IR 05000313/1987032
| ML20236V520 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 11/12/1987 |
| From: | Jaudon J, Madsen G, Mckernon T, Greg Pick, Tapia J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20236V488 | List: |
| References | |
| 50-313-87-32, 50-368-87-32, GL-82-21, GL-83-28, NUDOCS 8712040350 | |
| Download: ML20236V520 (12) | |
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APPENDIX B
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U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
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NRC Inspection Report: 50-313/87-32 Licenses:- DPR-51-50-368/87-32 NPF-6 Dockets: 50-313~
50-368 Licensee: Ar. kansas: Power & Light Company (AP&L)
P. 0.-Box 551 Little Rock, Arkansas 72203 Facility Name: ArkansasNuclearOne(ANO)
Inspection At:
Russellville, Arkansas Inspection Conducted:
September 14-18, 1987-oa2 An
// 5~87 Inspectors: h.
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Project Enginesr,' Project Date I
ection
, Division of Reactor Projects
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. L/. Mafsen, Preject Engineer, Project Date Sectidn C, Division of Reactor Projects
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T( 0. McKernon, Reactdr/ Inspector, Plant Date /
Systems Section, Engineering Branch f
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G. A. {}cf Reactor Inspector, Operational Date Progrhmt Section, Operations Branch i
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. Approved:
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,j fis f.' Ofudq6, BT,Profect Section A Date vision of Reac Projects kDR DO O
313 G
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l Inspection Summary Inspection Conducted September 14-18, 1987 (Report 50-313/87-32; 50-368/87-32)
Areas Inspected:
Routine, announced inspection of nonroutine event reporting, records retention program, and followup on generic letters and previously identified items.
Results: Within the areas inspected, one violation as identified (failure to include the provisions of an Order for Modification of License and the accompanying Technical Evaluation Report in the development of a procedure, paragraph 8).
The results of the inspection in the area of ILRT will be forwarded under a separate cover..In addition, three open items (paragraphs 4.b, 5, and 6) were identified.
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DETAILS
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Persons Contacted d
- J. Levine, Executive Director of Site Operations
- S. M. Quennoz, General Manager of Plant Operations
- E. C. Ewing, General Manager of Plant Support
- L. "umphrey, General Manager of Nuclear Quality
- R. Lane, Manager of Engineering
- D. Howard, Special Projects Manager J. Vandergrift, Manager of Training
- J. Grisham, Administration Manager
- J. McWilliams, Manager of Maintenance D. B. Lomax, Plant Licensing Supervisor
- P. L. Michalk, Plant Licensing Engineer A. B. Cox, Unit 1 Operations Superintendent R. Tucker, Superintendent of Electrical Maintenance
- H. Greene, QA Superintendent
- D. Graham, QC Engineering Supervisor
- J. Davis, Office Services Supervisor
- S. McGregor, Engineering Services Supervisor
- D. Crabtree, Engineering Services Engineer R. West, Instrumentation & Control (I&C) Supervisor
- D. C. Eichenberger, Maintenance Coordinator 8. Austin, Assistant Office Services Supervisor D. Bennett, Senior Production Plant Engineer S. Copehart, I&C Maintenance Engineer W. Garrison, Operations Technical Support Staff W. Perks, Head Classroom Trainer E. Wentz, Head Classroom Trainer R. C. OMv, Engineering Supervisor, I&C
- Attended exit interview on September 18, 1987.
2.
Followup on Previously Identified Items (Units 1 and 2)
(Closed) Violation 313/8615-01; 368/8615-01:
Licensee couV ot provide documented evidence that the volumetric admixture dispent-utlized at the concrete supplier's facility, was calibrated as presci,oed by National Ready Mixed Concrete Association (NRMCA), nor evidence of inspection requirements performed as stipulated in AP&L Specification APL-C-2401.
When the admixture dispenser calibration could not be documented, an inspection and calibration program was conducted.
This inspection by AP&L determined that the admixture dispensers were within the required tolerance of the NRMCA checklist and calibrated by a consulting engineering firm.
This provided for compliance with the NRMCA certification requirements.
