IR 05000267/1978016

IE Inspec Rept 50-267/78-16 on 781016-20 During Which No Items of Noncompliance Were Noted.Major Areas Inspected Incl:Organization & Admin,Qa Prog review,70% Pwr Level Data Review & Review of Startup Rept & Ler'S

ML20062F461
Person / Time
Site:	Fort Saint Vrain
Issue date:	11/03/1978
From:	Cupp E, Dickerson M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20062F459	List: IR 05000267/1978016 ML20062F456
References
50-267-78-16, NUDOCS 7812180331
Download: ML20062F461 (7)

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U. S. NUCLEAR REGULATORY COMMISSIO*1 0FFICE OF INSPECTION AND ENFORCEMENT '

REGION IV

IE Inspection Report No. 50-267/78-16 Docket No. 50-267 License No. DPR-34

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Licensee: Public Service Company of Colorado P. O. Box 840 Denver, Colorado 80201 Facility Name: Fort St. Vrain Nuclear Generating Station Inspection At: Fort St. Vrain Site, Platteville, Colorado Inspection Conducted: October 16-20, 1978 Inspectors: b //khcLn~!

'M. W. Dickerson, Reactor Inspector ll/3l7Y Date lmb

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E. A. Cupp, Readtdr Inspector (Training)

//7!78 Date'

Approved By: yd ///5/#

'G. -L. Madsen, Chief, Reactor Operations and Date Nuclear Support Branch Inspection Sumary Inspection on October 16-20, 1978 (Report No. 50-267/78-16)

Areas Inspected: Routine, unannounced inspection of organization and admin-istration; QA program review; 70% power level data review; review of startup report; review of event reports; review of IE Bulletins and Circulars; follow-up on inspector identified and unresolved problems; and followup on items of noncompliance. The inspection involved seventy-two (72) inspector-hours on-site by two (2) NRC inspector Results: No items of noncompliance or deviations were identifie o33\

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'-2-DETAILS Persons Contacted Public Service Company of Colorado M. Block, Senior Results Engineer

• L. Brey, Manager, QA W. Franek, Results Supervisor

• J. Gamm, Supervisor, Technical Services E. Hill, Supervisor Operations

• W. Hillyard, Superintendent, Administrative Services F. Mathie, Superintendent, Operations M. McBride, Engineering Coordinator H. Revie, Senior QA Technician J. Solakiewicz, QA Engineer

• C, Tracy, Superintendent, Operations QA R. Wadas, Training Supervisor

• D. Warembourg, Nuclear Production Manager

• Attended exit intervie . Licensee Action on Previous Inspection Findings (Closed)InspectorFollowupItem(50-267/77-10): Surveillance of pipe

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supports and restraints. The licensee approved surveillance procedure SR 5.3.8.d-R, " Hydraulic Snubber Functional Test," on October 16, 1978 for inspection and testing of snubber (Closed) Unresolved Item (50-267/78-06): Completion of CN 763. Change Notice 763 has been completed and functional testing of Level Switches 21385, 21440, 21441, 21451, 21452, 21453, 2498-1, 2498-2, 21499-1, 21499-2, 21500-1, 21500-2, 21501-1 and 21501-2 has been complete (Closed) Item of Noncompliance - Infraction (50-267/78-13): Failure to follow procedure. The adherence to Administrative Procedure ADM-28 has been discussed by the licensee with the Operating Supervisor and Shift l Supervisor who were instructed to ensure that all requirements are being met on a continuing basi . Review of Quality Assurance Program The inspector reviewed the licensee's QA Program for changes that had been made since the previous QA Program inspection and verified that any changes made were in confonnance with 10 CFR 50, Appendix B, and applicable codes, standards and regulatory guides. Any changes were reviewed with personnel having responsibilities for implementing the changes and any procedures which required changes were also reviewe During the inspection, it was noted that temporary changes to the Quality Assurance Inspections (QAI) and Quality Assurance Procedures (QAP) were not being utilized in accordance with the requirements of QAI-1, " Guide-lines for Preparation and Issuance of Quality Assurance Documents and

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-3-General QA Department Administrative Controls." These rcqsivements included bi-monthly review of changes and incorporation within 30 days of any changes determined to be permanent. A change to QAI-l was procured and approved during the inspection which now requires that temporary changes be reviewed quarterly and those changes that are to be made permanent be incorporated into the document No items of noncompliance or deviations were identifie . Organization and Administration The inspector reviewed records, shift schedules, the FSAR, Technical Specifications and conducted discussions with representatives of the licensee to verify that: The licensee's on-site organization structure is as described

in the Technical Specification Personnel qualification levels, authorities and responsibilities are in conformance with applicable codes and the Technical Speci-fication Minimum shift crew composition and licensed personnel require-ments are as required by the Technical Specification , The Plant Operations Review Committee and the Nuclear Facility Safety Committee are as described in the Technical Specification Recent organizational changes have resulted in the preparation of a Technical Specification change which is in the process of being sub-mitted to the NRC. The change will reflect appropriate modifications to the Plant Operations Review Committee (PORC) and the Nuclear Facility Safety Committee (NFSC).

The changes included the elimination of the position of Superintendent of Maintenance and the creation of the positions of Superintendent Administrative Services, Scheduling /QC Supervisor, and Senior Maintenance Superviso The Superintendent Administrative Services will be responsible for the i scheduling QC, training, security, stores and clerical areas. Now included in the Superintendent of Operations responsibilities is main-tenance. This is accomplished through the Senior Maintenance Super-visor who reports directly to the Superintendent of Operation No items of noncompliance or deviations were identifie . .

