IR 05000455/1986040

Safety Insp 50-455/86-40 on 861016-31.No Violations Noted. Major Areas Inspected:Licensee Action on Previous Insp findings,50.55(e) Repts & Part 21 Repts.Concern Re Potential Failure in Auxiliary Feedwater Sys Discussed

ML20214G150
Person / Time
Site:	Byron
Issue date:	11/17/1986
From:	Forney W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20214G124	List: IR 05000455/1986040 ML20214G120
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-***, TASK-TM 50-455-86-40, IEB-83-05, IEB-83-5, IEB-86-006, IEB-86-6, NUDOCS 8611250591
Download: ML20214G150 (25)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION III

Report No. 50-455/86040(DRP)

Docket No. 50-455 License No. CPPR-131 Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facili ty Name: Byron Station, Unit 2 Inspection At: Byron Station, Byron, IL Inspection Conducted: October 16 - October 31, 1986 Inspectors: J. M. Hinds, J P. G. Brochman J. A. Malloy R. M. Lerch B. H. Little M. M. Holzmer C. H. Scheibelhut Approved By:

RFl)km.ak &

W. L. Forney, Chief Reactor Projects Section 1A

//// 7/#4 Date Inspection Summary Inspection on October 16 - October 31, 1986 (Report No. 50-455/86040(DRP))

Areas Inspected: Routine, unannounced safety inspection by the resident inspectors of licensee action on previous inspection findings; 50.55(e)

reports; IEBs; Part 21 reports; SERs; NUREG-0737 items; comparison of as-built plant to the FSAR; Technical Specification Review; housekeeping; independent ir.spection; Region III requests; Commissioners Tour; and Management meeting Results: No violations cr deviations were identified nor were any items identified which could affect the public health and safety. However, a concern with a potential failure in the auxiliary feed system is discussed in paragraph 1 DR 861118 ADOCK 05000455 PDR

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pETAILS Persons Contacted Commonwealth Edison Company

1. T. Maiman, Vice President and Manager of Projects

• 1. R. Querio, Station Manager

1. V. Schlosser, Project Manager, Byron

• 1. R. Pleniewicz, Production Superintendent

• 1. R. Ward, Services Superintendent

• 1. R. Tuetken, Startup Superintendent

• 1. E. Martin, Quality Ass 4cance Superintendent

• W. Burkamper, Quality Assurance Supervisor, Operations

1. A. Rosenbach, Quality Assurance Supervisor, Construction

• L. Sues, Assistant Superintendent, Operating

• G. Schwartz, Assistant Superintendent, Maintenance

• T. Joyce, Assistant Superintendent, Technical Services D. St. Clair, Assistant Superintendent, Work Planning

1. P,. Moravec, Project Construction Assistant Superintendent

• R. Klinger, Project Quality Control Supervisor W. Blythe, Operating Engineer, Unit 0 T. Tulon, Operating Engineer, Unit 1 D. Brindle, Operating Engineer, Unit 2

• J. Schrock, Operating Engineer, Rad-Waste A. Chernick, Regulatory Assurance Supervisor

• 1. E. Falb, Unit 2 Testing Supervisor

• F. Hornbeak, Technical Staff Supervisor R. Flahive, Radiation / Chemistry Supervisor P. O'Neil, Quality Control Supervisor

• G. Grabins, Assistant Unit 2 Testine Supervisor

• W. Kouba, Assistant Technical Staff Supervisor

• E. Zittle, Regulatory Assurance Staff

1. D. Milroy, Startup Engineer

• 1. W. Pirnot, Station Regulatory Assurance

• J. Pausche, Regulatory Assurance Staff

• M. Whitemore, GSEP Coordinator

• J. Snyder, Quality Assurance Inspector i * K. Yates, Onsite Nuclear Safety

• J. Langan, Regulatory Assurance Staff The inspector also contacted and interviewed other licensee and contractor personnel during the course of this inspectio # Denotes those present during the management meeting on October 23, 198 * Denotes those present during the management meeting and exit interview on October 31, 198 '

. Action on Previous Inspection Findings (92701 & 92702) (Closed) Unresolved Item (455/85027-01a(DRP))

FSAR and construction code commitments concerning splicing of Class 1E wiring in panels not translated into appropriate procedures. The licensee submitted a change to the FSAR, Amendment No. 47, which enables splices to be used on individual conductors of external field run cables within switchboards for the purpose of extending individual conductors to their point of termination. The NRC Staff has reviewed and approved FSAR Amendment No. 47; and therefore, this item is considered close (Closed) Violation (455/86031-02(DRP)): Flammable material discovered in cable spreading room and not stored in approved containers or properly supervised. The licensee's immediate corrective action was to remove the flammables from the cable spreading room to remove the immediate threat to equipment important to plant safety. The licensee's action to prevent further violation includes short range and long range measure The short range actions have been completed and included: distribution of a memorandum on October 30, 1986, from the Station Manager to all station and contract personnel on control of flammable liquids in the station; performance of documented training on October 30, 1986, of all craft personnel on the safety rules which apply to flanmables used in the plant; and

, the Security Administrator addressed the security and fire watch personnel on the subject matter on October 29, 1986. A separate short range measure includes conduct of awareness sessions for all contractor personnel by the Services Superintendent from October 29, 1986, through November 14, 1986, on the subject of control of flammable liquids in the plant. The long range measures, which will be on-going, include a revised Nuclear General Employee Training (NGET) lesson plan (reviewed by the inspector and found acceptable)

which emphasizes the control of flammable liquids in the plant. It will be incorporated into the NGET training beginning with the first class on November 3, 198 Based on the actions completed and the awareness of personnel resulting from future NGET training (both initial and re-qualification) the inspector has no further concerns and this item is considered close . Followup on 10 CFR 50.5S(e) Reports (92700)

