IR 05000455/1986031
| ML20215J571 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 10/20/1986 |
| From: | Forney W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20215J566 | List: |
| References | |
| 50-455-86-31, NUDOCS 8610270092 | |
| Download: ML20215J571 (9) | |
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I U. S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-455/86031(DRP)
Docket No. 50-455 License No. CPPR-131 Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name: Byron Station, Unit 2 Inspection At: Byron Station, Byron, IL Inspection Conducted:
September 13 - October 15, 1986 Inspectors:
J. M. Hinds, Jr.
P. G. Brochman R. M. Lerch J. A. Malloy Ld. U52EkWW Approved By: W.Lt'Forney,hief
/o to (c Reactor Projects Section 1A ate Inspection Summary Inspection on September 13 - October 15, 1986 (Report No. 50-455/86031(DRP))
Areas Inspected: Routine, unannounced safety inspection by the resident inspectors of 10 CFR 21 reports; SERs; comparison of as-built plant to the FSAR; operating procedures; emergency procedures; housekeeping; licensee actions concerning suspected drug use; management meetings; operational readiness inspection; and other activities.
Results: Of the seven areas inspected, no violations or deviations were identified in six areas; one violation was identified in the remaining area:
(failure to control flammable materials - Paragraph 7). This violation is of more than minor safety significance and had it gone undetected, had the potential to affect the public health and safety.
8610270092 861020 PDR ADOCK 05000455 G
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DETAILS 1.
Persons Contacted Commonwealth Edison Company
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C. Reed Vice President for Nuclear Operations
$# T. Maiman, Vice President and Manager of Projects
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W. Shewski, Quality Assurance Manager
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K. Graesser, Division Vice President Byron, Braidwood, LaSalle
$# R. Querio, Station Manager
$# V. Schlosser, Byron Project Manager
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K. Hansing, Director of Quality Assurance (Engineering / Construction)
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D. Farrar, Director of Nuclear Licensing
$# R. Pleniewicz, Production Superintendent
$ *R. Ward, Services Superintendent, Byron
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C. Schroeder, Services Superintendent, Braidwood
$# R. Tuetken, Startup Superintendent
$#*E. Martin, Quality Assurance Superintendent
$ *W. Burkamper, Quality Assurance Supervisor, Operations
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J. Woldridge, Quality Assurance Supervisor (Construction)
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K. Ainger, Nuclear Licensing Administrator
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L. Sues, Assistant Superintendent, Operating
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G. Schwartz, Assistant Superintendent, Maintenance T. Joyce, Assistant Superintendent, Technical Services D. St. Clair, Assistant Superintendent, Work Planning
- R. Moravec, Assistant Construction Superintendent
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D. Elias, Project Engineer, Byron, Braidwood
- R. Klingler, Project Quality Control Supervisor
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G. Toleski, Nuclear Security Administrator
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H. Campbell, Production Engineering Supervisor W. Blythe, Operating Engineer, Unit 0'
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T. Tulon, Operating Engineer, Unit 1
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D. Brindle, Operating Engineer, Unit 2 J. Schrock, Operating Engineer, Rad-Waste
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A. Chernick, Regulatory Assurance Supervisor
$#*E. Falb, Unit 2 Testing Supervisor
- A. Rosenbach, Quality Assurance Supervisor F.'Hornbeak, Technical Staff Supervisor R. Flahive, Radiation / Chemistry Supervisor P. O'Neil, Quality Control Supervisor
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D. Robinson, OnSite Nuclear Safety Supervisor
$#*M. Snow, Assistant Regulatory Assurance Supervisor
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J. Pizzica, Staff Assistant
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J. Pausche, Regulatory Group Leader
- E. Zittle, Regulatory Assurance Staff
$# D. Milroy, Startup Engineer
- J. Langan, Regulatory Assurance Staff
- J. Snyder, Quality Assurance Inspector
- S. Kraus, Quality Assurance Inspector
- E. Cope, Operations Staff
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- R. Linboom, Station Fire Marshal
- C. Diaz, Fire Protection Engineer Illinois Department of Nuclear Safety
$B. Metrow, ASME Code Engineer / Division of Engineering The inspector also contacted and interviewed other licensee and contractor personnel during the course of this inspection.
