IR 05000029/1981014
| ML20011B019 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 10/08/1981 |
| From: | Bettenhausen L, Chung J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20011B015 | List: |
| References | |
| 50-029-81-14, 50-29-81-14, NUDOCS 8111030537 | |
| Download: ML20011B019 (14) | |
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e U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND_ENFORCEMELT Region I
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Report No. 50-29,yl-14 Docket No. 50-29 License No. DPR-3 Priority Category C
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Licensee:
Yankee Atomic Electric Company
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Star Rouce Rowe, Massachusetts 01367 Facility Name:
Yankee Rowe s
Inspection At:
Rowe, Massachusetts i
Inspection Conducted:
August 31-September 4, 1981 Inspectors:
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3fn W. Churfg, Reactor Inspector date signed
date signed
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Approved By:
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Lee H. Bettenhausen, Chief, Test Program date signed i
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Section, EIB, DETI s
Inspection Summary:
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Inspection on August 31-September 4, 1981 (Report No. 50-29/81-M)
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Areas Inspected:
Routine, unannounced inspection of' Core XIV post r
refueling startup test report; Core XV post refueling startup testing,
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including pre-critical checks and tests; zero power tests; power escalation i
tests; and control room and plant tours. The inspection involved 31 inspector-hours onsite by one region-based inspector.
Results: Noncompliance - None l
Region I Form 12 (Rev. April 77)
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DETAILS 1.
Persons Contacted Principal Licensee Employees
- H.A. Autio, Plant Superintendent
- G. Babineau, Reactor Engineer /STA
- L.D. French, Technical Assistant T. Henderson, Reactor Engineering Supervisor D. Long, Reactor Engineer W.A. Loomis, I&C Supervisor N. St. Laurent, Assistant Plant Superintendent
- J.L. Staub, Technical Assistant to Plant Superintendent R. Williams, Reactor Engineer /STA The inspector also contacted other licensee employees during the inspection, including Reactor Operators, and other Technical Support personnel.
- denotes those present at the Exit Interview.
2.
Core XIV Startup Summary Report Review The inspector reviewed Core XIV Startup Testing Summary Report of YAEC Yankee Rowe Nuclear Power Generating Station, issued February, 1979.
The test results include:
Rod Drop times,
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Control Rod Group and Rod Reactivity Worths,
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Power Plus Xenon Reactivity Worth,
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Temperature and Power Coefficients,
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Just Critical Boron Concentrations, and
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Parameters associated with the Power Distribution.
The inspector determined that the test results and conclusions were consistent with the predicted values of the performance analysis and were within the limits specified in Technical Specifications.
The inspector has ro further questions.
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3.
Core XV Startup Testing - Precritical Tests a.
Functional & Calibration Test Review The inspector reviewed functional and calibration tests and their results to verify that:
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Procedures were provided with detailed stepwise instructions;
Instruments and calibration equipment used were traceable
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l to the National Bureau of Standards;
"As Found" and "As Lef t" conditions were recorded;
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Acceptance and operability criteria were observed in accordance with the Technical Specifications; Technical content of procedures was sufficient to result
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in satisfactory component calibration and test;
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Work Order was issued and corrective actions were taken if the test was not acceptable.
b.
Specific Test Revinw The following tests were reviewed:
(1) Administrative Controls The inspector reviewed the administrative control documents to verify that the post-refueling sequences and te:ts were conducted in compliance with the station procedures and Technical Specifications (TS), and that the test p,'ogram was consistent with the requirements specified in the ANSI N18.7-1976.
The documents reviewed were:
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AP-0001, Plant Procedures and Instructions, Revision 8;
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AP-0223, Document Control, Revision 3;
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AP-0227, Corrective Action, Revision 1;
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Yankee Rowe Core XIV-XV Refueling Manual, Copy 3, Revision 3, June 16, 1981;
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OP-1700, Shim Rod Inspection, Attachment G, Revision
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Core XV Performance Analysis, Yankee Nuclear Power
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Station, March 1981.
No unaccettable conditions were identified.
