IR 05000455/1986021
| ML20203D055 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 07/15/1986 |
| From: | Forney W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20203D038 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-***, TASK-TM 50-455-86-21, NUDOCS 8607210129 | |
| Download: ML20203D055 (14) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-455/86021(DRP)
Docket No. 50-455 License No. CPPR-131 Licensee: Commonwealth Edison Company Post Office Box 767
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Chicago, IL 60690 Facility Name: Byron Station, Unit 2 Inspection At: Byron Station, Byron, IL Inspection Conducted: June 23 - July 8, 1986 Inspectors:
J. M. Hinds, Jr.
J. A. Malloy Approved By:
Nr
> /,th4 Reactor Projects Section 1A D(te /
Inspection Summary Inspection on June 23 - July 8, 1986 (Report No. 50-455/86021(DRP))
Areas Inspected: Special, unannounced safety inspection by the resident inspectors of licensee actions on TMI-Action Plan Items contained in NUREG-0737.
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Results: No violations or deviations were identified nor were any items identified which could affect the public health and safety.
8607210129 860716 PDR ADOCK 05000455 G
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DETAILS 1.
Persons Contacted Commonwealth Edison Company
- 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. 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, Compliance Supervisor F. Hornbeak, Technical Staff Supervisor R. Flahive, Radiation / Chemistry Supervisor
- D. Elias, Project Engineer
- E. Falb, Unit 2 Testing Supervisor
- J. Langan, Regulatory Assurance Staff
- M. Snow, Regulatory Assurance Assistant Supervisor
- E. Zittle, Regulatory Assurance Staff
- K. Yates, Nuclear Safety Staff
- S. Nosko, Quality Assurance Engineer The inspector also contacted and interviewed other licensee and contractor personnel during the course of this inspection.
- Denotes those present during the exit interview on July 8,1986.
2.
Purposes of Inspection
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This inspection concerned licensee actions relative to certain NRC requirements developed as a result of the Three Mile Island Unit 2 accident. These requirements were contained in NUREG-0737,
" Clarification of TMI Action Plan Requirements," and were assigned item numbers for identification.
Completion of certain licensee actions required by the items reviewed during this inspection were previously verified by-NRC inspection personnel. This inspection was conducted to establish a correlation between individual items (by number and title assigned in NUREG-0737)
and the NRC Inspection Reports which document these verifications.
This inspection was also conducted to supplement previous inspection activities as necessary to verify completion of required licensee actions for all items reviewed.
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5.
I.A.1.1 Shift Technical Advisor (STA)
The NRC Staff found the licensee's training program for individuals performing the STA function acceptable in Section 13.2.2.2 of the Byron Safety Evaluation Report (SER) [NUREG-0876]. Training of designated individuals was verified complete as part of Region III inspections of operitting staff training documented in NRC Inspection Report Nos. 455/84042(DRS), 455/84048(DRP) and 455/84053(DRP). The inspector established by review of operating staff work schedules, discussions with licensee personnel and direct observation during routine plant tours that the STA position for both units has been filled on each shift since the beginning of Unit 1 fuel load.
4.
I.A.1.2 Shift Supervisor Administrative Duties - Implementation The Staff found that procedures issued and implemented at the time of an NRC interoffice evaluation team site visit (see Item I.B.1.2 below)
satisfied this item. This finding was documented in Section 13.5.1.2 of the Byron SER.
5.
I.A.1.3 Shift Manning - Implementation The inspector determined that the licensee established and implemented Byron Administrative Procedure (BAP) 100-7, " Overtime Guidelines for Personnel That Perform Safety Related Functions", and BAP 320-1, " Shift Manning". BAP 100-7 incorporated the guidelines on the use of overtime provided in NUREG-0737 as well as supplemental guidance provided in NRC Generic Letter 82-12. BAP 320-1 requires that shift staffing be in
accordance with licensee commitments discussed in Section 13.1.2.2 of the Byron SER and the approved technical specifications. The minimum
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shift crew exceeds the requirements of NUREG-0737. These procedures l
were previously reviewed during inspections documented in NRC Inspection i
Report Nos. 455/84011(DRP), 455/84014(DRP) and 455/84053(DRP) and found acceptable.
