IR 05000302/1980032
| ML19345D554 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 10/30/1980 |
| From: | Burnett P, Falconer D, Ford E, Quick D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19345D553 | List: |
| References | |
| 50-302-80-32, NUDOCS 8012160041 | |
| Download: ML19345D554 (4) | |
Text
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101 MARIETT A ST., N.W., SUITE 3100 U
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ATL ANTA, GEORGI A 30303 Report No. 50-302/80-32 Licensee: Florida Power Corporation 3201 34th Street, South St. Petersburg, FL 33733 Facility Name:
Crystal River Unit 3 Docket No. 50-302 License No. DPR-72 Inspection at Crystal River plant site near Crystal River, FL Inspectors:
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Approved by:
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/0 - 3C "[0 D'. R. Quick, Section Cniel, RONS Branch Date Signed SUMMARY Inspection on September 8-12, 1980 Areas Inspected This routine, unannounced inspection involved 96 inspector-hours onsite. The areas inspected included pre-critical activities and initial criticality for cycle 3, zero power physics tests, power escalation tests, and LER followup.
Results Of the four areas inspected, no items of noncompliance or deviations were identi-fled.
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DETAILS 1.
Persons Contacted Licensee Employees
- M. E. Collins, Reactor Specialist
- W. A. Cross, Operations Engineer
- S. W. Johnson, Maintenance
- K l. Lancaster, Nuclear Compliance Supervisor
- E. K. Neushaefer, Nuclear Compliance Auditor
- G. M. Williams, NQA/QC Supervisor J. E. Barrett, Engineer NRC Resident Inspector
- T.
Stetka B. Smith
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summarized on September 12, 1980 with those persons indicated in Paragraph 1 above.
3.
Licensee Action on Previous Inspection Findings Not inspected.
4.
Unresolved Items Unresolved items were not identified-during this inspection.
5.
Pre-Critical Activities and Initial Criticality The inspector reviewed the following completed procedures:
PT-100, "Controling Procedure for Pre-Critical Testing". The requirement a.
in this procedure to perform PT-101, "RC Flow and Coastdown Test", was deleted pending later installation of pump-power monitors.
b.
SP-102,_ " Control Rod Drop Times". Review of Data Sheet I of this proce-dure indicated that all rods met the requirements of Technical Specifi-cation 3.1.3.4.
Detailed review and analysis.of recorder traces for the drop-time tests of four of the seven rod groups confirmed the values determined by the licensee for the fastest and slowest rods in each group.
6.
Zero Power Physics Testing The inspector reviewed the following procedures for technical adequacy:
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PT-110, " Controlling Procedure for Zero Power Physics Testing", Rev.
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This included a review of the data and charts recorded during the deboration to criticality as well as the reactimeter checkout and its calibration.
b.
PT-116, " Determination of Sensible Heat", Rev.1.
c.
PT-111, "All-Rods-Out-Boron-Concentration-Determination".
d.
SP-110, " Reactor Protective System Functional Testing", Rev. 28. This consisted of a review of procedural instructions and data on pages 35 through 42 and pages 79 through 82.
PT-112, Rev. 4, " Hot Zero Power Regulating Rod Group Worth and Differ-e.
ential Boron Worth Measurement".
f.
PT-114, lev. 7, " Moderator and Temperature Coefficients Determination at Hot Zeco Power".
g.
SP-42, Rev.12, " Reactivity Balance Calculations".
h.
SP-420, Rev. 1, " Control Rod Worth and Shutdown Margin Physics Testing Surveillance".
No questions arose from the technical review of these procedures.
I Zero-Power Physics Test The zero power physics test was performed to verify the nuclear design parameters used in the safety analysis and the technical specifications.
Performance testing procedure PT-110, "Zero-Power Physics Testing", pre-scribed the order in which various cycle 3 zero power physics tests were performed and the procedures by which they were performed. This inspection confirmed that the tests specified were performed in the sequence specified, given initial review, and met the established acceptance criteria. The following completed zero power physics tests were reviewed:
Control Rod Calibration Performance Testing Procedure PT-112, " Hot Zero-Power Regulating Rod Group Worth and Differential Boron Wo-th Measurement" was used to determine the hot zero power, cycle 3 integral and differential reactivity worth for Control Rod Assembly (CRA) Groups 5, 6 and 7.