(Closed) 50.55(e) Report (455/86004-EE): Two engine failures occurred on the 2A diesel generator. These two failures were caused by cylinder liner and head misalignment. These failures have been satisfactorily evaluated by the licensee and the vendor (Cooper-Bessemer). The maintenance and corrective action procedures developed by the vendor have been shown to be effective. These procedures address the proper installation and alignment of the diesel engine cylinder liners, cylinder heads, and rocker arms. The procedures were followed by the licensee and vendor personnel in servicing the 2A diesel generator engine. The ver. dor also recommends using an engine analyzer, which can detect abnormal conditions in checking diesel engine rocker arm components. An En-Spec 1000 analyzer is available for use at the

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station. The use of the analyzer has been incorporated into the diesel generator surveillance procedures, BVS 8.1.1.2E-1 and BVS 8.1.1.2E- The corrective actions taken and the maintenance procedures incorporated by the licensee are satisfactory and this item is close . IE Bulletin (IEB) Followup (92703) (0 pen) IEB (455/83005-BB): Problems with Hayward Tyler Nuclear Code Pumps. This bulletin required the licensee to perform testing on pumps manufactured by Hayward Tyler Company. Byron Unit 2 utilizes one of these pumps as an Essential Service water booster pump 2SX04 In a letter from P. L. Barnes to J. G. Keppler, dated August 12, 1983, the licensee committed to testing 2SX04P as part of the bait 2 Auxiliary Feedwater (AF) preoperational test 2.3.60. During a review of AF 2.3.60, the inspector identified that the testing method was inadequate to meet the requirements of the IEB, discussed in Inspection Report 455/86041(DRS). AF 2.3.60 only required obtaining data at " normal" flow, not at normal, runout, and minimum flows, as required by the lEB. This IEB will remain open pending licensee completion of retesting of the pump and submission of the supplemental response to this IE In discussion with licensee management, the licensee committed to complete the retesting as part of AF retest R-2088 and submit the retest results prior to initial entry into Mode This commitment was documented as Item C.1 of Attachment 1 to the Facility Operating License for Unit 2, NPF-6 (Closed) IEB (455/86006-8B): Engineered Safety Features reset controls. The bulletin documented instances of improper response of Engineered Safety Features (ESF) equipment following reset of a Safety Injection (SI) signal. The licensee committed to verify proper ESF equipment response to SI reset conditions. The item was closed for Unit 1 in Inspection Reports 50-454/84070 and 50-454/8500 In the preparation of preoperational test 2.26.60, " Engineered Safety Features", the licensee included steps to demonstrate proper ESF equipment response following an SI signal reset. The licensee prepared a summary of the test steps which indicated for each component where compliance with the bulletin's requirements was verifie The inspector reviewed the licensee approved test results for test 2.26.60 and the summary. The review indicated a successful demonstration of proper ESF equipment response to a SI signal rese This item is close No violations or deviations were identifie ; '.

5. Followup of 10 CFR Part 21 Reports (92716)

(Closed) Part 21 Report (455/84005-PP): Failure of Ruskin Fire Dampers to close under design airflow conditions. The inspector reviewed the licensee's evaluation of this matter with respect to dampers supplied to Byron Unit 2. The licensee identified dampers installed in Unit 2 and plant common systems potentially affected by the reported failure Ten of these dampers had been tested under air flow conditions with no failures. The remaining 11 dampers were successfully tested under no airflow. The 10 dampers tested under airflow were shown by analysis to bound the service conditions of the remaining 11 in that the remaining 11 would be subject to lower air velocities and were of smaller sizes than the largest of the 10 tested under construction airflow condition Preoperational tests were performed by the licensee to verify that the dampers cycled properly and were properly aligned. Based on the inspector's review, this item is considered close . Followup of Safety Evaluation Report (SER) Items (92719) (Closed) SER Item (455/830000-20): Operator training for safe operation of the facility and restoration of AC power following a station blackout. The inspectors reviewed the lesson plans and training records for several operators and determined that all had received the required training. This item is considered closed, Licensees are required to provide a Safety Parameter Display System (SPDS). The objective is to improve the ability of nuclear power plant control room operators to prevent accidents or cope with accidents if they occur by improving the information provided to them (NUREG-0660, Item I.D.1, "NRC Action Plan Developed as a Result of the TMI-2 Accident," USNRC, Washington, D.C., May 1980; Revision 1, August 1980). The need for an SPDS was confirmed in NUREG-0737 (Requirements for Emergency Response Capability,"USNRC, Washington, D.C., November 1980), and in Supplement 1 to NUREG-0737. The SPDS requirements in Supplement 1 to NUREG-0737 replaced those in earlier documents. Supplement 1 to NUREG-0737 requires each licensee or applicant to implement an SPDS on a schedule negotiated with the NRC. Human factors guidelines for SPDS design are currently provided in NUREG-0800 " Standard Review Plan for the Review of

Safety Analysis Reports for Nuclear Power Plants" Section 18.2,

! Revision 0, " Safety Parameter Display System (SPDS)," and Section j 18.2, Appendix A, " Human Factors Review Guidelines for the Safety Parameter Display System," November 1984, and NUREG-0700 " Guidelines i

for Control Room Design Reviews," September 198 The results of NRR's audit of the Byron SPDS wcre centained in an October 30, 1985 letter from B. J. Youngblood (NRC) to Dennis L. Farrar (Commonwealth Edison). Three human engineering

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discrepancies were noted in the letter.

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On June 25, 1986, the inspector reviewed the SPDS to verify that these three discrepancies had been corrected.

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The first discrepancy noted that there was no clear way of determining whether the wide-range or narrow-range display was on the screen. The licensee attempted to correct this discrepancy by solid coloring the center of the wide-range display. The inspector, after discussions with NRR's human factors reviewer, did not find this acceptable. The licensee agreed to have its human factors consultant review the issue. Pending completion and review of the consultant's response to this issue, this matter is considered an Open Item (455/86040-01(DRP)).