The inspector determined by discussions with licensee personnel, review of the licensee's organizational chart, Byron Operating Department work schedules, and NRC operator license examination records that the licensee has established an adequate staff prior to Unit 2 fuel load to meet commitments, procedural requirements, and the approved technical specifications.
6.
I.A.2.1 Immediate Upgradinc of Reactor Operator (RO) and Senior Reactor Operator (SRO) Training anc; Qualifications - Implementation In Section 13.2.1.2 of the Byron SER the Staff concluded that the training program established by the licensee at the time of the review met the requirements of this item. Completion of required training for both license and nonlicense operating staff personnel was verified and documented in NRC Inspection Report Nos. 455/84014(DRS),455/84042(DRP)
and 455/84048(DRP). The inspector determined that 7 individuals licensed as SR0s on Unit 1 and 5 individuals licensed as R0s on Unit I will be examined for Unit 2 licenses in September 1986.
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I.A.2.3. Administration of Training Programs for Licensed Operators ~
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The inspector verified by discussion with Byron Training Department personnel and review of operator license records.that licensed operator training was conducted by instructors who have either held an SR0 i
license or have been certified by the NRC as SR0 instructors. After the administration of NRC cperator exams at Byron, the Training l
Department has four persons licensed at the SR0 level. -These people conduct the licensed operator training program..The training department
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has contracted additional instructors all of which have either held an SR0 license or have been certified by the NRC as SR0 instructors.
All of the operator training is conducted under the supervision of an individual who presently holds an SR0 license on Byron Station. The portion of training that is conducted at the Production Training center is under the supervision of an individual who presently holds an SR0
license on Byron Station.
8.
I.B.I.2 Evaluation of Organization and Management Improvements of OL
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Applicants
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An evaluation team consisting of personnel from the NRC Offices of Nuclear Reactor Regulation and Inspection and Enforcement visited-the i
Byron site on September 8 through 11, 1981. The team's observations and findings from this visit were documented in Chapter 13 of the Byron SER.
l The inspector reviewed the Organization and Administration Manual for the licensee's Nuclear Safety Department which described and proceduralized the functioning of the Independent Safety Engineering Group (ISEG)
established in accordance with the requirements of this item. The
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inspector determined by interview with individuals assigned to the Byron
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ISEG that the group was fully staffed and functioning in accordance with i
licensee commitments and the approved technical specificatic,ns.
9.
I.C.1 Short Terms Accident and Procedure Review Licensee action on this item will be completed during the third quarter-
of 1986.
10.
I.C.2 Shift Relief and Turnover Procedures I
The inspector reviewed Byron Administrative Procedure BAP.335-1, j'
conduct of turnover and relief for the shift engineer, shift control
" Operating Shift Turnover and Relief." This procedure details the t
room engineer, shift foreman, radwaste foreman, nuclear station-operator, and equipment operator.
Implenentation.of this procedure has been
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observed by the resident inspectors on various occasions with no adverse
findings, i
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11.
I.C.3 Shift Supervisor Responsibilities The Staff found that procedures issued and implemented at the time of an NRC interoffice evaluation team site visit (see Item I.B.1.2. above)
satisfied this item. The Staff's relevant findings concerning this item were included in the discussion of Item 1.A.1.2 in Section 13.5.1.2 of the Byron SER.
The inspector verified that, per licensee commitment, the licensee has continued to reissue management directives which emphasize primary management responsibility and clearly establish the command authorities of the shift supervisor.
Resident inspector observations made during routine inspections have indicated that these procedures and directives were effectively implemented.
12.
I.C.4 Control Room Access The inspector reviewed Byron Administrative Procedures (BAP) 900-10.
