IR 05000454/1986025

Insp Rept 50-454/86-25 on 860701-31.Violation Noted:Failure to Provide Acceptance Criteria or Conduct Adequate Evaluation of Test Results for post-mod Testing & Failure to Comply W/Tech Spec Action Statements

ML20206N740
Person / Time
Site:	Byron
Issue date:	08/15/1986
From:	Forney W, Lerch R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20206N715	List: IR 05000454/1986025 ML20206N711 ML20206N730
References
50-454-86-25, IEB-86-002, IEB-86-2, NUDOCS 8608260372
Download: ML20206N740 (14)

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U. S. NUCLEAR REGULATORY COPNISSION

REGION III

Report N /86025(DRP)

Docket N License N 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: July 1 - July 31, 1986 Inspectors: J. M. Hinds, J P. G. Brochman R J}o'lrla Da~te'

J. A. Malloy Approved By: ne //////4>

Reactor Projects Section IA Date'

Inspection Summary Inspection on July 1 - July 31, 1986 (Report No. 50-454/86025(ORP))

Areas Inspected: Routine, unannounced safety inspection by the resident inspectors of licensee action on previous inspection findings; LERs; IEBs; operations summary; headquarters requests; surveillance; maintenance; operational safety; event followup; licensee actions concerning suspected drug use; and other activitie Results: Of the nine areas inspected, no violations or deviations were ioentified in six areas; two violations were identified in the following areas: (failure to provide acceptance criteria or conduct an adequate evaluation of test results for post-modification testing - Paragraph 2.f; failure to comply with Technical Specification Action Statements, 4 examples -

Paragraphs 3.b, 3.c and 10.b) Violation number one was of minor safety significance with minimal potential to affect the public health and safet Violation number two is of more than minor safety significance and had the potential to affect the public health and safet PDR

ADOCK 05000454 PDR

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DETAILS 1. Persons Contacted Commonwealth Edison Company

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• R. Querio, Station Manager

• R. Pleniewicz, Production Superintendent

• R. Ward, Services Superintendent-L. Sues, Assistant Superintendent, Operating G. Schwartz, Assistant Superintendent, Maintenance

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

• W. Burkamper, Quality Assurance Supervisor, Operations W. Blythe, Operating Engineer, Unit 0 T. Tulon, Operating Engineer, Unit 1 D. Brindle, Operating Engineer, Unit 2

• J. Schrock, Operating Engineer, Radwaste A. Chernick, Regulatory Assurance Supervisor

• F. Hornbeak, Technical Staff Supervisor R. Flahive, Radiation / Chemistry Supervisor

• P. O'Neil, Quality Control Supervisor T. Higgins, Training Supervisor

• C. Anderson, General Foreman, Mechanical Maintenance

• B. Vivan, General Foreman, Mechanical Maintenance A. Javorick, Assistant Technical Staff Supervisor

• K. Weaver, Station Health Physicist

• J. Miller, Foreman, Mechanical Maintenance

• J. Langan, Regulatory Assurance Staff

• E. Zittle, Regulatory Assurance Staff

• D. Robinson, Nuclear Safety Group, Onsite R. Williams, Technical Staff L. Wehner, Technical 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 July 31, 198 . Action on Previous Inspection Findings (92701 & 92702) (Closed) Violation (454/84071-01(DRS)): Calculations which provided original justification for PHI factor were not retrievabl Based on discussions with the inspector responsible for this item, the licensee's corrective actions were acceptable and the issue was considered resolved for Byron at the time of issuance. Also, see Section I. Introduction to Inspection Report 454/84071. Therefore, no further inspection is required and this item is close (Closed) Violation (454/84071-02(DRS)): Failure to meet code requirements for slenderness ratio (KL/R) for HVAC duct support r