The second discrepancy stated that the red alarm bars at the end of each Iconic spoke were difficult to detect. The inspector verified that these bars had been made longer and were now easy to detec The third discrepancy noted that the wide range steam generator level spoke did not cover the full range at plant operation. The licensee corrected this by changing the wide range steam generator reference level when the plant is at power. The inspector found this acceptabl No violations or deviations were identifie . Followup of NUREG-0737 Items (25401) (Closed) I.C.1 Short Term Accident and Procedure Review As discussed in the Byron SER, Supplement 4, Section 13.5, the licensee committed to establish emergency operating procedures based upon the staff approved Westinghouse Owners Group (WOG) guidelines, Revision 0. The requirements of this item were updated by NRC Generic Letter 82-33, " Supplement 1, to NUREG 0737 - Requirements for Emergency Response Capability". In the Byron Unit 1 Operating License, the licensee committed to established emergency operating procedures based on the staff approved WOG guidelines, Revision Selected emergency operating procedures were previously reviewed during inspection 455/86031 (DRP). This inspection included a review of procedure verification and validation activities as well as certain aspects of operator training. Inspector comments resulting from this review are being tracked as an open item 455/86031-01(DRP). (0 pen) I.C.7 NSS Vendor Review of Procedures Licensee action on this item will be completed after Startup Testing for Unit (Closed) I.C.8 Pilot Monitoring of Emergency Procedures for NTOLs In Section 13.5.2.3 of the Byron SER the Staff stated that based

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upon the applicant's adoption of the Westinghouse Owners Group Guidelines (WOGs Guidelines), Revision 0, for Emergency operating

[ procedures and the Staff's anticipated approval of the WOG's j Guidelines, the Staff was not going to require application of i tem to Byron. In Section 13.5 of Supplement 4 to the Byron SER the Staff reported its approval of the WOGs Guidelines, Revision 0.

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. (0 pen) I.D.1 Control Room Design Reviews Licensee action on this item will be completed in December, 1986, (0 pen) Plant Safety Parameter Display Console This requirement provides a Safety Parameter Display System (SPDS)

to improve the ability of nuclear power plant control room operators to prevent accidents or cope with accidents if they occur by improving the information provided to the operators. NRR's audit of the Byron SPDS in an October 30, 1985 letter, (B. J. Youngblood (NRC) to D. L. Farrar (Commonwealth Edison )), identified three discrepancies. On June 25, 1986 the inspector verified that two of these discrepancies had been corrected. The remaining discrepancy noted that there was no clear way of determining whether the wide-range or narrow range display was on the screen and was noted as an Open Item (455/86040-01(DRP)). (See Paragraph 6.b).

The operation of the SPDS was successfully tested in System Demonstration 2.20.76. Licensee action on this item will be completed in December, 198 (0 pen) I.G.1 Training During Low Power Testing Licensee action on this item will be completed during the first refueling outage of Unit (Closed) II.B.1 Reactor Coolant System Vents The inspector verified in the course of emergency operating procedure reviews discussed in Paragraph 7a of this report that the licensee had included procedures for operation of the reactor vessel head vents consistent with the Westinghouse Owners Group Guidelines, Revision 1. Specifically, the inspector reviewed Byron Functional Restoration (BFR) procedure 2BFR-I.3, " Response to Voids in Reactor Vessel Unit 2", Revision 1 dated August 7, 1986. Nu deficiencies were identified as a result of this review.

Byron Technical Specification 3/4.4.11, " Reactor Coolant System Vents", required at least one reactor vessel head vent path consisting of two valves in series powered from emergency buses

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to be operable in Modes 1, 2, 3 and 4. This specification also included suitable surveillance testing requirements to establish vent path operability once per 18 months. The inspector verified

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that these requirements were contained in Byron Operating Procedures 2B05 4.11.A-1, 2B0S 4.11.B-1 and 2B05 4.11.C-1.

l (Closed) II.B.3 Post Accident Sampling Verification of licensee actions relative to this item was previously performed and documented in NRC Inspection Report No. 455/83026 (DRSS). The Unit 1 cperating licensee has been

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conditioned (License Condition C. (8)) to require the licensee

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complete actions to account for sample line losses and other phenomenon to demonstrate that radiciodine and particulate samples are representative prior to Unit 1 startup after the first refueling. These actions will be tracked by Item II.F.1.2, Iodine Particulate Sampl (See Paragraph 7.n). (Closed) II.D.1 Relief and Safety Valve Test Requirements In lieu of conducting safety / relief and block valve performance tests, the licensee adapted the results of the full scale valve testing program performed by the Electrical Power Research Institute (EPRI) on behalf of the PWR Owners Group. The Staff reported i Supplement 5 to the Byron SER that based on a preliminary review .

of the licensee's submittals the licensee's approach to meeting the performance test requirements of this item was acceptabl The inspector reviewed inservice test procedures and the latest test results for: safety valves 2RY8010A, 2RY8010B, 2RY8010C; power operated relief valves (PORV's) 2RY455A and 2RY456 and; PORV block valves 2RY8000A and 2RY8000B. These test results included lift setpoint verification for the safety valves performed on September 18, 1986, and full stroke exercising of the PORVs and PORV block valves performed on October 11, 1986. Specified acceptance criteria established in conformance with the technical specifications were met for all valve k. (Closed) II.E.3.1 Emergency Power for Pressurizer Heaters In Section 8.4.6 of the Byron SER the Staff found that the Byron design conformed to the requirements of this item. The inspector reviewed approved Technical Specification 3/4.4.3, " Pressurizer".

