IR 05000313/1977009
| ML19317H082 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 06/06/1977 |
| From: | Dickerson M, Johnson E, Madsen G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML19317H075 | List: |
| References | |
| 50-313-77-09, 50-313-77-9, NUDOCS 8004140567 | |
| Download: ML19317H082 (8) | |
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U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
REGION IV
Report No. 50-313/77-09 Docket No. 50-313 License No. DPR-51 Licensee:
Arkansas Power & Light Company Sixth & Pine Streets Pine Bluff, Arkansas 71601
Facility Name: Arkansas Nuclear One, Unit 1 Inspection At: Arkansas Nuclear One Site, Russellville, Arkansas Inspection Conducted: May 16-20, 1977 Inspectors:
. M). OM dhr/77 M. W. Dickerson, Reactor Inspector
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c/A E. H. Johnson, Reactor Inspector Date
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&bYO A. B. Rosenberg, Reac,te? Inspector, Date Engineering Support Section
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wW Approved By:
'G. L. Madsen, Chief, Reactor Operations and Date Nuclear Support Branch Inspection Summary Inspection on May 16-20,1977 (Report No. 50-313/77-09)
Areas Inspected:
Routine, unannounced inspection of organization and administration training; requalification training; core power distribution limits; startup testing-new or modified system; review of plant operations; follow up on inspector identifie and unresolved problems; follow up on items of noncompliance; and tour of plant areas.
The inspection involved 72 inspector-hours on site by three NRC inspectors.
Results: Of the eight areas inspected, no items of noncompliance or deviations were found in six areas; two apparent items of noncompliance were found, each in different areas of the inspection (infraction 1 - failure to follow a precedure
. which required complete data recording - paragraph 5; infraction 2 - failure to maintain current a procedure which is necessary for perio'dic calibration of Incore Detectors - paragraph 6)
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-2-DETAILS 1.
Persons Contacted
- J. W. Anderson, Jr., Plant Superintendent
- G. H. Miller, Assistant Plant Superintendent
- B. A. Terwilliger, Supervisor Plant Operations
- P. Jones, Instrumentation & Controls Supervisor
- R. L. Turner, Assistant Instrumentation & Controls Supervisor R. T. Elder, Assistant Instrumentation & Controls Supervisor
- S. J. McWilliams, P1anning & Scheduling Coordinator D. A. Reuter, Manager Licensing
- L. Alexander, Quality Control Engineer
- S. S. Strasner, Quality Control Inspector M. Leister, Instrument Technician B. T. Moon, Shift Supervisor G. M. Dupriest, Shift Supervisor
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D. Trimble, Training Coordinator T. Green, Assistant Training Coordinator
- P. Rogers, Reactor Engineer D. Bullington, Reactor Technician J. A. Albers, Plant Operator O
B. L. Garrison, Assistant Plant Operator
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G. D. Mitchell, Maintenance Helper
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G. G. Young, Production Engineer W. R. Pool, Radio Chemistry L. Long, Assistant Plant Operator W. C. Childres, Plant Helper M. W. Rogers, Instrument Technician P. D. Hawk, Clerk Typist
R. Lewis, Clerk
- S. A. Alleman, Assistant Station Superintendent, Waterford-3 (LP&L)
E. B. Hyatt, Louisiana Power & Light Company
- Attended exit interview.
2.
Licensee Action on Previous Inspection Findings
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(0 pen) Noncompliance (76-15) Inadequate drawing control. The licensee has now indicated that a revised drawing control procedure will be completed and approved for use June 20, 1977.
(0 pen) Unresolved Item (76-17) Procedure conflict between Quality Assurance Procedure 1004.01 and Nuclear Services Procedure NSP-5.
The licensee indicated that QC Procedure 1004.01 will be completed and approved for use August 15, 1977.
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-3-(0 pen) Unresolved Item (77-02) Procedural requirements not clear.
Maintenance procedure 1401.8, "CROM Removal and Installation," will be issued and approved for use May 31, 1977.
