IR 05000454/1986040

Safety Insp Rept 50-454/86-40 on 861001-31.Violations Noted: Failure to Properly Store safety-related Component to Prevent Damage or Deterioration & Failure to Follow Mod Program Requirements

ML20214T001
Person / Time
Site:	Byron
Issue date:	11/25/1986
From:	Forney W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20214S969	List: IR 05000454/1986040 ML20214S966
References
50-454-86-40, IEB-83-05, IEB-83-5, NUDOCS 8612080445
Download: ML20214T001 (15)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION III

Report No. 50-454/86040(DRP)

Docket No. 50-454 License No. NPF-37 Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name: Byron Station, Unit 1 Inspection At: Byron Station, Byron, IL Inspection Conducted: October 1 - October 31, 1986 Inspectors: J. M. Hinds, J P. G. Brochman J. A. Malloy R. M. Lerch H. A. Walker W ILl & W Approved By: W. L. Forney, Chief Reactor Projects Section IA

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Inspection Summary Inspection cn October 1 - October 31, 1986 (Report No. 50-454/86040(DRP))

Areas InspecteG: Routine, unannounced safety inspection by resident and regional inspectors of licensee action on previous inspection findings; LERs; IEBs; operations summary; surveillance; maintenance program implementation; maintenance; operational safety and ESF walkdown; Region III requests; event followup; and other activitie Results: Of the nine areas inspected, no violations er deviations were identified in seven areas; two violations were identified in the following areas: (failure to properly store a safety related component to prevent damage or deterioration - Paragraph 9; failure to follow modification program requirements - Paragraph 9). These two violations were of more than minor safety significance. Additionally, two unresolved items were identified which have the potential to be a common mode failure of both trains of the Auxiliary Feedwater System (Paragraph 9).

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-DETAILS Persons Contacted Connonwealth Edison Company

• R. Querio Station Manageb

• R. Pleniewicz, Production Superintendent

• R. Ward, Services Superintendent

• W. Burkamper, Quality Assurance Supervisor, Operations

• L.-Sues, Assistant Superintendent, Operatin * G.-Schwartz, Assistant Superintendent, Maintenance

• T. Joyce, Assistant.Saperintendent,. Technical Services

- D. St. Clair, Assistant Superintendent, Work Planning W. Blythe, Operating Engineer, Unit 0 T. Tulon, Operating Engineer, Unit 1 D. Brindle, Operating Engineer, Unit * J. Schrock, Operating Engineer, Rad-Waste A. Chernick, Regulatory Assurance Supervisor

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

• E. Zittle, Regulatory Assurance Staff

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• R. Tuetken, Startup Superintendent

• G. Grabins, Unit 2 ATS

• E. Falb, Unit 2 Testing Supervisor

• W. Kouba, Assistant Technical Staff Supervisor

• J.'Pausche, Regulatory Assurance-Staff-

• W. Pirnot, Regulatory Assurance Staff

• M. Whitemore, GSEP Coordinator

• R.' Klinger, Project Quality Control Supervisor

• E. Martin, Quality Assurance Superintendent

• J. Snyder, Quality Assurance Inspector

• 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 exit interview on October 31, 198 . Action on Previous Inspection Findings (92701 & 92702)

(Closed) Unresolved Item (454/86012-02(DRS))

Elimination of requirement for a copy of the purchase order to be included with

material transferred between stations. The inspector reviewed the current issue of Attachment A to Q.P. No.10-54 entitled " Receiving

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i Inspection Criteria". Revision 6 of Q.P. Form 10-54-2, " Request For Interstation Material Transfer" issued August 12, 1986, which is a part of Attachment A, now contains the requirement for a copy of the 1 purchase order to be included with material transferred between

stations. The inspector has no further' concerns in this area.

