was in welding supervision not in quality control'. The alleger also stated that the Site Quality Manager did not know how to handle people and this was causing many problem The alleger cited an example of a situation where the Site Quality Manager's handling of an individual was questionabl The alleger identified a potential witness who could substantiate his concern On April 7, 1987, the Region III Investigation and Compliance Specialist interviewed te witness via teiephone. The witness, a former Field Quality Co c ol Supervisor stated he terminated his employment because of the Site Qus;11ty Manager's management style. The witness identified his concerns as: The Site Quality Manager was not qualified for his position. The witness stated during a meeting the Site Quality Manager could not j define the term " Safety-Related" and stated that only ASME code work has an effect on plant safety. The witness furnished an individual's name who also attended the meeting and therefore could substantiate the Site Quality Manager's statement j The witness stated that the Site Quality Manager affected the I documentation and that the document reviewers were not experienced enough to perform detailed document review .

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The witnessistated that he provided the licensee's Quality First organization with his concerns prior to his departure from the Braidwood sit '

On April 8,'1987,.the alleger' contacted'NRC Region III by telephone for the' purpose of providing more.information and concerns. The alleger stated the fol. lowing: n The hardware installed by the mechanical contractor was.in good 1 shape and could not provide a specific example of where hardware was affected by actions of the Site Quality-Manage A review of documentation in the quality department will show that !

the paperwork was inadequat The paperwork would not stand alone-and that three people have to explain why something was satisfactor The alleger did state the document reviewers were qualified and traine l The inspector reviewed the alleger's and.the witness's concerns with the'

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following resuits: The Site Quality Manager was not qualified for his positio '

The. inspector reviewed the site mechanical contractor's Quality Assurance Manual (QAM) to determine if there were specific qualification requirements for the positions of Site Quality Manage There was no. specific qualification requirements; .

however, in.Section 1 of the QAM, the responsibilities of the l Site Quality Manager were delineate The QAM identified his responsibilities as:

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• Implementation of the Quality Program, including Quality Assurance, Quality Control, and Non-Destructive Examination activitie * Organizes and directs the quality staf * Assures the availability of suitable manpower and skill * Interfacing with the Authorized Inspection Agenc The responsibilities described above are typical in the industry for a Quality Manager in a mechanical contractor's organization. At present, there are no regulatory requirements which stipulate minimum qualification requirements for such an individual. The inspector reviewed the resume of the Site Quality Manager. Prior to becoming the Site Quality Manager for the Braidwood site mechanical contractor, the individual had approximately 16 years experience in nuclear welding supervision (Byron, LaSalle, and Zion) and four years as a nuclear welder.(Dresden and Kewaunee). Even though the Site Quality Manager's experience is in nuclear welding and weld supervision, the inspector does not consider this a liability in j

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fulfilling the responsibilities delineated in the QA The inspector considers welding experiences as an asset in implementing the Site Quality Manager's responsibilities since one of the most important functions of a nuclear site mechanical contractor is their ability to comply with the welding requirements of the nuclear industry. In conclusion, the inspector could not substantiate that the Site Quality manager was not qualified for his positions. This conclusion is based on:

• There is no regulatory requirement for the qualifications of a Quality manage * The Site Quality Manager had approximately 16 years nuclear welding experience. Welding is an integral part of a nuclear site mechanicals contractor's wor * There have been no significant mechanical hardware problems identified by the NRC in the last two years. Also, the alleger did not identify any hardware problems during his conversations with Region III personnel, b. The Site Quality Manager could not define the term " safety-related" and stated that only ASME Code work has an effect on plant safet The inspector interviewed the individual the witness identified as being present when the Site Quality Manager stated only ASME Code Work had effect on plant safet This individual substantiated that the Site Quality Manager had made the statement concerning ASME code wor However, the individual believed the Site Quality Manager's statement concerning the ASME Code Work was pertaining to the amount of ASME Code Work required prior to licensing the plant and not a reflection of the Site Quality Manager's knowledge of regulatory requirements. In conclusion, even though the inspector has substantiated this concern, the statements of the Site Quality Manager were subject to different interpretations of the meaning by two individuals who were present when the Site Quality Manager made the statemen c. The Site Quality Manager affected the documentation and the document reviewers were not experienced enough to perform detailed document review The inspector reviewed certification folders for ten document reviewers. These document reviewers were certified as Level II in accordance with Phillips Getschow Procedure, QAP/BR-QCT-20.1 These individuals had 48-hours of on the job training, approximately 8-hours of classroom training, and were given a general and practical examination prior to their certificatio A recent NRC inspection (April through June 1987) of Phillips Getschow's documentation associated with the N-5 Data Report did not identify any significant problems. This inspection is documented in Inspection Reports 456/87015; 457/87015.

