IR 05000338/1979049
| ML19323E846 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 01/29/1980 |
| From: | Moon B, Upright C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19323E818 | List: |
| References | |
| 50-338-79-49, NUDOCS 8005270316 | |
| Download: ML19323E846 (8) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION o
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f 101 MARIETTA ST., N.W., SulTE 3100
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ATLANTA, GEORGIA 30303
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Report No. 50-338/79-49 Licensee: Virginia Electric and Power Company P. O. Box 26666 Richmond, VA 23261 Facility Name: North Anna Docket No. 50-338 License No. NPF-4 Inspection at North Anna site near Mineral, Virginia Inspector:
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K T. Moon
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/Date $igned Approved by: [
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C. M. Upri
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SUMMARY Inspection on December 17-21, 1979 Areas Inspected This routine, announced inspection involved 32 inspector-hours on site in the areas of calibration of in plant instrumentation required by technical specifications; calibration of mesuring and test equipment used on safety-related systems; and surveillance of in plant equipment required by technical specifications.
Results Of the three areas inspect'ed, four apparent items of noncompliance were found in three areas (Infraction - failure to implement surveillance test paragraph 6.b;
?.nfraction - failure to conduct proper review of surveillance test results -
paragraph 6.c; Deficiency - failure to maintain records for test and measuring equipment paragraph 5; Infraction - failure to follow surveillance test proce-dures paragraph 6.d).
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DETAILS 1.
Persons Contacted Licensee Employees
- W. R. Cartwright, Station Manager
- J. R. Eastwood, Senior Engineering Tech. (Mechanical)
- E. W. Harrell, Superintendent Maintenance
- S. L. Harvey, Supervisor Operations
- J. D. Kellams, Superintendent Operations
- W. R. Madison, NRC Coordinator - VEPCO
- F.
P. Miller, QC Inspector
- J. W. Orgen, Supervisor Administrative Services
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- R.
E. Sidle, Maintenance Coordinator
- E.
R. Smith, Jr.,
Superintendent Technical Services i
- D. E. Thomas, Electrical Supervisor
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Other licensee employees contacted included technicians, operators, mechanics, and office personnel.
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NRC Resident Inspector M. S. Kidd
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- Attended exit interview 2.
Exit Interview i
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The inspection secpe and findings were summarized on 12/21/79 with those persons indicated in paragraph 1 above. The noncompliance items discussed in paragraphs 5 and 6, vare acknowledged by the licensee. With regard to the issue of making sure that safety-related periodic tests are done by properly indoctrinated / qualified individuals, which is discussed in paragraph 6.c.(1), the licensee agreed to give close supervision for such an indivi-dual. With regard to the issue of revising PT-59.3 discussed in paragraph 6.a, the licensee stated that they will revise the procedure. However, they could not provide a commitment date for NRC to followup.
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Licensee Action on Previous Inspection Findings Not inspected.
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Unresolved Items Unresolved items were not identified during this inspection.
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5.
Calibration The inspector conducted a review of the calibration procedures and results as follows:
1-PT-26.2, Axial Power Distribution Monitoring System Calibration 1-PT-30.3.1, NIS Power Range Level Channel Calibration 1-PT-31.2.1, Reactor Coolant System Temperature Instrumentation Cali-bration (T-412) Protection Channel 1 1-PT-32.3.2, Loop Steam Flow and Feedwater Flow Protection Channel III (F-484, F-487) Functional Test 1-PT-32.7.1, Safety-Related Instrumentation Calibra*. ion The inspector monitored on 12/19/79, the performance of portions of the i
calibration conducted in accordance with 1-PT-32.3.1, Loop 1 Steam Flow and Feedwater Flow Protection Channel III (F474) (F479) Functional Test.
The calibration above were reviewed to verify:
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Reviews and approvals were being conducted, and data was recorded and evaluated.
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Test instruments were listed and ident!fied in test results.
Adherence to required technical specification and limiting conditions for operation.
Technical contents of the procedure were correct.
Verification of return of equipment to service and removal of test
equipment was accomplished.
The inspector conducted an inspection of instrument calibration shop and instrument storage areas. The facilities were inspected to verify:
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That standard test instruments were labeled and included in the cali-bration program.
i Instruments were being calibrated on schedule and labeled with calibra-tion status.
Storage facilities were proper for protecting the equipment.
Certification and calibration records were available and being maintained for test equipment.
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The inspector used one or more of the following acceptance criteria for evaluating the above items in the calibration program:
Technical Specifications.
Nuclear Power Station Quality Assurance Manual, Section 11.5.3 - Periodic Test Program.
Nuclear Power Station Quality Assurance Manual, Section 12 - Control of Measuring and Test Equipment.
ANSI N18.7 (1972), Administrative Controls for Nuclear Power Plants.
ANSI N45.2 (1971), Quality Assurance Program Requirements for Nuclear Power Plants.
Topical Report Quality Assurance Program, Operations Phase, VEP-1-3A, Section 17.2.12 - Control of Measuring and Test Equipment.
ANSI N18.1 (1971), Selection and Training of Nuclear Power Plant Personnel
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Industry Practice Within the areas inspected., one apparent item of noncompliance was identified.
