IR 05000285/1989002
| ML20245J177 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 02/15/1989 |
| From: | Bundy H, Mckernon T, Seidle W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20245J170 | List: |
| References | |
| 50-285-89-02, 50-285-89-2, NUDOCS 8903060195 | |
| Download: ML20245J177 (7) | |
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APPENDIX B'
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
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NRC Inspection Report:
50-285/89-02 Operating License:
DPR-40 Docket:
50-285 Licensee: Omaha Public Power District (OPPD).
1623 Harney Street..
Omaha, Nebraska 68102.*
Facility Name:
Fort'Calhoun Station (FCS).
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Inspection At:
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Inspection Conducted: January 30 through February 3; 1989
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Inspectors:
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H.F.Bundyf)ivisionofReactorSafety Reactor Ins ~pector, Test Programs Date Section, D bD I
s/tr/r 7 T. O. McKernb6, Reactor Inspector, Test Date Programs Section, Division of Reactor Safety Accompanied By:
W. C. Seidle, Chief, Test Programs Section Division of Reactor Safety, (February 2-3, 1989)
Approved:
b 7/////7 W. C. Seidle() Chief, Test Programs Section Date Division of Reactor Safety Inspection Summary Incpection Conducted January 30 through February 3, 1989 (Report 50-285/89-02)
Areas Inspected:
Routine, unannounced inspection of low power physics testing, a complex surveillance, and reactor thermal shield sway test results.
8903060195 890227 PDR ADOCK 05000295 G
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Results: The licensee's nuclear engineering staff exhibited a high level of expertise. The low power physics test results appeared to satisfy acceptance criteria. However, the NRC inspectors' review did reveal a calculational error in moderator temperature coefficient and another inconsistency between the calculations and the procedure review requirements for control element assembly (CEA) worth. The two examples cited above constitued an apparent violation (285/8902-01) indicative of deficient low power test control, in that sufficient result evaluations were not performed and a procedure was not strictly followed (paragraph 2). Records of the licensee's complex surveillance of offsite power loss system (OPLS) indicate the licensee is pursuing the testing and monitoring of this system. A review of the licensee's reactor thermal shield testing and monitoring program showed the-licensee to be monitoring the thermal shield for sway indications and has maintained contact with the vendor and the NRC on this matter.
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DETAILS
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/ Persons-Cont! acted
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10 PPD
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- . J.,H. MacKinnon, Acting Division Manager, Nuclear Operations
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- C. F Simmons, Licensing Engineer o
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- K? R. Henry,; Lead Engineer,. Systems; Engineering.
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- G. Gates,; Manager,'FCS
- M. Guinn,' Supervisor, Reactor Physics
- K.-Holthaus, Manager, Nuclear Engineering-C. Sterba,- Assistant Engineer,s ystems Intervals S
' *L. T;' Kusek, Manager, ' Safety Review Group
- J. T. Smith,' Alternate Manager, Security Service
- F. C. Scofield, Manager, Nuclear Planning-
- R. L L. Jaworski, Manager, Station Engineering
- A. W. Richard, Manager,-Quality Assurance & Quality' Control
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- R.HLewis,-Supervisor, Mechanical Engineering
- J. K. Gasper, Manage'r, Training
- G.-Peterson, Assistant Plant Manager r
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- M. A. Yerdign Nuclear-Management Development
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'W.' Weber,-Reactor Physics Engineer-
- S. K. Gambhir,' Division: Manager, Production Engineering NRC.
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- P. Harre11'LSenior Resident Inspector
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- T Reis,nResident-Inspector $
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- Denotesth'ose individuals' attending the exit inte"rview.
- 2.
St$rtup Testing '- RefuelingV(72700)
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This port, ion of the inspection involved the review of data from specific core physics. test associated with Refueling 11~ (Fuel Cycle 12) for FCS.
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Core physics test data.and other applicable tests were reviewed to verify compliance with NRC. requirements and licensee procedures.
The NRC inspectors re' viewed the foll) wing documents"and associated data:
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Transmittal No. < LIC-88-709, from -R.
L. Andrews to NRC - Subject:
" Application'for Amendment of Operating License," (FCS, Unit 1, Cycle 12 Relcad Evaluation attached) dated September 2, 1988 Letter from NRC to OPPD, "FCS,. Unit No. 1 - Amendment No. 117 to Facility
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Operating License No. DPR-40 (Safety Evaluation Report for Fuel Cycle <12 attached)," dated December 14, 1988
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Startup Procedure (SP)-PRCPT-1, Revision 28, "FCS' Unit No. 1, Post?
