IR 05000334/1982009
| ML20053D010 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 04/27/1982 |
| From: | Bettenhausen L, Chung J, Troskoski W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20053D007 | List: |
| References | |
| 50-334-82-09, 50-334-82-9, NUDOCS 8206030270 | |
| Download: ML20053D010 (6) | |
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U. S. NUCLEAR REGULATORY COMMISSION Region I Report No. 50-334/82-09 Docket No. 50-334 License No. DPR-66 Priority
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Category C
Licensee:
Duquesne Light Company P. O. Box 4 Shippingport, Pennsylvania 15077 Facility Name:
Beaver Valley Power Station, Unit No. 1 Inspection At:
Shippingport, ?ennsylvania Inspection Coi cted:
April 5-9, 1982 Inspectors:
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g_h [ w J - 2 n W. Chung, Rea t Inspector ditepgned
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. M.' Troskoski, R de Inspector date ned Approyed y:
[_;2.s2 M Y/I/[8'-
L.11. BettenhausWn, Chief, Test Program Section date signed Inspection Summar_y:
Inspection on April 5-9, 1982 (Report No. 50-334/82-09)
Areas Inspected:
Special, arr.ounced inspection of followup on prior identified items; cycle 2 startup testing; proposed cycle 3 startup testing; and cycle 3 reload evaluation report.
The inspection involved 24 inspector-hours onsite and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> off-site by one region-based and one resident NRC inspector.
Results: Noncompliance:
None 8206030270 820512 PDR ADOCK 05000334
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DETAILS 1.
Persons Contacted On-site M. Coppula, Superintendent, Technical Services
- K. D. Grada, Superintendent, Licensing W. S.-Lacey, Chief Engineer F. J. Lipchick, Senior Compliance Engineer
- J. P. Sieber, Manager, Nuclear Safety and Licensing
- H. P. Williams, Station Superintendent R. T. Zabewski, Technical Supervisor
- G. F. Zepsic, Coordinator, Station Study Projects DLC, Analytical R. Scherer, Core Analysis Engineer
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- L. J. Winker, Core Operations Engineer
- E. Yue, Sr. Core Operations Engineer
Westinghouse, Nuclear Fuel Division M. G. Arlatti, Manager, Fuel Licensing i
K. Forcht, Nuclear Safety
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8. McKenzie, Fuel Projects
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- W. M. Troskoski, Resident Inspector The inspector also interviewed licensee employees during the inspection.
- denotes those present at the exit interview.
- denotes licensee representative during the exit interview.
2.
Previous Inspection Item (0 pen) Unresolved Item (50-334/82-02-01):
t leak rate test procedure be revised, and the RCS leak rate testing performed during cycle 3 startup testing using the revised procedure.
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The inspector verified by review of Procedure OST 1.6.2, Revision 17, that the procedure was revised for adequate performance of the leak rate testing by a technically valid testing method.
However, a conversion factor, from the mass leak rate to the volumetric leak rate, in instructional step 12.a was incorrect. The specific volume used was not the value for the RCS operating temperature, thus failing to reflect the operational RCS leak rate.
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This item remains open pending NRC:RI review of the procedure change and subsequent completion of the test using the revised procedure.
3.
Cycle 2 Startup Testing
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The inspector reviewed selected test programs and the results to verify that:
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Procedures were provided with detailed stepwise instructions, and the technical content was sufficient to result in satisfactory tests;
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Instrument calibration was performed prior to the test, and "As Found" and "As Left" conditions were recorded; Acceptance and operability criteria were observed in accordance with
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the Technical Specifications (TS);
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Methods and calculations were clearly specified and the tests were performed accordingly.
3.1 Approach to Criticality The inspector reviewed Procedure BVT 1.2-2.2.1, Initial Approach to Criticality after Refueling, Revision 0, September 24, 1980, and verified that the Reactivity Computer was calibrated and that the criticality was observed employing the Inverse Count Rate Ratio-(ICRR) method.
The procedure specified that the reactivity computer would be calibrated using an exponential signal generator by comparing the input-to-output
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i response, and be checked against the reactivity calculated from the doubling time.
The inspector noted that the calibration record using the exponential signal generator was not included in the test
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documentation and "As Found" and "As Left" calibration data were not recorded in accordance with the station administrative procedure and ANSI N18.7-1972 requirements. A licensee representative acknowledged this inspection finding and stated that the calibration data would be maintained properly during the cycle 3 startup testing.
The inspector had no further questions in this area.
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3.2 Startup Tests The inspector reviewed procedure BVT 1.2-2.2.2, Core Design Check Test - Cycle 2, Revision 0, September 30, 1980, and verified that the test results, performed November 11 - December 18, 1980, were in conformance with the requirements specified in TS and station procedures.
No unacceptable conditions were identified.
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4.
Cycle 3 Startup Testing No changes or modifications from the cycle 2 startup testing have been proposed for upcoming cycle 3.
The inspector recognized that the licensee performed Control Bank Worth Measurements by both Boron Dilution and Rad Swap methods during the cycle 2 testing.
However, the rod swap method was not_ incorporated into the cycle 3 startup test program at this time.
The inspector had no further questions.
5.
