IR 05000311/1987005
| ML20207S296 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 03/10/1987 |
| From: | Eselgroth P, Wen P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20207S291 | List: |
| References | |
| 50-311-87-05, 50-311-87-5, NUDOCS 8703190248 | |
| Download: ML20207S296 (10) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-311/87-05
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Docket No.
50-311 License No. DPR-75 Priority
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Category
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Licensee: Public Service Electric and Gas Company P. O. Box 236 i
Hancocks Bridge, New Jersey 08038 Facility Name: Salem Unit 2 Inspection At: Hancocks Bridge, New Jersey Inspection Conducted:
February 3-6, 1987 c
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3/9/37 Inspectors:
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P. C. Wen, Reactor Engineer date i
Approved by:
/# 77 P. W. Eselg h, Chief, Test Programs date Section B, DRS Inspection Summary:
Inspection on February 3-6, 1987 (Inspection Report No. 50-311/87-05)
Areas Inspected:
Cycle 4 startup physics testing program, pre-critical tests, low power physics tests and power ascension tests.
Results: No violacions were identified.
NOTE: For acronyms, not identified, refer to NUREG-0544, " Handbook of Acronyms and Initialism".
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DETAILS 1.
Persons Contacted Public Serv' ice Electric and Gas Company R. Dulee, QA Principle Engineer
- M. Gray, Licensing Engineer D. D. Glassic, Engineer J. L. Jones, Jr., Senior Staff Engineer
- L. K. Miller, Assistant General Manager - Salem Operations
- J. P. Ronafalvy, Technical Manager - Salem Operations F. Safin, Lead Engineer U.S. Nuclear Regulatory Commission
- K. H. Gibson, Resident Inspector T. J. Kenny, Senior Resident Inspector
- Denotes those present at the exit interview on February 6,1987.
The inspector also contacted other licensee employees in the course of the inspection.
2.
Cycle 4 Reload Safety Evaluation The cycle 4 reactor core is comprised of 193 fuel assemblies. During the cycle 3/4 refueling, 84 fuel assemblies were replaced with Region 6 fresh fuel. Since cycle 4 uses fuel essentially the same as that used in the cycle 3, the cycle 3/4 refueling was conducted under 10 CFR 50.59. The Reload Safety Evaluation (RSE) performed to support this cycle's operation concluded that there was no unreviewed safety question involved. The result was presented to the Station Operations Review Committee (SORC) (Meeting No.86-098) and received its approval on November 26, 1986.
The inspector reviewed the RSE performed by the fuel vendor (Westing-house), and noted that the cycle 4 core loading pattern change did not alter the previous reference cycle's safety analysis results. The inspector also verified that the actual cycle 3 burnup to be 15,470 MWD /MTE which is consistent with the value used in the cycle 4 RSE basic assumption.
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No unacceptable conditions were identified.
3.
Cycle 4 Start-up Testing Program The start-up test program was conducted in accordance with Salem 2, Cycle 4 Refueling Test Sequence, Reactor Engineering Manual, Part 200, Revision 4, dated November 26, 1986. The test sequence outlined the steps in the test program, set initial conditions and prerequisites, specified
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calibration or surveillance procedures at appropriate points in the sequence, and referenced detailed tests and data collections in separate test procedures.
Initial criticality of cycle 4 was achieved on December 18, 1986. The Zero Power Physics Testing (ZPPT) was completed on December 19, 1986.
All ZPPT test results met test acceptance criteria. Subsequently, the licensee conducted the Power Ascension Tests, and completed tests on January 27, 1987.
The inspector independently verified that the predicted values and acceptance criteria were obtained from "The Nuclear Design of Salem Unit 2 Nuclear Power Plant, Cycle 4, "WCAP-11218. The inspector reviewed test results and documents described in this report to ascertain the start-up testing was conducted in accordance with technically adequate procedures and as required by Technical Specifications (TS). The details and findings of the review are described in Sections 4 and 5.
4.
Cycle 4 Start-up Testing - Precritical Tests The inspector reviewed calibration and functional test results to verify the following:
Procedures were provided with detailed instructions;
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Technical content of procedure was sufficient to result in
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satisfactory component calibration and test; Instruments and calibration equipment used were traceable to the
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National Bureau of Standards; Acceptance and operability criteria were observed in compliance with
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TS.
The following tests were reviewed:
4.1 Control Rod Drop Time The rod drop measurement was performed in accordance with procedure 2IC-5.2.001, Rev. 3.
Drop times were measured at hot full flow conditions. The inspector verified by review of the test results performed on December 16 and 17, 1986, that Rod Cluster Control Assemblies (RCCA) were tested for drop times and the individual RCCA drop times were all less than 2.2 seconds as required by the T.S.
