IR 05000271/1986019

Insp Rept 50-271/86-19 on 860811-15.No Violations Noted. Major Areas Inspected:Cycle 12 Startup Physics Testing Program & Power Ascension Tests

ML20214N979
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	09/03/1986
From:	Eselgroth P, Wen P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20214N951	List: IR 05000271/1986019 ML20214N949
References
50-271-86-19, NUDOCS 8609170199
Download: ML20214N979 (7)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /86-19 Docket N License N DPR-28 Priority -

Category -

Licensee: _ Vermont Yankee Nuclear Power Corporation RD 5, Box 169 Ferry Road Brattleboro, Vermont 05301 Facility Name: Vermont Yankee Inspection At: Vernon, Vermont Inspection Conducted: August 11-15, 1986 Inspectors: C. I de 9A /fb P. C. Wen, Reactor Engineer ~ date

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Approved by: p/ /

P. W. gffelgroth Ch f7 S

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9/J/f4 date Inspection Summary: Inspection on A_ugust 11-15, 1986 (Inspection Report No. 50-271/86-19)

Areas Inspected: Cycle 12 startup physics testing program and power ascension test Results: No violations were identifie NOTE: For acronyms not identified, refer to NUREG-0544, " Handbook of Acronyms and Initialisms".

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DETAILS Persons Contacted Vermont Yankee J.-Brooks, Reactor Engineer

• B. Buteau, Reactor & Computer Supervisor P. Carusl, QC Engineer

• J. Pelletier, Plant Manager T. Stetson, Reactor. Engineer

• R. Wanczyk, Technical Services Superintendent U.S. NRC W. J. Raymond, Senior Resident Inspector T. B. Silko, Resident Inspector

• Denotes those present at the exit meeting on August 15, 1986.

The inspector also contacted other licensee staff members in the course of the inspectio . Cycle 12 Reload Safety Evaluation and Startup Test Program The Cycle 12 reload contains 120 fresh fuel bundles and 248 irradiated fuel bundles ir. the core. Since Cycle 12 uses fuel essentially the same as that used in the Cycle 11, the Cycle 11/12 refueling was conducted under 10 CFR 50.59. The inspector reviewed " Vermont Yankee Cycle 12 Core Performance Analysis," YAEC-1507, dated November, 1985. The safety analyses performed to support this cycle's operation concluded that there was no unreviewed safety question involved. The result was presented to the Plant Operation Review Committee (PORC) (Meeting No. 86-08) and received its approval on February 11, 198 The startup test program was conducted according to test procedure , Reactor Engineering Beginning of Cycle Startup Testing, Revision The test program outlined the steps in the testing sequence, set initial conditions and prerequisites, specified calibration or surveillance procedures at appropriate points, and referenced detailed test procedures and data collections in attachments. Initial criticality of Cycle 12 was achieved on June 30, 1986. The startup tests were completed aoout July 25, 198 The results of tests: Control Rod Friction Test, Shutdown Margin Test, Core Verification and Reactor Initial Startup were reported in NRC Inspection Report.50-271/86-10. The results of the remaining startup tests were reviewed during this inspection and the details are discussed in Section 3.

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3 Cycle 12 Startup Testing The inspector reviewed selected test programs and their results to verify the following:

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Procedures were provided with the detailed stepwise instructions, 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;

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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 TS and the licensee's administrative control The following tests were reviewed:

3.1 Control Rod Drive Scram Time Test The control rod drive (CRD) scram time test was performed in accord-ance with procedure 0.P. 4424, Control Rod Scram Testing and Data Reduction, Revision 10. The inspector verified by review of the recorder traces and data obtained on July 5 and 6, 1986 that the average scram times at various insertion levels were all within the TS limits. The maximum scram time for 90% insertion of 2.85 seconds was well within the TS limit of 7 second No unacceptable conditions were identifie .2 Reactivity Anomaly Check The inspector reviewed test procedure 0.P. 4430, Reactivity Anomal-ies, Revision 10 and test results obtained since the beginning of this cycl The inspector noted that all critical rod configurations were in good agreement with analytically predicted values, and well within the 1% delta K/K TS limit No unacceptable conditions were identified.

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3.3 Core Thermal Power and AP.RM Calibration The licensee's procedure 0.P. 4400, Calibration of the Average Power

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Range Monitoring System to Core Thermal Power, Revision 9 was reviewed for technical adequacy. The inspector reviewed the calibra-tion results of July 3 through August 11, 1986, and verified that the Core Thermal Power was determined by the on-demand program 00- The final APRM readings wpre all within 1% of adjusted rated Core Thermal Powe The inspector also reviewed procedure 0.P. 2410, Revision 10. This procedure provides alternative methods to calculate core thermal power. These methods consist of (i) hand calculation and (ii) a backup computer method using an on-line Time-Share System (TSS)

program "CTP". During the startup testing period, at the 75% and 100% power plateaus, the licensee performed heat balance comparisons between the process computer 00-3 and TSS method. The inspector '

independently calculated a heat balance at the 100% power level using 0.P. 2410 Hand Calculation Method. All comparisons were in reason-ably good agreement and within 20 MWt licensee imposed acceptance criteria as shown below:

Test Date Power Plateau Method Results (MWt)'

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7/9/86 75% 00-3 1228.69

.TSS 1246.04

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7/16/86 100% 00-3 1577.88 TSS 1583.10 ,

hand calculation- 158 (by inspector) '