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" Control Room Access", and BZP 100-01, " Supervision of Emergencies, Exercises and Drills." These procedures assign authorities and i
responsibilities for access control during normal operations and emergency conditions. Routine access is limited to designated essential management and operating personnel. Access is granted to other personnel
only when necessary for them to perform essential tasks. The procedures incorporated the requirements of Section 2.2.2.a of NUREG-0578, "TMI-2
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Short Term Lessons Learned."
Resident inspector observations made during this inspection and previous routine inspections indicated that these procedures were effectively implemented.
13.
I.C.5 Procedures for Feedback of Operating Experience to Plant Staff The Staff reviewed Byron Administrative Procedure (BAP) 300-8, " Operating Experience Feedback", and the Licensee's Organization and Administration Manual for the Department of Nuclear Safety and found these documents acceptable for implementing the requirements of this item. This finding was documented in Section 13.5.1.2 of the Byron SER.
BAP 1260-1 " Operating Experience Feedback" has replaced BAP 300-8.
The inspector reviewed BAP 1260-1 and the Licensee's Organization and i
Administration Manual for the Department of Nuclear Safety. The inspector concluded, based upon previous reviews of licensee actions on Licensee Event Reports, Deviation Reports, NRC Bulletins and Circulars
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and by attendance at the Licensee's daily planning meetings that the j
requirements of BAP 1260-1 were effectively implemented.
The inspector also verified that the offsite review-group chartered to review and evaluate operating experiences for applicability to. Byron was in place and fuqctioning in accordance with the approved Technical Specifications.
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I.C.6 Verify Correct Performance of Operating Activities Inspection of this item for Unit I was performed and documented in NRC Inspection Report No. 454/84-41(DRS). The inspection identified that provisions for independent verification were not always included in surveillance procedures as required.
In response to this finding the license issued Byron Administrative Procedure (BAP) 100-13, " Guidelines for Performance of Independent Verification of Proper Equipment Alignment." This procedure provides administrative guidelines for independent verification based upon the requirements of this item. The licensee reported on January 30, 1985, that all affected procedures had been reviewed and that, where necessary, procedures were permanently revised for conformance to the requirements of this iten. The inspector reviewed BAP 400-9, " Maintenance Alterations" and found that provisions for independent verification were included.
15.
I.C.7 NSS Vendor Review of Procedures Licensee action on this item will be completed after Startup Testing for Unit 2.
16.
I.C.8 Pilot Monitoring of Selected Emergency Procedures for NT0Ls Licensee action on this item will be completed during the third quarter of 1986.
17.
I.D.1 Control Room Design Reviews
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Licensee action on this item will be completed in December 1986.
18.
I.D.2 Plant Safety Parameter Display Conscle Licensee action on this item will be completed during the third quarter of 1986.
19.
I.G.1 Training During Low Power Testing Licensee action on this item will be completed during the first refueling outage of Unit 1.
20.
II.B.1 Reactor-Coolant System Vents The inspector reviewed the licensee's position in Appendix E of the Byron FSAR, " Requirements Resulting From TMI-2 Accident". The inspector verified by test procedure review that vent valve actuation and position indication and vent line temperature monitoring instrumentation were i
preoperationally tested as part of preoperational test 2.63.62, " Reactor Coolant Instrument and Component Check".
Verification of emergency procedures for operation of the reactor vessel head vents will be completed after the Unit 2 Emergency Procedures have been approved (during the third quarter of 1986).
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Byron Technical Specification 3/4.4.11, " Reactor Coolant System Vents",
requires at least one reactor vessel head vent path consisting of two valves in series powered from emergency buses to be operable in Modes 1, 2, 3 and 4.
This specification also includes suitable surveillance testing requirements to establish vent path operability once per 18 months.
21.
II.B.2 Plant Shielding Inspection of licensee actions relative to this item, documented in NRC Inspection Report No. 455/84027(DRMSP), did not identify any outstanding issues. Subsequently Amendment 47 to the FSAR was issued which amended the description of liquid source term assumptions used in the licensee's earlier review of plant shielding. The amenoment indicated that exception was taken to the requirement of this item for including Noble gases in liquid source terms.
22.