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. . , (Closed) Violation (454/86025-02(DRP)): Four examples of failure to follow Technical. Specification Action Statements. The inspector reviewed the licensee's response to these violations and verified that the corrective actions had been implemented as stated. Based on the corrective actions taken this item is considered close (Closed) Open Item (454/85016-01(DRP)): Source Range NIs susceptibility to noise causing spurious actuations of the Boron DilutionProtectiveSystem(BDPS). During the initial fuel load and startup of Unit 1, numerous noise spikes were received on Nuclear Instrument Source Range Monitors N31 and N32. These noise spikes caused the actuation of the BDPS. The licensee has completed an extensive program to reduce the noise sources. Additionally, procedure changes have been made and additional guidance given to operators when blocking and unblocking the BDPS systems. These corrective actions are described in LERs 454/84010,454/84019,and 454/84031 and are discussed further in Paragraph 3.a. Based on these corrective actions taken, the inspector has no further concerns regarding this matter and this item is considered close . Licensee Event Report (LER) Followup (90712 & 92700) (Closed) LERs (454/84010-LL, 454/84019-LL, 454/84031-LL, 454/86007-LL,454/86026-LL): An in-office review was conducted for the following LERs to determine that the reportability requirements were fulfilled, immediate corrective action was accomplished and corrective action to prevent recurrence had been accomplished in accordance with Technical Specification LER N Title 454/84010-01 Boron dilution protection system blocked indicatio /84019-01 Boron dilution protection actuatio /84031-01 Boron dilution protection actuation 454/86007-01 Auto start of train "B" control room ventilation make-up fan due to a noise spike on the radiation monito /86026 Control room ventilation actuation due to lightning induced distribution system voltage transien (Closed)LERs(454/86012-LL): Through direct observation, discussions with licensee personnel, and review of records the fellowing LERs were reviewed to determine that the reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with Technical Specification ... \

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LER N Title 454/86012-01 Rod position indication surveillance not performed within required periodicity due to personnel erro With regard to revision 1 of this LER, on October 21, 1986, the inspector interviewed shift operating personnel to verify their understanding of the revised corrective actions for this LER. In paragraph E of this LER the licensee comitted to three separate corrective actions and stated that they had all been implemente Corrective action #3 of this LER required that a log be established to track the performance of certain non-routine Technicci Specification surveillances and that this log would be kept by shift operating personnel. The shift engineer and other operating supervisory personnel had no knowledge of this commitment nor how it was to be implemented. The revision to the LER was submitted to the NRC on October 16, 1986, and information regarding the coannitment to establish a log nor the revision to the I.ER had been disseminated to any shift operating personne The inspector discussed this concern with licensee management and the necessary information was provided to shift operating personnel by October 22 and included in the required reading program by October 24, 198 The inspectors interviewed the supervisor who prepared the LER regarding the implementation of the corrective actions. The inspectors and NRC regional management subsequently met with licensee management to express the NRC's concern regarding the necessity for complete and accurate submission of information to the NRC. Based on the interview and discussions with station senior management, the inspectors consider this an isolated event, and this issue is considered close . Followup of IE Bulletins (IEB) (92703)

During a review of Unit 2 preoperational test AF 2.3.60, discussed in Inspection Report 455/86041(DRS), the inspector identified that the test measurements made on the Essential Service Water Booster Pump 2SX04P did not meet the requirements of IEB 83005. As preoperational test AF 2.3.60 was based on Unit 1 testing of Essential Service Water Booster Pump ISX04P, the inspector again reviewed the licensee's response to IEB 83005 for Unit 1 to determine if the same problem also existed. This additional review indicated that all of the test measurements required by IEB 83005 had not been obtained for 1SX04 IEB 83005 Action 1.c required that the licensee identify any usage and conduct a pump performance test for ASME Nuclear Code Pumps manufactured by the Hayward Tyler Pump Company and that the results of these tests should be provided to the NRC. Action 1.c required that as a minimum the

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i performance test should contain the criteria of Attachment 2 to the IE Attachment 2, Paragraph B required that five measurements be obtained for

'the pumps in question: head vs. flow, vibration, temperature, motor current, and leakage; these measurements are to be taken at normal, minimum, and runout flow conditions. Attachment 2 Paragraph C required that pump rundown be observed, when the pump is stoppe The licensee stated in a letter from P..L. Barnes to J. G. Keppler, dated-August 12, 1983, that only two Hayward Tyler pumps were used at Byron Station. The pumps were identified as Essential Service Water Booster Pumps ISX04P and 2SX04P. These pumps are driven by an accessory shaft on the diesel engines which power the IB and 28 AF pumps. The purpose of the SX04P pumps is to circulate cooling water through the AF diesel and its accessories during a complete loss of all AC. power at Byron Station, when normal Essential Service Water is not running. Part V of the Attachment to the August 12, 1983, letter stated that the results of tests on the Unit 1 and Unit 2 pumps would be provided after testing was completed. The Unit 1 test results were provided in a letter from P. L. Barnes to J. G. Keppler, dated January 10, 1984, and were reviewed as acceptable in Inspection Report 454/83061(DRP).