The specification required that once per 18 months the cross-tie for the pressurizer heaters to the Engineered Safety Features (ESF)

power supply be demonstrated operable by energizing the heater The inspector reviewed Byron Operating Surveillance Procedure 2B05-4.3.3-1, Revision 51, dated September 30, 1986. The procedure crossties ESF electrical bus 241 with non-ESF electrical bus 243 from which heater groups A and D are powered. With each heater group separately energized from the ESF supply bus voltages and heater currents are measured and heater capacity (KW) calculate The same is done for heater groups B and C utilizing ESF electrical bus 242 and non-ESF electrical bus 244. Calculated heater capacity for at least two of the heater groups is required to be greater than or equal to 150 KW each. This 18 month surveillance is required to be performed in Modes 1, 2 or . (Closed) II.E.4.1 Dedicated Hydrogen Penetrations The inspector verified that the licensee had satisfactorily completed testing of the two redundant on-line hydrogen analyzers in accordance with preoperational test procedure 2.61.61, " Process

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Sampling Hydrogen Monitors". The analyzers were demonstrated capable of measuring hydrogen concentrations with an accuracy of 2.5%. Approved Technical Specifications 3/4.6.4.1, " Hydrogen Monitors", require the redundant hydrogen monitoring and control systems to be operable in Modes 1 and 2 and specify suitable surveillance test requirements for demonstrating operabilit (Closed) II.E.4.2 Containment Isolation Dependability In Section 6.2.4 of the Byron SER the NRC Staff found that the Byron design met the requirements of this item except that the licensee had not adequately demonstrated that the containment setpoint pressure that initiates containment isolation had been reduced to the minimum compatible with normal operating condition Inspector verification that the installed containment isolation system functioned in accordance with design was accomplished by witnessing of test performance and results evaluations for preoperational tests 2.26.60 " Engineered Safety Features", 2.26.61, ,

"ESF (ECCS Full Flow)", 2.26.62, "ESF-Logics and Time Response",

2.58.60 " Local Leakrate", 2.58.61, " Containment Purge".

Collectively, these tests demonstrated: (1) containment isolation actuation from diverse monitored parameters, (2) acceptable isolation times below those values specified in the approved technical specifications, (3) no valves would reopen upon reset of the containment isolation signal and, (4) isolation of the containment ventilation system on high containment atmosphere radiation levels. These inspector verifications were performed during inspections documented in NRC Inspection Reports No. 455/86022 (DRS), 455/86034(DRS), and 455/86041(DRS).

The inspector reviewed approved Technical Specification 3/4.3.2,

" Engineered Safety Features Actuation System", and verified that operability and surveillance requirements were included for each

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functional unit employed, by design, to actuate containment isolation. The inspector noted that the licensee and NRC Staff had reached agreement on the containment high pressure setpoint for Phase A isolation which was established at 5.8 psig. The inspector reviewed Technical Specification 3/4.6.3, " Containment Isolation Valves", and verified that operability and surveillance requirements for identified containment isolation valves specified maximum permissible isolation times and included periodic testing to demonstrate that required valves closed within these isolation time limits upon receipt of a simulated Phase A or Phase B isolation signal. For the containment ventilation isolation valves, periodic testing is required which verifies that the normal and mini-flow

, purge supply and exhaust valves close within 5 seconds upon receipt of a simulated high radiation signal from radiation monitors 2RE-AR011 and 2RE-AR012 (1/2 coincidence logic). The operability and surveillance test requirements for the radiation monitors were contained in Technical Specification 3/4.3.3.1, " Radiation Monitoring for Plant Operations". The inspector reviewed Technical

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Specification 3/4.6.1.7 and verified that it required that the 48

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inch containment shutdown purge and exhaust isolation valves be closed with power removed in Modes 1, 2, 3, 4 and verification of same at leait once per 31 days. The inspector verified by review of Byron 0- ating Surveillance (B0S) procedure 2B05-6.1.7.1-1, Revision t, dated September 26, 1986, that this surveillance requirement was procedurally require (0 pen) II.F.1.2 Iodine / Particulate Sampling Licensee action on this item will be completed prior to startup after Unit 2 first refuelin (0 pen) II.F.2 Instrumentation for Detection of Inadequate Core Cooling Licensee action on this item will be completed after Unit 2 achieves full power.

. (Closed) II.K.2.13 Thermal Mechanical Report and II.K.2.17 Potential for Voiding in the Reactor Coolant System During Transients

None of the requirements assigned to the NRC Region III Office for followup inspection applied to Byro (Closed) II.K.3.5 Automatic Trip of Reactor Coolant Pumps During Loss of Coo' ant Accident In Section 15.5 of the Byron SER the NRC Staff concluded that based on satisfactory progress of this issue and the licensee's commitment to comply with the Westinghouse Owners Group generic resolution of the issue, the staff found this issue acceptabl In Amendment 34 of the FSAR, Appendix E, the licensee stated that response to this item was made by the Westinghouse Owner's Group in the June, 1981 letter from R. W. Jurgensen to P. S. Chec Additional submittals concerning this item, if necessary, will be made in accordance with the schedule provided in the June 1981 lette The inspector reviewed this letter and subsequent correspondence and concluded that the licensee complies with the generic resolution of the issue. The licensee intends to amend Appendix E of the FSAR to expand the discussion of the resolution of this issue, (0 pen) II.K.3.31 Plant Specific Calculations to Show Compliance with 10 CFR 50.46 Licensee action on this item will be complete prior to initial criticality.