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Based on discussions with the inspector responsible for this item, the licensee's corrective actions were acceptable and the issue i was considered resolved for Byron at the time of issuanc Also, see Section I., Introduction to Inspection Report 454/8407 Therefore, no further inspection is required and this item is close c. (Closed) Violations (454/85042-04(DRP); 454/85043-01(DRP)): These two items were composed of six examples which were categorized in the aggregate as a Severity Level III Problem and a $50,000 civil penalty was propose In a letter from B. Thomas to J. M. Taylor, dated July 3, 1986, the licensee acknowledged the six violations which were categorized as a Severity Level III Problem in a letter from J..M. Taylor to J. J. O'Connor, dated May 6, 1986 (EA 86-48),

and remitted payment for the $50,000 civil penalt The inspector reviewed the licensee's immediate corrective action and the corrective actions taken to avoid further violations for the six examples and verified through review of records, observations and discussions with licensed reactor operators and operations department supervisors and management that these actions had been accomplished as stated in the licensee's response. Based on the corrective actions taken this item is considered closed, and the inspector has no further concerns regarding this matte d. (Closed) Open Item (454/85055-02(DRP)): Licensee to explain the differing tolerance values in the expected values section for 1500 gpm point in Unit 1 and Unit 2 RHR preoperational tests. The licensee explained that the Unit 2 tolerance (1500+500-0 gpm) was obtained from the NSSS Vendor (Westinghouse) who provided acceptance criteria for Residual Heat Removal (RHR) test procedures on December 23, 1981. Westinghouse later provided a revision to these acceptance criteria on July 12, 1984 and provided the setpoint of 1500 gpm with no tolerance listed. Project Engineering assigned a standard 10%

tolerance. Based on these explanations, the inspector has no further concerns regarding this matter and this item is considered close e. (Closed) Violation (454/85056-01(DRP)): Failure to include Feedwater Tempering Line Flowrate in Calorimetric Surveillanc The inspector reviewed Revision 52 to the calorimetric surveillance, IBOS 3.1.1-2, and verified that the typographic errors of Revision 51 had been correcte The inspector observed licensed operators utilizing the computer calorimetric program and through discussions verified the operators understanding of the surveillance and its limitations. The inspector reviewed the licensee's evaluations for Deviation Reports DVR 6-1-86-040 and DVR 6-1-86-078 and verified that licensed power limits were not exceeded. Based on these actions this item is considered closed, and the inspector has no further concerns regarding this matte f. (Closed) Unresolved Item (454/86018-02(DRP)): Four concerns identified with post-modification testing of safety related system The licensee's staff met with the inspector and provided additional

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s information on these concern Concern number (2): lack of valve stroke timing for ISD054 valves, and (3): acceptability of stroke times for valves 0AS013 and 0AS167; were resolved by the licensee's additional information and are considered close Regarding concerns number (1) and (4) the licensee agreed that no acceptance criteria were specified further for the ISD054 valves and 1TS-SD054 temperature switches in test procedures SPP 85-76,85-111 and 86-18 nor had a test results evaluations been completed for the first four sections of SPP 86-1 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. Section 5. requires that design activities associated with modifications to safety-related structures, systems, or components shall be accomplished in accordance with ANSI N45.2.11-1974. 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 N45.2.11-1974, " Quality Assurance Requirements for the Design of Nuclear Power Plants," Section 6.3.3 requires that qualification testing shall be performed in accordance with written test procedures which incorporate or reference the requirements or acceptance limits contained in applicable design documents and that the test results shall be documented and evaluated to assure that the test requirements have been me The failure to specify an acceptance criteria and subsequently testing the 150054 valves and the 1TS-SD054 temperature switches in post-modification test procedures SPP 85-76,85-111 and 86-18 and the failure to complete a test results evaluations for the first four sections of SPP 86-18 is a violation of ANSI N18.7-1976/ANS-3.2 and ANSI N45.2.11-1974 (454/86025-01(DRP)).

As corrective action the licensee reviewed the affected test procedures to verify that the testing did actually verify the design basi Subsequent to the performance of these tests the licensee initiated a corporate review of modification testing and verification at all of the Commonwealth Edison facilities. Based on these corrective actions this item is considered closed and consequently no reply to this violation is require The inspector will review the Byron Station's implementation of the results of the corporate program as part of an annual inspection of design changes and modifications.