The review of Unit 2 AF.2.3.60 indicated that the measurements had only been taken at normal flow, not minimum and runout flow condition Additionally, pump rundown was not observed. Another review of the January 10, 1984 (Unit 1) response indicated that; (1) the required measurements had only been taken at normal flow, not at minimum or runout flow conditions; and (2) no statement of pump rundown was mad The inspector discussed these concerns with the licensee management and the' licensee agreed with the inspector that the response of-January 10, 1984, did not meet the requirements of IEB 8300 The acceptance by the NRC of the licensee's response of January 10, 1984, does not obviate the licensee's responsibility to comply with the requirements of the IEB. The licensee committed to perform the additional testing on ISX04P and and to revise the response submitted for this IEB. The licensee committed to perform this testing in conjunction with the next regularly scheduled surveillance test for the 1B Auxiliary Feedwater (AF) pump and to submit the revised response within 30 days of completion of the testing. This supplemental response will be tracked as an Unresolved Item (454/86040-01(DRP)).

During discussions with the licensee's staff the licensee expressed concern with performing all of the measurements required by the IE The licensee stated it was not practicable to observe rundown on a diesel driven pump as (the diesel engine, i.e. pump) stops immediatel Additionally, running the pump at minimum flow has the potential to inadequately cool the diesel engine resulting in damage and potential failure of the diesel. The inspector acknowledged these concerns and informed the licensee that any measurements which could not be made should be documented, with the reasons why, in the supplemental respons ...

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Additionally, the inspector identified a question related to the requirements to periodically test ISX04P to the ASME Boiler and Pressure Vessel Code,Section XI, Winter 81 Addenda.Section XI, Subsection IWP-1100 defines the scope of the subsection as providing the rules and requirements for inservice testing of Class 1, 2, and 3 centrifugal pumps that are installed in light water cooled nuclear power plants and that are required to perform a specific function in shutting down a reactor or mitigating the consequences of an accident and are provided with an emergency power source. The inspector believes tht the pump ISX04P falls within the scope of Section XI Subsection IWP-1100 since providing water to cool the AF diesel during a complete loss of all AC will be mitigating the consequences of an accident and since the AF diesel is an energency power source for the pum The inspector discussed this concern with the licensee's staff and was informed that 1SX04P is not presently part of the licensee's inservice inspection plan. The NRC has not yet completed final review of the plan, but has completed an interim review of the plan. The inspector discussed this concern with the NRC staff and the staff agreed to review this concern. Followup of this concern will be followed as Unresolved Item (454/86040-02(DRP)). Summary of Operations The unit was shutdown at the beginning of the period and was taken critical at 0609 on October 2, 1986. At 0650, an urgent rod failure occurred. During trouble shooting, the rods of Group 2 of Control Bank C (CBC) dropped into the core and the reactor was then manually tripped at 1150. The reactor was taken critical at 2149 on the same day, connected to the grid at 0103 on October 3, and returned to powe At 1515 on October 26, indications of main condenser tube leakage were received. Reactor power was reduced to less than 70% and the leaking tubes were isolated, plugged, and welded. The unit was returned to power at 0151 on October 31 and operated at power levels up to 94%

for the remainder of the month. These events are discussed further in Paragraph 1 . MonthlySurveillanceObservation(61726)

The inspector observed Technical Specifications required surveillance testingontheMainSteamIsolationValves(MSIV)andverifiedthat testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that limiting conditions for operation were met, that removal and restoration of the affected components were accomplished, that test 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 No violations or deviations were identifie . .. l

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l Maintenance Program Implementation (62700)

The inspectors continued a detailed review of the maintenance program to determine whether the program was being implemented in accordance with regulatory requirements; to determine the effectiveness of the maintenance program on important plant equipment; and to determine the ability of the maintenance staff to conduct an effective maintenance program. This review is an ongoing inspection and its completion will be documented in a subsequent inspection repor No violations or deviations have been identified to dat . Monthly Maintenance Observation (62703)