4.2 Reactivity Computer Setup / Verification The reactivity computer was setup and calibrated according to procedures 2IC-16.4.019. The inspector independently verified that the reactivity computer was adjusted with the correct inputs of
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delayed neutron fractions (betas) and decay constants (lambdas), and noted that the results of this " cold" calibration check were satisfactory.
The reactivity computer was further checked when the reactor reached criticality. Comparisons of predicted and measured reactivities based on doubling time measurement were acceptable.
5.
Cycle 4 Start-up Testing - Post-Critical Tests 5.1 The inspector reviewed selected test programs to verify the following:
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The test programs were implemented in accordance with Cycle 4 Refuleing Sequencing Procedures; Step wise instructions of test procedures were adequately
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provided including Precautions, Limitations and Acceptance Criteria in conformance with the requirements of the TS; Provisions for recovering from anomalous conditions were
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provided; Methods and calculations were clearly specified and the tests
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were performed accordingly; Review, approval, and documentation of the results were in
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accordance with the requirements of the TS and the licensee's administrative controls.
5.2 Boron Endpoint Determination The licensee measured the just critical boron concentration in accordance with Reactor Engineering Manual, Part 15, Revision 2.
The inspector reviewed the data and noted the following results.
Predicted Value Measured Value Configuration (ppm)
(ppm)
All Rods Out (AR0)
1682 1 4%
1640 Test result met acceptance criteria.
5.3 Isothermal Temperature Coefficient Isothermal temperature coefficient (ITC) was measured and documented in accordance with the procedure specified in Reactor Engineering Manual, Part 16, Isothermal Temperature Coefficient Determination, Revision 2.
The inspector noted the following result.
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Predicted Value Measured Value Configuration (pcm/ F)
(pcm/ F)
ARO, Hot Zero Power-5.65 i 3.0-4.65 The measured ITC (AR0) was based on an RCS boron concentration of 1640 ppm, while the predicted value is based on 1682 ppm. The following is the inspector's independent calculation to verify the test acceptance.
The AITC boron adjustment from 1682 ppm to 1640 ppm at RCS average temperature of 547*F is:
AITC = 0.008 pcm/ F/ ppm (1682 - 1640) ppm
= 0.34 pcm/*F Therefore, the predicted ITC value based on the actual test condition is -5.65 - 0.34 = -5.99 pcm/ F.
The difference between the measured and predicted values is -4.65 -
(-5.99) = 1.34 pcm/ F, and this value is within the acceptance criteria.
The calculated Moderator Temperature Coefficient (MTC) from the ITC measurement is:
MTC TS Limit Conditions (pcm/ F)
(pcm/*F)
AR0/HZP/BOL-1.85
<0 Test results were within the acceptance criteria.
5.4 Control Rod Worth Measurement The control rod reactivity worth measurements were performed in accordance with Reactor Engineering Manual, Part 20, Revision 2.
The rod swap technique was used for measuring the bank rodworth rather than the traditional boron dilution method. The following
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results were noted:
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Measured Integral Predicted worth worth Rod Bank (pcm)
(pcm)
D 865.2 798 1 15%
C 1092.9 1082 15%
B 1112.9 1072 15%
-A 857.5 826 1 15%
SD 693.0 621 15%
SC 845.9 744 1 15%
SB 1280.5-1192 i 15%
SA 189.7 205 100 Total Banks 6937.6 6540 i 10%
Test results were within acceptance criteria.
5.5 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 personnel.
The data taken by the Moveable Incore Detector System was digitized and stored by the plant computer.
This information was then fed into a large scale CYBER Computer at Corporate Headquarters which performed the core power distribution calculation using licensee's version of the Westinghouse "Incore" code.
The result from flux maps which were taken to support the cycle 4 power ascension testing are tabulated below:
Flux Map Number:
2401 2402 Power Level (% RTP):
28.3 47.5 Measured TS Limit Measured TS Limit Hot Channel Factor F
2.07 4.54 1.86 4.50 g
Nuclear Enthalpy Hot Channel Factor R
0.80 1.00 0.82 1.00 R
0.80 1.04 0.82 1.01
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m Radial Peaking Factor Top 1.585 1.62 1.433 1.62 Middle 1.573 1.67 1.414 1.67 Bottom N/A 1.63 N/A 1.63 Quadrant Power 1.029 1.049*
1.018 1.043*
Tilt Ratio
- Westinghouse recommended values (No TS limits at these power levels)
Flux Map Number:
2403 2404 Power Level (% RTP):
95.1 99.6
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Measured TS Limit Measured TS Limit Hot Channel Factor F
1,74 2.43 1.70 2.27 q
Nuclear Enthalpy Hot Channel Factor R
0.91 1.00 0.92 1.00 R
0.91 1.00 0.92 1.01
Radial Peaking Factor Top 1.481 1.62 1.475 1.62 Middle 1.494 1.67 1.493 1.67
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Bottom N/A 1.63 N/A 1.63 Quadrant Power 1.011 1.02 1.01 1.02 Tilt Ratio All measured power distribution parameters were within TS limits.