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.: : (Closed) III.A.2 Emergency Preparedness In Section 13.3 and Appendix D of the Byron SER, the Staff concluded that the Commonwealth Edison Company's Generating Stations Emergency Plan and the Byron site-specific annex meets the requirements of 10 CFR 50, Appendix E. The plan also describes an offsite dose calculation system which meets the design objectives of the NRC Class A Model. The Byron Station A Model has been tested and approved in the system demonstration test 2.20.7 (0 pen) III.D.1.1 Primary Coolant Sources Outside Containment Licensee action on this item will be completed after Unit 2 achieves full powe No violations or deviations were identifie . Comparison of As-Built Plant to the FSAR (37301)

> Purpose The inspectors continued a review of selected safety related mechanical and fluid systems to verify that the as-built plant

conforms to commitments contained in the Byron FSA This inspection consisted of a review of seven safety related systems:

Component Cooling (CC), Diesel Fuel Oil (D0), Main Steam (MS),

Pressurizer (RY), Reactor Floor Drains (RF), Residual Heat Removal *

(RH), and Safety Injection (SI).

The as-built drawings of these seven systems were reviewed against their respective FSAR drawings and descriptions. A field

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verification focusing on system configuration was then performed,

to ascertain that the installed system was in agreement with the

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as-built drawings, including valve and instrument identificatio The systems were also reviewed to verify that installed instrumentation, including control room indicators and controls, was adequate to allow performance of the proposed Technical Specifications surveillance Results l

Component Cooling Water The following documents and drawings were reviewed:

P& ids M-139 Sht 1, 2 M-66 Sht 3a, 3b, 4b, 4c, 4d, I

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M-2006 Sht 2, 3

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M-2139 Sht 1, 2, 3, 4

Isometrics There were 156 small bore and 32 large bore isometrics reviewe Valve Lineup BOP CC-M2 Revision 4, " Component Cooling Systems Valve Lineup".

Technical Specification Surveillance Procedures 2BOS 7.3.i.A-1 2803 ,'.3.3.A-1

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2BOS 7.3.3.B-1 2BVS 0.5-2 C BVS 0.5-3.CC.1-1 2BVS 0.5-3.CC.1-2 Documents Engineering Change Notice (ECN) 33874 Comments One discrepancy with the P&ID was identified (the batch corrosion inhibitor feed valve for the chemical addition tank). This valve change was the subject of an ECN and the P&ID is under revisio Conclusion The inspector determined that the component cooling system has been completed and is in physical conformity with the FSAR description.

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The following documents and drawings were reviewed:

P&ID M-130 Sht 1A, IB, 2 C&ID M-2130 Sht 1, 2, 3 l

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M-2556A Sht 101, 102, 103, 104, 105, 106, 108, 109, 110, 111, 112, 113, 114, 115, 116, 117, 118, 119, 120, 121, 122, 123, 124, 125, 126, 127, 128, 129, 130, 131, 132, 133, 134, 135, 136, 137, 138, 139, 140, 141, 142, 143, 144, 145, 146, 149, 150, 151, 152, 153 Isometric Spools D0-3, D0-24, D0-50 Surveillances BOP DG-12 280S 8.1.1-la LOCAR 2BVS 0.5-3.D BVS 4.10-5.1

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2BOS 7.1.2.3.a-1 2BOS 7.1.2.3.b-1 280S 8.1.1.2.a-1 280S 8.1.1.2.a-2 280S 8.1.1.2.a-3 280S 8.1.1.2.a-4 2BOS 8.1.1.2.c-1 2805 8.1.1.2.d-1 2B05 8.1.1.2.e-1 2B05 8.1.1.2.f-3 280S 8.1.1.2.f-4 ,

2BVS 8.1.1.2. BVS 8.1.1.2.h.1-2 i

Comnents

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The inspector identified the below listed concerns during the

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field walkdown and procedure revie These concerns and other minor typographical errors were discussed with the licensee's I

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staff. Correction of concern 1 will be followed as a Open Item (455/86040-03(DRP)). Correction of concern 3 will be followed as an Open Item (455/86040-05(DRP)). Correction of concerns 4, 5, and 6 will be followed as an Open Item (455/86040-04(DRP)).

Concern #7 will be followed as Open Item (455/86040-02(DRP)). Valves 200259, 2002060, and 2002096 were found to be locked open; however, the valves were not indicated as locked on either the P&ID or the valve lineup sheets. The valves were in their correct positio . Valve 200?117, AF Diesel Day Tank overflow return isolation valve, is required to be locked open by the P&ID and valve lineup sheets. The valve was discovered not locked, but was

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E open. This information was given to the Shift Engineer and the valve position was verified and the valve was locked per approved procedure . Valve 200001A, the 2A storage tank fill valve, was oriented such that it was not readily accessible to an operator. The inspector recommended that an additional access platform be built or the presently installed platform modifie . The following procedures specified the performance, or referenced, Unit 1 mechanical and electrical lineups rather that Unit 2: 2BVS 0.5.3.00.1, 2BVS 8.1.1.2.P-3, 2BVS 8.1.1.2.F-4, 2BVS 8.1.1.2.L.1-1, and 2BVS 8.1.1.2.h.1- . 2BVS 8.1.1.2.f-3, Paragraph F.1.25 does not specify INDEPENDENT VERIFICATION of valve 200129 when it is returned to the locked shut position at the end of the procedure, as required by the locked valve progra . 2B05 LOCAR 8.1.1-la, LCO 3.8.1.lc.2, specified that the Opposite Units A Diesel Generator is capable of being manually started and crosstied, rather than being OPERABLE per Technical Specifications. This change of Technical Specification wording is applicable after October 31, 198 . The Carbon Dioxide fire suppression piping in 2A Diesel Generator Day Tank Room, 20002TA, was supported by wire from

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electrical conduit, not properly supporte Conclusion The inspector determined that the diesel fuel oil system has been completed and is in physical conformity with the FSAR descriptio Main Steam System The following documents and drawings were reviewed:

P& ids M-120 Sht 1 C& ids M-2120 Sht 1 l

Isometrics MS-56A, MS-56B, MS-57A, MS-57B, MS-60A, MS-60B, MS-61A, MS-61B Electrical 6E2-4030-MS06, 6E2-4031-MS39

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Surveillance Procedures 2B05 3.2.1-910 2B0S 6.3-la BOP MS-6 2BVS 6.3.3-19 2BVS 7.1.5-2 BMP 3114-003 2 BIS 3.2.1-001 BIP 2500-038 2BVS 7. BAR 2-01-A5 BAR 2-15-A3 BAR 2-15-Al Conclusion The inspector determined that the installation of the main steam i

system is in physical conformity with the FSAR descriptio Pressurizer The following documents and drawings were reviewed:

P& ids M-135 Sht 5, 6, 7, 3 M-679 M-535 l C& ids M-2135 Sht-6, 7, 8, 9, 10, 12 Isometrics (Small Bore)

S-RC-001 Sht 237, 238, 239, 240, 241, 242, 253, 278, 319, 320,

!