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3. Licensee Event Report (LER) Followup (90712 & 92700)

, (Closed) LERs (454/86020-LL; 454/86022-LL): Through direct

. observation, discussions with licensee personnel, and review of i records, the following LERs were evaluated to determine that the

reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence j had been accomplished in accordance with Technical Specification '

LER N Title 454/86020 Unusual Event Declared due to Excessive Reactor Coolant System (RCS) Leakage 454/86022 Invocation of 10 CFR 50.54(x) due to l Conflict Between Technical Specifications for RCS Leakage and Source Range Inoperability LERs 454/86020 and 454/86022 are discussed further in Paragraph 10.b.

(Closed) LER (454/86019-LL): This LER described an event on June 30, 1986, while in Mode 1, when a Technical Specification Action Statement

for a Power Range (PR) Nuclear Instrument (NI) was exceeded.

On June 30, 1986 at 0905, a surveillance was begun on PR channel i N41. A channel of instrumentation is considered to be inoperable j -

when it is in test (having a surveillance performed on it). At 1049

the channel was declared inoperable. Technical Specification 3.3.1, i Table 3.3-1, Action Statement No. 2 requires that with a PR NI channel 4 inoperable that the respective bistables be placed in the tripped j condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and the Quadrant Power Tilt Ratio (QPTR)

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~ monitored once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, while in Modes 1 and There are

-six protective trips associated with each channel of a PR NI:

l Neutron Flux, high and low and Neutron Flux Rate High, positive l and negative; additionally, each PR channel provides input to the 4 Overtemperature Delta T (OTDT)'and Overpower Delta T (OPDT)

protective trip The bistables associated with Neutron Flux protective trips are activated by removing the control power fuses for the PR channel and the OTDT and OPDT by tripping two bistable

! test switches. The test switches for the OTDT and OPDT bistables l were tripped at 1049 but the control power fuses for PR channel N41

! were not removed as it was necessary to have power to the instrument drawer to perform the surveillance. Operating department personnel i believed that the surveillance procedure would remove the fuses, not I

recognizing the conflict. At 2045 the QPTR was being monitored when

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it was realized that the control power fuses had not been removed i from N41 and consequently the bistables were not tripped. Operations i department personnel decided it would be more appropriate to complete i

the surveillance, since only a short time remained to complete the surveillance and the surveillance had to be completed to declare the

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channel operable. At 2245 the surveillance was completed and channel N41 was declared operable; thereby exceeding the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> time limit by 5.9 hour1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> The failure to trip all the bistables associated with a PR channel in test, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is a violation of Technical Specification 3.3.1, Table 3.3-1, Action Statement No. 2 (454/86025-02a(DRP)).

. (Closed) LER (454/86021-LL): This LER described an event on June 14, 1986, while in Mode 1, when a Technical Specification Action Statement for Reactor Coolant System (RCS) leakage detection l monitors was exceeded. At 1525 on June 13, 1986 the licensee was notified by its Architect / Engineer that the solenoid for Process Radiation (PR) containment isolation valve IPR 066 was not environmentally qualified. Followup of questions relating to the environmental qualification of IPR 066 will be accomplished by the Division of Reactor Safety, RIII, and is tracked by Unresolved

! Item (454/86025-03(DRS)). Technical Specification (TS) 3.6.3,

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Action No. a.2 requires that an inoperable containment isolation valve be secured in the isolation position within the required time.

With valve 1PR066 shut, two of the RCS Leakage Detection systems:

Containment Atmosphere Particulate Radioactivity Monitor and

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Containment Gaseous Radioactivity Monitor, were inoperable. TS t 3.4.6.1, Action No. c.3 requires that a RCS water inventory balance be performed once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> while in Modes 1, 2, 3, and The first surveillance was performed at 2008 on June 13. The next performance was at 0535 on June 14, 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> over the eight hour limi The cause of this event was a cognitive error by the licensed Shift Control Room Engineer (SCRE). The SCRE incorrectly interpreted the action requirement to be once per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shift, L with a maximum interval of 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> between surveillances, instead l of once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The licensee identified the error when the l oncoming shift was reviewing the list of TS action requirements in

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, The surveillance interval for this same surveillance was exceeded

, before on August 20, 1985. This violation was documented in

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Inspection Report 454/85039-0 As corrective action for that i violation the licensee reiterated to its senior reactor operators j the necessity to perform surveillances in a timely manner.