Although the TS has no specific limit for the quadrant power tilt ratio when the power level is below 50% RTP, the measured values from flux maps 2401 and 2402 met fuel vendor recommended values.
5.6 Incore - Excore Calibration Incore/excore calibrations were performed at 47% RTP and again at 96% RTP during power ascension test.
In each case, the licensee used the results from one full core flux map to develop axial flux difference relationship between the incore and excore nuclear instrumentation. The calibrations were completed by I&C personnel as requested in Work Order #8608011964 and #870113064 r
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The licensee's "1-Point" calibration process differed from the normal practice which uses 3 full core flux maps or 1 full core flux map and multi quarter core flux maps. The revised procedure has been reviewed by the SORC (Meeting No.85-124) and an outside independent reviewer, and was found acceptable. However, due to the complexity of this problem and the potential for impact on safe plant operations, the NRR and Region I inspector will further evaluate its technical adequacy. This is an unresolved item (50-311/87-05-01).
5.7 Shutdown Margin Determination (SDM)
The inspector reviewed the licensee's shutdown margin determination procedure Reactor Engineering Manual Part 3 and surveillance rtsults performed during Mode 2 Special Test Exception on December 18 and 19, 1986. The inspector noted that the frequency of evaluation was performed within the requirements as prescribeo by the TS and all calculated SDM values met TS requirement of 2 1.6% AK/K.
However, one of four calculations used an old work sheet. The inspector verified that the rod worth was correctly being used in this calculation, and therefore had no impact on the final calculational result. Through the entire inspection period, the inspector did not have any similar finding and determined that this was an isolated case. Nevertheless, this matter was brought to the licensee management attention during the exit interview.
6.
Fuel Performance Followup Higher than normal reactor coolant activity was synpled in the previous cycle (Cycle 3) operation. Data from radiochemistry trending indicated that one high burnup fuel rod failed early in the cycle life and some more lower burnup fuel -failed later in the cycle.
Based on the visual inspec-tion and ultrasonic examination using the Failed Fuel Rod Detection System (FFRDS) performed during the Cycle 3/4 refueling outage, a total of 5 fuel rods failed in 3 fuel assemblies (P51, P-61, and R-64).
Fuel Assembly R-83 has no failed fuel rod but contained damage in a grid structure.
All damaged fuel assemblies as identified in the fuel inspection were discharged to the spent fuel pit.
Therefore, the beginning of Cycle 4 core contains no known leaking fuel. This is further substantiated by the favorable RCS radiochemistry trending data. The dose equivalent I-131 at the end of Cycle 3 was 5 x 10 2 uCi/cc, while at beginning of this cycle (Cycle 4) it was about 2 x 10 3 uti/cc.
The licensee and fuel vendor (Westinghouse) are still investigating the root cause of the cycle 3 fuel failure. The licensee committed to inform the NRC of the investigation finding _
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At the end of this inspection period, on February 6, 1987, the trended dose equivalent I-131 data showed a significant increase to about 1.5 x 10 2 uC1/cc level. This indicated that some fuel failure had just occurred. The NRC resident inspector will followup this event and review I
the licensee's corrective actions.
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7.
Independent Calculations / Verifications The inspector performed independent calculations / verifications of Cycle 4 startup physics testing related activities. These included the following:
Test acceptance criteria verification as described in Section 3.
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Independent engineering calculations as described in Section 5.
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QA/QC Interface
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Through document review and discussion with a Station QA engineer, the inspector noted that QA/QC had performed surveillance in reactor engineering activities during the Cycle 4 startup testing period.
This included: fuel shuffling witnessing, fuel reconstitution, final core verification, and rod swap reactivity measurement witnessing.
No unacceptable conditions were identified.
9.
Unresolved Item Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations or deviations. The unresolved item identifi q during this inspection is discussed in paragraph 5.6.
10.
Licensee Action on Previous Inspection Findings (Closed) Inspector Follow Item (311/83-29-01): Completion of some I&C works are necessary prior to power ascension.
However, no tracking mechanism within the Refueling Test Sequence procedure could verify the completion of these work orders.
The inspector reviewed the current Refueling Test Sequence procedure, Reactor Engineering Manual Part 200, Revision 4, and noted that the procedure had been revised.
The current procedure contains the provision to verify the completion of these I&C work orders. The cycle 4 startup tests were conducted in accordance with this approved procedure.
This item is closed.
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11. Management Meeting Licensee management was informed of the scope and purpose of the inspection at an entrance meeting conducted on February 3,1987.
The findings of the inspection were discussed with licensee representatives during the course of the inspection. An exit meeting was conducted on February 6,1987 at the conclusion of the inspection (see paragraph 1 for attendees).
At no time during this inspection was written material provided to the 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 information subject to 10 CFR 2.790 restrictions.
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