IR 05000312/1976005
| ML19309A630 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 07/16/1976 |
| From: | Crews J, Dodds R, Malmros M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML19309A618 | List: |
| References | |
| 50-312-76-05, 50-312-76-5, NUDOCS 8003310540 | |
| Download: ML19309A630 (9) | |
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NUCLEAR REGULATORY CONMISSIOP '
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.1FFICE OF INSPECTION AND ENFORCEMEh.
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4II Inspection Report No. 50-312/76-05 Licensee Sacramento Municioal Utility Ofstrict Docket No.
50-312 P. O. Box 15830 License No.
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Sacramento, California 958i3 Priority
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- Facility Rancho seen category C
Location Clay Station, California
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Type of Facility PWR. B&W, 913 MWe (2772 MWt)
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Type of Inspection Routine. Announced and Unannnunced
'l Dates of Inspection Jure 8-10 and Juiv 6-8. 1976
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Datas of Previous Inspection March 29-Aoril 1,1976
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Principal Inspector,
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/Dach R. T. Dodds, Reactor Inspector
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M M.H.Malm'ro's,ReactorIn(pector
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Date Other* Accompanying Personnel:
M. H. Strayer, Reactor Inspector
Trainee
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d Eaviewed by i
J. L. Cr
, Chief, Rea'ctor Operations and Nuclear 7D***
Support Branch
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SUM 4ARY.
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Enforcement Action
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None.
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Status of Previously Identified Enforcement Action
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A.
It was verified by examination of Management Safety Review Consnittee records that the corrective measures and consnit.wnts pertaining to certain review and reporting functicas had been effected as-stated in the r!censee s letters of April 22 and May 3,1976.
(Item Closed)
8.
Procedu c have been prepared and the required annual visual examination of accessible interior and exterior
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surfaces of the containment structure completed on June 9, 1976 in conformance with the licensee's letter of April 22, 1976.
(ItemClosed)
Nonroutine Events A.
The circumstances and corrective actions described in O
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Licensee Event Report Nos. 76-03 through 76-08 were confinned.
(See Paragraph 2 of Details)
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Design Changes
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A.
The surveillance specimens and holder tube; have been removed pursuant to the licensee's application dated April 8,1976 to revise the technical specifications and b.e exempt from the
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requirements of 10 CFR 50, Appendix H, for the remainder of fuel cycle no. 1.
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Other Significant Findings Current Findings A.
The reactor was shut down on April 4 to remove the reactor vessel test specimens and holders.
Examination of the turbine-generator during the outage disclosed that the stator wedges in the generator had come loose and that a number of the stator coils were faulted. SMUD has elected to replace all 94 stator coils. This work was scheduled for completion by August 21 with startup predicted for September 1976.
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It was verified that the licensee had indeed reviewed and detennined that IE Bulletin Nos. 76-02, 76-03 and 76-05 were not applicable to Rancho Seco operations.
(Items Closed),
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C.
The program for calibration of test instrumentation did not require an evaluation be made and documented concerning the validity of previous test results and the acceptability of devices previously tested from the time of previous calibra-tion when calibrations or testing devices were found out of calibration.
(See Paragraph 8 of Details)
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Status of Previously Identified Unresolved Items and other Items of Interest A.
The flip chart drawings in the reactor control room have been labeled to indicate the potential existence of outstanding design changes.
(ItemClosed)
B.
A position was being developed requiring the reactor operator to promptly adjust the reactor power level to a maximum of 2772 megawatts thermal should the hourly heat balance data from
the computer during steady state operations indicate this value was exceeded (IE Inspection Repcrt No. 50-311/76-04).
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Management Interview The results of the inspection were discussed with Messrs. Mattimoe, '
Rodriguez, Oubre and other members of SMUD management as appropriate after each part of the inspection. The following items were i
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' specifically addressed and connitments made by licensee representa ~
tives as indicated:
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A.
The program for the calibration of test equipment will be
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examined and upgraded to include the eva7uation of items previously tested should test equipment be found sybstantially out of calibration.
(SeeParagraph8ofDetails)
B.
The Quality Assurance Program will be modified to include the
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assignment of specific responsibility for review of the QA procedures.
(See Paragraph 7 of Details)
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C.
A test of the manual scram feature of the safety control system will be included as a regular startup test.
(See Paragraph 11.c of Details)
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The surveillance log sheets and accompanying procedures will be examined and modified, if appropriate, to include acceptance criteria.
(See Paragraph ll.a.(5) of Details)
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DETAILS 1.