- Denotes those present during the management meeting on September 25, 1986.
$ Denotes those present during the management meeting on October 9, 1986.
- Denotes those present during the exit interview on October 15, 1986.
2.
Followup on 10 CFR Part 21 Reports (92716)
(Closed) 10 CFR 21 Report (455/85004-PP): Problems with terminal blocks in Anchor / Darling Main Steam Isolation Valves (MSIV).
In a 10 CFR 21 report submitted to the NRC, Anchor / Darling identified a problem with the environmental qualification of some replacement terminal strips which could be used in MSIVs, nylon verses polysulfore. To eliminate the problem the licensee decided to remove all terminal strips and to splice the field wires together using environmentally qualified RayChem splice kits. The inspector reviewed wiring diagrams for the 2A, 2B, 2C, and 2D MSIVs (drawings 6E2-4382E - H) which directed that the terminal strips be removed and that the wires be spliced together using RayChem qualified splices. The inspector also reviewed Construction Work Records (CWR)
M56007 and MS60011 and verified that the CWRs indicated that the wires had been spliced, inspected, and tested satisfactorily. Based on this review this item is considered closed.
3.
Followup on Byron Safety Evaluation Report (SER) Items (92719)
(Closed) SER Item (455/83000-22):
Installation of P-4 voltmeters at reactor trip switchgear.
In Supplement No. 2 of NUREG-0876, Paragraph 7.3.2.9, " Byron SER " the licensee committed to the installation of permanent' voltmeters at the reactor trip switchgear to facilitate testing of the P-4 interlock. The inspector reviewed drawings 6E2-4030EF22, 6E2-4030EF66, 6E2-4030RD06,-6E2-4030RD07, 6E2-4208A, and 6E2-4208B which provide schematic and wiring details for the P-4 voltmeters. The inspector verified that the voltmeters were installed at the reactor trip switchgear 2RDOSE as described in the drawings. Based on this review this item is considered closed.
No violations or deviations were identified.
4.
Comparison of As-Built Plant to the FSAR (37301)
a.
Purpose The inspectors commenced a review of selected safety related mechanical and fluid systems to verify that the as-built plant
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conforms to commitments contained in the Byron FSAR. This inspection consisted of a review of the as-built drawings of six safety related s Diesel Fuel Oil (D0), Main Steam (MS),
Pressurizer (RY)ystems:, Reactor Floor Drains (RF), Residual Heat Removal (RH), and Safety Injection (SI), against their FSAR drawings and descriptions; and a field verification that the installed system was in agreement with the drawings, the proposed Technical Specifications, Technical Specification surveillances, and the FSAR.
b.
Results This. inspection is an ongoing effort and its completion will be discussed in a subsequent report.
5.
Operating Procedures Review (42450)
The inspector reviewed general plant operating, system operating, chemistry surveillance, operating surveillance and technical specification surveillance procedures.
Procedures were reviewed for compliance and consistency with' proposed Technical Specifications, Regulatory Guide 1.33 and ANSI N18.7-1976/ANS-3.2.