The concrete supplier, although unable to provide documentation of calibration, indicated that the admixture dispensers were regularly checked and maintained by a factory-trained
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1 technician during the period from November 1985 to May 1986.
AP&L conducted field tests on the quality of concrete prior to use.
Actions
~taken' demonstrated that the concrete for the condensate storage tank and tie-ins (foundation,' valve pits, pipe trenches, and missile shield) meet specifications and acceptance requirements.
This item is closed.
(Closed) Violation 368/8626-02:
ANO QC had not performed surveillance and/or inspected Combustion Engineering ISI inspection activities.
AP&L revised the following procedures to define more clearly the role and involvement of QC in ISI and other contractor work activities:
l Procedure No.
Procedure Title Revision No.
Date 1000.12, S5.8 & 6.7.3 Control of Site
12/5/86 Contractors 1009.003, Attachment B, Contract Administration
11/5/86 93.4 Procedure The above two procedures implement inspection report commitments.
1032.07, Step 5.4.23 ISI Program
7/14/86 This procedure implements commitments for ISI Coordinator to notify QA to commence NDE work.
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1000.23 Quality Control Program
10/28/86 The above four procedures verify and reference the information required to close this item.
3.
Records Program This inspection was conducted to ascertain that the licensee is implementing a program for the control of records that is in conformance with regulatory requirements, FSAR commitments, and industry guides and standards.
The NRC inspector reviewed the following licensee procedures:
Number Title Revision Date 1000.17 Records Management
01/08/87
'1013.02 Control of Procedures
06/02/87 1013.03 Correspondence Control
05/12/87 1013.04 Technical Manual Control
01/23/87 1013.05 Indexing and Storage
01/29/87 1013.06 Micrographics
05/06/87 1013.09 Document Retention
01/15/87
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i The review of the above procedures and discussions with licensee personnel indicated that the licensee has established an acceptable records program.
A review of the monthly audit checklist for calendar year 1987 revealed that, in February 1987, the Halon system was inoperable for the storage vault. Maintenance activity was in process. The condition was corrected and future audits revealed no repetition.
The licensee maintains dual storage of documents. The primary storage is located at the ANO site and the secondary facility is located at the AP&L corporate offices in Little Rock, Arkansas. During a tour of the ANO site vaults, it was noted that boxes of radiographs for the Unit 2 reactor
- coolant pumps were in the archives vault awaiting permanent storage.
No violations or deviations were identified in this inspection area.
4.
Nonroutine Reporting Program The NRC inspector reviewed the licensee's system for tracking and reporting.nonroutine occurrences internally and to the NRC. The scope of this review included examining procedures for assigning responsibilities to evaluate the deportability of events and to assure conformance with regulatory requirements.
In addition, a review of procedures was conducted to determine whether vendor bulletins and circulars were l
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reviewed for applicability to the facility.
During this inspection, the NRC inspector reviewed the following docunents:
1000.01, Revision 16, Plant Administration Organization and Responsibilities Procedure 1000.08, Revision 23, NRC Reporting and Communications Procedure 1000.09, Revision 17, Surveillance Test Program Control Procedure A review sampling of the following Reports of Abnormal Conditions (RAC) was conducted.
RAC 2-86-062 RAC 1-84-254 RAC 1-86-265 RAC 2-86-071 RAC 1-85-208 RAC 1-86-399 RAC 1-87-089
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a.
The NRC. inspector found that the licensee had not submitted a licensee event report (LER) relative:to elevated temperatures in the Unit I containment previously reported in NRC' Inspection Report 50-313/87-27; 50-368/87-27 dated September 21, 1987 and to be subsequently' addressed in the special inspection team report (Inspection Report 50-313/87-29).
10 CFR 50.73 requires that the holder of an operating license for a nuclear power plant shall submit an LER for any event described in 10 CFR 50.73 within 30 ' days after discovery.
In particular, 10 CFR 50.73(a)(2)(ii) requires the licensee to submit an LER for a nuclear power _ plant-operating:
(a)inanunanalyzedconditionthat significantly compromised plant safety or (b) in a condition that was outside the design basis of the plant.