(2) Nuclear Instrumentation (NI) and Reactor Protection System (RPS) Precritical Checks The inspector verified that functional tests of NI and
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RPS had been performed in accordance with the test procedure OP-4611, Nuclear Instrumentation and Reactor Protection System Precritical check, Revision 10, and that the test results were consistent with the requirements specified in TS.
The following test data were reviewed:
July 26, 1981: Functional tests of Main Coolant
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Flow, Pressurizer, PORV Bistable and Relays, and Pressurizer High Level Scram Bistable.
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July 27,1981: Source and Intermediate Range Channels, Power Range and Intermediate Power Range Channels, Pressurizer Narrow Range Level Channel, and Reactor Trip Circuits.
No unacceptable conditions were identified.
(3) Fuel Sipping Test An In-core Fuel Sipping Test was conducted on May 22, 1981 through June 3,1981 to identify damaged fuel pins, which had been suspected from the preliminary chemistry sampling of the Reactor Coolant System. The procedure employed was OP-7202, Fuel Sipping, Revision 1 and Advanced Change Notice No. 1.
The change had been reviewed by the
Plant Operations Review Committee (PORC), May 16, 1981.
An excess of radioactivity in C-9 assembly of Core XIV was observed.
Nine damaged fuel pins in the assembly were replaced with the solid Zircaloy rods, and the reconstituted assembly (8574) was reshuffled to a new core location, H-4, during the XIV-XV refueling outage.
This move was to prevent any further damage to the assembly.
Findings The Fuel Sipping test required that the gross chemistry test results of the activity checks be recorded on the summary data sheet OPI-7202.1. This enabled Reactor
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Daeritors to verify the conditions of the fuel assemblies.
spector verified that the chemistry results of nine-ssemblies sipped were not recorded in the data during the fuel shuffling operations of June 3, These assemblies were:
B530 B546 A547 A519 B544 A531 AS49 B518 A523 The inspector independently verified by review of the chemistry test data sheet, DP-9102, and the Iodine activity analysis charts that the 9 assemblies did not show any evidence of damaged fuel pins and that their activity analyses were within the acceptable level.
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A subsequent discussion with a licensee representative revealed that the data recorder on the refueling bridge neglected to record the analysis results.
However, the inspector determined that the basic problem was the review process of the sipping test procedure and data sheet. The procedure did not specify that supervisory personnel review the final results.
Consequently, the incomplete data recording was not identified until this inspection.
A licensee representative acknowleged the procedural inadequacy and stated that a means to review the data package would be established and be incorporated into the procedure by October 1, 1982.
This is an unresolved item pending a NRC:RI review of the revised procedure (50-29/81-14-01).
(4) 50.59 Review of Reconstituted Fuel The damaged C-9 recycled assembly in Core XIV was reconstituted by replacing the 9 fuel pins with 9 solid Zircaloy fuel rods, and was placed in core location H-4 in Core XV.
Furthermore, to prevent damage to new fuel assemblies and to redistribute the power in the region, 24 Stainless Steel inert rods with Zircaloy cladding were placed in 8 fresh fuel assemblies.
These assemblies are positioned at the western edge of the core baffle location to prevent power peaking.
Four A-type fresh assemblies in the Northwest and four B-type in the Southwest core position had two inert rods
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and one inert rod in each assembly respectively.
These assemblies were located in the opposite side of the core from the reconstituted fuel assembly which occupied H-4 of the first quadrant (Northeast) la core XV.
This was done to counter balance power peaking and resulted in a slight power redistribution, but without creating abnormal conditions.
The reconstituted fuel assembly and the fresh assemblies with the inert rods are:
Location in Fuel Number of Inert Rod Core XV Assembly Inert Rods Type H-4 B 57*'
Zircaloy l
A-4 A615
S.S./Zr clad l
B-3 A593
S.S./Zr clad C-2 A619
S.S./Zr clad D-1 A625
S.S./Zr clad A-7 B604
S.S./Zr clad B-8 B622
S.S./Zr clad C-9 B618
S.S./Zr clad D-10 B596
S.S./Zr clad The inspector reviewed the safety evaluation summary, power distribution analysis and other related documents in the Design Modification Package. The documents reviewed were:
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Plant Design Change Report (PDCR) No. 81-5, Repair Fuel Assembly #B574 and install Dummy Pins in New
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Fuel, July 28, 1981.