The following procedures were selected for review:
2BGP 100-1, " Plant Heatup" 2BGP.100-2, " Plant Startup" 2BGP 100-5, " Plant Shutdown" 2BGP 100-7, " Reference Reactivity Data Calculation and Estimated Critical Condition Calculation" 2B0P AB-1,
" Recycle Evaporator Startup" 280P AF-3b, " Filling and Venting the Auxiliary Feedwater System (Uint 2)"
2B0P AF-4b, " Draining the Auxiliary Feedwater System (Unit 2)"
2B0P FW-5b, " Operation of a Startup Feedwater Pump Unit 2" 2BCS 1.2.5.a.1-1, " Unit 2 Borated Water at Shutdown Weekly" 2BCS 1.2.6.a.1-1, " Unit 2 Borated Hai.er While Operating - Weekly" 2BCS 4.7-1, " Unit 2 Reactor Coolant Chemistry Once Per 72 Hours" 2B0S 1.1.4.A-1, " Reactor Coolant System Minimum Temperature for Criticality Surveillance" 2B05 3.2.1-12,
" Reactor Trip P-4 Contacts" 2B0S 8.1.1.2.a-2, "2B Diesel Generator Operability Monthly Surveillance" 280S MS-Q1, " Steam Dump Operability Quarterly Surveillance" 2BVS 4.5.0-1, " Eddy Current Testing of Steam Generator U-Tubes" 2BVS 0.5-2.RY.1, "ASME Quarterly Surveillance Requirements for PORV's and PORV Block Valves" 2BVS 7.1.2.1.a-2, " Diesel Driven Auxiliary Feedwater Pump Monthly Surveillance" 2BVS 8.2.1.2.d-1, "125 Volt Battery Bank and Charger Operability -
Battery Charger"
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As a result of this review, the inspector identified items of concern and provided comments to the licensee for consideration. No regulatory concerns were identified and the inspector concerns were resolved satisfactorily.
No violations or deviations were identified.
6.
Emergency Procedure Review (42452)
Byron Unit 2 Emergency Operating Procedures (E0P) have been written based upon the Westinghouse Owners Group (WOG) Emergency Response Guidelines, Revision 1.
The inspector reviewed the E0Ps for compliarse and consistency with Technical Specifications, Regulatory Guide 1.33, ANSI N18.7-1976/ANS-3.2, and Byron Administrative Procedure 1310-A3, " Byron Emergency / Abnormal / Critical Safety Function Procedures-Writers Guide".
The emergency procedures are broken down into several classes of procedures and these are: Byron Emergency Procedures (BEP), Event Specific Emergency Subprocedures (BEP ES), Byron Status Trees (BST).
Byron Functional Restoration Procedures (BFR), and Byron Emergency Contingency Actions (BCA). The following procedures were selected for review:
2BEP-0, " Reactor Trip or Safety Injection - Unit 2" 2BEP ES-0.1, " Reactor Trip Response - Unit 2" 2BEP-1, " Loss of. Reactor or Secondary Coolant - Unit 2" 2BEP ES-1.1, "SI Termination - Unit 2" 2BEP-3, " Steam Generator Tube Rupture - Unit 2" 2BEP ES-3.3, " Post SGTR Cooldown Using Steam Dump - Unit 2" 2BST-1, "Subcriticality Status Tree" 2BST-2, " Core Cooling Status Tree" 2BST-3, " Heat Sink Status Tree" 2BST-4, " Integrity Status Tree" 2BST-5, " Containment Status Tree" 2BST-6, " Inventory Status Tree" 2BFR-5.1, " Response to Nuclear Generation /ATWS - Unit 2" 2BFR-C.1, " Response to Inadequate Core Cooling - Unit 2" 2BFR-H.1, " Response to loss of Secondary Heat Sink - Unit 2" 2BFR-H.2, " Response to Steam Generator High Level - Unit 2" i
2BFR-H.3, " Response to Steam Generator Low Level - Unit 2" l
2BFR-P.1, " Response to Imminent Pressurized Thermal Check Condition i
- Unit 2" 2BFR-Z.1, " Response to High Containment Pressure - Unit 2" 2BFR-Z.3, " Response to High Containment Radiation Level - Unit 2" 2BFR-I.2, " Response to Low Pressurizer Level - Unit 2" 2BCA-0.0, " Loss of AC Power - Unit 2" 2BCA-0.2, " Loss of AC Power Recovery with SI Required - Unit 2" 2BCA-3.1, "SGTR with Loss of Reactor Coolant-Subcooled Recovery Desired - Unit 2" 2BCA-3.2, "SGTR with Loss of Reactor Coolant-Saturated Recovery Desired - Unit 2" During this review the inspector identified several concerns related to these procedures.
These concerns were discussed with the licensee's staff and are listed below:
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a.
2BEP-3, Step 23.b specifies a pressurizer level of 21%/(50% Adverse Containment). WOG Procedure E-3, Step 17, which is the equivalent step, specifies a pressurizer level which equates to 4%/(Adverse Containment 38%) for Byron Unit 2 and no documentation of deviation from WOG procedure E-3 exists.
b.