Cycle 3 Reload 50.59 Review The inspector reviewed the Reload Safety Evaluation (RSE) for Beaver Valley Unit 1, Cycle 3, to verify that:
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Proposed changes, modifications, and tests for cycle 3 did not decrease the margin of safety as defined in the Technical Specifications; Proposed changes, modifications, and tests for cycle 3 did not
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involve unreviewed safety questions; Cycle 3 RSE was in conformance with the requirements specified in 10
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CFR 50.59.
The documents reviewed included:
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Technical Specification Change Request No. lA-66, February 23, 1982.
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Technical Review of Reload Safety Analysis Checklist and Reload Safety Evluation (RSE) for Beaver Valley Unit 1 Cycle 3, February 16, 1982.
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Reload Safety Evaluation, Beaver Valley Nuclear Plant, Unit 1 -
Cycle 3, January, 1982, Westinghouse Nuclear Fuel Division.
The inspector also discussed the safety evaluation with the licensee's analytical group and their contractors, Westinghouse Nuclear Fuel Division, to clarify points of the evaluation on the RSE report for Beaver Valley
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Unit 1, cycie 3.
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5.1 Reload Safety Evaluation (RSE)
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The RSE included safety evaluation of the cycle 3 reload for two and three loop operations.
Its objective was to demonstrate that the reload would not adversely affect the safety of the plant nor decrease
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the margin of safety within the safety bounds reported in the FSAR.
The inspector determined by discussions with the licensee representatives and Westinghouse Nuclear Fuel Division personnel that the RSE was not performed using the existing TS conditions, and that several
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assumptions upon which the RSE was based were proposed TS changes submitted to NRC for approval under change request No. IA-66,. February 23, 1382. Furthermore, the safety evaluation explicitly stated that it was assumed in the RSE that the TS changes noted above are adopted and TS are adhered to in operation.
a A licensee representative acknowledged this and stated that the plant would not start up without an approval of the proposed TS
changes by NRC.
5.2 Cycle 3 Core Modifications Fifty-two new fuel assemblies of 3.0 w/o U-235 enrichment would be loaded in the peripheral locations of the core.
These fuels are standard design assemblies with the same mechanical design as the fuel already in the inner core locations. However, the inspector was informed that the grid corners were slightly modified to minimize potential grid-to grid interaction during fuel handling. The grid corners were ground off by 0.0035 lb. each.
Each grid weighs 2.3 lbs., and there are seven grids per assemblies.
Total contribution of the removed Inconel material from the fifty-two assemblies would not be a significant amount.
This change would not contribute more than 1 ppm of boron compensation due to the reactivity increase.
The inspector was further informed that the modified grids were crush-tested, and total core flow change due to the altered flow fricton would be less than 1%, if any.
Two optimized fuel demonstration assemblies (OFDA) were loaded in edge locations R-8 and A-8 during the previous cycle 2.
For the cycle 3 core, the two 0FDA were loaded in region 4A at core locations L-8 and E-8.
Their relatively lower U-235 enrichment and the locations would prevent their being lead power production assemblies-during normal operation or transient conditions.
The inspector did not have any further questions in this area.
5.3 -Accident Evaluation - Ejected RCCA The RCCA ejection analysis is dependent on ejected rod worth, power peaking, and delayed neutrons after the rod is ejected. The inspector noted that all the kinetics parameters were within their limits, except.for the delayed neutron beta-effective for rod ejections at BOC and EOC for two and three loop operations.
The inspector was informed that the reanalysis of the RCCA ejection analysis, accounting the beta-effectives found, verified the following:
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Maximum clad temperature was less than 220 F in conformance
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with 10 CFR 50.46(b)(v);
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Maximum stored energy was 225 cal /gm and 200 cal /gm for irradiated
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and unirradiated fuels respectively.
Under the proposed TS changes, the High Positive Rate-of-Power Range Neutron Flux trip setpoint would be returned to the pre-cycle 2 value of < 5% with a time constant > 2 seconds. This more restrictive
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trip setpoint under the proposed TS change would provide additional protection for the RCCA ejection accident.
The inspector did not have any further questions in this area, provided that the proposed TS change is adopted.
5.4 Accident Evaluation - Dropped RCCA The inspector noted that the DNBR would be maintained above its limit if Control Bank D was withdrawn equal or more than 215 steps in the automatic control mode above 90% power. The inspector noted that the automatic control mode above 90% power would be the most restrictive initial condition for dropped RCCA incidents. With the control rods in automatic response mode, the rods would respond instantaneously to compensate the reactivity insertion caused by the dropped rod, resulting in an additional power tilt. Again, a more restrictive reactor trip setpoint for the high negative power rate under the proposed TS changas would provide the added protection in event of dropped RCCA incidents.
The inspector did not have any further questions, provided that the TS change is adopted.
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Entrance and Exit Interviews Licensee management was informed of the purpose and scope of the inspection at the entrance interview, and the findings of the inspection were period-ically discussed with the licensee representatives as summarized in the following:
Date Reports Details Covered April 5, 1982 Entrance Interview April 6, 1982 morning Item 5, Pittsburgh afternoon Item 5, Monroeville (W-NFD)
April 7, 1982 Items 2, 3, 4 April 8, 1982 Exit Interview, participants identified in Item 1.
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