IR 05000334/1986022

Insp Rept 50-334/86-22 on 860915-19.No Violations Identified.Major Areas Inspected:Cycle 6 Startup Physics Testing Program,Precriticality Tests,Zero Power Physics Tests & Power Ascension Tests

ML20211C473
Person / Time
Site:	Beaver Valley
Issue date:	10/06/1986
From:	Briggs L, Wen P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20211C464	List: IR 05000334/1986022 ML20211C459
References
50-334-86-22, NUDOCS 8610210378
Download: ML20211C473 (10)

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U.S. NUCLEAR REGULATORY COMMISSION REGION'I Report N /86-22 Docket N License No. DPR-66 Priority -

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Licensee: Duquesne Light Company Post Office Box 4 Shippingport, Pennsylvania 15077 Facility Name: Beaver Valley 1 Inspection At: Shippingport, Pennsylvania Inspection Conducted: September 15-19, 1986 Inspector : w eer9 JJ ff78

,p[, P. C. Wen, Restifor Engineer date Approved.by: [ ges > /B[6 84 L. E. Briggs #Cliief date Test Programs Section, 08, DRS Inspection Summary: Inspection on September 15-19, 1986 (Inspection Report No. 50-334/86-22).

Areas Inspected: Cycle 6 startup Physics Testing Program, precritical tests, zero power physics tests and power ascension tests.

. Results: No violations were identifie Note: For acronyms not identified, refer to NUREG-0544,

" Handbook of Acronyms and Initialisms".

8610210378 861014 PDR G

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DETAILS Persons Contacted Duquesne Light Company A. Brunner, Refueling Supervisor

• A. Burger,' Core Performance Engineer A. Dulick, Chemistry Supervisor L. Freeland, Nuclear Operations Supervisor F. Lipchick, Compliance Engineer A. Mizia, QA Operation Supervisor

• B. Sepelak, Licensing Engineer

• G. Sovick, Senior Licensing Supervisor

• G. Zupsic, Testing Supervisor U.S. Nuclear Regulatory Commission

• L. Prividy, Resident Inspector W. Troskoski, Senior Resident Inspector

• Denotes those present at the exit interview on September 19, 198 The inspector also contacted other licensee _ employees in the course of the inspection.

2. Cycle 6 Reload Safety Evaluation and Core Verification The Cycle.6 reactor core is comprised of 157 fuel assemblies (FAs).

During the Cycle 5/6 refueling, 69 FAs were replaced with 68 Region 8 fresh fuel-and one once-burned Region I fuel. The reload safety evalua-tion (RSE) performed to support this cycle's operation concluded that there was no unreviewed safety question involved. The result was presented to the Onsite Safety Committee (Meeting No. 36-86) and received its approval on July 17, 1986. The basic assumption used in the RSE was Cycle 5 burnup of 14,140 - 16,140 MWD /MTU. The inspector verified the actual Cycle 5 burnup to be 15,942 MWD /MTU. The assumption is thus vali Other safety evaluations performed to support the Cycle 6 startup were:

increasing the steam generator tube plugging level limit to 2.5% in each steam generator, and utilizing of the reconstituted fuel assembly of G-1 These results were also reviewed by the Onsite Safety Committee (Meeting Nos. 38-86 and 36-86) which concluded that there was no unreviewed safety question involve The inspector reviewed the core verification videotapes and verified that the core loading agreed with the intended core loading plan. No inadequacies were identifie .