During the inspection of measuring and test equipment storage area in electrical shop, the inspector found that four torque wrenches (NQC561, 562, 585 and 586)
were past due for a scheduled calibration. The devices were calibrated last on 4/18/79, 2/1/79, 4/9/79, and 4/3/79, respectively, according to the cali-bration labels attached to the devices. Although the required recalibration
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interval for these devices was three months, it was established during discussion with the licensee that the devices had been calibrated prior to use every time.
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The inspector, however, found that the licensee failed to have proper handling of associated documents, in that no usage information had been established and maintained for these torque wrenches. Also, complete history records of cali-brations had not been established for torque wrenches used by the e?ectrical shop.
10 CFR 50, Appendix B, Criterion V, requires activities affecting quality be accomplished in accordance with written instructions. Section 17.2.12 of Topical Report Quality Assurance Program (VEP-1-3A) as implemented by the Nuclear Power Station Quality Assurance Manual (Sections 12.4.3 and 12.4.4) requires that the cognizant supervisor and the supervisor of the individual performing the calibration are responsible for establishing and maintaining usage information and maintenance history records for each device in the measuring and test equipment calibration program. The inspector found the licensee failed to control the torque wrenches in accordance with their quality assurance procedures in that individual log sheets assigned i
to each device were not established and appropriate calibration test data and maintenance history were not recorded.
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The licensee was informed that the failure to control the torque wrenches in accordance with their procedures was an item of noncompliance (338/79-49-01). The inspector verified that the licensee transferred the torque wrenches to the VEPC0 maintenance shop as of December 18, 1979 and the devices were to be maintained and calibrated by the maintenance shop.
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Surveillance Test Procedure and Results Verification The inspector conducted a review of the following surveillance procedures and results:
1-PT-11, Core Reactivity Balance 1-PT-15.1, Boric Acid Transfer Pump (1-CH-P-2A) Test 1-PT-20.2, Axial Flux Difference (Every 31 EFPD)
1-PT-23, Quadrant Power Tilt Ratio d
lept-24, Calorimetric I-PT-31.1.1, Delta T/Tavg Protection Channel I (T-412) Functional Test
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1-PT-31.3.1, Reactor Coolant Flow Channel I, Loop 1 (F-414)
1-PT-31.5.1, Pressurizer Pressure Channel I (P-455) Functional Test 1-PT-31.1.3, Containment Pressure Channel III (P-LM-100C) Functional Test 1-PT-36.3, Reactor Protection System and Engineered Safety Features Response Tine Testing - Senors 1-PT-44.1, Post-Accident Monitoring Instrumentation - Channel Check 1-PT-57.2, Valve Inservice Test (SI)
1-PT-71.2, Auxiliary Feedwater Pump (1-FW-P-3A) and Valves Test 1-PT-75.1, Service Water System - Valves 1-PT-75.2B, Service Water Pump (1-5W-P-1B) Test 1-PT-82A,1H Diesel Generator Test 1-PT-100.3, Fire Suppression Water System - Valve Position Verification The inspector monitored, on 12/18/79, the performance of portions of surveil-lance testing conducted in accordance with PT-59.3. Heat Tracing.
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-5-The test procedures and test results were reviewed to verify:
That test prerequisites and plant conditions were specified and reviewed, and approvals had been performed; That contents of the procedures were correct and the tests were performed on schedule; That test instruments were listed in the procedures and instruments used were identified in the results; That test results were recorded and compared with acceptance criteria; and i
The return of the system / equipment to service and removal of test equipment was verified upon completion of the test.
The inspector conducted a review of training records for two individuals who performed the PT's to verify qualification requirements are met.
The inspector used one or more of the following acceptance criteria for the above items:
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ANSI N18.7 (1972), Administrative Controls for Nuclear Power Plants.
ANSI N45.2 (1971), Quality Assurance Program Requirements for Nuclear Power Plants.
ANSI N18.1 (1971), Selection and Training of Nuclear Power Plant Personnel.
Technical Specifications, 4.0.2.a, 4.7.14.1.1.c and 6.8.1.c.
Regulatory Guides,1.33.
Nuclear Power Station Quality Assurance Manual, Section 11.5.3 -
Periodic Test Program Industry Practice.
Within the areas. inspected, three apparent items of noncompliance and one inspector followup item was identified.
During the monitoring of PT-59.3, the inspector noted that an acceptance a.
criteria was not clearly defined in the procedure and references used for the test were not specified in the procedure as follows:
(1) Ampere readings of heat tracing circuits were recorded in data sheets when the readings exceeded certain reference values. The j
reference values were not specified in the procedure, but written on the test circuit identification plates attached to the test panel _ _ _ - - - _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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r (2) Ampere readings of redundant channels were compared to ensure that no substantial differences existed in between them. In the procedure, a criteria for " substantial difference" was not defined.
The inspector stated that an appropriate acceptance criteria should be developed in comparing the " Normal" and " Redundant" circuit ampere readings and also the reference values should be, if e.pproved, incor-porated in the procedure.