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 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 the status of outstanding jobs and to assure that priority is assigned to safety related equipment maintenance which may affect system performanc The following maintenance activities were observed / reviewed:

Installation of Temporary Cable EF007 between IPA 10J and IPA 12J Following completion of maintenance on temporary cable EF007, the inspector verified that this system had been returned to service properl No violations or deviations were identifie . Operational Safety Veritication and Engineered Safety Features System Walkdown (71707 & 71710)

General The inspectors observed control room operation, reviewed applicable logs and conducted discussions with control room operators during the month of October 1986. During these discussions and observations, the inspectors ascertained that the operators were alert, cognizant of plant conditions, attentive to changes in those conditions, and took prompt action when

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appropriate. The inspectors verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected components. Tours of the auxiliary, turbine, and rad-waste buildings were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, excessive vibration, and to verify that maintenance requests had been initiated for equipment in need of maintenanc Battery Chargers During a tour of the turbine building on October 29, 1986, the inspector observed a battery charger lying on the floor in an electrical contractor eating area. The batter (tools, hardhats, etc.) y were charger was lying oncovered top of it.with Thedust and other battery items charger was not protected in any way. The attached label identified the battery charger as 1AF01EA-1, which is the intended replacement for battery charger 1AF01EA. Battery Charger 1AF01EA is one of two chargers used to maintain the batteries for the IB Auxiliary Feedwater (AF) pump diesel in a charged condition. The Safety Related Component List (SRCL) for Byron Station states that replacement battery charger 1AF01EA-1 is a safety related, Class 1-E, electrical componen On September 29, 1986, the inspector had observed two battery chargers, labeled as 1AF01EA-1 and 1AF01EB-1, lying on the floor outside the door to the 1B AF pump room. The chargers were not covered nor protected in any wa Further investigation by the inspector revealed that these battery chargers were new chargers, Model FS 32/50 made by Power Conversion Products, to replace the two existing chargers, IAF01EA and 1AF01EB, Model BC1-1212 made by Custom Power In The inspector met with station management and questioned whether the storage method and location for battery chargers 1AF01EA-1 and 1AF01EB-1 was appropriate for long term storage (the modification was scheduled to be completed during the Unit 1 outage which is scheduled to start in February 1987). On a subsequent tour of the Auxiliary Building the battery chargers were no longer present and the inspector believed they had been placed in an approved storage locatio After again finding 1AF01EA-1, on October 29, in an unapproved storage area, not protected against damage or deterioration, and in apparently worse condition than before, the inspector met with senior station management to request that action be taken to resolve this matte Additionally, the inspector requested the licensee determine the location of the other battery charger IAF01EB-1. The licensee's investigation determined that the electrical contractor had been directed on approximately September 30, 1986 to store the battery chargers in an approved location. 1AF01EB-1 was taken to an approved storage warehouse and 1AF01EA-1 ended up in a contractor eating are CFR 50, Appendix B, Criterion XIII implemented by Commonwealth Edison C,mpany's Quality Assurance Manual, requires that measures be established to control the handling and storage of equipment to prevent damage or deterioration. ANSI N18.7-1976/ANS-3.2, " Quality Assurance for the Operational Phase of Nuclear Power Plants," Section 5.2.13.4 requires that measures shall be provided to control handling, storage and

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shipping, including cleaning, packaging of preservation of materials and equipment to prevent damage, deterioration and loss. ANSI N18.7-1976/ANS-3.2 is endorsed by Regulatory Guide 1.33, Revision Regulatory Guide 1.33, Revision 2, is committed to in Appendix A of the Byron FSA ANSI N18.7-1976/ANS-3.2, Section 5.2.10 requires that housekeeping practices shall be utilized recognizing requirements for the control of important parts of the nuclear plant. Housekeeping encompasses all activities related to the control of cleanliness of facilities and material. Housekeeping practices shall assure that the quality of items is not degraded as a result of housekeeping practices or technique Particular attention should be given to housekeeping in work and storage areas where important items are handled and stored to prevent damag The failure to ensure that replacement battery charger IAF01EA-1 was not stored in a manner to prevent damage or deterioration; and the failure to ensure that the housekeeping practices in the area where the battery charger was actually stored did not degrade its quality is a deviation from ANSI N18.7-1976/ANS-3.2, Sections 5.2.13.4 and 5.2.10 and a violation of 10 CFR 50, Appendix B, Criterion XIII (454/86040-03(DRP)).