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. Cycle 6 Startup Testing Program The startup test program was conducted according to test procedures BVT 1.6-2.2.1, Initial Approach to Criticality After Refueling, Issue 1, Re O, and BVT 1.6-2.2.2, Core Design Check Test, Issue 1, Rev. The test sequence outlined the steps in the testing program, set initial conditions and prerequisites, and specified calibration or surveillance

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procedures at appropriate points in the test sequenc Initial criticality of Cycle 6 was achieved on August 24, 1986. The Zero Power Physics Testing (ZPPT) was completed on August 25, 1986, and the Power Ascension Tests (PAT) were completed on September 8, 198 The inspector independently verified that the predicted values and acceptance criteria were obtained from "The Nuclear Design and Core Management of the Beaver Valley Unit 1 Power Plant Cycle 6", WCAP-1112 The inspector reviewed test results and documents described in this report to ascertain that startup testing was conducted in accordance with

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technically adequate procedures and as required by TS. The details and J

findings of the review are described in Sections 4 and . Cycle 6 Startup Testing --Precritical Tests The inspector reviewed calibration and functional tests results to verify the following:

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Procedures were provided with detailed instructions;

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Technical content of procedure was sufficient to result in satisfactory component calibration and test;

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Instruments and calibration equipment used were traceable to the National Bureau of Standards;

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Acceptance and operability criteria were observed in compliance with TS.

, The following tests were reviewed:

i l 4.1 Control Rod Drop Times The rod drop measurement was performed in accordance with procedure BVT 1.1-1.1.1, Issue 2, Rev. 2. The inspector verified performance by review of the test results obtained on August 18, 1986. The drop

, times for all 48 rods were less than 2.2 seconds as required by the TS. The inspector also reviewed several visicorder traces and

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verified that the drop times had.been interpreted correctly from these selected trace No unacceptable conditions were identifie ~ ._ _ _ _ . .- _ ___- _ -

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4.2 Reactivity Computer Setup / Verification The reactivity computer was setup and calibrated according to procedure BVT 1.6-2.2.1, Appendix A. The inspector independently verified that the reactivity computer was adjusted with the correct inputs of delayed neutron fractions (betas) and decay constants (lambdas), and noted that the results of this " cold" calibration check were satisfactor The reactivity computer was further checked when the reactor reached criticality. Comparisons of predicted and measured reactivities based on doubling time measurement were acceptable with a maximum deviation of only 0.67%.

5. Cycle 6 Startup Testing - Post-Critical Tests

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3 5.1 The inspector observed and reviewed selected test programs to verify the following:

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The test programs were implemented in accordance with Cycle 6-Core Design Check Test Procedure;

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Step-wise instructions of test procedures were adequately provided including Precautions, Limitations and Acceptance Criteria in conformance with the requirements of the TS;

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Provisions for recovering from anomalous conditions were provided;

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Methods and calculations were clearly specified and the tests were performed 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 .2 Baron Endpoint D_etermination The licensee measured the critical boron concentration in accordance with BVT 1.6-2.2.2. The inspector reviewed the data and noted the following results:

Predicted Value Test Result Configuration (ppm) (ppm)

All Rods Out (AR0) 1504 50 152 Control Bank B In 1325 15% 137 Test results were within acceptance criteri _ _ _

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5.3 Isothermal Temperature Coefficient Isothermal temperature coefficient (ITC) was measured in accordance with procedure BVT 1.6-2,2.2. The inspector noted the following result:

Predicted Value Measured Value Configuration (pcm/ F) (pcm/ F)

ARO -2.59 3 -2.96 The calculated Moderator Temperature Coefficient (MTC) from the ITC measurement is:

MTC TS Limit Conditions (pcm/ F) (pcm/ F)

AR0/HZP/BOL -0.84 <0 Test results were within acceptance criteri ~

5.4 Control Rod Worth Measurement The control rod reactivity worth measurements were performed in accordance with BVT.1.6-2.2.2. The following results were noted:

Predicted Worth Measured Worth Rod Bank Test Conditions (pcm) (pcm)

Control Bank B Dilution 1349 1 10% 1249 Control Bank D . Interchange 1154 15% 107 Control Bank C Interchange 833 15% 79 Control Bank A Interchange 624 15% 59 Shutdown Bank B Interchange 1014 i 15% 96 Shutdown Bank A Interchange 1080 1 15% 100 Total Banks -----------

6054 i 10% 568 Test results were within acceptance criteri .5 Core Power Distribution The procedure and method used by the licensee to verify that the plant is operating within the power distribution limits defined in TS were reviewed and discussed with cognizant licensee personne The data taken by the Movable Incore Detector System is obtained on the plant P-250 process computer. This information was then fed into a main frame CDC computer as input to the "INCORE" analysis code and analyzed results transmitted back to BVP .

