IR 05000271/1981009
| ML20038B555 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 11/12/1981 |
| From: | Bettenhausen L, Chung J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20038B539 | List: |
| References | |
| 50-271-81-09, 50-271-81-9, NUDOCS 8112080394 | |
| Download: ML20038B555 (11) | |
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I Report No. 50-271/81-09 Docket No. 50-271 License No. DPR-28 Priority
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Category C
Licensee: Vermont Yankee Nuclear Power Station Vernon, Vermont Facility Name: Vermont Yankee Inspection at:
Framingham, Massachusetts and Vernon, Vermont Inspection cond cted:
September 30 - October 9, 1981 Inspectors:
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//-/4 -<f W.'Chung, Reactor nspector date signed /
Approved by:
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L. Bettenhausen, Chief, Test date signed Program Section, Engineering Inspection Branch Inspection Summary:
Inspection on September 30- October 9,1981 (Report No. 50-271/81-09)
Areas Inspected:
Routine, unannounced inspection of cycle 9 performance analysis; cycle 9 post refueling test procedures; cycle 8 post refueling startup testing; review of cycle 8 startup test summary report; and centrol room observation. The inspection involved 47 inspector-hours onsite by one region-based inspector.
Results: Noncompliance: None 9112080394 811118 PDR ADOCK 05000271
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DETAILS
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1.
Persons Contacted
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Engineering, Framingham YAEC
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F. Ansari, Titled Engineer, BWR Transient Analysis
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J. T. Cronin, Senior Engineer, BWR Safety Analysis
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J. M. Halzer, Titled Engineer
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D. Kapitz, Senior Engineer
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M. A. Sironen, Titled Engineer B. Slifer, Manager, Nuclear Engineering Department
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D. Ver Planek, Specialist Engineer
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R. A. Woehlke, Senior Engineer D. E. Vandenburgh, Vice President
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Vermont Yankee
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R. Bobua, Nuclear Safety Engineer
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F. Burger, QA Coordinator
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K. W. Burke, Operations Engineer, G.E.
- B. Buteau, Reactor and Computer Supervisor
- M. Hyster, Operations Superintendent
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- S. Jefferson, Technical Services Superintendent
- R. Kenny, Engineer
- D. Reid, Engineering Systems Superintendent
- J. Pelletier, Assistant Plant Manager The inspector also interviewed other licensee employees during the inspection, including Engineering Support, Reactor Operators, and Performance personnel.
- denotes those present at the Exit Interview on October 8, 1981 2.
Cycle 9 Performance Analysis - Yankee Atomic Electric Company (YAEC)
a.
Inspection Objective The objective of this portion of the inspection was to ensure that the YAEC analyses and the results were compatible with the methods applied to the-previous reloads, and the safety margins were equivalent to or better than those in the Final Safety Analysis Report.
The inspectoi reviewed the analysis report and other supporting documents to verify the following:
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Overall plant safety margins,
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Consistency of the operational parameters,
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Differences between General Electric (GE) Company and YAEC methodology, analysis, and results,
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Computer codes used in the analyses
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Proposed amendment to Technical Specifications (TS),
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Implementation of the proposed TS changes in station procedures.
b.
Documents reviewed:
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Vermont Yankee Cycle 9 Core Performance Analysis, YAEC-1275, August 1981.
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Reload 8 Licensing Submittal, September 2, 1981.
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One-dimensional Core Transient Model for Boiling Water Reactors, Volume 1&2, NED-24154, October 1978.
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Licensing To.nical Reports: Analytical Method of Plant Transient Evaluations for the General Electric Boiling Water Reactor, NEDO-10801-1: NED0-10802-2; NED0-10802.
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Vermont Yankee FSAR, Section 15, Safety Analysis
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Methods for the Analysis of Boiling Water Reactors Lattice Physics, YAEC-1232, December 1980.
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Methods for the Analysis of Oxide Fuel Rod Steady -State Thermal Effects (FROSSTEY) Code Qualification and Application, YAEC-1265P, June 1981.
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FIBWR: A Steady - State Core Flow Distribution Code for Boiling Water Reactors - Code Verification and Qualification Report, EPRI NP-1923, July 1981.
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Methods for the Analysis of Boiling Water Reactors: A Systems Transient Analysis Model (RETRAN), YAEC-1233, April 1981
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RETRAN - A program for One-dimensional Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems, EPRI CCM-5, December 1978.