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In conclusion, the. inspector could not substantiate the Site Quality manager had affected the documentation. Also, a review of ten document. reviewer ~ certification files did not reveal any improperl certified individuals, The replacement for the witness was not qualified for the positio of a QC Superviso The witness had given this concern to Braidwood's Quality First organization prior to'his departure:from the Braidwood site. After-the witness departed the Braidwood site, the Field QC Supervisor position was expanded to. include-ASME N-S' activitie The title was then subsequently changed to Field QC Supervisor / Field N-S Coordinator. The Quality First investigation concluded that.the individual who replaced the' witness as the Field QC Supervisor / Field N-5 Coordinator was not performing any quality acceptance function !

These functions were performed by certified Level II:. inspector In conflict situations requiring an interpretation, the Phillips Getschow.QC Manager Office / Field would be the ultimate interpretation authority. On June 29, 1987, the position of Field

.QC Supervisor / Field N-5 Coordinator was eliminated and the individual who held that position was transferred;to the Production Welding Engineering group. Since Quality First investigated this ,

concern, the inspector reviewed the results of their investigatio The inspector. determined the Quality First investigation to be satisfactory and with the elimination of the position, Field QC Supervisor / Field N-5 Coordinator, determined no further inspection activities were deemed necessar No violations or deviations were identifie . Unit 2 Fuel Receipt, Inspection, and Storage During the inspection period,'a number of shipments of new fuel were received by the licensee. The inspector. selected a' sample of shipments and verified that the licensee'had appropriate documentation and record control,'e.g. shipping documents, D0T and NRC required documents, and Quality Assurance document The inspectors also monitored the licensee's receipt of several shipments of fuel for shipping container external damage, security integrity, shock indicator integrity, and loose material or part The inspectors also monitored the opening of some of the containers, initial fuel inspection, handling, radiation controls, cleanliness control, and storage of the fue No violations or deviations were identifie . Comparison of As-Built Plant to FSAR Description This inspection was conducted in order to ascertain that selected <

mechanical and electrical system installations are in agreement with current P& ids, electrical drawings, and FSAR commitments. Also, that

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L .as-built ~ plant, and_ control and logic instrumentation of selected l systems conform to the' descriptions contained in the FSAR..

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For the systems reviewed, .the inspectors verified that the cognizant test:

and/or system engineer was aware of system changes or discrepancies that-by field observation that component installation, including control'and

, instrumentation,~is as described in current system drawings, Technical

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Specifications, and FSA With Unit 2 construction greater.than 90%' complete and the sample of Technical Specification related systems selected, the inspectors were-able to determine that'the facility is constructed essentially in conformance with the FSAR. 'Since Braidwood (BWD) Unit 2.is a replicate of BWD Unit 1 and Byron Units 1 and 2, the FSAR versus Technical

, Specification comparison for systems selected was not required for

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systems that were reviewed during the other units' review. Minor discrepancies, such as valve and breaker tagging were identified to the licensee for resolutio The following is a list of the systems and specifications reviewed:

-3/4.5, " Emergency Core Cooling System" i-

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3. . /4.6, " Containment Systems" 3.6. .6. .6. .6. .6. .6.2.2-i 3.6.4.1- 3.6.3 L 3/4.7.1, " Turbine Cycle" 3.7. .7. .7. .7. '

3/4.8, " Electrical Power Systems" 3.8. .8.3.1

! 3.8. No violations or deviations were identifie . ' Followup on TMI Action Items The fuel load related items listed below for NUREG-0737 are considered closed for Unit 2. These items were reviewed and closed in previous '

Unit 1 inspection reports. The administrative controls and equipment installations for Unit 2 are identical to those used in Unit 1: l