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! The failure to perform a RCS water inventory balance every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />

! is a violation of TS 3.4.6.1, Action No. c.3 (454/86025-02b(DRP)).

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4. IE Bulletin (IEB) Followup (92703)

! (Closed) IEB (454/86002-BB): For this IEB the inspector verified that l the written response was within the time period stated in the bulletin, j

that the written response included the information required to be reported, that licensee management forwarded copies of the written l

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response to the appropriate onsite management representatives, and that information discussed in the licensee's written response was accurat This IEB requested the licensee to identify any uses of Static "0" Ring (SOR) differential pressure switches, SOR Series 102 or 103, as electrical equipment important to safety, as defined by 10 CFR 50.49(b).

The licensee's response stated that no SOR Series 102 or 103 differential pressure switches are installed at Byron as important to safety electrical equipmen Based on this response, this IEB is considered close No violations or deviations were identifie . Summary of Operations The unit operated at power levels up to 74% until 1950 on July 2,1986 when an Unusual Event was declared due to Reactor Coolant System unidentified leakage exceeding the Technical Specification limit of 1 gp The unit was taken to Cold Shutdown (Mode 5). The leak was repaired and other naintenance activities were completed, including replacement of the 1C RCP seal package and replacement of the 1RY8010C pressurizer code safel During heatup to normal operating conditions 1RY8010C actuated at 1750 psis and began leaking excessively. The unit was taken to Mode 5 and valves 1RY8010A and 1RY8010C were both replace The unit was taken critical at 2357 on July 23, 1986 and connected to the grid at 0400 on July 24. The unit operated at power levels up to 99.5%

for the remainder of the month. These events are discussed further in Paragraph 1 . Followup of Headquarters Request (25580)

On July 3, 1986 the inspector received Temporary Instruction (TI)

2515/80, which requested the collection of data for a trial program to establish a Performance Monitoring Indicator Program. This trial program was established by the NRC to develop and evaluate indicators for monitoring the performance of licensees. Byron 1 was one of eight plants selected in Region III to be involved in this trial progra The information requested by TI 2515/80 was forwarded to Region III on July 30, 1986 and this TI is considered close No violations or deviations were identifie . Monthly Surveillance Observation (61726)

The inspector observed Technical Specifications required surveillance testing on the Solid State Protection System-Train B, Containment Air Lock Local Leak Rate Test, Reactor Vessel Head Vent Valve Stroking and verified that 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

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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 . 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 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:

Reactor Coolant Pump 1RC01PC Seal replacement Pressurizer Code Safety Valve 1RY8010C replacement Volume Control Tank Level Transmitter 1LI-212 repair Following completion of maintenance on the IRC01PC seal, 1RY8010C valve, and ILI-212 level transmitter, the inspectors verified that these systems had been returned to service properly. Maintenance activities associated with the replacement of 1RY8010C are discussed further in Paragraph 1 and in Inspection Report 454/86029(DRP).

9. Operational Safety Verification (71707)

The inspectors observed control room operation, reviewed applicable logs and conducted discussions with control room operators during the month of July 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 containment, auxiliary, l turbine and radwaste buildings were conducted to observe plant equipment I conditions, including potential fire hazards, fluid leaks and excessive

vibration and to verify that maintenance requests had been initiated for i equipment in need of maintenance.