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Exxon Drawings, Yankee Rowe-Inert Replacement Rod Specifications.
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Evaluation of Changes affecting the Yankee Core XV-Performance Analysis, July 29, 1981.
Yankee Rowe XV Fuel Reconstitution, July 21, 1981.
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Based on the document review and discussions with the cognizant licensee engineers, the inspector determined that there were no unreviewed safety questions regarding to the reconstituted fuel assembly.
No unacceptable conditions were identified.
4.
Post-Critical Tests a.
Test Program Review
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The inspector reviewed selected test programs to verify the following:
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The test programs were implemented in accordance with
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Core XV Refueling and Post-refueling Sequencing Procedures; Step-wise instructions of test procedures were adequately
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provided, including Precautions, Limitatioss and Acceptance Criteria in conformance with the requirements of the Technical Specifications;
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Provisions for recovering from anomolous conditions were provided; Methods and caiculations were clearly specified and the
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tests were pevformed accordingly;
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Review, Approval, and Documentation of the results were in accordance,with the requirements of the Technical Specifications and the licensee's administration controls.
b.
Zero Power Physics Tests The inspector verified by review of the test sequencing procedure, OP-1701, Core XV BOL Zero Power Physics Tes;, Revision 5, that the test sequences prior to power escalation were specified in the procedure and the tests were performed July 27-29, 1981 in accordance with the specified sequences.
(1) Control Rod Exercise The control rods were exercised from 0" to 90" and then from 90" to 0" to verify the rod op'erability and the position indications.
The inspector noted by review of the functional test data that #1 and #10 rods in Group B were either 6 inches ahead of the other rods in the group or 6 inches lagging behind the others. However, the inspector agiced with the licensee's conclusion that the above discrepar.cies
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were within the operability criteria specified in TS 3.1.3.1.
The inspector had no further questions.
(2) Functional Checks of the NI Power Range The functional tests of the Nuclear Instrument Power Range Channels were performed July 26, 1981.
Procedure OP-4643, Revision 3, was used.
No taacceptable conditions were identified.
(3) Critical Boron Concentration Just critical boron concentration was 1852 ppm with All Rods Out (ARO) and an average coolant temperature of 528'F.
The acceptance criterion was 1900 ppm i 10%.
Findings The zero power physics testing requires utilization of the data in the Core XV Performance Analysis Report, Yankee Nuclear Power Station, March, 1981, in which Table 5-1 listed the acceptance value of the predicted just critical boron concentration as 1920 ppm 1 10%.
A subsequent discussion with the cognizant engineer revealed that the predicted boron concentratior in Table 5-1 was incorrect and was not used for the test. The predicted values for the Core XV performance in Table 5-1 were the best estimated results obtained from the estimated end-of-cycle parameters of the Core XIV rather than using the final observed end-of-cycle values of Core XIV. The new updated values were later issued in the Data Reference Manual, dated July 22, 1981, and were used for the physics tests.
A licensee representative acknowledged the confusion as tc which data sets were correct and should have been used, since procedure OP-1701 neither specified the table nor referenced the Data Reference Manual nor the Performance Analysis Report.
The licensee representative stated that it was due to an oversight during the course of procedure writing and procedure OP-1701 would be revised by October 1, 1982, to include the proper references. The irspector determined that the Reactor Engineers used the correct data and references for the tes *
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This item is unresolved pending NRC:RI review of the revised procedure (50-29/S1-14-02).
(4) Main Control (MC) System Flow The total flow rate for the 4 MC loops was determined from the Steam Generator pressure drops using procedure OP-4715, Main Coolant System Total Flow Rate Determination, Revision 2.
Total flow Rate of 4.2 x 10' lb/hr was greater than the required 3.83 x 10' lb/hr. Measured data of July 27, 1981 were reviewed.