Security The inspectors verified by observation and direct interviews that the physical security plan was being implemented in accordance with the station security plan. The inspectors observed plant housekeeping /

cleanliness conditions and verified implementation of radiation protection control System Lineup During the month of October 1986, the inspectors walked down the accessible portions of train 1B of the Auxiliary Feedwater (AF) system to verify operability. During the walkdown the inspector identified several discrepancies: (1) The power supply breaker for Battery Charger IAF01EB was not listed on the electrical lineup sheet (Byron Operating Procedure) B0P AF-El, Revision 4; (2) The power supply breaker for valve IAF022A was not listed on the electrical lineup sheet B0P AF-E1; and (3) valve 1AF015B, "1B AF pump discharge line high point vent", was found to be open when the valve lineup sheet B0P AF-M1, Sht 3, Revision 5, specified it to be close Further investigation by the inspector revealed that piping and an additional valve had been welded to 1AF015B; however, it was not indicated on the inspector's copy of the AF Piping

& Installation Drawing (P&ID) M-37, Revision AE. The same situation regarding valve 1AF015A aho existed for train 1A. Adding to the confusion, the new additional valves had construction labels on them indicating that the valves were 1AF015A and 1AF015B. The open valves, which were shown on M-37 Revision AE, also had station approved labels attached indicating they were 1AF015A and 1AF015B, making two sets of valves with the same number ..

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The inspector went to the control room to review the Unit I abnormal valve position log. IAF015A and 1AF015B were not listed on the log; therefore, the control room operators did not have any knowledge that the valves were not positioned in accordance with the normal valve lineup, B0P AF-M1. The inspector reviewed the control room drawing of M-37, which did not have any indication of the new piping and valves downstream of IAF015A and 1AF015B. The inspector then informed the shift engineer of these facts. The shift engineer directed that the positions of the breakers for 1AF022A and 1AF01EB be verified and annotated in the Unit 1 Shift Log and that the positions of 1AF015A and 1AF015B be listed in the abnormal valve position log for Unit The licensee's investigation determined that the new piping and valves had been installed as Modification M6-1-84-065 and installation work had been completed on March 31, 1986. However, drawing M-37 had not been updated nor had the necessary post-modification testing been performed when the inspector identified this problem to the licensee. Modification M6-1-84-065 specified that a visual examination test be performed on the newly installed piping and valve CFR 50, Appendix B, Criterion VI, as implemented by Comonwealth Edison Company's Quality Assurance Manual, Quality Requirement 6.0, required that methods shall be established to control the issuance of documents, such as drawings, including changes thereto, which prescribe activities affecting quality. Quality Procedure (QP) 6-52, " Document Control for Operations," implements this requirement and Section B. requires that drawings which are utilized by Control Room staff to effectuate their decision making, be updated to reflect the as-built condition of the plant and its equipment. Byron Administrative Procedure BAP 1340-3 implements these requirements. Paragraph C.I.h, defines Critical Drawings as "A group of drawings identified in BAP 1340-A5 which are maintained as-built for Operating Shift decision making." BAP 1340-A5 lists P&ID M-37 as a Critical Control Room Drawing. The failure to update a critical control room drawing, P&ID M-37, to reflect the as-built condition of the plant following the installation of Modification M6-1-84-065 is a violation of 10 CFR 50, Appendix B, CriterionVI(454/86040-04a(DRP)).

10 CFR 50, Appendix B, Criterion XI, implemented by Commonwealth Edison Company's Quality Assurance Manual, requires preopertional tests of systems and components. ANSI N18.7-1976/ANS-3.2, " Quality Assurance for the Operational Phase of Nuclear Power Plants," Section 5.2.7 requires that modifications which may affect the functioning of safety-related structures, systems, or components shall be performed in a manner to ensure quality at least equivalent to that specified in the original design bases and requirements. Section 5.2.19.3 requires that tests shall be performed following plant modifications to confirm that the modifications reasonably produce expected results and that the change does not reduce safety of operations. ANSI N18.7-1976/ANS-3.2 is endorsed by Regulatory Guide 1.33, Revision 2. Regulatory Guide 1.33, Revision 2, is committed to in Appendix A of the Byron FSAR. The failure to perform the required post-modification visual inspection test

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on the piping and valves installed by Modification M6-1-84-065 is a deviation from ANSI N18.7-1976/ANS-3.2, Section 5.2.19.3 and a violation of 10 CFR 50, Appendix B, Criterion XI (454/86040-04b(DRP)).