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(Closed) Unresolved Item (77-01) The Local Leak Rate Test Procedure OP-1304.23 did not meet the requirements of 10 CFR 50, Appendix 0.
Change Notices Nos.1 through 4 to Procedure OP-1304.23, Rev. 2, were reviewed for conformance to Appendix J requirements. Several concerns previously identified have been clarified, including pressure testing between the air lock doors and isolating the air supply prior to start-ing pressure decay tests.
This matter is considered resolved.
(0 pen) Unresolved Item (77-03) The licensee was unsble to provide
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documentation to show the acceptability of leak ratt testing of isolation valves from the reverse direction of accident pressure. The inspector was informed that this matter has been referred to the licensee's Nuclear Projects Section for review and evaluation.
The review by the Nuclear Projects Section was not complete.
(0 pen) Unresolved Item (77-03) The licensee was u'nable to measure the local leakage at Penetration No. 51 (type C test) due to limitations of the test equipment. Review of the records indicated two other Penetrations (Nos. V-1 and V-2) where leakage was greater than the test O
equipment could measure.
The licensee informed the inspector that this matter had been referred to the licensee's Nuclear Projects Section for review and evaluation.
Discussions with the Nuclear Projects Section indicated that various means of quantifying the actual leak rates are
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being considered.
3.
Training
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The inspector reviewed the " Arkansas Nuclear One Training Plan," dated September 23, 1975, training records, and interviewed eight individuals to verify that the training had been conducted and recorded.
L.'iew of the training plan established that it does not contain a description of training for the Quality Control Staff or portions of the technical staff, i.e., the Nuclear Engineer. Moreover, the plan does not appear to provide for training relative to administrative procedures or for systems training for groups other than reactor operators.
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At present, responsibility for administering the training program is not specified in writing other than that the Training Coordinaty has the responsibility for " Development and conduct of all operator training and qualifications." No responsibility is specified with regard to other plant disciplines ~ Discussion with a representative of the applicant indicated that each individual supervisor was responsible for the train-ing of his group. However, this does not appear to be specified in I
writing.
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-4-Each of the for.egoing points were discussed with a representative of the applicant.
It was indicated that a request to augment the training group by three individuals has been submitted to Little Rock and these individuals could be used to provide additional training in the area of administrative controls and systems traini_ng, The representative also indicated that the plan would be augmented to describe the quality control and technical staff tr:ining.
The matters of overall responsibility for plant personnel training, 'a description of the training provided in the areas of quality control and technical staff, and provision for training in the areas of administrative controls and systems (for other than operators) will remain as an open item.
No items of noncompliance or deviations were identified.
4.
Requalification Training The inspector reviewed the licensee's operator requalification training l
program, training records and requalification records.
The requalifi-l cation program reviewed was " Arkansas Nuclear One - Unit 1 Requalification
Program" which was accepted for use by NRC Licensing by letter dated
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March 22, 1974. Also reviewed was a lesson plan consisting of ten l
O lecture groups which included 34 lessons.
Records reviewed for these
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\\,j individuals included required simulation of emergency, and abnormal conditions, supervisory evaluation, yearly examinations and manipulations
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j required by Appendix A to 10 CFR 55.
No items of noncompliance or deviations were identified.
l 5.
Review of Plant Operations During the inspection, the conduct of plant operations was reviewed to determine that facility operations conform to requirements of the facility license and the licensee's administrative procedures.
Included in the review were observations of control room activities, logs, records and a tour of accessible areas.
During a review of the Reactor Protection System Log for April 29, 1977 it was observed that a portion (the entire records for the laet half of the 4:00 p.m. - 12:00 midnight shift) had not been recorded. However, additional discussion with the Supervisor Plant Operation and his subse-quent investigation of the omission established that the data that was missing was from the last half of the 8:00 a.m. - 4:00 p.m. shift and that the subsequent readings taken during the 4:00 p.m. - 12:00 midnight shift had been misplaced by one-half a shift.