The inspector had no further questions in this area.
(5) Control Rod Decp Time The inspector verified by review of the recorder traces and data obtained July 27-28, 1981, that Control Rod (CR)
drop times from the fully withdrawn position to the 6 inch coil entry were all less than 2.5 seconds as required by T.S.
Findings The drop time measurements required that a Visicorder, YAEC No. 2368, be connected to the rod position indicating coils in order to trace the drop times.
The calibration record of the Visicorder OP-6728 for January 5,1981, showed that the calibration data sheet did not require the Visicorder identification (ID) on the calibration sheet. The inspector expressed concern about the possible confusion regarding the proper identification of the instrument calibration.
The licensee acknowledged the inspector's concern and subsequently revised calibration procedure UF-6728 and DP-6703 prior to the exit meeting.
The inspector had no further questions in hi, area.
Test procedure OP-4703 requires that the CR drop time measurements be reviewed by the Plant Reactor Engineer
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and that the engineer entar his signature on the data sheet. Test data performed July 27-28, 1981 did not have any objective evidence that the Plant Reactor Engineer reviewed the results, and his signature was missing.
A licensee representative stated that the results would be reviewed by September 15, 1981 and steps would be taken to prevent future recurrenc r
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The inspector determined that this was an isolated case and this item would be reinspected by an NRC:RI inspector (50-29/81-14-03).
(6) Reactor Criticality Criticality prediction during the startup testing was conducted employing the "1/M curve" method as given in the procedure OP-2103, Reactor Startup and Shutdown, Revision 8.
The reactor achieved criticality at 0320 hours0.0037 days <br />0.0889 hours <br />5.291005e-4 weeks <br />1.2176e-4 months <br />, July 30, 1981, with the control rod group C in 11 2/8".
The predicted position was 16", and the difference of 4 3/8" was within the specified requirement of 1% a,p.
Findings The inspector determined that the operating procedure OP-2103 did not require a record of the critical boron concentration on the data sheet and Reactor Engineers were recording the value separately in their engineering logs.
The licensee stated that the procedure would be revised by October 1, 1982 to record the critical boron concentration on the data sheet.
This is an inspector follow item (50-29/81-14-04).
c.
Power Escalation Tests Power ascension was initiated at a rate of 2%/hr on July 31, 1981, 1948 hour0.0225 days <br />0.541 hours <br />0.00322 weeks <br />7.41214e-4 months <br />; Xenon was stabilized in an equilibrium state after the plant attained 100 MWe. However, due to a failure of an 0-Ring in the #3 Non-Return (NR) Dump Solenoid Valve, the reactor was manually scrammed, and subsequently restarted after replacing the damaged 0-Ring.
The power ascension tests were sequenced in accordance with the procedure OP-1702, Core XV Zero to Full Power Physics Test Procedures, Revision 7.
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(1) Core Thermal Powe, The inspector reviewed test procedure OP-7302, Calorimetric, Revision 6, and verified that-the tests performed August 15, 1981 through August 25, 1981 were conducted in accordance with the test procedure.
The inspector noted that the calorimetric estimation of the core thermal power was calculated from the secondary-
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side enthalpy balances in the Steam Generator (SG) without taking account of the Reactor Coolant Pump (RCP) inputs and the letdowr line losses. A licensee engineer stated that the gain by the RCP was offset by the loss through the letdown lines.
The calculated core powers were evaluatad against the previous Back-Pressure (BP) curve for u.2 operation, and a new BP ce ve (#21) was calibrated as summarized in the following:
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Test Date Curve #
Calorimetric BP Curve Difference 8/25/81
599.4 600.5-1.09997 8/22/81
597.9 596.5 1.40002 8/19/81
589.9 586.3 2.69995 8/17/81
583.9 581 2.3005 8/15/81
579.5 577.1 2.40002 The operating power curves are:
new curve #21: MWe=3.4 + 15 (BP) +30
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old curve #20: MWe=3.4 + 15 (BP) + 26.6
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The inspector had no further questions.