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- X Refueling CoretPhysicsTestingandPowerfAscension" k
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., Surveillance; Test Procedure (ST)-CEA-1,' Revision 69, " Control ElementL
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- Assemblies,", dated December 30, 1988
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TheNRCins'pectors'verifiedthit'datafromtheLlowhowerphysics-
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H tests.(LPPT) and the core' power distribution logs taken between 0-30 percent rated power levels were within the stated procedural-acceptance. criteria.
For those control element assemblies tested'in-
.accordance with ST-CEA-1 (normal operating temperature and fullLRCS flow'
, conditions) all' group indication lights'were operable and the! control
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.. elements' dropped within'the Technical Specification.(TS) requirement of.
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2.5-seconds.
In addition to the physics' test discussed in moreldetail-below, the NRC-
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. inspectors noted that the licensee is incorporating.a new vendor computer
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system, Combustion Engineering Reactivity Analysis.(CERA),Linto the LPPT program. The system processes plant signals from reactor cooling system-(RCS) temperature and the. reactivity computer to evaluate and determine
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an isothermal temperature coefficient and control element assembly group worths.
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,a.
Surveillance of Core Power Distribution Limits (61702)
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During'this portion of the inspection, the NRC inspectors verified that the plant was operating within the~ licensed power distribution
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< limits. -The NRC inspectors reviewed the licensee's data analysis code and verified.that the package has been reviewed and approved for
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use. A review of the reactor core flux map, as produced.by the CECOR snapshot, revealed that the licensee was accounting for control element assembly insertions, core power levels, burnups at time of flux mapping,-incore detector sensitivity'as a factor of burnup,
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various uncertainty factors, and flux peaking factors.
The licensee accounted for. anomalous incore detector readings-and, where' applicable, declared specific instrument strings inoperable.
Furthermore, a review of the CECOR snapshot indicated that-the axial shaping index, hot channel factors, azimuthal power tilts, maximum linear heat' rate, total integrated radial peaking factor, and total planar radial peaking factors were within the allowable TS limits and correlated closely with predicted values.
The-NRC inspectors reviewed core-power distribution snapshots for 10,.15, 20, 25, and 30 percent rated power levels'and found.'no' anomalous power distributions.
In addition, the'NRC inspectors conducted interviews with the licensee's key reactor physics personnel.
The personnel appeared very. knowledgeable about the plant's analysis code, the core's fuel cycles data history, capabilities of the code, and improvements currently being implemented, i
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b.
Isothermal and Moderator Temperature Coefficient Determinations.
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(61708)
During this portion of the inspection, the NRC inspectors verif'ie'd?
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.the licensee's determination of moderator: temperature'
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coefficient (MTC) and reviewed isothermal temperature.
coefficient (ITC) comparisons'with predicted values andLTS
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requirements.
The NRC. inspectors verified'that the licensee-satisfactorily fulfilled ITC measurement test precautions = and P
> prerequisites. Actual' plant conditions. established during the test-
were the~same as those assumed in the analytical predictions. :The
licensee correctly calculated ITC from reactivity and RCS temperature-traces taken for-dual'RCS heatup:and cooldown cycles, performed.at hot-
.zero power (HZP).
The ITC value determined was consistent with:the
. predicted value and within the TS requirements.
During. review of:
manual' calculations'of MTC, the NRC inspectors noted that the licensee.had used an incorrect fuel temperature coefficient'(FTC).
The FTC used in the MTC at HZP calculation was'-1.47 X;10 5 Ak/k/ F (Full power:FTC) in lieu of -1.97 X 10 5' Ak/k/*F (HZP FTC).
This calculational error resulted in the nuclear engineering. staff.
presenting the. plant review committee (PRC) a' maximum positive.MTC
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'value.which erroneously represented the margin of safety to the TS3
. limit: of <.5 X 10 4 Ak/k/ F...The correct MTC value was.
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.493 X 10 4 Ak/k/ F instead of the reported.44 X 107 Ak/k/ F.
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. calculational error represented a change in margin of safety to the TS' limit ofL10 percent (i.e.,'~11.4 percent' reported versustan" actual 1.4 percent margin). 'In.further review of the' licensee'siprocedure-(SP-PRCPT-1)', the NRC inspectors notedsthe procedure did not require
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an, independent verificatiori of input data 'used in either manual it. calculations' or a computer software data reduction program.
The failure to-sufficiently evaluate' test 'results to ~ detect' the MTC error is L an example ~ of an apparent ' violation of regulatory requirements, 10 CFR 50, Appendix B, Criterion XI, " Test Controls.".(285/8902-01)
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Control Rod Worth Measurements (61710)
The control element assembly (CEA) group' worth measurements were performed utilizing the rod group exchange technique in accordance.
with Section 6 of SP-PRCPT-1.
Integral to this method was determina-tion of Reference Groups A and 2 worths by boration-dilution.
Co11ectinn'and reduction of data conformed to the requirements of SD-PRCPT-1.
The NRC inspectors verified selected portions of the results by independent data reduction and evaluation.