II.B.3 Post Accident Sampling Verification of licensee actions relative to this item was previously performed and documented in NRC Inspection Report Nos. 455/83026(DRMSP),
455/84008(DRMSP)and 455/84044(DRSS). Resolution of certain inspector concerns was accomplished by a variance from the requirements of this item. The variance delayed empirical determination of sample line losses and included a commitment to install appropriate sample line heat tracing in the Containment Atmosphere Sample Panel (CASP). The inspector verified that heat tracing was installed in the CASP and tested during preoperational test 2.61.61, " Process Sampling Hydrogen Monitoring".
The Byron Unit 1 operating license has been conditioned [ License Condition C.(9)] to require the licensee complete actions to account for sample line losses and other phenomenon to demonstrate that radioiodine and particulate samples are representative prior to startup after the first refueling.
The NRC Staff has accepted a variance with the requirement of this item for backup sampling capability for the on-line reactor coolant hydrogen analyzer.
The license condition and the variance involving the reactor coolant hydrogen analyzer was not verified during this inspection.
23.
II.B.4 Training for Mitigating Core Damage
The Staff documented its acceptance of the licensee's training programs to address this item in Section 13.2.1.2 of the Byron SER. Training of
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both licensed and non-licensed personnel was verified complete as part of j
NRC inspections of operating staff training documented in NRC Inspection Report No. 455/84053(DRP). During this inspection, the inspector reviewed personnel records to verify that all licensed personnel as well as non-licensed personnel performing the Shift Technical Advisor (STA)
function have successfully completed the training required by uiis item.
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24.
II.D.1 Relief and Safety Valve Test Requirements In lieu of conducting safety / relief of block valve performance tests, the licensee adapted the results of the full scale valve testing program performed by the Electric Power Research Institute (EPRI) on behalf of the PWR Owners Group. The Staff reported in 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 acceptable.
The inservice testing of safety valves 2RY8010A, 2RY8010B, and 2RY8010C, power operated relief valves (PORVs) 2RY455A and 2RY456, and power operated relief block valves 2RY8000A and 2RY8000B is scheduled prior
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to full power Unit 2 testing.
25.
II.D.3 Valve Position Indication In Section 7.5.2.3 of the Byron SER the Staff found that the Byron design conformed to the requirements of this item by inclusion of safety grade environmentally qualified positive valve position indication for the pressurizer code safety valves, power operated relief valves (PORVs), and PORV block valves as well as control room annunciator alarms for PORV or safety valve actuation. Previous NRC inspections related to this item included witnessing portions of preoperational test 2.69.60, " Reactor Coolant-Pressurizer." These inspections were documented in NRC Inspection Report No. 455/86016(DRP).
During this inspection the inspector verified, by test results review that the installed safety PORV and PORV block valve position indication and control room annunciation and PORV and safety valve actuation were satisfactorily demonstrated to function per design in preoperational test 2.69.60.
26.
II.E.1.1 Auxiliary Feedwater System Evaluation (AFWS)
Section 10.4.9 of the Byron SER documented the Staff's receipt and review of the licensee's evaluation of the AFWS conducted pursuant to this item.
No hardware modifications were required for the AFWS as a result of the Staff's review; however, several commitments, involving technical specifications and administrative controls to enhance AFWS reliability, were requested from the licensee and reviewed by the staff.
The licensee was to submit technical specifications which state that one AFW pump may be inoperable for no more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and, that if this time limit was exceeded, the affected unit must be placed in hot standby within six hours and hot shutdown within an additional six hours. The licensee was to provide technical specifications which require that if the opposite unit's "A" diesel generator was inoperable for seven days, immediately restore it to an operable condition or place the unit in hot shutdown within six hours and in cold shutdown within an additional six hours. The licensee committed to adding a full AFWS flowpath
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verification of proper AFWS valve position following restoration of an
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AFWS train to service after periodic maintenance or testing. The licensee also comitted to performing 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> endurance tests of the AFW pumps.