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O-5-U The 'icensee was informed that this was an item of noncompliance and was contrary to 10 CFR 50, Appendix B, Criterion V and to Admi' %trative Control Procedure #1005.01, Section 6.1.5.A.1 which stater in part that,
"All log entries, data sheet entries and chart notes must be legible, complete and understandable." Section 6.1.5.A.2 states that, "The Supervisor of Operations is responsible for a review of all logs and records. He will advise Shift Supervisors of any shortcomings of
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operator entries and will initiate measures to eliminate or minimize the shortcomings."
Relative to Section 6.1.5.A.2 the data sheet had been reviewed, however, no apparent effort was made to advise the Shift Supervisor of the short-coming or to initiate measures to eliminate or minimize the shortcoming.
The following shift logs and Operating Records were reviewed by the inspector:
a.
Auxiliary [.og Sheets - 3/1/77 to 5/16/77 b.
ANO-1 Station Log: entries made in this log from 3/10/77 to 5/17/77 c.
Bypass and Jumper Log to 5/17/77 O
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Trouble Reports to 5/17./77
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RCS Leak Rate Determinations - 3/1/77 to 5/17/77 f.
E.S.A.S. Log - 3/1/77 to 5/17/77 g.
NSS and Safeguards Log - 3/1/77 to 5/17/77
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R.P.S. Log - 3/1/77 to 5/17/77 1.
NSS Data Point Log - 3/1/77 to 5/17/77 j.
BOP and Turbine Log - 3/1/77 to 5/17/77 The inspector had no further questions in this area.
6.
Surveillance of Core Power Distribution Limits The inspector reviewed plant procedures, data sheets, operating logs and records and computer printouts associated with the determinations of reactor power distributions.
This review was performed to determine that Technical Specifications requirements for power distributions and
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frequency of calibrations were met.
The licensee's method of handling computer software changes was also reviewed.
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The inspector identified one item of noncompliance in this area.
Plant procedure 1302.03, " Periodic Calibration of Incore Detectors,"
revision 0, dated August 30, 1974, was reviewed.
This procedure is-used to measure the instrument lead leakage from the self-powered
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neutron detectors (incore instrumentation). These measurements are necessary to correct the detector outputs for this leakage current, otherwise the readings would indicate lower than actual power. The leakage correction factors determined in this procedure are named CALFI factors and are used in'the SPND program in the plant computer.
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Step 6.1.18 and 6.2.12 of this procedure contains a formula for manually calculating the detector CALFI. This formula is:
F = 1 + I/V R*10E-9 where I/V = nanoamps to volt leakage as measured, and R = 50,000 ohms, dropping resistor value.
The inspector independently calculated CALFI factors from the leakage measurements taken in January 1977, using the above formula, however, could not duplicate the CALFI factors as determined by the licensee.
The inspector discussed this discrepancy with a licensee representative who indicated that the 50,000 ohm dropping resistor in each detector O
circuit had been replaced with a 5,000 ohm resistor in August 1975 upon
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the recomendation of the NSSS vendor. The CALFI factors determined by the licensee are generated by a computer which automatically
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performs the calculations by the above formula.
This computer has the proper dropping resistor value of 5,000 ohms in the program and the inspector determined that the SPND computer program was using the proper CALFI factors. The manual means of reducing the data must be available in case of a computer failure.
I The inspector deterinined that revision 0 to Procedure 1302.03 was the current revision to this procedure.
This procedure was reviewed by the Plant Safety Comittee on September 14, 1976 for correctness as part of the periodic review of plant procedures.
The failure to update this procedure following the plant modification is in noncompliance with Section 5.6.2 of the Quality Assurance Manual and section 6.8.1 of the Technical Specifications which requires that i
written procedures be implemented and maintained to cover the surveillance, calibration and testing activities for equipment such as the incore detectors.