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Methods for the Analysis of Boiling Water Reactors:
Transient Thermal Margin Analysis Code (MAYUO4-YAEC), YAEC-1235, December 1980.
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Methods for the Analysis of Boiling Water Reactors Transient Core Physics, YAEC-1239P, August 1981.
The inspector also discussed the scope and the characteristics of the methodology with the iAEC engineering staff members.
However, the details of the algorithms and the mathematics were not reviewe.
c.
Specific areas discussed and reviewed were:
(1) YAEC Methodology and Analysis The overall approach and methodology employed in the design and the operational information for the Reload 8 Analysis appeared to be the same.as those employed for the previous reload analyses.
The analytical tools and the computer codes were developed by the YAEC in conjunction with the previous performance analysis methods and in cooperation with the Electric power Research Institute (EPRI). YAEC is a member of the EPRI code user's group.
The steady state thermal effects of oxide fuel rods'were calculated using the FROSSTEY computer code.
FROSSTEY provides the pellet-to-clad gap conductance and fuel temperatures under the normal and peak Linear Heat Generation Rate (LHGR) conditions.
These are input to the one-dimensional transient thermal-hydraulic code RETRAN and the modified transient thermal margin analysis code MAYUO4-YAEC.
RETRAN and MAYUO4-YAEC replace GE's REDY/0DYN-
/MAYUO4/GEXL for the cycle 9 fuel and thermal evaluations.
The core thermal-hydraulic analyses were performed using the steady state core flow distributor code FIBWR. An All-Rods-Out (AR0)
Haling depletion method using SIMULATE code was developed for the power distribution.
Accident analyses and hot channel evaluations were performed employing RETRAN and MAYUO4-YAEC codes.
The EPRI void model and the GE GEXL correlations were incorporated into the single hot channel code, MAYUO4.
This modified MAYUO4-YAEC code requires fewer assumptions for the core density distribution and employs an iterative technique for the critical heat flux.
In essence, RETRAN and MAYU04-YAEC require the same input information as the GE ODYN/REDY one-dimensional codes and provide the same output information but can accomodate higher power peaking.
The EPRI qualified RETRAN code is an advanced version of RELAP-4 code and includes an additional inertia term in the steamline momentum balance, which effectively describes the steamline reverse flow when the turbine stop valves are bypassed.
The RETRAN had been tested with the Peach Bottom turbine trip tests with and without the bypasses.
SIMULATE is a three-dimensional neutronic computer code developed by YAEC and tested on the Vermont Yankee cycle 1-7 exposure data and gamma scan data.
It is claimed that SIMULATE is relatively effective in dealing with albedo problems, and to
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provide more stable eigen values as opposed to GE's " saddle" Keff values. The 3-D simulation predicts the region-wise reactivity coefficients accurately and includes void history effects on the neutron cross-section determination, for which 25 energy groups are used from the Engineered Nuclear Data File (END F).
(2) Cycle Management Information Shutdown Margin was calculated for a symmetric octant from the cold Keff and with ARO and All-Rods-In (ARI) using SIMULATE.
The R-value was 0.66% ak/k for core 8, higher than the value given by GE. This was probably due to the Haling model employed by YAEC.
The Standby Liquid Centrol System margin was calculated under cold, Xenon-free, ARD conditions at the most reactive time.
The margin was more than 5.5% ak/k with 720 ppm boron injection.
The Rod withdrawal sequence would be the same as the reload 7 (cycle 8).
The misloaded bundle analysis was performed for the misoriented and the mislocated bundles.
The inspector discussed the cycle 9 core management parameters with the cognizant licensee engineers.
They stated that the information required for the proper conduct of the post refueling startup testing would be provided to the site reactor engineers prior to the restart, provided that the proposed licensing submittal to NRR is approved.
The licensee also stated that the formats and the content of the cycle management information would be basically same as those for the previous cycles.
The inspector had no further questions.
(3) Cycle 9 Technical Specification Changes and procedures In order to have additional thermal margin at the end of the operating cycle, a change to the TS control rod drive scram time is proposed.
The proposed change was a trade-off between the scram time requirements and the Minimum Critical Power Ratio (MCPR) limits.
The previous experiences in the scram time measurements indicated that the proposed shorter scram times would be still well within the plant capability.
However, the proposed change would require two sets of TS requirements for the scram times and MCPR at the beginning and the end of the operating cycles.