Following the identification of these deficiencies by the ins ectors, the licensee accomplished the following corrective actions: 1)

updated P&ID M-37 to show the as-built system configuration; 2)

completed the re and valves; (3) quired placedpost-modification approved stationtesting on the labels identification new piping on the new valves; and (4) relabeled valves IAF015A and 1AF015B with new labels as specified by Modification M6-1-84-6 Valve 1AF024 Additionally, during the walkdown the inspector observed that both of the control circuits for 1AF024 were located in the IB AF pump roo Further review by the inspector identified two potential common mode failures of valve IAF024. The failure of 1AF024 has the potential in certain circumstances to result in damaging and failure of both AF pumps and resultant complete loss of AF to the steam generator Both concerns deal with the failure of the IAF024 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, Sesmic 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 occu The AF pumps auto-start on three signals: Safety Injection, LO-LO Level in any Steam Generator (SG), and Low Bus Voltage on the Reactor Coolant Pump Electrical Busses. 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,1AF006B and 1AF017B reposition open when a low suction pressure is sensed at the IB AF pump suction concident with one of the three auto-start signals. This provides a water supply from the SX system when water from the CST is not available. See Enclosure When 1AF006A and 1AF017A both 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 ope .. .

Similarly, when 1AF006B and 1AF017B go open, 1AF022B 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 IAF024 are all physically located in the IB AF pump room. Both 1AF022A and 1AF022B are physically located outside the IB AF pump room. The positions of 1AF022A, IAF0228, and 1AF024 are not indicated remotely, so the control room operator does not have any indication of what the position of the valves is, unless an operator is dispatched to determine valve positions locall As a potential failure situation, the inspector po:*"b+ H the following scenario, wherein one of the three AF auto actuations is reuived 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 reposition open. Consequently, the recirculation capability of both AF pumps is limited to a two inch line which runs from the pump discharge to 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 SGs, 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, and resultant loss of the safety grade makeup water source for the secondary heat sink, occurs; anc'

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 control's 1AF024 such that the valve will not repositio The IB AF pump is presumed lost due to the fire. The 1AF022A valve functions normally and 1AF024 stays shut. Following the pump discharge throttling scenario, the 1A AF pump is damaged and fail 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 1AF022 A&B, to verify that the valves are capable of moving. With no periodic testing, IAF024 freezes in its normal position; and does not reposition open when called upon to do s Following the scenario, both pumps overheat and fai u.___--__--_-._--_____-----------_-------------------__-----------_--------_-----__

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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 valve, as with 1AF022A or B shut, and no need for recirculation flow, the CST would be protected from contaminatio 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 minimum recirculation / cooling flow necessary during operation to prevent damage to and failure of the AF pumps. The licensee committed to provide the analysis within 15 days of receipt of this report. As an interim 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 during an AF actuation. The inspectors questions relating to the fire hozard aspect of the potential failure of 1AF024 will be followed as Unresolved Item (454/86040-05(DRP)). The inspectors questions relating to the performance of periodic testing on 1AF024 to verif will be followed as Unresolved Item (454/86040-06(y its operability DRP)).