The inspector verified that the approved technical specifications 3.7.1.2, " Auxiliary Feedwater System," and 3.8.1 "AC Sources -
Operating," satisfied the licensee's commitments with one exception; in lieu of requiring the Unit 2 "A" diesel generator OPERABLE for the first two years of operation of Byron Unit 1, the Unit 2
"A" diesel generator was required to be " capable of being manually started and crosstied to electrical bus 141." After two years of Unit 1 operation the Unit 2 "A" diesel generator OPERABILITY will be required per the licensee's commitment. The inspector verified that the Staff has approved this deviation from the licensee's original commitment.
Implementation of commitments concerning AFWS valve position verification was verified by Unit I surveillance test procedure reviews documented in NRC Inspection Report Nos. 455/84038(DRP)and 455/84048(DRP). The Unit 2 surveillance test procedures had not yet been approved.
Inspector verification of the performance of the AFW pump endurance tests was accomplished as part of AFWS preoperational test activities documented in NRC Inspection Report No. 455/86004.
27.
II.E.1.2 Auxiliary Feedwater System Initiation and Flow In Section 7.3.2.8 of the Byron SER the Staff found that the Byron AFWS design conformed to the requirements of this item. Subsequent to this Staff review, the NRC Office of Inspection and Enforcement performed an Integrated Design Inspection (IDI) which examined the AWFS design in considerable detail. This inspection included instrumentation and control aspects of the AFWS design input through the installed instrumentation and control equipment at the Byron plant. The inspection was conducted against the requirements of this item as well as other NRC requirements and licensee commitments contained in the Byron FSAR. The results of the IDI were documented in NRC Inspection Report No. 454/83032. Concerns identified by the IDI in the area of AFWS instrumentation and control have been satisfactorily resolved.
28.
II.E.3.1 Emergency Power to 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.
Byron Operating Surveillance Procedure (BOS) 4.3.3-1 will test the cross-tie for the pressurizer heaters to the Engineered Safety Features (ESF) power supply once per 18 months. Unit 2 B05 4.3.3-1 has not yet been approved.
29.
II.E.4.1 Dedicated Hydrogen Penetrations In Section 6.2.5 of the Byron SER and Supplements to the Byron SER the Staff found that the Byron design met the requirements of this item.
Preoperational testing of the hydrogen recombiners was performed for both units to verify that the installed system performed in accordance with design. NRC inspections verified that applicable test procedures were
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developed and approved in accordance with the licensee's administrative controls and that results were properly evaluated and that identified deficiencies were appropriately resolved, including completion of any required retesting. This inspection effort was documented in NRC Inspection Report Nos. 455/84017(DE) and 455/84033(DRS).
Installation of redundant hydrogen analyzers was reviewed during inspections discussed under. Item II.B.3 in Paragraph 22 of this report.
Inspector review of operating procedures used for the operation of the hydrogen recombiners and the hydrogen monitoring system were performed during an inspection documented in NRC Inspection Report No. 455/84038(DRP).
Preoperational test 2.61.61, " Process Sampling Hydrogen Monitors" has not yet been completed.
30.
II.E.4.2 Containment Isolation Dependability
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In Section 6.2.4 of the Byron SER the NRC Steff 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 conditions.
Verification that the installed containment isolation system functioned in accordance with the design will be accomplished by witnessing of test performance and results evaluations for preoperational test 2.26.61
"ESF(ECCS Full Flow)", 2.26.62 "ESF - Logics and Time Response", 2.58.60
" Local Leakrate" and 2.58.61 " Containment Purge". These tests have not been fully completed.
Operation and surveillance requirements will be verified when surveillance procedures for Unit 2 are issued.
31.
II.F.1.1 - II.F.1.6 Accident Monitoring Instrumentation Licensee action relative to those portions of this item dealing with the noble gas effluent radiological monitor, iodine / particulate sampling and the containment high range monitor were previously verified during inspections documented in NRC Inspection Report Nos. 455/84008(DRMSP)
and 455/84044(DRSS).
Resolution of inspector concerns identified during these inspections required several variances from requirements of this item. These variances have been granted by the Office of Nuclear Regulation.