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(4) Findings Since the proposed TS change requires two sets of scram times and MCPR, the associated test procedure, OP4424, should be revised to reflect the changes, and both the scram time and its corresponding MCPR limits have to be satisfied.
The on-demand computer programs, such as P-1 and 00-6, normalize their thermal parameters with the limiting MCPR, and thus, the computer files should also be updated with the appropriate set of MCPR limits under the new TS.
The licensee stated that the procedure and the computer files would be updated by November 30, 1981, provided that the proposed changes are approved.
This is an inspector follow item (30-279/
81-09-01).
3.
Cycle 8 Startup Summary Report Review The inspector reviewed the cycle 8 Startup Testing Summary Report of YAEC Vermont Yankee Nuclear Station, issued March 31, 1981.
The test report included:
Core Verification,
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Process Computer Data Checks,
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Shutdown Margin Testing,
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In-Sequence Critical Testing,
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Rod Scram Time Testing,
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Thermal Hydraulic Limits and Power Distribution and
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TIP calibration checks.
The inspector determined that the test results and conclusions were consistent with the predicted performance analysis results and met the requirements specified in TS.
The inspector identified that the thermal parameters in tabies II and III were the normalized ratio of the limiting valves, and that the orrect
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parameters in the tables were CMFCP and MAPRAT, instead of MCPR and MAPLHGR respecoively. The inspector also noted that the acceptance values of the measured versus the predicted values were not clearly specified in the report, and that TIP traces and the power distribution were compared with the calculations performed by SIMULATE code, which was neither qualified nor approved by NRC.
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The licensee stated that the SIMULATE results were for information and demonstration purposes, and the future test reports would be carefully reviewed to eliminate any erroneous misrepresentation.
This is an inspector follow item.
(50-279/81-09-02)
4.
Cycle 8 Startup Testing The inspector reviewed selected test programs and their results to verify that:
Procedures were provided with the detailed stepwise instructions,
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including Precautions, Limitations, and Acceptance Criteria;
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Technical content of the procedures was sufficient to result in satisfactory calibration and test;
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Test programs were implemented in accordance with test sequencing procedures; Provisions for recovering from anomalous conditions were provided;
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Methods and calculations were clearly specified and tests were conducted 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 administrative control procedures.
The following test programs and results were reviewed:
a.
Administrative Controls The inspector reviewed the administrative control documents to verify that the post-refueling sequences and testing were conducted in conformance with the station procedures and TS requirements, and that the test procedures were consistent with ANSI N18.7-1976.
The documents reviewed were YAEC OA Manual,Section II, August 15, 1977, and AP4000, Surveillance Testing Control, Revision 6, June 6, 1980.
b.
Core Thermal Power and APRM Calibration APRM readings were calibrated against Core Thermal Power and the final adjusted values were all within 1%.
The inspector reviewed the calibration procedure, OP 4400, Revision 7, December 19, 1980 and the calibration results of December 29, 1980 through January 12, 1981, and verified that the Core Thermal Power was determined by the on-demand program 00-3, Option 2.
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The inspector also reviewed core thermal evaluation procedures by a hand calculation and by a backup computer method using an on-line Time Sharing System (TSS).
The inspector determined that the on-demand 00-3, T!S, and hand calculations were all heat balarcing methods and their results were in good agreement as shown in the following:
Test Date Method Core Thermal Power MWt 12/28/80 00-3 400.82 TSS 400.21 11/29/79 00-3 1590.94 Hand 1593.20 The thermal output of the reactor core was estimated from the enthalpy
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balance of the steam and feed lines, the control rod hydraulic units, two recirculation pumps, and the reactor water cleanup system loss.
The circulation pump input was calculated from the electrical energy requirements to drive the pump, multiplied by the electric-to-thermal energy conversion factor, i.e., Recirculation Pump Effective Power Transfer Coefficient (EPTC).
The inspector verified by review of TP 75-01 and startup procedure ST-20 that the EPTC was an exponential function of the pump speed.
However, the core thermal power calculation in the procedure OP4400 specified the pump EPTC as 0.93 regardless of the pump speed, and the pump performance curve was not referenced in the procedure.
The inspector determined that the EPTC should be applied separately to each pump, and that 0.93 was applicable only near the fullflow region. Although the recirculation pump inputs were small compared with the total thermal power, the application of the highest EPTC value was not conservative particularly at the lower pump speed.