The licensee's. failure to report this event'will be addressed'in the findings of the special inspection team report (Inspection Report 50-313/87-29).
b.
The NRC inspector reviewed the above referenced reports of abnormal conditions and, where applicable, the associated LERs to verify that reporting requirements had been met, causes had been. identified, appropriate corrective actions implemented, generic applicability considered, and forms completed. During this review, the NRC inspector found corrective actions to RAC-1-84-254 lacking in documentation and-followup actions, in that notation in the RAC committed the licensee to obtaining an advisory letter from the B&W Owners' Group and issuance of an information letter by the applicable plant supervisor. No corrective followup actions were taken by management to obtain an advisory letter from the B&W Owners' Group, issue an information letter, nor revise procedures for the addition of a more adequate quantitative criteria or inclusion of an explanatory note addressing the anomaly.
The licensee's corrective actions relative to RAC-1-84-254 were.not comprehensively and fully implemented. Documentation committed to by the licensee was not available and filed with the package or had referenced procedures been revised to address the anomaly.
This is an open item pending licensee action to complete documentation as previously committed to and where applicable to revise procedural requirements (313/8732-01).
Within the scope of this inspection, no violations or deviations were found.
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5.
Instrumentation Used to Assess Plant Conditions During and Following an Accident (Implementation of Regulatory Guide 1.97) - lnit 1 The NRC. inspector verified that AP&L had instrumentation systems for assessing plant conditions during and following the course of an accident that met the licensee's commitments to Regulatory Guide (RG) 1.97, Revision 3.
The NRC inspector compared installed plant instrumentation with the commitments contained in an AP&L letter dated June 25, 1984, and the SER concerning that letter dated December 3, 1986.
The NRC inspector utilized the following documents and correspondence related to Unit 1 during the conduct of the inspection:
'" Environmental Qualification Master List," Revision 4, dated April 15, 1987
" Alphabetical Instrument Index List," M-501, Revision 41, dated August 19, 1986
" Environmental Qualification Equipment List," Revision 10, dated August 24, 1987 Environmental Qualification
" Regulatory Guide 1.97 Interface,"
0LC-125-29, dated December 13, 1985
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Submittal, "Nureg 0737 Supplement 1
" Regulatory Guide 1.97," from J. R. Marshall, Manager, Licensing, AP&L, to D. G. Eisenhut, Director, Division of Licensing, NRR, dated June 25, 1984 Safety Evaluation Report, "Conformance to Regulatory Guide 1.97,"
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from J. F. Stolz, Director, PWR PD-6, NRR, to T. G. Campbell, Vice
President, Nuclear Operations, AP&L, dated December 3, 1986 j
j In ten Category 1 systems (as defined in RG 1.97), the NRC inspector verified:
that the environmental qualification (EQ) master list addressed environmental qualification of instrumentation; that the licensee's Q list confirmed seismic qualification and confirmed that the system was covered by the licensee's quality assurance (QA) program; that redundant l
instrument circuits were utilized with the required separation and that
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power supplies were safety-related and redundant; that one indication per
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i division (two per system) and one recording of readout information existed in the control room as determined by visual observation of the instrumentation monitors; that the range of the instrumentation, including overlapping for extended ranges, met the commitments; that the parameters
of interest were being monitored by a direct sensing device and that all interfaces with unqualified portions of the circuit were isolated; and that the testing / calibration frequency met the requirements in RG 1.97.
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is o e.
In seven Category 2 systems, the NRC inspector verified the following items:.that the EQ master list confirmed environmental qualification and that the Q-list confirmed the instrumentation was covered under the QA program; that the instrument signal was displayed in the control room or processed for display on demand; that the range met commitments including l
overlap, if required; that the parameters of interest were directly sensed l
and any interfaces within the instrument circuit between qualified /
unqualified or sensing circuitry / monitoring circuitry had isolation devices; and that the testing / calibration frequency met RG 1.97 requirements.