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II.K.1-21 II.E.4.1-3 I.C.3-1 II.K.1-5 II.E.1.2-1B I.C.2-1 II.K.1-10 II.E.1.2-2 I.C.1-1 II.K.1-17 II.E.1.2-2B I.C.4-1 II.K.1-20 II.E.1.2-1A I.B.I.2-1 II.F.2-1 II.F.1-1 I.A.1.2-1 II.F.2-2 II.D.3-1 I.A.2.1-4 II.E.4.1-2 II.G.1-1 II.E.1.2-2A II.G.1-2 l

I.A.1.1 Shift Technical Advisor (STA) i The staff has found that the licensee's program for STA on shift is acceptable. The licensee uses the Station Control Room Engineer (SCRE),

a licensed SRO, to comply with the STA requiremen No violations or deviations were identifie . Operational Safety Verification The inspectors. conducted routine plant tours during the inspection period to make an independent assessment of equipment conditions, plant conditions, construction activities, security, fire protection, general personnel safety,' housekeeping, and adherence to applicable regulatory requirements. During the tours, the inspectors reviewed various logs, daily orders,. interviewed personnel, attended shift briefings and plan of the day meetings, witnessed various construction work activities, and independently determined equipment status. During the shift changes, the inspector observed operator, shift control room engineer, and shift l engineer turnovers and panel walkdown These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under technical specifications, 10 CFR, and administrative procedure No violations or deviations were identifie ;

10. Radiological Protection The inspectors selected portions of the licensee's radiological program to verify conformance with facility policies, procedures, and regulatory  ;

requirement Various functions observed were health physics managers i awareness of any unusual conditions or challenges, implementation of the ALARA program, use of Radiological Work Permits (RWPs), control and monitoring of radiation exposured, including work in high radiation areas 4 if applicable, and control of radioactive materia No violations or deviations were identified. l 13 l

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11. Engineered Safety Feature (ESF) Systems During the inspection, the inspector selected accessible portions of several ESF systems to verify their status. Consideration was given to the plant mode, applicable Technical Specification, Limiting Conditions for Operating Action Requirements (LC0ARs), and other applicable requirement Various observations, where applicable, were made of hangers and supports; housekeeping; freeze protection, if required, was installed and operational; valve positions and conditions; potential ignition sources; major component labeling, lubrication, cooling, etc.; interior conditions of electrical breakers and control panels; instrumentation was properly

' installed, functioning and significant process parameter valves were consistent with expected values; instrumentation was calibrated; necessary support systems were operational; and breaker and valve positions concurred ,

locally and at the control panel '

During the inspection, the following ESF systems / components were walked down:

Unit 1 1A Emergency Diesel Generator 1A and 1B Residual Heat Removal Systems IA Charging and Volume Control System 1A Safety Injection System -

Unit 2 2A Emergency Diesel Generator No violations or deviations were identifie ,

12. Physical Security At various times throughout the inspection period, the inspectors monitored compliance with the Physical Security Plan (PSP). Observations were made of selections of manning levels and collateral duties of  ;

assigned personnel; access control equipment and processes, such as l x-ray machines, metal detectors, explosive detectors, and other search mechanisms; PA and VA barriers are properly maintained; procedures are properly followed; compensatory measures are appropriately used when required; persons in the PA and VA are properly badged and escorted if required; and various detection / assessment aids are operable, such as ,

fences, illumination of the PA and TV monitors have sufficient clarity '

and resolutio No violations or deviations were identifie a

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13. Monthly Maintenance Observation and Modification Installations j

Station maintenance activities of safety-related systems and components listed below were observed / reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with Technical Specification The following items were considered during this review: the limiting conditions for operation were met while components or systems were removed from and restored to service; approvals were obtained prior ;

to initiating the work; activities were accomplished using approved {

procedures and were inspected as applicable; functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and, fire prevention controls were implemented. Work requests were reviewed to determine status of outstanding jobs and to assure that priority is assigned to safety-related equipment maintenance which may affect system performanc Maintenance activities on the following equipment were observed and reviewed: ,

Unit 1 Nuclear Instrumentation - Source Range Channel 32 Noise Evaluation j Digital Rod Position Indication (DRPI) Trouble Shooting and Repair Result Area Radiation / Process Radiation Monitors Trouble Shooting and Taskforce Evaluations Unit 2 2A Emergency Diesel Generator - Rocker Arm Replacement i 2B Emergency Diesel Generator - 18 month inspection and head replacement  !