(2) Power Distribution The inspector reviewed 13 traces of the Incore Flux Maps taken during August 2-19, 1981 and determined that the test was conducted in accordance with the procedure OP-
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7105, Normal Operation of the Incore Flux Mapping System, Revision 7.
The inspector determined by review of the data sheets and the computer calculations that the measured reaction rates were all within 4.8% of the predicted values in the high powered assemblies.
The reaction rate is a measure of the Incore detector outputs as compared with the predicted values for the assemblies located near the high power center region.
The measurements of incore traces were performed with
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control rod group C withdrawn over 80".
The acceptance reaction rate criterion requires less than 5% deviation.
Hot channel factors and peak linear heat generation rates (LHGR) were calculated as per procedure OP-4704, Hot Channel Calculation, Revision 9, from data taken on August 19, 1981. The inspector verified that instrument and engineering errors were applied by taking geometrical average of the errors as specified in TS.
The results are:
Limits
_ Fresh Fuel Burned Fuel F
<2.798 2.304 2.248 q
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<1.805 1.592 1.597 H
Max. LHGR
<10.669 10.427 10.572 kw/ft The inspector noted that some entries in the data sheet were not consistent with the procedure, in which the limits for the Hot Channel Factors be corrected for the power level by normalizing the power level with the rated thermal power. However, the entries in the data sheet
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were more conservative than the proccdure limits.
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The inspector had no further questions in this area.
(3) Ejected Rods The ejected rod worths were calculated on July 29, 1981 as summarized in the following table:
Ejected Rod Worth, %A (Rod Insertion)
Measured Acceptance J
Rod 16 (Group C)
0.644 0.63 1 15%
Rod 16 (Groups C and A)
0.839 0.84 1 15%
No unacceptable conditions were identifiec'.
(4) Shutdown Margin The inspector reviewed test procedure, data, and reactivity computer traces, and determined that the shutdown margin was determined in accordance with the procedure OP-4708,.
Determination of Shutdown Margin, Revision 5.
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The measured value was 2324 ppm at 194'F average main coolant temperature. The required value was 1600 ppm or larger.
No unacceptable conditions were identified.
(5) Reactivity Coefficients The inspector determined that the Moderate Temperature Coefficient (MTC) was measured with ARO on July 28, 1981 in accordance with procedure OP-7306, Moderator Temperature Coefficient, Revision 2.
The inspector reviewed the computer traces and the calculations.
A total of six runs were performed by repeating heatup and cooldown. The average of the six runs was obtained.
The following table summarizes reactivity coefficients:
REACTIVITY COEFFICIENT Measured Predicted Difference Acceptance MTC,
-5.02-4.6 0.42
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pcm/'F Sum of Xenon
& Power Defect 3.03 3.5927 0.0373
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power,
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Power Coeff.0-2.1 x 165-1.68 x 10 5-0.42 x 105
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90% power,
%A P/'F The inspector had no further questions.
5.
Contral Room Observation and Plant Tour The inspector observed control room operations for manning, shift turnover, daily logging, and facility operation.
The inspector also discussed the continuous indication of low Safety Injection (SI) tank temperature on the Control Room alarm board with operational staff members. The inspector observed the local temperature indications for the SI tank and verified that the SI tank temperature was above 123'F as indicated on the local temperature indicators and the auxiliary operator's log, and that the Control Room Alarm was caused by an alarm circuit p"-klem, for which the corrective maintenance work was in progress.
The lo temperature alarm setpoint for the SI tank was 122'F.
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The inspector also toured Emergency Diesel Generator Room, and witnessed a portion of the SI pump surveillance testing and in-service inspection. The inspector verified that the surveillance was being performed in accordance with the procedure, and the SI pump vibration amplitudes were within the acceptance limits.
No unacceptable conditions were identified.
6.
Unresolved Items Unresolved items are those items for which further information is
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7.
Exit Interview Licensee management was informed of the purpose and scope of the inspection at the entrance interview.
The findings of the inspection were per'odically discussed and were summarized at the conclusion of the inspection on September 4, 1981. Attendees at the exit interview are denoted in Paragraph 1.
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