Procedure PT-112 uses the boron-swap method to determine rod group reactivity worth ~. This method
sets a deboration rate and compensates for the change in reactivity by small step changes in rod group position. The calculation of reactivity is
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made on a continuous basis by a Babcock and Wilcox reactimeter.
The measured reactivity worth of CRA Groups 5, 6 and 7 each met the accept-ance criterion of i 15% of the predicted valve. The total reactivity worth l
of the sum of CRA Groups 5, 6 and 7 met the acceptance criterion of i 10% of the predicted value.
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Boron Worth Measurement Procedure DT-112 was also used to determine the hot, zero power, cycle 3 differential boron worth. The differential boron worth was calculated by dividing the change in the measured reactivity worth of the inserted CRA groups between initial and final critical positions of the group worth measurements by the corresponding change in boron concentration.
The measured differential boron worth met the acceptance criterion of i 15%
of the predicted value.
Temperature Coefficient Determination Performance testing procedure PT-114, " Moderator and Temperature Coefficients Determination at Hot Zero Power", was used to determine the cycle 3 isothermal temperature coefficient at hot zero power. The moderator temperature coef-ficent was determined by subtracting the calculated dop,nler coefficient of-2pcm/*F from the measured isothermal temperature coefficient. The measured hot zero power temperature coefficients met the established acceptance criteria.
Shutdown Margin Calculations Surveillance Procedure SP-421, " Reactivity Balance Calculations", was used to determine the shutdown margin prior to raising reactor power above 5%
following cycle 3 hot, zero power physics testing in accordance with Tech-nical Specification 4.1.1.1.1.1.
.Two shutdown margin calculations were perforned per procedure SP-421 section 6.1.3, " Shutdown Margin Calculation" The measured shutdown margins met the conditions and actions prescribed by Technical Specification 3.1.1.1.1.
7.
Power Escalation Tests PT-120, " Power Escalation Testing", was used to control the post uro-power tests. The portion of this procedure and results addressed to calibration of excore detectors versus incore detectors was reviewed in detail and acceptable results confirmed. No recalibration of detectors was required.
Those data sheets entitled " Power Doppler Coefficient Determination" of PT-120 were used to determine the 100% full power power doppler coefficient, calculations.
This method derives - the coefficient by multiplying the change in control rod assembly positions from power level at time 1 to j
power level at time 2 times the average differential road worth. This inspection verified the results and computations. No items of non-com-pliance were identified.
Surveillance Procedure SP-104, " Hot Channel Factors" was reviewed for power distribution maps obtained at 40, 74, and 98 percent of rated themal power. Some of the heat balance results, SP-312, for this period of testing were also reviewed. No items of noncompliance were identified.
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LER Followup LER 80-005 was reported on February 8, 1980, for an event dientified on January 16, 1980.
The event was a missed, eighteen-month incerval, sur-veillance test.
Calibration of the incore neutron detector system should have been performed no later tht.4 July 30, 1979, but was not comp eted
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until January 14, 1980.
The required calibration is accomplished by performance of SP-140, "Incore Detector Surveillance" The inspector reviewed the completed copies of SP-140 in the licensee's files. For the version completed on January 14, 1980, the initial conditions were signed-off as satisfied on October 29, 1979. Other steps in the procedure and enclosed computer printout sheets were dated between November 2, 1979 and November 13, 1979. It follows then that the test is reflective of system conditions on October 29, 1979 not January 16, 1980.
During the exit interview the inspector informed the licensee that the next surveillance should be within eighteen months of the earlier date. That is no later than April 29, 1981.
9.
Miscellaneous Activities The feasibility of nondestructive examination of the grippers on the fuel handling machines was discussed with licensee personnel. Af ter a discussion with the machine vendor their initial opinion was that such tests were pos-sible; however, the machine in the spent fuel pool will be inaccessible for such examination once the pool is flooded.
In the context of ANSI B.30.2 the grippers are analogous to books on the special purpose fuel handling crane.
Other discussions addressed the search for dropped control rod rodlets in an operating core. Licensee personnel in this discussion were not aware of such occurrences at other facilities of a different NSSS vendor. Nor were they aware of any guidance from their vendor.
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