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The first flux map of the Cycle 6 operation (DLW1-06-01) was taken at about 30% power level on August 28, 1986. The following results were noted:

Measured Value Acceptance Criteria N

F AH 1.5077 < 1.8755 T

.F q 2.1348 (At Axial < 4.6168 Point 29)

Quadrant Power Tilt Ratio 1.0207 <. 1.02 The measured quadrant power tilt ratio (1.0207) slightly exceeded the acceptance criterion of 1.02, however, the provisiont of TS 3.2.4 (Quadrant Power Tilt) are not applicable below 50% RT The second flux map (DLW1-06-02) was taken at 74.1% RTP. The measured F A H Fg T and quadrant power tilt ratio were all within N

the acceptance criteria. The small tilt measured on the 30%.RTP fiux map was, therefore, not a problem. This conclusion was further substantiated by the 100% RTP flux map (DLW 1-06-09). All measured power distribution parameters were within the acceptance criteri The inspector independently reviewed the DLW 1-06-09 flux map and noted that the measured and expected values of F N AH were in g d agreement with a maximum deviation of 3.6% at core location J-1 The inspector also verified that engineering and measurement uncertainties were included in the licensee's power distribution limits chec Due to the baffle jetting problem (see Section 6), fuel assembl G-15 (17 x 17, Region 7 fuel) which is now located at core location D-13 was reconstituted with 7 stainless steel rods at the rod positions susceptible to damage. The licensee's fuel vendor (Westinghouse) performed an engineering calculation prior to the Cycle 6 startup, and determined that this fuel changeout in FA G-15 had very minor impact on the core power distribution. The results from all flux maps which were taken during the Cycle 6 startup physics testing essentially supported this conclusio The inspector had no further question .6 Incore/Excore Calibration The incore/excore correlation was determined at approximately 74% ,

RTP in accordance with procedure BVT 1.3-2.2.3, " Nuclear Power Range I Calibration", Issue 1, Revision Seven (7) axial flux differences obtained from one full core flux map (FCFM 0602) and six quarter core flux maps (QCFM 0603-0608) were analyzed and compared to

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responses of the excore detectors to develop a calibration curve for each power range detecto Reasonably good linear relationship for all eight detectors was observed. The excore detectors were then calibrated by I&C. Group personnel prior to reaching full power level.

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The inspector reviewed the full core flux map (FCFM-0609) performed on September 8,1986, and noted that the measured axial offsets from excore detectors agreed fairly well with the measured incore axial offset and were within 3% of licensee established acceptance criteri .7 Target Axial Flux Difference Determination Target Axial Flux Difference Determination was performed by the methods detailed in the test procedure BVT 1.3-1.49.1, " Delta Flux Target Update", Issue 1, Rev. 3. The inspector reviewed the calculation performed on September 8-10, 1986 and verified that this new set of target flux sensitivity factors was entered into the plant process computer. The inspector toured the control room and noticed that this revised information was in use by the reactor operator .