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-4- Power Acension Program Scope of Inspection The inspector reviewed selected test data associated with the power ascension program. Specifically reviewed was the test data developed at 70% power level (Sequence 61, 65 and 68).

Included in the reviews were the licensee's evaluation of the data, test changes, test deficiencies, data sheets, QA audits, and test data approva Reviewed were sequences 53, 61, 65, 68, the approval for deletion of sequences 47-52, 54-60, 62-64 and 66-67, Technical Review Committee Meeting Minutes for meeting Nos. 18 and 19 and the records for QA surveillance conducted in accordance with QA Surveil-lance Procedures 201, 401 and 601: Inspection Findings (1) QA Surveillance QA surveillance of the test data is being conducted on a 100%

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basis. While some administrative problems are being identified, timely corrective action is generally resolving the problem (2) Steam Generator Performance Steam generator performance at 70% power was considered acceptable. During the sequence 61, 65 and 68 tests (70%

power), an average reheat temperature of approximately 10110F was achieved with an average main steam temperature of 9800F. Maximum measured crossover tube temperature was 8490 (3) Helium Purification System Detectable levels of nitrogen continue to be observed at the discharge of the purification syste (4) Gas Chromatograph Calibration Calibration of gas chromatograph continues to exceed acceptance valves of + 3 (5) PCRV Heat Loads Based upon the PCRV heat load data taken during the test, the heat load experienced in the lower barrel section was 107% of the specified acceptance criteri _ _

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-5-(6) PCRV Liner Cooling

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PCRV liner cooling tubes continue to run at delta T's greater than expected. Tube delta T's of 330F and 420F were experienced in the core support floor and 250F in the bottom barrel. The high PCRV liner cooling tube delta T's will be followed closely during future testin (7) Helium Leak Rate -

The helium leak rate continues to be higher than desire The average helium loss during this test (at 70% power) was 63 pounds per day. Excessive helium flow to the PCRV inter-space was indicated. Subsequent investigation indicated the primary closure seal on C circulator was the prime sourc (8) Huclear Instrument Decalibration Decalibration of the nuclear instrumentation as a function of the control rod positions continue to be observe (9) Temperature Coefficient / Temperature Defect The temperature coefficient was measured during sequence 61 (70% power) over a load change from 69% of rated to 32% of rated and back to 69% of rated. During the temperature decrease,5the measured temperature coefficient at 12350F was-4.0lX10- apfoF compared to the predicted value of-3.93X10-5 ap/oF. During the return of 69% power, the measure-ed temperature coefficient at 12250F was -4.10X10-5 apfor compared to the predicted -3.95X10-5 apfoF. The temperature defect from 800F to 13000F was 0.062 ao vice the predicted value of 0.067 a (10) Iodine Probe Analysis The measured circulating I-133 level was 5.1% of anticipated and circulating I-135 level was 1.4% of anticipated. The results were considered acceptabl (11) Circulater Trip The comparison between the predicted and measured plant per-formance following a circulator trip (sequence 65) was con-sidered acceptabl (12) Loop Shutdown The loop shutdown test was performed during sequence 68 (4/14/78) at which time the main steam temperature control was not in full automatic and the automatic adjustment of the i

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main steam and reheat' steam temperature set pointt with '{

load were not in use. Following the shutcom, the throttle i pressure experienced a transient of 28 psi acd -314 psi. A droop in throttle pressure resulted from latc% up of the feed-water flow controller at e 50% flow. This resulted in loss of throttle pressure control'.~

Data available from a loop shutdown which occurred on June 6, 1978, when the main steam temperature controller was in full

' automatic and with automatic adjustment of the main steam and hot reheat steam temperature set points with load, indicated an initial overshoot in throttle pressure of 86 psi. It then returned to its original value under automatic control of the operating loop feedwater flow. At this time the feedwater flow had dropped to 50% of rated and the feedwater controller latched. This resulted in a 314 psi droop similar to that experienced on April 14, 197 The predicted variation in main steam temperature following a loop shutdown was +180F and -200F. During the test on April 14, 1978 a +460F was experienced while following the loop shutdown on June 6, 1978 a 250F increase followed by a 700F decrease in main steam temperature occurred. Thus, it appears that during a loop shutdown that using the first stage pressure to adjust the reheat steam temperature 5et points results in excessive adjustments. Additionally, the feedwater flow controller latc up results in loss of control of the main steam pressur No items of noncompliance or deviations were identifie . S, tar, tup, Repo,r,ts The inspector reviewed the seventh-Fort St. Vrain Startup Report. The report covering the period February 22, 1978 through May 22, 1978, was

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reviewed in the Region IV offic No items of noncompliance or deviations were identifie . _ Review of BulletinsfCirculars

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The inspector verified by record review, observation and discussion with representatives of the licensee, the action taken in response to IE Circular 78-13, Limitorque Valve Actuator No items of noncompliance or deviations were identifie . Review of Licensee Event Reports The inspector verified that licensee event reporting activities were in accordance with Technical Specification, Section 7, including

identification of details, corrective action, review and evaluation, aspects of operations and accuracy of reporting.

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-7-The following reports.were reviewed by the inspector:

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R0 77-22A R0 78-23 R0 78-03- R0 78-24 R0 78-15 RO 78-26 RO 78-21'

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No items of noncompliance or deviations were identifie ' Exit Interview An exit interview was conducted on October 20, 1978. At the' interview 1 the inspectors discussed the findings indicated in the previous para-graphs. The _ licensee acknowledged these finding .

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