! However, further review indicated that the 4.1 MWt difference between E the TSS method and inspector's result was due te a difference in steam enthalpy calculation. The inspector found that the steam enthalpy at the rated conditions, determined by the TSS methed, is about 1 Btu /lbm less than the value from the ASME Steam Table. At the exit meeting, a licensee representative stated that the TSS steam table algorithm will be evaluated.

l 3.4 Thermal Hydraulic Limits The . inspector reviewed procedure 0.P. 4401, Core Thermal Hydraulic *

l Limits Evaluation, Revision 12, and the results of tests performed on July 7 - August 11, 1986. The inspector verified by review of the '

i P-1 and OD-6 computer outputs that the thermal limits (LHGR, MAPLHGR, -

and MCPR) were all within the TS Limits during this perio ,

The inspector compared the core limit data produced by the process computer P-1 output to the data obtained by the backup computer BUCLE i program at vayious power levels daring the startup testing period, ,

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and verified that they were al' n excellent agreement.

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3.5 Local Power Range Monitor (LPRM) System Calibration The inspector reviewed test' procedure 0.P. 4406, LPRM Calibration and Functional Check, Revision 6 for technical adequacy. During the startup testing period, an LPRM Calibration was performed at 25%

rated thermal power (RTP) per procedure 0.P. 4406 on July 7,198 LPRM gain were also updated by using the TIP system and process computer function 00-1 at 25%, 55%, 75% and 100% RT .6 Core Power Distribution The inspector reviewed test result of TIP Calibration 997 which was taken on July 31, 1986, and noted that the measured Critical Power

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Ratio, Maximum Average Power Ratio, Maximum Fraction of the Limiting

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Power Density, and all 20 TIP traces were in good agreement with the

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SIMULATE Code calculated results. The inspector also independently verified that the measured average axial power shape agreed closely with the analytical predicted value as described in " Vermont Yankee Cycle 12 Core Management Report," YAEC-1540, dated May 198 . Thermal-Hydraulic Stability The possibility of BWR thermal-hydraulic instability in certain operating regions, especially the high power / low flow corner of the power / flow map has been previously identified through testing and operating experience in the industr To prevent and mitigate the consequence of operatino the unit near these regions, the fuel vendor (GE) had issued SIL-380, DBWR Core Thermal Hydraulic Stability," to provide guidelines to the affected BWR owners. The inspector reviewed procedure 0.P. 2427, " Monitoring Reactor Stability," dated June 28, 1986. This procedure describes thermal-hydraulic stability monitoring requirements for various power / flow operating conditions in the core and provides methods for detecting and suppressing conditions of instabilit The inspector noted that the recommendations as specified in GE SIL-380, Revision 1 have been incor-porated in this procedure (0.P. 2427). Baseline data had also been collected during the startup testing perio No unacceptable conditions were identifie . Recirculation Flow Anomaly

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During plant startup at rated power and flow conditions, flow anomalies in the recirculation system were observed. The anomaly consists of l

periodic increases in Loop 'A' recirculation flow (by about 1%) for a short time and then decreases back to its original value. Consequently, the core flow (summation of jet pump flows) and APRM readings increase approximately 0.5% and 0.2%, respectively. This flow oscillation is j

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random in nature. It occurs with a frequency of about 3-4 times in a one hour period, and each time lasts for anywhere from to 5 minutes. A similar problem was previously observed at the Pilgrim Nuclear Power l

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6 Station following its startup from the recirculation pipe replacement outage in January 198 In the Pilgrim case, based on the amplitude and frequency information from the operating data, it has been determined that this flow oscillation phenomenon has no safety impact on the plan During the Cycle 11/12 outage, the licensee replaced the recirculation loop piping. The cross-tie between the two loops was eliminated in the new piping configuration which is similar to Pilgrim's piping arrangemen The inspector noted from the licansee's trended data, that the amplitude and frequency of the phenomenon are no worse than the Pilgrim cas Nevertheless, the licensee is continuously trending and evaluating the ,

problem, and the NRC inspector will follow it u ' Independent Calculations / Verifications

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The inspector performed independent calculations / verifications of Cycle 12 startup physics testing related activities. These included the following:

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Independent core thermal power calculation as described in Section Core power distribution. verification as described in Section ,

! Quality Assurance (QA) Involvement in Startup Testing The inspector reviewed the licensee's QA involvement during the post refueling startup test and noted that many audits, including two survell-lances directly related to the testing program, had been performed. Oper-ation Quality Group (0QA) surveillances86-168 and 86-177 covered the Reactor Engineering activities of Shutdown Margin Demonstration and

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Reactivity Anomalies Check. In addition, the 00A group had a detailed QA

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coverage plan for the entire refueling outag Based on the document review and a discussion with a cognizant QA repre-  :

sentative, the inspector determined that QA/QC was actively involved in the Cycle 12 startup testing activitie ..

i No unacceptable conditions were identified.

l Exit Interview Licensee management was informed of the purpose and scope of the inspec-tion at the entrance intervie The findings of the inspection were <

periodically discussed with licensee personnel and were sm.warized at the '

l conclusion of the inspection on August 15, 1986. Attendees at the exit l

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interview are denoted in paragraph 1.

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At no time during this inspection was written material provided to the i licensee. Based on the NRC Region I review of this report and discussions

held with the licensee representatives at the exit it was determined that this report does not contain informattan subject to 10 CFR'2.790 restric- .

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