Radwaste The inspectors also witnessed portions of the radioactive waste system controls associated with rad-waste shipments and barreling. Facility operations observed were verified to be in accordance with the requirements established under Technical Specifications,10 CFR, and administrative procedures, 1 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 inspectors determined that Byron Station does not use steam driven Auxiliary Feedwater pumps. This information was forwarded to RIII and this item is considered close . Onsite Followup of Events at Operating Reactors (93702) General The inspector performed onsite followup activities for events which occurred during October 1986. This followup included reviews of operating logs, procedures, Deviation Reports, Licensee Event Reports (where available) and interviews with licensee personne For the event, the inspector developed a chronology, reviewed the functioning of safety systems required by plant conditions, reviewed licensee actions to verify consistency with procedures, license conditions and the nature of the event. Additionally the inspector

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verified that licensee investigation had identified root causes of equipment malfunctions and/or personnel error and had taken appropriate corrective actions prior to plant restart. Details of the events and licensee corrective actions developed through inspector followup are provided in Paragraphs b and c belo b. Manual Reactor Trip on October 2, 1986 While in Mode 2, with reactor power at 2%, the reactor was manually tripped when four control rods of Control Bank C (CBC), Group 2, dropped into the core. At the time of the trip, instrument mechanics were troubleshooting a Control Rod Urgent Failur At 0147 CDT on October 2, 1986, a reactor startup was commence The reactor was being restarted following a trip on September 30 (See Inspection Report 454/86033(DRP)). Before the reactor was critical, at 0155 a Control Rod urgent Failure was received. This caused current orders to be sent to both the stationary and moveable grippers of the control rods; thereby, holding them in plac Instrument mechanics (IM) began troubleshooting and identified the failure as a logic error in power cabinet 2AC. The IMs replaced six circuit cards and by 0550 the Urgent failure had been cleared and at 0600 control rod withdrawal was recommenced. At 0609 the reactor was critical. At 0650 another Control Rod Urgent Failure Was received and Was again indicated as a logic error in p0wer cabinet 2AC. The IMs resumed their troubleshooting and following discussions with the manufacture, Westinghouse, the licensee determined that four additional circuit cards would be replaced.

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When a firing card was removed, the four control rods of Group 2 of CBC dropped into the core. The shift supervisor then directed that the reactor be manually trippe The IMs replaced the remaining cards, and tested all 10 cards on a test bench. Four cards were determined to be defective: firing circuit, alarm circuit, multiplex error detector, and slave cycle decoder. The licensee determined that the defective firing card, and an intermittent failure on the alarm card caused the rods to drop. The firing card caused part of the logic error and the alarm card caused the rod drop by not sending current orders to the movable and stationary grippers during the Urgent Failure, so when the firing card was removed, all current orders for the four control rods went to zero, dropping the rods. The licensee contacted the

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manufacture and verified that the troubleshooting methodology which l had been used were correct. Westinghouse suggested three additional l steps to remove the Urgent Failure and the logic error that caused the rods to drop. These steps consisted of checking for bent pins, l

loose screws and other poor connections and replacing the alarm l card, which cannot be replaced when any control rods are withdrawn.

i Following satisfactory completion of the repairs, the reactor was

! taken critical at 2149 and was connected to the grid at 0103 on i October 3, 1986.

l l 14

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c.. Load Reduction due to Condenser Tube Leak on October 26, 1986  !

L At 1515 on October 26, 1986, Steam Generator Cation Conductivity was measured at greater than 0.8 micromhos/cm. In accordance with Byron Abnormal Operating Procedure 1 BOA SEC-2, Action Level I was then-entered. Action Level 1 allows operation for up to one week and then Action Level 2 must be followed, which is to reduce turbine power to less than 25%. The licensee identified the leak as being

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in the A water box of the main condenser. At 0938 on October 27, l the licensee reduced power to 70% and isolated the A waterbox.

l The water box was drained and the leaking tubes were identified,

, plugged, and welded. Additional cleaning of the A waterbox was j also perfonned. The unit was returned to rated power at 0151 on 1 October 31, 1986.

l l No violations or deviations were identified, i.

l 12. Unresolved Items i Unresolved items are matters about which'irore information is required l in order to ascertain whether they are acceptable items, items of

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noncompliance, or deviations. Unresolved items disclosed during the inspection are discussed in Paragraphs 4 and 9.

l 13. Exit Interview (30703) i The inspectors met with licensee representatives denoted in Paragraph 1 at the conclusion of the inspection on 0ctober 31, 1986. The inspectors

sunnarized the purpose and scope of the inspection and the findings.

l The inspectors also discussed the likely informational content of the

! inspection report with regard to documents or processes reviewed by the '

l Inspectors during the inspection. The licensee did not identify any such l

documents / processes as proprietary.

l Enclosure 1: Simplified Diagram of Auxiliary Feedwater l

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