Persois Contacted J. Mattimoe, Assistant General Manager and Chief Engineer _
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D. Rausch, Manager, Generation Engineering l
R. Connoley, Manager, Engineering Department A. Schwieger, QA Director
J. Jewett, QA Engineer R. Rodriguez, Manager, Nuclear Operations R. Columbo, Technical Assistant i
P. Oubre Assistant Superintendent, Plant Operations J. McCulligan, Assistant Superintendent, Technical Support i
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F. Eisenhutt, Nuclear Engineering Assistant
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A. Locy, Electrical Maintenance Foreman O. Cass, Mechanical Maintenance Foreman i
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H. Heckert, Surveillance Scheduler
J. Wheeler, Electrical Engineer J. Nymen, Nuclear Engineer W. Garrett, Mechanical Engineer R. Low, Instrumtnt and Control Engineer
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M. Carter, Shift Supervisor J. Elliot, Instrument Control Foreman A. Frazer, Reactor Operator
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J. Wilkinson, Reactor Operator L. Stevenson, Supervising Instrument and Control Engineer
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J. Man, Shift Supervisor D. Anderson, Maintenance Inspector - Foreman C. Long, Maintenance Inspector - Sub-Foreman
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D. Yount, Electrical Maintenance Foreman D. Abbot, Nuclear Engineer
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R. Byers, Senior Control Operator J. Sullivan, Quality Assurance Engineer
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R. Schimp, Scheduling Clerk D. Whitney, Nuclear Engineer J. Field, Mechanical Engineer i
2.
Nonroutine Event Reports
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Licensee Event Report Nos. 76-03 (leaktge in weld of bypass line to suction isolation valve of the prima:y makeup pump); 76-04 (reactor coolant pressure sensors out of calibration); 76-05 (high i
pressure injection pump trip).; 76-06 (failure to recalibrate power
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range channels when indicated power deviated from thermal heat l
balance by 2%); 76-07 (operation of crane over fuel transfer
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canal without an approved procedure); and 76-08 (operation of crane over open reactor vessel) were submitted by the licensee since the previous inspection. Based on an examina-
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tion of selected facility records and discussions with licensee representatives, the following were verified.
a.
The cause of each event was identified and the
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details reported to facility management.
b.
Limiting safety system settings for the pressure sensors were verified to be out of tolerance as described in Licensee Event Report No. 76-04.
c.
Limiting conditions for operations were confimed to have been exceeded as stated in Licensee Event Reports Nos. 76-06, 76-07 and 76-08.
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d.
The corrective actions described in the reports to prevent recurrence of the events were being implemented by the licensee.
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The events had been reviewed by the Plant Review Connittee and were scheduled for review by the Management Safety Review Conmittee.
3.
0A Program
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Changes and revisions to the QA procedures were examined and verified to conform to the program as described in Appendix I of the FSAR.
Implementing procedures such as the " Engineering
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Configuration Manual" were found to have been revised to reflect new requirements. Significant changes have been
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implemented, as appropriate, by those persons responsible.
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While the QA procedures do not specifically address the subject of periodic review of the QA manual, this was being accomplished by management's independent auditor as part of his annual review of the QA program for confomance to applicable Appendix B and ANSI criteria.
4.
Design, Design Changes and Modifications Design, design changes and modifications are controlled by the procedural requirements of SMUD's quality assurance program.
The records that document the review, approval, installation and testing of the three design changes and modifications to safety related systems or components reported in the annual report were examined, with the following results:
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Design changes were reviewed by the Plant Review Committee and the Management Safety Review Committee in accordance with the requirements of the technical specifications.
The reviews addressed the criteria described in 10 CFR 50.59.
b.
Design change documentation contained 'the results of -
engineering reviews for the specification of applicable codes, standards and acceptance test methods which define the acceptance criteria.
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c.
The results of the tests performed for acceptance cf the design changes and modifications had been reviewed and approved in accordance with the QA program requirements.
Test records indicated that the test results of completed modifications met the acceptance criteria.
d.
Procedure revisions were made and approved when the completed modifications changed the operating methods or characteristics of the affected system or component.
e.
Drawing change notices or vendor drawing change notices had been issued to reflect design changes on the O,
applicable as-built drawings. However, the examination of working copies of drawings revealed that Drawing No.
23.01-59 SH50A, on file in the I&C shop, had not been reyised in accordance with the drawing change notice.
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This. was subsequently accomplished, during the course of
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the inspection, after a check of the as-built wiring.
a -5.
Procedures Procedures were examined for ccmpliance to the requirements of the technical specifications and conformance to the guidance in
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Procedures examined consisted of safety
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related procedures for system operations (7), abnormal alarm condition procedures associated with each of these systems (7),
i em'ergency conditions (4), maintenance procedures associated with the systems (7), and administrative procedures (2). The
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results of the examination were as follows:
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a.
The review and approval of procedure revisions were in accordance with the technical specifications.
b.
The procedures included appropriate technical specifica-l tion limitations and, as applicable, required the use of
checklists with provisions for signatures and dates.