322, 323, 324, 325, 326, 327, 328, 330, 331, 332, 333, 334, 335, 336, 337, 354, 355, 356, 357, 358, 359, 360, 361, 362, 363, 403, 404, 405, 609, 612, 623, 629, 630, 632, 633, 634, 635 S-CV-001 Sht 418, 420, 421 S-RY-100 Sht 4, 302, 303 i S-RC-100 Sht 3A t

l

,

l

- . - _ , . ,- - , , . .-- . - . - - - - _ , - - _. - . . . . - _ . - - . - - - - - - - - - - - . . . , - . - - - . . - , . - -

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.

Iseretric Spools

_

RY-9, 10, 11, 12, 13, 14, 15, 16, 20, 21 RC-19 RE-15 2H-CBE-Surge

/

Surveillance Procedures 2B0S 0.1-1, 2, 3 2805 4.3.2-1 2B0S 4.3.3-1 2BOS 4.4.lb-1 2B05 4.4.2-1 2BOS 6.3.3-10 s s .

B0P RY-M2 Re Conclusion The inspector determined that the pressurizer system has been completed and is in physical conformity with the FSAR descriptio Residual Heat Removal

<

The following documents and drawings were reviewed:

'i P&ID M-137 Sht 1 M-679 M-535 C&ID M-2137 Sht 1, 2, 3

! Isometrics (Small bore)

S-RH-100 Sht 23, 24, 25, 26, 27, 28, 29, 30, 31, 32, 33, 34, 48, 49, 50, 51, 52, 53, 55, 56, 57, 58, 59, 60, 61, 61, 63, 64, 67, 68, 105, 106, 201, 202, 203, 204 Isometrics (Large Bore)

RH Sht 9, 10, 11, 12, 13, 14, 15 SI Sht 21, 37, 55 Surveillances B0P RH-M2 l

l

- . . _ . _ _ . _ _ _ . _ _ _ - -__ - _ _ _ . _ . . _ . _ _ _ _ . _ _ . . _ . . _ . _ _ _ _ . _ _ . . _ . _ . _ _ _ , . _ _ . _ . - - - _ _ _ - .- _ _

s' !

Technical Specification Required Surveillance Procedures; BHS, BIS, BVS, Serie Conclusion The residual heat removal system has been completed and is in physical conformity with the FSAR descriptio Reactor Floor Drains The following documents and drawings were reviewed:

P& ids M48 Sht 6A, 6B C& ids M-2048 Sht 27 Electricals 6E-2-4030RF04 6E-2-4030RF06 56 Isometrics RF 2, 4, SA, SC, 8 Surveillances BIS 4.6.1.b-200 2B0P RF-01 280P RF-02 280P RF-E2 280P RF-M2 2B05 4.6.1.c-1 2BOS 0.1-4

,

2B05 0.1-1,2,3 Conclusion The reactor floor drains system has been completed and is in physical conformity with the FSAR descriptio Safety Injection System i

The following documents and drawings were reviewed:

,' P&ID M-136, Sht 1, 2, 3, 4, 5, 6

- _ _ - - _ . - - . ,_ - - . - - - - ~ . , , . - . , , . . . , _ , , - . _ _ . _ _ - - - . - - - - ,-- --.

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l Isometrics (Large Bore)

SI-22, 51-24, SI-26, SI-27, SI-38, SI-38, SI-39, SI-40, SI-53, SI-55, SI-56, RC-14 Isometrics (Small Bore)

S-SI-001, Sht 60, 61, 62, 63, 105, 107, 201, 202, 217, 218, 221, 222, 223, 249, 250 Procedures:

2BCS 5.1.1.b-1 2B0S 0.1-1,2,3 2B05 5.1.1.c-1 2B05 5.2.c-1 2B05 5.2.g.1-1 2BVS 5.2.d.2-1 2BVS 5.2.e.1-1 2BVS 5.2.f.2-1 Comments No significant discrepancies between the field installation, P&ID

, and isometric drawings were noted, and no significant procedural deficiencies were noted. In general, for about one third of the isometric drawings examined there were minor discrepancies which consisted of a disagreement between the isometric drawings and the field or the P&ID. These discrepancies are listed below and are collectively considered to be an open item (455/86040-06(DRP)):

ISOMETRIC DWG. N DISCREPANCY SI-22 The P&ID shows the two RWST temperature element wells in series, whereas the isometric and field installation have these wells at the same pipe locatio COMMENT:

RWST temperature measurements are periodically required by Technical Specifications. The RWST temperature element wells are located about 5 feet below the RWST in an underground pipe tunnel. The RWST's are located outside the auxiliary building in a thick concrete enclosure directly above the pipe

,

tunnel. The temperature at the wells may not be representative of RWST bulk temperature unless some means of circulation is provided to assure flow past i the wells and to produce even mixing of the RWST content , y + w --, - -