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The inspectors verified by observation and direct interviews that the physical security plan was being implemented in accordance' with the station security pla The inspectors observed plant housekeeping / cleanliness conditions and verified implementation of radiation protection controls. Several poor radiological work practices were identified during these observation These poor work practices were identified to licensee managemerit during routine weekly meetings with the inspector The inspectors reviewed Radiation Occurrence Reports 06-86-10 and 06-86-11 which detail entries into restricted areas without the proper personnel monitoring equipment being wor CFR 20.202 requires that I

each individual who enters a restricted area where he is likely to receive a dose in excess of 25% of the limits of 10 CFR 20.101(a), shall use appropriate personnel monitoring equipment. On July 16, 1986, a contractor individual crossed a posted radiation barrier and was not wearing the required film badge. The area was posted to require a film badge and self-reading dosimeter (SRD) for entry. The individual was observed to be wearing only a SRD and was dire::ted to leave the are On July 17, 1986 a contractor individual was escorting a visitor in a controlled area and the visitor was observed to be not wearing any_

dosimetry. The area was posted to require a film badge and SRD for all individuals for entry. The visitor was observed by licensee staff to not be wearing any dosimetry and was directed to leav the controlled are Calculations of the exposure received by both individuals was estimated to be less than 1 mrem. Based on these calculations and discussions with Region III, the inspector believes that it was unlikely for these individuals to receive a dose in excess of 25% of the limits of 10 CFR 20.101(a) and therefore no further action is contemplated regarding this matte As preventative measures the licensee retrained the two individual These reports were also discussed with all the employees of the affected contractor stressing the need to follow posted hazard warnings and reviewing the escort responsibilities for visitors in radiation area During the month of July 1986, the inspectors walked down the accessible l portions of the Auxiliary Feedwater system to verify operabilit The

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inspectors also witnessed portions of the radioactive waste system controls associated with radwaste shipments and barreling.

l Facility operations observed were verified to be in accordance with the

requirements established under Technical Specifications, 10 CFR, and i administrative procedure No violations or deviations were identified.

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e 10. Onsite Followup of Events at Operating Reactors (93702) General The insp6ctor performed onsite followup activities for an event which occurred during July 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 ifcensee actions to verify consistency with procedures, license conditions and the nature of the event. Additionally the inspector verified that licensee investigation had identified root causes of equipment malfunctions and/or personnel error and had taken appropriate corrective actions prior to plant restar Details of the event and licensee corrective actions developed through inspector followup are provided in Paragraph b belo Unusual Event on July 2, 1986 An Unusual Event was declared on July 2, 1986 due to Reactor Coolant System (RCS) unidentified leakage rate being greater than 1 gallon per minute (gpm).

At 1551 on July 2, 1986 RCS unidentified leakage rate was calculated to be 1.07 gpm. TS 3.4.6.2.b requires that RCS unidentified leakage be limited to 1 gpm while in Modes 1, 2, 3,and 4 or else with RCS unidentified leakage greater than 1 gpm, reduce the leakage to less than 1 gpm in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in ut least Hot Standby (Mode 3) within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown (Mode 5) within the following 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> At 1950 the leakage had not been identified and reduced to less than 1 gpm; consequently, the licensee declared an Unusual Event and began to reduce load to place the unit in Mode At 0140 on July 4 Mode 3 was entered and both Source Range (SR) Nuclear Instruments, N31 and N32, were declared inoperable. TS 4.3.1.1, Table 4.3-1, Functional Unit No. 6 requires that an Analog Channel Operational Test be performed on N31 and N32 once per 92 day It is not possible to perform this test while at power; consequently, since the unit had been at power for 121 days the surveillance interval was exceede TS 4.0.3 requires that surveillances be performed within their time interval or else the component is inoperable; consequently, both N31 and N32 were inoperable. This fact was recognized by the licensee prior to the shutdown and preparations had been made to perform the surveillances as soon as possible. With both N31 and N32 inoperable, TS 3.3.1, Table 3.3-1, Functional Unit No. 6.b requires that two SR channels be operable in Modes 3, 4, and 5 or else follow Action Statement No. 5. Action Statement No. 5 requires with both SR channels ineperable, that the shutdown margin be verified to be greater than or equal to 1.3% Delta K/K and within one hour, open