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.c.
The procedures have been updated to include revisions to the technical specifications (Nos 1 through 3).
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Procedures were being revised as applicable to 1-incorporate Revision No. 4 to the technical specifica-tions.
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i d.
Temporary procedure changes did not conflict with the technical specification requirements.
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Cleanliness
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The areas of housekeeping and cleanliness were examined by (1) a review of procedures, (2) discussions with maintenance
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and operations personnel, and (3) inspection during tours of
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the facility. The practices employed by the licensee were found to be consistent with the guidance provided in Draft No.
5 of Revision 1 of ANSI N18.7-1972, September 11, 1975.
7.
Defueling and Refueling Activities Activities related to the defueling and refueling of the reactor core necessitated by the removal of vessel specimen holder i O tubes were examined. Specific procedures had been prepared and were approved for these activities.
It was verified that surveillance testing had been completed for pre-fueling
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activities in accordance with Section 3.8 and Table 4.1-2 of the technical specifications. The review of records and direct
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observation of (1) fuel handling (activities, (2) containment 4) housekeep
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integrity,(3)coremonitoring,
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vis S n and manning of the refueling deck and the control room di(: not disclose any deviations from regulatory requirements or guides.
8.
Type B & C Local Leakage Rate Testing The facility records pertaining to containment building type
B & C local leakage rate testing were examined.
It was verified that test methods and results conformed to technical specifica-i tion requirements, except for the one occasion reported by the
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licensee (Licensee P/ent Report No. 75-12) relating to the
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Reactor Building purge outlet valve. Test data. sheets for ten
~different penetrations were examined. The only deviation noted was that a record was not made to identify the pressure gauges used for pressure decay type tests. When questioned, the licensee stated that the calibration program did not require
an evaluation of the validity of previous test results should
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test instruments be found to be significantly out of calibration.
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9.
Safety Limits, Limiting Safety System Settings, Limiting Conditions for Operations Records of facility operations were examined to verify conformance to the following technical specification requirements. No anomalies were identified.
a.
. Reactor Coolant System - T.S. Nos. 2.1.1,2.2.1,2.22, 3.1.1 and 3.1.2.
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b.
Reactivity and Power Control - T.S. Nos. 3.1.7,3.1.8 and 3.5.1.
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c.
Power Conversion and Auxiliary System - T.S. No. 3.4.1.
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Containment Systems - T.S. No. 3.6.1.
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Emergency Core Cooling Systems - T.S. No. 3.3.1.A.
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Other Engineered Safety. Features - T.S. No. 3.5.3.
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Electrical Systems' - T.S. No. 3.7.1.
10. Maintenance-Refueling Maintenance activities during the present o'.*sge were examined and found to have been conducted by qualiffeo personnel in accord-ance with approved procedures. The procedures used had been
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properly reviewed and approved and included applicable inspection and testing requirements prior to the return of the component to an operational status. Specific maintenance activities examined
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included reactor vessel head removal and replacement, let down filter replacement, vent valve ark on the "B" high pressure injection pump, and the removal of the SFAS isolation signal to
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11. Plant Operations a.
Shift Logs and Operating Records
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Operating logs and records were selectively examined for the period April 1 through July 6,1975. Observations from this examination were as follows:
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(1) The log sheets were complete and contained appropriate signatures and initials of personnel responsible for log entries and log review.
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o (2) Abnonnal conditions described in the logs contained sufficient information to conmunicate equipment or system status and results of corrective action taken.
(3) No conflicts with' technical specification requirements l
were identified during the review of "Special Orders" issued by the Assistant Superintendent, Plant Operations.
i (4) The outstanding jumper and bypass pennits examined conformed to the requirements of the technical specifications considering plant status.
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(5) The daily and weekly surveillance log did not always contain acceptance limits (i.e., concentrated boric
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acid water storage tank, daily check of reactor l
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coolant system leakage rate, etc.).
b.
Site Tour Tours of the facility. including controlled access areas in
'the reactor building and auxiliary building, were conducted with the plant in a cold shutdown condition for turbine-generator repairs. No signs of abnormal piping vibrations
or fluid leaks from facility power operation were observed.
Proper position of valves in the nuclear service cooling
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Theclearancetagsexaminedwerefoundtohavebeeninstalled
with the appropriate installation and approval signatures
consistent with the requirements of AP-4, " Tagging System."
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Pipe hangers and supports examined appeared to be in i
satisfactory condition.
The control room and refueling deck manning was in confor-mance with the facilit.y technical specifications.
c.
Outage Plans
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It was verified that plans existed to test and check for operation those systems and components disturbed and/or
tested during the outage.
It was determined, however, that the manual scram has not been tested per se as part of the routine startup. test or surveillance programs. Licensee representatives stated that this circuit would in the future be tested during each regular startup test.
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