The inspector reviewed the licensee's response to the portions of this item dealing with wide range containment pressure and containment water level instrumentation. The inspector visually observed the installed
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wide range containment pressure indicators, containment water level indicators, and containment sump level indicators all located on main control board 2PM06J. All of the instruments had been preoperationally tested and were determined to have been calibrated within the intervals required by Technical Specification 3/4.3.3.6.
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Containment. hydrogen monitoring was discussed under Item II.E.4.1 in i
Paragraph 29 of this report. During this inspection the inspector observed the installed containment hydrogen monitoring-indication on main control board 2PM06J. The inspector verified 'that the indicators had been calibrated within the interval required by Technical Specification 3/4.3.3.6 32.
II.F.2 Instrumentation for Detection of Inadequate Core Cooling
In Section 4.4.7 of Supplement 5 to the Byron SER the Staff found that
the Byron design met the requirements of this item; however, the Staff required'that prior to startup after the first refueling the licensee
provide the ability to trend subcooling utilizing the backup subcooling
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j margin monitor displays. The Staff also required that the licensee
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develop emergency operating procedures and identify and explain
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deviations from the Westinghouse Owners Group ~(WOG) generic technical
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guidelines prior to exceeding 5% power. The Byron SER indicated that the Inadequate Core Cooling Instrumentation (ICCI) System would be operational prior to fuel load.
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During an NRC audit of inadequate core cooling instrumentation at Byron l
Units 1 and 2 on March 26, 1986, the audit team determined that a program
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has been implemented thct meets the requirements of NUREG-0737. The i
audit team determined.that the licensee will not be required to provide the temperature trending discussed in Supplement 5 to the Byron SER.
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The ICCI system utilized the process computer to calculate degrees of subcooling/superheat based upon input from wide-range RCS pressure
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instrumentation and the Incore Thermocouple (IT) subsystem. Additionally, the ICCI system included a Reactor Vessel Level Indication Subsystem
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(RVLIS).
Subcooling margin, incore thermocouple readings, and reactor-
vessel level were to be displayed on the Safety Parameter Display -
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System (SPDS). Safety grade backup displays were provided for the IT
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subsystem and the RVLIS subsystem.
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l Preoperational testing of the RVLIS and the IT subsystems, and the system demonstration for the SPDS has not yet been completed.
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The licensee has made the required submittal which identified and explained deviations from the WOG technical guidelines for emergency
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procedures utilized for detection and mitigation of inadequate core
cooling. These emergency operating procedures will be completed during
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the third quarter of 1986 and will be discussed under Item I.C.1 j
(Paragraph 9) of this report.
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II.G.1 Power Supplies for Pressurizer Relief Valves, Block Valves and Level Indicators
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In Section 8.4.7 of the Byron SER the NRC Staff found that the Byron
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design conformed to the requirements of this item. ' Verification by
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NRC inspection that installed equipment performed as designed was accomplished by inspector witnessing of portions of preoperational
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test 2.69.10.
" Reactor Coolant-Pressurizer". These inspections were documented in NRC Inspection Report No. 455/86016(DRP). This preoperational test and associated retest verified functioning of the Class IE powered pressurizer PORVs and PORV block valves as well as pressurizer level instrumentation.
The inspector verified that the Technical Specification 3/4.3.3.6 and 3/4.4.4 included requirements for PORVs, PORV block valves, and pressurizer level instrumentation operability and specified appropriate surveillance testing to be conducted to establish equipment operability.
34.
II.K.2.13 Thermal Mechanical Report
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Licensee action on this item will be completed during the third quarter of 1986.
35.
II.K.2.17 Potential for Voiding in the Reactor Coolant System During Transients Licensee action on this item will be completed during the third quarter of 1986.
36.
II.K.2.19 Benchmark Analysis of Sequential Auxiliary Feedwater Flow In a letter dated June 26, 1981, S. A. Varga (NRC) to J. S. Abel (Commonwealth Edison Company), the NRC Staff concluded that concerns expressed in this item are not applicable to NSSS with inverted U-tube steam generators such as those designed by Westinghouse and installed at Byron.