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SI-39 The P&ID shows SI pump discharge temperature indicator 2TI-SIO43 to be located on the SI pump discharge line between the discharge pressure gage and the mini-flow recirculation line. The isometric and field installation have the tap for 2TI-SIO43 between valve 2SI 8821B and the pump discharge pressure transmitte The P&ID and field installation show a capped test connection (piping line number 2SIF4A-3/4) between valve M0V 2SI-8802B and mechanical penetration M-7 The isometric did not show this test connectio SI-38 A valve numbering discrepancy exists in that the P&ID refers to valve M0 25I 8821B as M0 2SI-8821B SI-55 and A valve numbering discrepancy exists in that the SI-56 P&ID refers to valve M0 2SI-8813-2, which is referred to as M0 25I-8813 on SI-55 and is referred to as M0 2SI-8813-1 or. SI-5 ,

S-SI-001-202 The B train SI loop 3 hot leg flow element is labelled 2FE 986 on the P&ID, but is labelled 2FE 936 on the isometric drawin S-SI-001-221 The train B loop 2 cold leg piping lines are labelled 2SI8EB-2 and 2SI8FB-2 on the P&ID, but are labelled 2SI18EB-2 and 2SI18FB-1 on the isometric drawin These discrepancies are collectively considered an Open Item pending NRC review of the licensee's corrective action (455/86040-06(DRP)).

Two minor procedural deficiencies were identified as follows:

PROCEDURE NUMBER DEFICIENCY 2BVS 5.2.e.1-1 A modification to Unit 2 added valves 2CV8114 and 2CV8116 to the charging pump mini-flow lin The valve lineups for testing and return to service in this procedure needs to be updated to

reflect the modificatio BVS 5.2.f.2-1 Step F 1.3 directs operators to " remove SI PA and PB from operation" as part of the pumps procedure. This step might be more clear if it were reworded in such a way as to prevent operators from placing both pumps in pull-to-lock, which would be a violation of Technical Specification . . _ - - . _ ._ _- __

I !

These discrepancies are considered a single Open Item pending NRC review of the licensee's corrective actions (455/86040-07(DRP)).

,

Conclusion The SI system has been completed and is in physical conformity with the FSAR descriptio Summary Of the seven systems reviewed, the inspectors did not identify any significant discrepancies between the FSAR description and the as-built plant. The minor drawing and procedure discrepancies were communicated to the licensee. Based on the satisfactory inspection of the seven systems and taking credit for the same reviews of other systems at Braidwood 1 and Byron 1, the inspectors have reasonable confidence that the as-built plant conforms to FSAR design and that Technical Specification surveillances can be performed with the plant as it is buil '

No violations or deviations were identified.

4 Technical Specification Review (71301)

The final issue of the Byron, Unit 1 and 2, Technical Specifications was reviewed for clarity and enforceability against the plant specific differences unique to Unit 2. Comments were provided to the licensee for clarification or resolutio No violations or deviations were identified.

. 10. Housekeeping (71302)

. The inspectors conducted plant tours of Unit 2 from October 16 through

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October 31, 1986. Areas of the Unit 2 plant observed during the tours included the containment, fuel handling and storage areas, auxiliary

building areas including the Unit 2 portion of the control room, and the turbine building. Areas were inspected for work in progress, state of cleanliness, overall housekeeping, state of fire protection equipment and methods being employed, and the care and preservation of safety-related

.

components and equipment. The inspectors were accompanied by licensee l personnel on portions of the tours for the purpose of identifying areas t

where additional housekeeping efforts should be concentrated to improve the overall cleanliness of Unit 2. Inspector concerns were related to

the licensee.

l No violations or deviations were identified.

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1 Independent Inspection (71710)

During a review and walkdown of the Auxiliary Feedwater (AF) system, the inspector observed that both of the redundant control circuits for valve 1AF024 were located in the 18 AF pump roo Further review by the inspector identified two potential failures of valve 1AF024. The failure of 1AF024 has the potential in certain circumstances to result in damage to and failure of both AF pumps and the resultant loss of AF flo Both failure mechanisms deal with the failure of the 1AF024 valve to reposition open following an AF actuation with switch of suction water supply to the Essential Service Water (SX) System. The SX system is the only safety grade, Sesmic Category I, water supply to the AF pumps. The normal water supply, the Condensate Storage Tank (CST). is a non-safety grade, Sehmic Category II structur The purpose of valves 1AF022 A&B and 1AF024, see Enclosure 1, is to change the recirculation flow path of the AF pumps from the CST to SX. This is designed to prevent river water in the SX system from contaminating the CST should a suction switchover occur.

. The AF pumps auto-start on three signals: Safety Injection, LO-L0 Level in any Steam Generator (SG), and Low Bus Voltage on the Reactor Coolant i

Pump Electrical Busses. Valves 1AF006A and 1AF017A reposition open when a low suction pressure is sensed at the 1A AF pump suction coincident with one of the three auto-start signals. Similarly, valves 1AF006B and 1AF0178 reposition open when a low suction pressure is sensed at the 1B AF pump suction coincident with one of the three auto-start signals. This realignment provides an alternate water source from the SX system when j the CST is exhauste When 1AF006A and 1AF017A bcth reach the intermediate position, the limit switches on each valve close a set of contacts in series which energizes a solenoid operated valve for 1AF022A. Air then enters the valve operator and causes 1AF022A to reposition closed. Additionally, the energized solenoid for 1AF022A closes a contact in one of two parallel circuits for valve 1AF024, which energizes one of two parallel solenoid

operated valves for 1AF024. Air then enters the valve operator and

causes 1AF024 to reposition open.