The ten Category 1 and seven Category 2 systems inspected are listed-below:
Category 1 RCS Hot Leg Water Temperature RCS Pressure Containment Hydrogen Concentration Steam Generator Level Steam Generator Pressure Condensate Storage Tank Level Borated Water Storage Tank Level Neutron Flux Coolant Inventory:
Bottom of Hot Leg to Top of Vessel
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l Containment Sump Water Level, Wide Range Category 2 Degrees of Subcooling Containment Sump Water Level, Narrow Range RHR (Decay Heat) System Flow RHR Heat Exchanger Outlet Temperature Core Flood Tank (Accumulator) Level Bottom to Top Primary System Safety Relief Valve Positions or Flow Through Relief Valve Lines Component Cooling Water Flow to ESF System The documentation utilized, during this inspection, included functional diagrams showing redundancy, electrical separation, related electrical power sources, and isolation interface points.
The diagrams consisted of the below listed Pipin and Instrument (P& ids) and Wiring Block Diagrams (electrical schematics.
P& ids M-204, Sheet 4, Revision 1, " Emergency Feedwater" M-206, Sheet 1, Revision 53, " Steam Generator Secondary System" M-230R, Sheet 1, Revision 5, " Reactor Coolant System" M-232R, Revision 10 " Decay Heat Removal System"
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M-236, Revision 31, " Reactor Building Spray and Core Flooding System" M-261, Sheet 1, Revision 17, " Air Flow Diagram HVAC Reactor Building"
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Wiring Block Diagrams E-186, Revision 4, "DH Removal Pumps Bearing Cooler Solenoid Valves" E-258, Sheet 1, " Emergency Feedwater Initiation and Control" E-258, Sheet 1A, " Emergency Feedwater Initiation and Control" E-258, Sheet 1B, " Emergency Feedwater Initiation and Control" E-260, Sheet 6, " Reactor Nuclear Instrument Source Range Detector" E-260, Sheet 7, " Reactor Nuclear Instrument Source Range Detector" E-265, Sheet 3, Revision 1, " Plant Auxiliary Control-Systems - Decay Heat" E-266, Sheet 2, Revision 0, " Core Flooding" E-281, Revision 6, " Service Water System - SVs" E-282, Revision 3, " Service Water System - MOVs" E-283, Sheet 1, Revision 10, " Service Water System - MOVs" E-284, Revision 9, " Auxiliary Building DH Unit Coolers - MOVs" E-296, Revision ll, " Emergency Feedwater Pumps Condensate and Service Water MOVs" E-331, Sheet 22, Revision 1, " Miscellaneous Instrumentation" E-331, Sheet 40, " Miscellaneous Instrumentation" E-331, Sheet 41, " Miscellaneous Instrumentation" E-414, Sheet 2, Revision 6, "RCS Relief Valve Position Indications" While determining the seismic qualification and determining whether the instrument system components were covered by the licensee's QA program, the NRC inspector found an apparent discrepancy between the definition of the "S" category in Procedure GTEP-217, " Control of Component QA Category Determination" and what the "S" category actually meant.
After discussions with AP&L, the NRC inspector ascertained that changes were already being made to the definition of "S" and "S-List."
A few
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This is an open item (313/8732-02;-368/8732-01)
awaiting final approval of Procedure GTEP-217 with the changes as discussed.
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No violations or deviations were identified.
6.
Actions Taken to Implement Generic Letter 82-21 Natural Circulation Cooldown The NRC inspector reviewed licensee actions taken to implement programs for the control of natural circulation cooldown in accordance with commitments.
Discussions with licensee personnel and a review of the following procedures indicated that procedural coverage meets the
. commitments of Generic Letter No. 81-21 for Units 1 and 2.
Unit 1 Unit 2 Emergency Operating Procedures 1202.01 2202.01
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d Natural Circulation 1203.13 2203.13
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Discussions with licensee training personnel and a review of classroom and simulator training records for three operators at ANO, Units 1 and 2, revealed the following:
Documentation exists indicating that the licensed operators had read and understood the Emergency Operating and Natural Circulation procedures.
Classroom training was conducted on the Emergency Operating and Natural Circulation procedures for Units 1 and 2.
The Unit 2 simulator has been programmed to demonstrate natural circulation cooldown.