37.
II.K.3.1 Installation and Testing of Automatic Power-Operated Relief Valve Isolation System In the Byron SER the applicant recommended against the installation of automatic PORV isolation in accordance with WOG generic submittal.
SER Supplement 5 closed this item based upon the determination that the requirements of NUREG-0737 Item II.K.3.2 are met with the existing power-operated relief valves, safety valves and high-pressure reactor trip setpoints.
38.
II.K.3.5 Automatic Trip of Reactor Coolant Pumps During Loss of Coolant Accident Licensee action on this item will be completed during the third quarter of 1986.
39.
II.K.3.9 Proportional Integral Derivative Controller Modification
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This item implements a Westinghouse recommendation to modify the PORV proportional integral derivative controller to prevent derivative action from opening the PORV. The derivative time constant has been set to zero.
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II.K.3.10 Proposed Anticipatory Trip Modification The applicant has not proposed a change in the interlock for the anticipatory reactor trip on turbine trip, therefore this item is not applicable to Byron.
41.
II.K.3.25 Effect of Loss of AC Power on Pump Seals The loss of offsite power has no effect on reactor coolant pump seals since emergency diesel generators rapidly restore the AC power to safety related pumps supplying cooling water.
42.
II.K.3.30 Small Break Analysis Methods In a letter dated May 24, 1985, B. J. Youngblood (NRC) to D. L. Farrar (CECO), the NRC approved the use of NOTRUMP, a program developed to respond to this item, which demonstrates the conservatism of the present small break LOCA analysis.
43.
II.K.3.31 Plant Specific Calculations to Show Compliance with 10 CFR 50.46 Licensee action on this item will be complete during the third quarter of 1986.
44.
III. A.1.1.1 Emergency Preparedness. Short Tenn Verification of licensee actions required by this as well as Item III.A.1.2 " Upgrade Emergency Support Facilities", was performed as part of Region III's emergency preparedness appraisal team inspection documented in NRC Inspection Report No. 455/83039(DRMSP).
Inspection concerns were satisfactorily resolved during followup inspections documented in NRC Inspection Report Nos. 455/84031(DRSS),455/84039(ORSS)
and455/84052(DRSS).
45.
III.A.1.2 Upgrade Emergency Support Facilities See discussion of Item III.A.1.1 in Paragraph 44 of this inspection report.
46.
III.A.2 Emergency Preparedness Licensee action on this item will be completed during the third quarter of 1986.
47.
III.D.1.1 Primary Coolant Sources Outside Containment In Sections 9.3.5.1 and 9.3.5.2 of Supplement 5 to the Byron SER the Staff described the licensee's submittals pursuant to this item and documented the Staff's finding of acceptability.
The Staff did, however, require that the licensee provide a report to the NRC Staff upon achieving full power operation detailing all recorded leakage and, as a direct result of the evaluation of this leakage, all preventive maintenance performed.
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'o The report will also identify general leakage criteria to be applied during the first fuel cycle as the basis for instituting corrective action in the form of preventive maintenance.
The inspector verified that licensee commitments relative to this item were reflected in approved Technical Specification 6.8.4.a, " Reactor Coolant Sources Outside Containment", requirements.
Submittal of the report required by the NRC Staff will be after Unit 2 achieves full power.
48.
III.D.3.3 Inplant Radioiodine Monitoring Verification of licensee actions required by this item was performed by NRC inspection personnel during inspections documented in NRC Inspection Report Nos. 455/83026(DRMSP) and 455/83040(DRMSP).
49.
III.D.3.4 Control-Room Habitability In Section 6.4 of the Byron SER, the Staff found that the control room habitability systems were adequate to provide safe habitable conditions within the control room under both normal and accident conditions. The Staff concluded that the control room satisfies the requirements of this item.
50. Exit Interview (30703)
The inspectors met with licensee representatives denoted in Paragraph 1 at the conclusion of the inspection on July 8, 1986. The inspectors summarized the purpose and scope of the inspection and the findings.
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 proprietary.
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