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During review of the power distribution results for a distribution l
calculation of May 4,1977, the inspector nted that the hot channel linear heat generation rate computations by the NOVA computer (an installed on-line backup to the process computer) and the process
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computer differed by an average of eight percent for the regicn III fuel cells for the same core conditions. A similar difference was noted in the calculation of the radial peaking factors for this
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region of the core by each computer. T;,e differences in calculated linear heat generation rate appears to be a reflection of the differences in calculated radial peaking factors. The licensee was aware of this anomaly but has not determined which compu*.er is incorrectly calculating these factors. This item will remain unresolved pending the licensee's evaluation of this item. This item L
is designated 7709-1.
In reviewing the May 4, 1977 power distribution calculatior.s, the inspector noted that the NOVA computer had six failed self-' powered neutron detector (SPND)' inputs (inputs bypassed)'whereas the plant computer printout showed only five failed SPND inputs. Of this field, only two of the failed SPND inputs were comen to both computers. The inspector determined that the licensee has no established mechanism to ensure that the plant computer and its backup (NOVA) have the same faild or bypassed inputs.
This item will remain unresolve1 pending the
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licensea's evaluation of this item.
This item is designated 7709-2.
The inspector determined that the licensee has not established any formal controls over plant computer hardware or software changes to O
ensure such changes are reviewed, approved and properly tested following implementation. This item will remain unresolved pending the licensee's
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evaluation of this area.
This item is designated 7709-3.
The inspector noted two additional items which were discussed with the licensee.
In reviewing the capabilities of the backup computer _(NOVA),
it was determined that this computer lacks the capability of perform-ing the calculation of DNB ratio. A demonstration of greater than a 1.3 nNBR for the core is required in the safety analysis.
The inspector also reviewed a draft of procedure 1302.15, " Core Performance Monitoring and Fuel Management Data Collection." This procedure draft has been under review since December 1976.
It contains the necessary
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controls to ensure the performance of the NOVA and plant computers are compared, which would resolve the unresolved item indicated above. The licensee acknowledged these additional items.
The inspector indicated that these items would remain open for further review during a future i
inspection.
7.
Organization and Administration
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The inspector reviewed the facility organization and manning (and1) the conducted interviews with key personnel to determine that:
facility organization is as described in the Technical Specifications;
(2)personnelauthorities,responsibilitiesandqualificationsarein (continued)
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-8-conformance with the requirements of the Technical Specifications; and (3) that shift composition is as required in the Technical Specifications.
No items of noncomplianc.e or deviations were identified in this area.
8.
Startup' Testing'of the'New Plant Computer The licensee has recently installed a new plant computer.. The purpose of this inspection effort was to review the test program for the
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computer to determine that the core physics pr.ograms yielded equivalent results as achieved by the replaced computer.
The specific test
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results were still under review at the corporate engiaeering offices and were not available at the site. The inspector indicated that these results would be reviewed during a future inspection.
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The inspector reviewed the " Integrated System Test Program" for the new plant computer which outlined the specific tests to be perfonned.
The inspector had no questions in this area.
The inspector also reviewed the records of the calibration selected l
inputs to the plant computer following changeout to ensure the new
computer was properly procsssing its inputs. The inspector had no O
l questions in th's area.
No items of noncompiance or deviations were identified in this area.
9.
Unresolved Items Unresolved items are matters about whd h more information is required in order to ascertain whether they are acceptable items, items of non-compliance, or deviations. The following unresolved items were identified during this inspection.
7709-1 Linear Heat Generation Rate Comoutation - paragraph 6 7709-2 Failed or Byoassed Inputs to Computer _- paragraph 6 l
7709-3 Controls for Review, Evaluation and Testing of Computer Hardware / Software Changes - paragraph 6 l
10.
Exit Interview An exit interview was conducted on May 17, 1977 following the completion of that portion of the inspection dealing with the follow up on items
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identified during the locai leak rate test. A second exit interview was
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conducted on May 20, 1977 to discuss the remaining portion of the inspection. At the exit interviews the inspectors discussed the findings indicated in the paragraphs above. The licensee acknowledged these findings.
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