The NRC inspector's observations, discussions with personnel, and review of records demonstrated utilization of the Unit 2 simulator for training purposes.
The Unit 1 simulator has been programmed to demonstrate attainment of l
natural circulation and is utilized for Unit 1 training.
However, to l
date, the Unit 1 simulator has not been programmed to demonstrate natural circulation cooldown and the associated opening of vents to prevent void formation in the reactor head.
This is an open item (313/8732-03).
j No violations or deviations were identified.
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Inspection Followup to Generic Letter 83-28, Item 4.1 - Vendor-Related Modifications to Reactor Trip Breakers The NRC inspector performed a post-implementation review of the licensee's records to verify that actions required in Item 4.1 of GL 83-28, Reactor i
Trip System reliability (vendor-related modifications), had been
implemente i
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'During the review, the NRC inspector examined the following licensee's documents:
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1405.18 and 2405.18, Reactor Trip Breaker Inservice Inspection and Maintenance Instructions
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1032.06,. Revision 8, Procurement Technical Assistance Procedure 1000.11, Revision 20, Requisition Preparation and Control Procedure 2304.37, Revision 11, Plant Protection System Channel Test Procedure
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A further sampling of 25 CRD Breaker inservice inspection procedures between 1984 and 1987 was reviewed.
As a result of this inspection, the NRC inspector found that the licensee provided evidence of a comprehensive and complete maintenance program for the modification of reactor trip breakers and a continued surveillance program.
Procedures reviewed were complete and detailed.
In those areas where adjustments to breakers were necessary, the component was corrected and the "as left" condition recorded.
Further, it was noted that a component case history was maintained and updated with correspondence from the B&W Owners Group and the vendor.
When appropriate, updated vendor technical guidance was adequately and expeditiously incorporated into applicable procedures.
No violations or deviations were identified.
8.
Verification of Compliance With Order For Modification of License:
RCS Pressure Isolation (Event V) Valves The Reactor Safety Study, WASH-1400, identified an intersystem loss-of-coolant accident in a PWR which is a significant contributor to risk from core melt accidents (Event V).
The design examined in WASH-1400
. contained two in-series check valves isolating the high pressure primary coolant system (PCS) from the low pressure injection system (LPIS) piping.
By letter dated February 23, 1980, the NRC, in accordance with 10 CFR 50.54(f), requested information concerning the existance of Event V isolation. valve configurations at both Units 1 and 2.
On March 24, 1980, AP&L provided the requasted information and on April.20,1981, the NRC issued an Order For Modification of License for both units.
This order contained Technical Specifications (TSs) with supporting Technical Evaluation Reports which require testing of the reactor coolant system pressure isolation valve leakage.
The monitoring of leakage across pressure interface check valves is accomplished by observing pressure indication in accordance with Attachment I to Unit 1 Procedure No. 1102.01, Revision 38, " Plant Preheat and Precritical Checklist," and Supplement 3 to Unit 2 Procedure 2102.01,
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L Revision 26, " Plant Pre-Heatup and Pre-Critical Checklist." The NRC inspector reviewed these procedures against the Technical Specifications with attached Technical Evaluation Reports.
Paragraph 2.2.2 of the Technical Evaluation-Report states that when leakage tests are made using pressures lower than function maximum pressure differential, the observed leakage shall be adjusted to function maximum pressure differential value.
The reviewed procedures did not include steps which would provide this adjustment. The procedures also included an instrument tolerance which could allow acceptance of a value that would not meet the TS requirements for minimum test differential not less than 150 psi.
The NRC inspector reviewed previous leak test data and found that no leaks have been previously identified.
The procedural concerns identified by the NRC inspector are considered a violation of Units 1 and 2, Technical Specification 6.8.1 for the establishment and implementation of written procedures for surveillance and test activities (313/8732-04; 368/8732-02).
9.
Exit Meeting The NRC inspector conducted an exit meeting on September 18, 1987, with the licensee personnel denoted in paragraph 1.
Both NRC resident inspectors and the Region IV Project Section Chief also attended the meeting.
The scope and findings of the inspection were summarized at this meeting.
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