Similarly, when 1AF006B and 1AF017B go open,1AF0228 closes, and the other parallel circuit energizes, which energizes the other solenoid operated valve and also allows air to enter the valve operator and causes 1AF024 to reposition ope Both control circuits, solenoid operated air valves and valve 1AF024 are

, all physically located in the 1B AF pump room. Both 1AF022A and 1AF022B are physically located outside the 1B AF pump room. The positions of 1AF022A, 1AF022B, and 1AF024 are not indicated remotely, so the control room operator does not have any indication of what the position of each valve is, unless an operator is dispatched to determine valve positions locall ; 1 As a potential failure situation, the inspector postulated a scenario wherein one of the three AF auto actuations is received coincident with a seismic event. The CST, being a seismic Category II structure, is destroyed. The loss of suction pressure is sensed by the suction pressure transmitters and 1AF006 A&B and 1AF017 A&B go open; 1AF022 A&B go closed; 1AF024, as the single active failure, fails to resposition open. Consequently, the recirculation capability of both AF pumps is limited to the two inch line which runs from each pump discharge to the pump suctio The shutoff head of the AF pumps is higher than the pressure in the steam generators; therefore, initially, adequate water will flow through the pumps and they will remain cool. However, when level begins to rise in the steam generators, the control room operator will begin to throttle down on the 1AF005 valves. Having no remote indication of 1AF024 and no procedural guidance, the inspector questioned whether it was possible to reduce flow to the point, or secure it completely, such that there would be no effective recirculation flow for the AF pumps. With no effective recirculation flow, the water rapidly heats up and begins to boil. With the water boiling the pumps cavitate, are degraded and failure result Consequently, a complete loss of AF flow, and resultant loss of the safety grade makeup water source for the secondary heat sink, occurs; and potential core damage result The active failure of 1AF024 was postulated to occur two ways. First, a fire could occur in the 18 AF pump room. This could damage the electrical wiring that controls 1AF024 such that the valve would not reposition. The 18 AF pump is presumed lost due to the fire. The 1AF022A valve functions normally; however, 1AF024 stays shut. Following the pump discharge throttling scenario, the 1A AF pump is damaged and fails. Consequently, both AF pumps are inoperable. There was no discussion of the 1AF024 valve in the Fire Hazard's Report regarding its potential to affect the opposite train of safe shutdown equipmen Second, no periodic testing is performed on 1AF024, or even 1AF022 A&B, to verify that the valves are capable of moving. With no periodic testing, 1AF024 freezes in its normal position; and does not reposition open when called upon to do s Following the scenario, both pumps overheat and fai The inspector discussed these concerns with the licensee's staff. The licensee's staff stated that it is not necessary to have a recirculation path for the AF pumps as adequate flow goes to the SGs; therefore, the pumps would be adequately cooled. The inspector acknowledged the licensee's position and questioned what then was the necessity of having an AF024 valv The inspector and NRC regional management met with licensee management, discussed these concerns and requested that an engineering analysis be provided to the NRC. This analysis would determine the recirculation /

cooling flow necessary to prevent damage to the pumps. The licensee committed to provide the analysis before entering into Mode 3. As an interin measure the licensee issued a standing order to the control room operators to verify that 1AF024 was open, following a suction switchover to SX, before throttling the AF005 valves below a specified value, when AF was actuate I I The inspector's questions relating to the fire hazard aspect of the potential failure of 1AF024 will be followed as Unresolved Item (455/86040-08(DRP)). The inspectors questions relating to the performance of periodic testing on 1AF024 to verify its operability will be followed as Unresolved Item (455/86040-09(DRP)).

12. Followup of Region III (RIII) Requests (92701)

The inspectors received a request from RIII, memorandum from E. G. Greenman, dated October 24, 1986, which requested that information be obtained on steam driven Auxiliary Feedwater pumps and forwarded to RIII. The inspector determined that Byron Station does not uses steam driven Auxiliary Feedwater pumps. This information was forwarded to RIII and this item is considered close . NRC Commissioner's Tour On October 8, 1986, NRC Commissioner Kenneth M. Carr, accompanied by M. V. Feterline, Technical Assistant, R. F. Warnick, Chief, Reactor Projects Branch 1, W. L. Forney, Chief, Reactor Projects Section IA, and the NRC resident inspector staff met with licensee management and toured portions of Byron Statio . Management Meetings (30702) On October 23, 1986, Messrs. R. F. Warnick, Chief, Reactor Projects Branch 1, W. L. Forney, Chief, Reactor Projects Section 1A, and the NRC resident inspector staff met with licensee management and supervisory personnel denoted in Paragraph 1 of this report. This meeting was held to assess overall facility status, in preparation for issuance of an operating license for Unit On October 31, 1986, Mr. W. L. Forney, Chief, Reactor Projects Section 1A, and the NRC resident inspector staff met with licensee management and supervisory personnel, in conjunction with the exit interview, denoted in Paragraph 1 of this report. This meeting was held to discuss the NRC's concerns related to verbatim compliance with written startup test procedures and documentation of all deficient conditions which are identified by the licensee during startup testin . Open Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action on the part of the NRC or licensee or both. Open items disclosed during the inspection are discussed in Paragraphs 6.b and ; . Unresolved Items Unresolved items are matters about which information is required in order to ascertain whether they are acceptable items, open items, deviations, or violations. Unresolved items disclosed during the inspection are discussed in Paragraph 1 . Exit Interview (30703)

The inspectors met with licensee representatives denoted in Paragraph 1 at the conclusion of the inspection on October 31, 1986. The inspectors summarized the purpose and scope of the inspection and the finding The inspectors also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspectors during the inspection. The licensee did not identify any such documents / processes as proprietar Enclosure: Diagram of Auxiliary Feedwater Simplified

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_ _ _ _ _ _ _

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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~ ~ Enclosure 1 for Inspection. Report 455/86040

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s Simplifed Diagram of Auxillary Feedwater

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AF005's b<; 'A/ AF pump

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AF006A h

AF032A AFOl7A

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AF022A s AF015A p[

SX To 4 Vent 45 SG 's 4 To From 4 X -_, CST w CST AF024 Vent iZ

AF015B From AF022B SX ir

{ AF032B AF017B h-O AF006B h

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O 4-Di 'B' AF pump AF005's t

__ _ __ --- - -