2BFR-H.1, Step 27 exits to 2BEP ES-1.1 Step 11. WOG procedure FR-1.1 specifies an exit which equates to Step 9 of 2BEP ES-1.1 and no documentation of deviation from WOG procedure FR-1.1 exists.
c.
Subsequent to entry into 2BEP ES-1.1 from 2BFR-H.1 no direction is provided to secure the Residual Heat Removal (RHR) pumps prior to completing the procedure. The inspector questioned why no direction was provided.
d.
2BEP-0, Step 19, Response Not Obtained (RNO) directs the operator to 2BFR-H.1. when less than 500 gpm of Auxiliary Feedwater is available and a Safety Injection has occurred.
If a Safety Injection has not occurred then Step 4 of 2BEP-0 exits the operator to 2BEP ES-0.1.
2BEP ES-0.1, Step 4.c, RNO, specifies that with less than 500 gpm of feedwater flow available to the Steam Generators (SGs) the operator
" Establish feed flow to SGs as necessary using: AF or Main FW."
2BFR-H.1 is a very explicit procedure which provides the~ operator with a significant amount of detailed guidance to respond to a loss of all feedwater accident.
The inspector questioned whether the case of a spurious Safety Injection with a loss of all feedwater was a similar enough event to the case of a simple reactor trip with loss of all feedwater so that it would be appropriate to direct the operator to the inadequate heat sink procedure 2BFR-H.1 rather than provide the simple guidance given of establish feed flow. The inspector recognizes that the status tree for inadequate heat sink 2BST-3 would eventually direct the operator to 2BFR-H.1 but questioned whether this action could not be accomplished more expeditiously.
Following discussions with the licensee's staff the licensee agreed to correct the concerns identified in paragraphs a and b.
Followup of these concerns will be tracked by the licensee in Action Item Record AIR 6-86-243. The licensee's staff agreed to discuss concerns c and d with the Westinghouse Owners Group to identify the reasoning behind the procedures and to determine if changes to the procedures were necessary).
The inspector will followup this action as Open Item (455/86031-01(DRP).
No violations or deviations were identified.
7.
Housekeeping / Care and Preservation of Safety-Related Components (71302)
The inspectors conducted plant tours of Unit 2 between September 13 through October 15, 1986. Areas of the Unit 2 plant observed during the tours included the containment, fuel handling and storage areas, auxiliary building areas including the Unit 2 portion of the control
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i room, and the turbine building. Areas were inspected for work in progress, state of cleanliness, overall housekeeping, state of fire protection equipment and methods being employed, and the care and preservation of safety-related components and equipment. The inspectors were accompanied by licensee personnel on portions of the tours for the purpose of identifying areas where additional housekeeping efforts should be concentrated to improve the overall cleanliness of Unit 2.
Inspector concerns were related to the licensee.
During an inspection of the Unit 2 general plant areas conducted on October 1, 1986, the inspector, accompanied by the Station Fire Marshal, was evaluating the state of cleanliness and fire protection measures in the lower cable spreading room at the 439 foot level in the vicinity of L-23.
At a location about 5 feet north of door D-409, the inspector discovered a one pcund coffee can with the lid loosely attached and filled about half full with a liquid which the inspector and the Fire Marshal, based on experience and knowledge of solvents, identified as acetone. Acetone is a moderately volatile, extremely flammable, and potentially toxic fluid solvent. The can of acetone was found unattended in an unapproved container sitting about 4 feet above the floor on a ventilation duct. The coffee can was without any form of hazard warning label and the acetone had deteriorated and deformed the lid such that it was no longer an effective vapor seal.
ANSI N45.2.3-1973, Section 1.1, " Housekeeping During the Construction Phase of Nuclear Power Plants," requires that housekeeping encompasses all activities related to control of cleanliness of facilities, cleanliness of material and equipment, fire prevention and fire protection including disposal of combustible materials and debris.