5.8 Core Th'ermal Power The procedure and method used by the licensee to calculate the core thermal power were reviewed and discussed with cognizant licensee personnel. The inspector performed an independent calculation. The plant parameters taken at 0115 on September 18, 1986 were used as inputs for compariso Licensee Calculated Inspector Calculated Test Date Result (Mwt) . Result (Mwt)

0115 261 .5 09/18/86

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Licensee calculated result (Cl-3 log sheet) agreed fairly well with the inspector calculated resul The inspector also reviewed the data from the measurements performed for the month of September 1986, and verified that the frequency of evaluation and excore power range channel calibrations were performed within the requirements as prescribed by the TS and plant operation procedur No unacceptable conditions were identifie l l

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5.9 Shutdown Margin Determination The inspector reviewed the licensee's shutdown margin determination procedure OST 1.49.1 and surveillance results performed during Mode 2 operation on August 24, 1986. The control rod worth and assumed maximum stuck rod worth used in the calculation were consistent with the nuclear design and core management report (WCAP-11122). The result based on insertion of control rods alone showed a shutdown margin of 6.20%~AK/K which met the TS requirement of 2 1.77% AK/ . Fuel Assembly Baffle Jetting Higher than normal reactor coolant activity was sampled in the previous

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cycle (Cycle 5) operation. Data from trending of I-131, I-131/I-133 ratio, and Cs-137/Cs-134 ratio indicated that fuel failed during mid-cycle life on Region 5 fue During the Cycle 5/6 refueling outage inspection of fuel assembly, E-37 (Region 5 thrice-burned fuel)

which was located in core location D-13 was found damaged due to baffle jettin The baffle jetting fuel failure is caused by a direct high pressure jet impingement on certain fuel rods. Tha high pressure jet is a result of the defective baffle joint and the "downflow" design of the reactor coolant flow between the core barrel and the baffle Figure 1 illustrates the high delta pressure jet flows through a small baffle ga As a result of this baffle jetting problem, a modification was made to FA G-15 for this cycle's operation as described in Section The inspector reviewed recent samples of the reactor coolant activity trending during the Cycle 6 operation and noted that the dose activity is much less than the Cycle 5 End of Cycle (E.0.C.) data, as shown below:

Dose Equivalent I-131 I-131/I-133 Date (% of TS Limit)

Cy5 E. ~2 ~0.12 Mid-Sept., 1986 ~ ~0.06 (full power operation)

Based on the result of licensee's -radiochemistry data, no degraded fuel was identified at the current fuel burnup (about 17 EFPD). The NRC will continue to follow the baffle jetting problem and its impact on plant operation .

, .. QA Role in Cycle 6 Startup Testing The inspector reviewed the licensee's QA involvement during the post refueling startup testing, and noted that QA surveillance group provided surveillance coverage TPP-06-86 and TPP-07-86 during Cycle 6 startup. QA auditors witnessed performance of BVT 1.6-2.'2.1 and BVT 1.6-2.2.2 related tests. An audit (BV-1-86-15) was also performed to evaluate the Test and Plant Performance Group's compliance with plant administrative procedure The inspector noted that the level of QA coverage had significantly increased since the previous Cycle 5 startup physics test progran,.

No discrepancies were identifie . Independent Calculations / Verifications The inspector performed independent calculations / verifications of Cycle 6 startup physics testing related activities. These included the following:

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Core loading verification as described in Section Test acceptance criteria were verified to be consistent with Westinghouse document "The Nuclear Design and Core Management of the Beaver Valley Unit 1 Power Plant Cycle 6," WCAP-1112 Core thermal power calculation as described in Section . Management Meeting Licensee management was informed of the scope and purpose of the inspection at an entrance meeting conducted on September 15, 1986. The findings of the inspection were discussed with licensee representatives during the course of the inspection. An exit meeting was conducted on September 19, 1986 at the' conclusion of the inspection (see paragraph I for attendees).

At no time during this inspection was written material provided to the license 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 information subject to 10 CFR 2.790 restriction _ _

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Core Barrel RCS Flow Directi~on (Downward) x

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RCS Flow Direction (Downward) x Baffle Plate Jet Flou l

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Gap in Daffle .

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Plate Joint Fuel Assembly b 5 80 :

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RCS Flow Direction (Upward) e

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Figure 1 Baffle Jetting

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