The cognizant licensee engineer acknowledged the inspector's concern and stated that the procedure would be revised by November 30, 1981 j
to include the reference figure and to apply the coefficient properly.
This is an unresolved item pending review by NRC:RI inspector (50-271/
81-09-03).
c.
Control Rod Scram Testing
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The inspector verified by review of the recorder traces and data obtained December 17-18, 1980 that the average control rod scram times were all within the TS limits, and the slowest scram time of
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2.8 seconds for 22-43 rod was well within the administrative limit of 5 seconds and TS limit of 7 seconds.
The inspector had no further questions.
d.
The inspector reviewed the test results of December 22, 1980, and determined that a minimum of 1.09% shutdown reactivity margin was demonstrated'with the strongest rod (30-23) and an adjacent 34-27 rod fully withdrawn and the moderator temperature of 90'F.
The TS requirement was a margin greater than R+0.25%Ak/k, where R was 0.274% Ak/k, including 0.07% Ak/k B C settling of possible q
inverted rod.
The SDM demonstration requires the R(=R + R ) value, the reactivity
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worths of the strongest and the adjacent withdrawn rods, and the reactivity correction of the moderator temperature.
The inspector determined that the stepwise instructions and'the corresponding references were not specified in the procedure OP 4426, and that the detailed calculations were actually performed based on perscnal notes and memory.
The inspector also identified that for the same core, the R value i
was specifiect as 0.5% Ak/k in the Cycle Management Report for Vermont Yankee Cycle 8, NEX: 80-198 and as 0.207% Ak/k in the Startup Information for Vermont Yankee Cycle 8, NEX:80-243, December, 1980. Also, the procedure OP 4426 was ambiguous as to when the R2value of BqC settling should be applied.
The licensee stated that the procedure would be revised by November 30, 1981 to correct the above findings.
This is an unresolved item pending revision of the procedure and subsequent NRC:RI inspection (50-271/81-09-04).
e.
Thermal Hydraulic Limits The inspector reviewed the test procedure OP 4401, Revision 8, December 19, 1980 and the results of January 1-29, 1981. The inspector also_ verified by review of the programs OD-6, Option 4, and P-1 that the thermal parameters, MAPLHGR, MCPR, and peaking factor, were all within the TS limits for this period.
The inspector determined from the core limit data of December 30, 1980 that the backup computer program BUCLE results and the P-1 results were identical.
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The inspector had no further questions.
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Exposure Surveillance The inspector verified by review of 00-59-30 and 00-6, Option 2, performed January 29, 1981, that the maximum nndal exposure and its thermal parameters were in conformance with the TS limits.
The assembly exposures for the different types of the fuel assemblies and their thermal parameters were also consistent with the TS require-ments.
The inspector had no further questions.
g.
Core Verification Selected fuel element locaton and process computer data were verified against the GE data files.
The Refueling Update Monitor Output (00-20) and I/O Type (0D-12, Option 1) were reviewed, for the following:
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Fuel assembly serial numbers were consistent with the core maps;
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The isotopic compositions were also consistent with the Nuclear Material Transfer Report.
The inspector also verified that the process computer inputs of the exposure (ELQ) and LHGR limits (FLQ) were consistent with the limits specified in TS.
h.
Critical Configuration and Anomaly Check The inspector reviewed test procedure OP 4430, Reactivity Anomalies, Revision 5, November 26, 1980 and the results of December 23, 1980, and verified that the critical rod configuration was within 1% Ak/k of the predicted critical pattern.
The inspector had no further questions.
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Traversing Incore Probes (TIP)
The inspector reviewed the TIP calibration data and traces of January 2,1981, and witnessed the calibration test performed October 8, 1981.
No unacceptable conditions were identified.
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Control Room Observation The inspector observed control room operations for manning, shift turnover, daily logging, and facility operation.
The inspector also discussed the spare Rod Worth Minimizer and the template utilization for the rod sequence control with the operational staff members.
Vermont Yankee is one of a few BWR plants which do not have the hard-wired Rod Sequence Control System and utilizes the Bankea Position Withdrawal Sequence.
No unacceptable conditions were identified.
6.
Unresolved Items Unresolved items are those items for which further information is required to determine whether they are acceptable or items of noncompliance.
Unresolved items are identified and detailed in Paragraphs 4.(b) and 4.(d).
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 periodically discussed and were summarized at the conclusion of the inspection on October 8, 1981. Attencees at the exit interview are denoted in Paragraph 1.
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