The licensee agreed to revise the procedure. This item will be followed in future inspections (338/79-49-02).
b.
During the review of PT's completed, the inspector found PT-100.3 (Fire Suppression Water System - Valve Position Verification) was not performed at its required frequency (monthly) between 8/12/79 and 10/13/79.
Technical Specifications, Sections 4.0.2.a an.1 4.7.14.1.1.c, require that the suppression water system shall be demonstrated operable at
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least.nce per 31 days with a maximum allowable extension not to exceed 25% of the surveillance interval by verifying that each valve in the flow path is in its correct position. The inspector found that no surveillance test was performed for PT-100.3 for about 61 days between 8/12/79 and 10/13/79.
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The licensee was informed that the failure to implement the surveillance test in accordance with the Technical Specifications was an item of noncompliance (338/79-49-03).
c.
During the review of PT's completed, the inspector found five PT's contain various errors / mistakes made by performers as follows:
(1)
PT-11, Core Reactivity Balance, performed on 2/22/79 and PT-20.2, Axial Flux Difference (AFD), completed on 3/1/79 contained calcu-lational/ recording errors in three different places where utili-zations of invalid engineering data were made resulting in not only an incorrect reactivity calculation but also an invalid computer AFD target value. These calculational mistakes were not identified during
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the licensee's review process.
The inspector found that both PT's were performed by the same performer who was under indoctrination. The inspector stated that safety-related PT's must be performed by properly indoctrinated /
qualified individual and close supervision must be given to those individuals undergoing indoctrination / training. The licensee agreed to give special supervision for those individuals so that errors can be corrected during their performance. This item will be followed in future NRC inspections.
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(2)
PT-31.1, Delta T/Tavg Protection Channel 1 (T-412) Functional Test, performed on 6/14/79 contained a miscalculated comparator trip setpoint error (TC-412C-2). This error was not identified during the review process.
(3)
PT-11, Core Reactivity Balance, performed on 5/29/79 was not expediously submitted for a review until 6/18/79 which was about 47 effective full power days (EFPD) since the last surveillance date (3/27/79). This exceeded the required surveillance interval (31EFPD) and was not identified in the review process.
(4)
PT-24, Calorimetric performed on 9/21/79 contained improper entries of Nuclear Instrumentation calibration data in four places, where original recordings appeared to be altered. Double entries were made for each recording without crossing out the original entry resulting in an illegible status. What appeared
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to be the original entry did not meet the acceptance criteria
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given in the procedure.
Futhermore, the alterations as such
cannot be considered proper without accompanying initials. These improper alterations were not identified in the review process.
Regulatory Guide 1.33 recommends an administrative procedure to
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handle log entries.
The PT's itemized (1) through (4) above had received two levels of
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review (cognizant supervisor and performance engineer) and were filed in the station file room with these inadequacies / errors unidentified.
This is contrary to the requirements stated in 10 CFR 50 and North Anna Nuclear Power Station Quality Assurance Manual (NPSQAM). 10 CFR 50, Appendix B, Criterion V, requires activities affecting quality be accomplished in accordance with written instructions. Topical Report Quality Assurance Program (VEP-1-3A) as implemented by the NPSQAM (Section 11.5.3.5) requires that the cognizant supervisor and perform-
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ance engineer are responsible for reviewing the test, including critique sheet and procedure for completness and acuracy.
Similar examples of calculational errors / inadequate reviews were
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brought to your attention in our inspection report No. 338/78-24 dated 9/23/78.
The failure to conduct reviews in accordance with procedures is considered an item of noncompliance (338/79-49-04).
d.
During the review of PT's completed, the inspector found that certain tests were not followed in accordance with the procedures in conducting the following PT's:
(1) PT-75.2B, Service Water Pump (1-SW-P-1B), performed on 8/24/78 noted on the critique sheet that pump flow was high and in the
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" Alert Range". The step 4.7.a of the procedure requires doubling
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the frequency of testing, if the flow is in the alert range. The licensee failed to double the frequency of test in the following month.
(2) The same PT performed on 5/24/79 did not complete step 4.1 of the procedure.
The step was marked as not-applicable, while the acceptance criteria (step 5.1) in the procedure requires an indication of flow in terms of gallons per minute. Also, step 5.3 of the procedure requires that the valve in step 5.1 shall be declared inoperable immediately if the valve does not meet the acceptance criteria. The reason why step 4.1 was not followed was not noted in the critique sheet. During the discussion with the licensee, it was established that there was an indication of flow through the valve.
Technical Specification 6.8.1.c requires PT Procedures be imple-i mented.
The inspectcr found that as itemized in (1) and (2)
above, the required instructions of PT 75.2B, Service Water Pump (1-SW-P-1B), were not followed in step 4.7.a, which was performed on 8/24/79, and in step 4.1, which was performed on 5/24/79.
Section 11.5.4 of NPSQAM also states in part that if for any reason test instructions are not followed, this must be recorded on the critique sheet.
Section 11.5.3.5 of the manual also
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requires the cognizant supervisor initiate corrective action if problems are encountered in performing a test.
The licensee was informed that this failure to follow procedures is considered an item of noncompliance (338/79-49-05).
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