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f-the reactor trip breakers, verify valves ICV 111B, ICV 8428, ICV 8439, 1CV8435, and 1CV8441 are secured in the closed position, and suspend all operations involving positive reactivity changes. These actions shall be verified at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. At 0237 on July 3 the surveillances on both SRs were completed and believing that all requirements had been met the licensee declared both N31 and N32 operabl At 0910 a cooldown was commenced to take the unit to Mode As an inherent design feature, the Byron reactor core has a negative moderator temperature coefficient of reactivity. Consequently, when the unit is cooled down positive reactivity is added to the reactor '

Cor At 1054 licensee's staff identified that the Boron Dilution Protection System (BDPS) function of both SRs was inoperable and '

consequently both SR channels N31 and N32 were inoperable. TS 4.3.1.1, Table 4.3-1, Note 9 requires that the Analog Channel Operational Test for SRs N31 and N32 include a verification that the i Boron Dilution Alarm Setpoint be less than or equal to an increase of 2.0 times the SR count rate within a 10 minute period. The purpose of the BDPS is to shift the charging pump suction source of i water from the Volume Control Tank to the Refueling Water Storage

, Tank when the Boron Dilution Alarm is received. This is intended i

to provide a water source with sufficient boron concentration (2000 ppm) to ensure that the reactor remains subcritical following

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a dilution accident.

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In LER 454/86015, dated June 2, 1986, the licensee identified that the BDPS setpoint was not being verified properly in Byron Operating

, Surveillance 180S 3.1.1-15, " Analog Channel Operational Test of l Source Range Channels N31 and N32," to less than or equal to a 2.0

times increase in the SR count rate, but to a 2.16 times increase in

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the SR count rate. As corrective action to this LER the licensee

{ stated that a Special Procedure (SPP 6-71) would be written to i

verify the correct BDPS setpoint and that 180S 3.1.1-15 would be revised to reflect the correct BDPS setpoin With the BDPS inoperable, (not verified as being set correctly),

l both SR channels N31 and N32 had been inoperable since the initial I

entry into Mode 3 at 0140 on July 3. The cooldown was secured at 1057, during which the RCS was cooled down from 557 to 535'F; Valves ICV 111B, 1CV8428, ICV 8439, 1CV8435, and ICV 8441 were secured in the closed position by 114 The failure to verify that valves ICV 111B, ,

1CV8428, ICV 8439, ICV 8435, and 1CV8441 were secured in the closed position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is a violation of TS 3.3.1, Table 3.3-1, Functional Unit No. 6.b (454/86025-02c(DRP)). The failure to secure t

the addition of positive reactivity within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is a violation of l TS 3.3.1, Table 3.3-1, Functional Unit No. 6.b (454/86025-02d(DRP)).

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At 1350 on July 3, the licensee believing that SPP 86-71 and IPOS 3.1.1-15 could not be performed quickly enough to demonstrate N31 and N32 operable and allow sufficient time to cool the RCS down to Mode 5 within the administrative limit on cooldown rate of 50 degrees F per hour, invoked 10 CFR 50.54(x).

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10 CFR 50.54(x) states, "A licensee may take reasonable action that departs from a license condition or a technical specification (contained in a license issued under this part) in an emergency when this action is immediately needed to protect the public health and safety and no action consistent with license conditions and technical specifications that can provide adequate or equivalent protection is immediately apparent."

The licensee believed that the prudent course of action was to continue the cooldown and not follow the Action Statement No. 5 of TS 3.3.1, Table 3.3-1 for the following reasons: 1) The licensee was unable to specifically identify the source of the leak and believed it to be RCS pressure boundary leakage; 2) Both SRs were indicating properly and all alarms and protective features other than BDPS were functional; and 3) the unit was in an emergency as defined by the Generating Station Emergency Plan. The licensee discussed their intended course of action with Region III management before resuming the cooldow Post event discussions between Region III, NRR, and the licensee have determined that the actions taken by the licensee were