The inspectors monitored the licensee's work in progress, verified that it was being performed in accordance with proper procedures, approved work packages, 10 CFR 50.59 and other applicable drawing updates were made and/or planned and that operator training was conducted in a reasonable period of tim No violations or deviations were identifie l l l

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... . Monthly Surveillance Observation The inspectors observed surveillance testing required by technical specifications for Unit 1 during the inspection period and verified that  ;

testing was performed in accordance with adequate procedures, that test i instrumentation was calibrated, that limiting conditions for operation were met, that removal and restoration of the affected components were accomplished, that results conformed with technical specifications and procedure requirements and were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate ,

. management personne J The inspectors also witnessed portions of the following test activities:

Unit 1 IB Diesel Generator - Monthly Operability Run Unit 2 2A Emergency Diesel Generator - 100 Hour Run i No violations or deviations were identifie . Unit 2 Plant Tour i The inspector observed work activities in progress, completed work and plant conditions during general inspections in Unit 2 work area Observation of work included cable trays, junction boxes, pipe support welding and mechanical equipment. Particular attention was given to material identification, nonconforming material identification and housekeeping. The-inspector reviewed work activities by reviewing travelers while touring the plant. These travelers pertained to electrical, piping, and hanger installatio No violations or deviations were identifie . Training Effectiveness l The effectiveness of training programs for licensed and non-licensed  ;

personnel were reviewed by the inspectors during the witnessing of I the licensee's performance of routine surveillance, maintenance, and operational activities and during the review of the licensee's response to events which occurred during the inspection period. Personnel appeared to be knowledgeable of the tasks being performed, and nothing was observed which indicated any ineffectiveness of trainin No violations or deviations were identifie a - - - - - - _-- ----_ - _ -- -

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1 Report Review l q

i During the inspection period, the inspector reviewed the licensee's 1 Monthly Operating Report for August and September 1987. The inspector confirmed that the information provided met the requirements of Technical Specification 6.9.1.8 and Regulatory Guide 1.1 The inspector also reviewed the licensee's Monthly Plant Status Report for August and September 198 No violations or deviations were identifie f 18. Meetings and Other Activities l Management Meeting - Ascension To Greater Than 50% Power On October 5, 1987, a management meeting was held in the Region III office in Glen Ellyn, Illinois between members of the Region III staff, NRR Headquarters staff, and Commonwealth Edison Corporate and Braidwood Station staff. The purpose of the meeting was to provide an opportunity )

for the licensee and the NRC to discuss the status of Braidwood Unit I !

prior to proceeding above 50% power. The licensee provided information on the Unit I history, current status, performance, strengths, and challenges. The NRC staff discussed the following areas needing management attention: surveillance program administration, events (with emphasis on personnel errors), the numbers of alarm annunciators illuminated in the control room, and responses to NRC inspection findings, addressing the problem versus the appropriateness of the findin The meeting concluded with each of the parties agreeing that there were no outstanding issues that would prevent the unit from proceeding above 4 50% power. It was also agreed that a similar meeting should take place l when Unit 2 achieves this platea Plant Status Meeting On October 23, 1987, a meeting was held at the Braidwood site to discuss the status of Unit 2 for fuel load. *he meeting was attended by the

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Region III Division of Reactor Projecin Branch Chief, Section Chief, Reactor Engineer, and the Resident Inspectors and Commonwealth Edison's Manager of Projects, Project Manager, Station Manager, plus other '

corporate and site personnel. The meeting included a discussion of the Unit 2 construction status, turnovers, testing program, release to operations, and other related subject The NRC provided discussion :

issues related to performance history and other timely topics such as; plant appearance, numbers of maintenance work requests - breakdown of numbers and scheduling, responses to notices of violation, reduction of numbers of illuminated annunciators in the control room, and minimizing

personnel errors. The activities also included a plant tour.

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k r . Exit Interview The inspector met with licensee and contractor representatives denoted in Paragraph I during and at the conclusion of the inspection on October 21, 1987. The inspector summarized the scope and results of the inspection and discussed the likely content of this inspection report. The licensee acknowledged the information and did not indicate that any of the information disclosed during the inspection could be considered proprietary in natur ;

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