Section 3.3 requires that the receiving, storage, and handling activities required by this standard shall be performed as specified in ANSI N45.2.2-1972. ANSI N45.2.3-1973 is endorsed by Regulatory Guide 1.39, Revision 2.
Regulatory Guide 1.39, Revision 2 is committed to in Appendix A of the Byron FSAR.
ANSI N45.2.2-1972, " Packaging, Shipping, Receiving, Storage, and Handling of Items for Nuclear Power Plants (During the Construction Phase),"
Section 6.2.3 requires that fire protection commensurate with the type of storage area and the material involved shall be provided and maintained.
Section 6.3.3 requires that hazardous chemicals, paints, solvents, and other material of a like nature shall be stored in well ventilated areas which are not in close proximity to important nuclear plant items.
The failure to store the acetone under supervision in an approved fire
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retardant container with appropriate hazardous warnings and allowing the acetone to be located in an enclosed area with low ventilation flow and in close proximity to electrical and ventilation components important to plant safety is a violation of ANSI N45.2.2-1972, Sections 6.2.3 and 6.3.3 (455/86031-02a(DRP)).
Additionally, the inspector discovered a quart paint-type can of construction adhesive lying in a cable tray. The adhesive was of the
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e type used by construction contractors working in the cable spreading room. The can's label contained a warning that the adhesive material was subject to spontaneous combustion under promoting conditions. The can of adhesive material was found inadequately sealed for the nature of it's contents, without apparent supervision and without appropriate controls.
The cable tray in which the can of adhesive was found was identified as safety-related and is in close proximity to other safety related and important to safety electrical wiring and components.
Failure to maintain adequate control and storage of this hazardous material in an inappropriate location, among safety related cables in a cable tray, without adequate ventilation in close proximity to components important to plant safety is a violation of ANSI N45.2.2-1972, Sections 6.2.3 and 6.3.3 (455/86031-02b(DRP)).
8.
Licensee Actions Concerning Suspected Drug Use (99014)
On September 24, 1986, the licensee notified the Senior Resident Inspector of a concern related to suspected drug use by a station employee. The individual in question was a non-licensed, non-supervisory employee.
In keeping with the licensee's established Drug Awareness Program and based on licensee observations and past discussions with the individual, related to job performance, the employee was relieved of all duties and site access was revoked pending the outcome of drug testing and further evaluation.
Following urinalysis testing which indicated drug usage, and interviews with station management and union representatives, the employee was terminated.
This concern is considered closed.
9.
Management Meetings (30702)
-On September 25, 1986, R. F. Warnick, Chief, Reactor Projects Branch, Mr. W. L. Forney, Chief, Reactor Projects Section 1A, and the NRC resident inspector staff met with licensee management and supervisory personnel denoted in Paragraph 1 of this report. This meeting was held to assess the facility status and readiness for issuance of the operating license.
10. Operational Readiness Inspection (30702)
On October 9,1986, Messrs. T. M. Novak, Acting Director, Division of PWR Licensing - A, NRR, C. E. Norelius, Director, Division of Reactor Projects, RIII, H. J. Miller, Deputy Director, Quality Assurance and Vendor Programs, IE, C. E. Rossi, Assistant Director, Division of PWR Licensing - A, NRR, S. A. Varga, Director, Projects Directorate #3, NRR, R. F. Warnick, Chief, Reactor Projects - Branch 1, RIII, Mark Ring, Chief, "est Programs Section, RIII, W. H. Swenson, Reactor Systems Engineer, NRR, D. E. Hickman, Training / Assessment Specialist, NRR, and the NRC resident inspector staff met publicly with licensee management
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and supervisory personnel denoted in Paragraph 1 of this report. This meeting was held to discuss the operational readiness of Byron Station for the licensing of Unit 2.
Following the meeting the participants toured the Unit 2 containment, auxiliary, and turbine buildings and the Unit 2 portion of the control room.
11. Open Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action on the part of the NRC or licensee or both. An open item disclosed during the inspection is discussed in Paragraph 6.
12.
Exit Interview (30703)
The inspectors met with licensee representatives denoted in Paragraph 1 at the conclusion of the inspection on October 15, 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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