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technically sound, but a more prudent option could have been followed other than invoking 10 CFR 50.54(x). These other options consisted of an Emergency Technical Specification change from NRR or the granting of " Enforcement Discretion" by the Region III Regional Administrator. Due to the untimely identification of the problem with the BDPS the licensee believed that their choice of options was constrained by the limited time remaining to reach Mode 5. However,

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the failure to address the problem which would occur with BDPS on a timely basis rests with station management. Station management had a reasonable basis (LER 454/86015) to believe that BDPS and consequently both SRs would be inoperable the next time the unit was shutdown; and failed to ensure that necessary actions were accomplished in a timely manner. Consequently, due to the late identification of this problem the licensee was forced to invoke 10 CFR 50.54(x).

At 1529 on July 3, 1986 the cooldown was resumed. At 0200 on July 4, 1986 Mode 5 was entered and the Unusual Event was terminate By 0245 on July 4, 1986, both N31 and N32 had been satisfactorily testad and were declared operabl The source of leakage was identified to be a failed diaphragm on a

"Kerotest" valve which is used as an isolation valve for the RCS Loop 1D Hot Leg resistance temperature detector manifold (IRC80360).

The leaking valve was replaced due to difficulties encountered

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during attempts to disassemble the valve. Additional work performed during this outage was to replace the seal package for_the 1C Reactor Coolant Pump, to replace-the 1C Pressurizer Code Safety Valve (1RY8010C), which was believed to be leaking, and to replace some terminal strips in the Main Steam Isolation Valves (MSIV) with environmentally qualified splices. Followup of questions relating to the environmental qualification of the MSIVs will be accomplished by the Division of Reactor Safety, RIII, and is tracked by Unresolved Item (454/86025-04(DRS)).

On July 17 after completion of repairs the licensee began to heat the unit up to normal operating conditions. In Mode 3 at approximately 1750 psig 1RY8010C actuated. The closed indicator light extinguished and an immediate 50 psig decrease in pressurizer pressure was observed as was an increase in the Pressure Relief Tank level, pressure, and temperature. However, the open indicator light for 1RY8010C did not illuminate nor did the " Pressurizer Safety Valve Actuated" annunciator alarm. Operations department personnel entered containment and identified that both valves 1RY8010C and 1RY8010A were leakin The unit was cooled down to Mode 5 and valves 1RY8010A and 1RY8010C were removed. 1RY8010C was inspected and it was discovered that the valve disk insert was not installe New valves were subsequently installed at 1RY8010A and 1RY8010 The unit was heated up to normal operating conditions and was taken critical at 2357 on July 23, 1986, connected to the grid at 0400 on July 24, and returned to rated powe .

The events surrounding the installation of a non operable pressurizer code safety valve are discussed in Inspection Report 454/86029(DRP).

11. Licensee Actions Concerning Suspected Drug Use (99014)

Concern: On July 17, 1986, the inspector was informed by the Byron i Station Manager that the licensee had received an anonymous statement from an individual who identified an employee at the Byron Station whom

the writer had reason to believe may be using drugs. The person named in this concern was a non-management employee performing safety-related F work.

i Findings: In keeping with the licensee's established drug awareness program, the employee was interviewed by Byron supervisors and managers

' and relieved of all duties at the Byron Station. The employee's photo

. identification. security badge and access key-card were revoked and the L individual's access was denied pending the outcome of an investigation.

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In keeping with the Commonwealth Edison established procedures, the

! employee provided an observed urine specimen to a licensee medical facility for analysis. The tests results of the urinalysis were

negative. Additionally, a review of the employee's performance by station management revealed no abnormal behavior during the employee's

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n tenure prior to this accusatio Based on the negative test results, recommendations of Byron Management and endorsement by the company medical staff, the individual was restored to security status and returned to full dut This concern is considered close . Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, items of noncorapliance, or deviations. Unresolved items disclosed during the inspection are discussed in Paragraphs 3.c and 1 . Exit interview (30703)

The inspectors met with licensee representatives denoted in Paragraph 1 at the conclusion of the inspection on July 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