However, in comparing the results to the review criteria listed in Appendix D of SP-PRCPT-1 as required by Step 11.a, the NRC inspectors noted that the licensee's
. calculations did not express the individual group and total regulating group worths as percentage variations from the predicted values.
In fact, all licensee calculations referenced to the review
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-6-criteria compared the deviations with respect to measured values in
accordance with a formula given in vendor document,E NPSD-366,
" Verification of Control Rod Integrity and Fuel Symmetry - Task 478, November 1987."
Acceptance of the test data whithout completed calculations to compare the percentage deviations from predicted values for CEA group.
worths, in accordance with Step 11.a and Appendix D of SP-PRCPT-1, is a second example of an apparent violation of 10 CFR 50, Appendix _B, Criterion XI, which requires evaluation of the test results, in accordance with the test procedure.
(285/8902-01)
Two examples.of one. apparent violation are discussed in subparagraphs 2.b and 2.c above.
No further violations'or deviations were identified.
3.
Complex Surveillance (61701)
During this portion of the inspection, the NRC inspectors reviewed the licensee's completed records for testing the offsite power. low system (OPLS). The purpose of this surveillance was to verify:
adequate station. voltage exists to safely shut down the reactor, correct operation of.the station voltage sensors, proper sensor actuation point, and OPLS functions.
The NRC inspectors verified that the surveillance was performed with a currently approved version of the procedure and was consistent'with licensee commitments and administrative controls.
The test personnel used test equipment calibrated and signed off on procedural steps _where required.
Quality control and independent verification signoffs were accomplished and any test anomalies were dispositioned.
The surveillance.was completed within a few days with all test objectives satisfied.
In addition, the NRC inspectors verified that: key test personnel training qualifications were adequate, r
No violations or deviations were identified.
4.
Reactor Thermal Shield Sway Test (92701)
During this portion of the inspection, the NRC inspectors reviewed the licensee's program for internal reactor vessel monitoring (RVM) and results from recent startup testing.
The following procedures and documents were reviewed:
Preventive Maintenance Procedure PM-VLPM-3,. Revision 1, " Thermal Shield / Core Support Barrel Motion," dated June 18, 1987 Transmittal No. LIC-86-421, from R. L. Andrews to NRC - Subject:
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" fort Calhoun Thermal Shield Support System Inspection Referral,"
dated August 28, 1986 Guidelines for Surveillance Monitoring for Reactor Intervals (CEOG Task 524), dated April 1986
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' Transmittal No. LIC-88-091,. from R. L. Andrews to NRC - Subject:
" Threshold; Levels for the Fort Calhoun Internals Vibration Monitoring System," dated February 25, 1988 During this review,;the NRC inspectors verified that the licensee's internals vibration monitoring (IVM) data did not exceed anticipated threshold levels.
The licensee has established a program. for quarterly monitoring of reactor vessel, internals vibration in addition to monitoring during startups after-refuelings. 0utput current = signals from various upper and lower level excore neutron detectors are inputted to separate channel amplifiers, filters, and output amplifiers with the resultant output signal characterized as a frequency modulated noise signal.- These noise signals are analyzed in'five separate spectral analysis tests, (power spectral density, cross power spectral density,. coherence, phase, and phase separated cross power spectral density).
The resultant spectral analysis plots compare a relative noise /
signal ratio to various frequencies which when analyzed can indicate thermal shield beam motion or cosine two-theta motion.
Thermal shield beam motion is indicative af shield motion along the analyzed signal path and is characterized by a:nosinal 7 hertz noise signal.
Cosine two-theta-thermal
~ hield motion is indicated by shield motion.which elongates the shield s
perpendicular to the signal's path and is characterized _by a nominal.12.5 hertz noise signal.
Should neutron detector noise signal increases show changes indicative of a loss of effectiveness of the thermal shield positioning pins, then the frequency of monitoring IVM data would be increased to one-month intervals.
Should subsequent data indicate the loss of effectiveness'of positioning pins and this data is corroborated by the loose parts monitoring data, then the plant will be shut down and a physical inspection of the thermal shield performed.
The licensee has submitted the IVM program to the NRC and anticipates that should analysis data show no appreciable movement of the thermal shield, then a relief to the 10 year inspection program may be requested.
To date, analysis has shown no noticeable motion of the thermal shield or other reactor' internals.
Discussions with the systems engineer responsible for IVM testing led NRC inspectors to conclude that he is knowledgeable of both the theory and application involved with IVM.
No violations or deviations were identified.
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Exit Interview The NRC inspectors met with the licensee representatives denoted in paragraph 1 on February 3, 1989, and summarized the scope and findings of this inspection.
The licensee did not identify as proprietary any of the materials provided to, or reviewed by, the NRC inspectors during this inspection.
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