IR 05000302/1976013
| ML19308D462 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 08/24/1976 |
| From: | Jape F, Robert Lewis, Rogers R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19308D459 | List: |
| References | |
| 50-302-76-13, NUDOCS 8002280774 | |
| Download: ML19308D462 (7) | |
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UNITED ' STATES
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NUCLEAR REGULATORY COMMISSION jy
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230 peAcwines sinasT, N. W. sulTE 01e g
AT1.ANTA, GEo nGI A 30303
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IE Inspection Report'No. 50-302/76-13
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Licensee: Florida Power Corporation 3201 34th Street, South P. O. Box 14042 St. Petersburg, Florida 33733 Facility Name: Crystal River 3 Dockat No.:
50-302 License No.:
CPPR-51 Category:
Location: Crystal River, Florida
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Type of License: B&W, PWR, 2452 Mut
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Type of Inspection: Routine, Unnnnnunced Dates of Inspection: August 3 6, 1976 Dates of Previous Inspection: June 29 - July 2, 1976 Inspectior-in-Charge:
R. F. Rogers, Reactor Inspector Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch Accompanying Inspectors:
T. N. Epps, Reactor Inspector Reactor Projects Seci: ion No. 2 Reactor Operations and Nuclear Support Branch
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Other Accompanying Personnel:
W.' J. Ross, Project Manager i
i Division of Operating Reactors Offica of Nuelaar Reactor Regulation Principal Inspector:
4( 6 Jo h 70[7d
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F. Jape, Reactor Inspe4 tor Date.
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Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch e de JN[o.
Reviewed by:
R. C. Lewic, Chief Date Reacton-Projects Section No. 2
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Reactor Cperations and Nuclear Suppo:t Branch
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IE Rpt. No. 50-302/76-13 *
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t SUMMARY OF FINDINGS -
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I.
Enforcement Matters None II.
Licensee Action on Previously Identified Enforcement Matters
Not inspected.
. III. New Unresolved Items
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I Nona e
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Status of Previously Identified Unresolved Items l
i Not inspected.
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Unusual Occurrences
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None
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Other Significant Findings
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None
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VII. Management Interview
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A management interview was held on August 6,1976, with J. Alberdi and a
members.of his staff. The inspection findings relating to preoperational and power ascension testing were discussed.
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IE Rpt. No. 50-302/76-13 I-1
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DETAILS I Prepared by:
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R. F. Rogers, Reactor pfspqdtor g Date'
, Reactor Projects Section No. 2
' Reactor Operations and Nuclear i
Support Branch
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Dates of Inspection: August 3-6, 1976
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Reviewed by: [.d.
M M78, R. C. Lewis, Chief Data
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Reactor Projects Section No. 2
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Reactor Operations and Nuclear Support Branch
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1.
Personnel Contacted
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J. Alberdi - Project Manager C. P. Beatty - Plant Superintendent J. C. Hobbs, Jr. - Manager Generator Testing
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D. W. Pedrick, IV - Compliance Engineer
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3. E. Froats - Manager, Site Survaillance J. Heilman - Reactor Operator 2.
Preoperational Test Results Evaluation Completed Category I preoperational casts were reviewed to assure that the licenses had performed ar. adequate review of test results,
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the test results were within tcablished acceptance criteria, a method of identifying and correcting test deficiencies was being followed, and administrative control practices were being followed in test results documentation.. The following emergency power srates tests were reviewed.
TF 71-451-5, Emergency Diesel Generators Functional Test
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TP-71-451-6, DC Power System Electrical Test TP-71-451-12,. SKY ES Switchgear Electrical and Functional Test The inspector had no further questions.
3.
Review of Initial Fuel Loadinst Procedure FP 202
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A revised drafe of this procedure was reviewed for cr dormance to regulatory guide 1.68, Appendix C, Section B, "Fue3 T.wding" and FSAR Section 13.4.5, " Initial Reactor Fuel Loadinr,." Radiation
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IE Rpt. No. 50-302/76-13 I-2
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control practices and precautions were defined and actions to be taken in event of damaged fuel were included. These areas were
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discussed in Report 76-6.
Step 9.3 of the procedure does not
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specify who has responsiblity for ordering new fuel assemblies lowered into the core. Also, the prerequisites Section 7.1.9 does not require that system valva line-ups be verified to support the fuel loading. RG 1.68, Appendix C, Section B.1.a requires that the
status of all systems should be specified and in readiness as
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specified prior to loading fuel. A licensee representative stated
that these concerns would ba addressed prior to approval of the
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peccedure.
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i, 4.. Review of Reactor Building Sump Flow Test TP 71-310-10 The reactor bn41r14ag sump flow test was performed on June 4, 1976,
.and the inspector reviewed the completed test package. Each spray and low pressure injection pump was operated approximately 15 minutes entrine a suction on the HB sump. A physical inspection of
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the RB sump. indicated that the mesh screens described in FSAR.
Section 6.2.2.1 were no't installed. A licensee representative stat;ed that the screens would be installed prior'to fuel loading.
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This item will be followed up on a future inspection.
5.
Plant Tour
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An inspection tour of the plant was conducted. It appeared that the licensee's programs and proceduras for routine housekeeping activities required further attention. Loose construction materials,
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debris, and dirt were evident in most areas. Rwm=4nscion of these
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areas will be conducted on future inspections.
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IE Ryt. No. 50-302/76-13 II-1
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!r DETAILS II Prepared by:
[- W M T. N. Epps, R g tor Inspector Dsta
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Reactor Projdcts Section No. 2 Reactor Operations and Nuclear Support Branch Dates of Inspection: August 3-6, 1976
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Reviewed by:
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R.,C. Lewis, Chief Data Reactor Projects Section No. 2
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Reactor Operations and Nuclear Support Branch 1.
Individuals Contacted
J. Alberdi - Project Manager G. P. Esatty - Nuclear Plant Superinter. dent r
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J. C. Hobbs - Manager, Generation Testing f
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D. W. Pedrick - Compliance Engineer W. R. Nichols - Operations Supervisor
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D. H. Ruzic - Results Engineer 2.
Power Ascension Test Procedures a.
Pseudo Ejected Rod Test
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Approved test procedure TP-7-1-800-19-0, " Pseudo Ejected. Rod
. Test," was reviewed to determine if the test procedure included appropriate review, approval, acceptance criteria, test conditions and initial conditions. The test procedure conforms to Regulatory Guide 1.68, Appendix C, " Preparation of Procedures."
Itam 11 of page 13-50 of the PSAR anecifies that this :sst will be run at 0 and 75 percent pot.et vF' % differs hon a stacament in TP-7-1-800-19-0 that specs.
. 40 t 2 percent.
A site representative stated that revuius to the PSAR has
.i been sent to the corporate office to correct this conflict.
Issus data of the FSAR was not determined. Followup on this
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,ccueene will be conducted on future inspections. The inspector discussed and resolved a question on TP-7-1-800-19-0 as'to whether the operating procedure for datam4*ime boron concentration,
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for shutdown margin, included having the most reactive rod withdrawn. A licensee representative stated that the Crystal River procedure for shutdown margin calculation ex udes the
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worth of the most reactive rod.
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IE Ryt. No. 50-302/76-13 II-2
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b.
Core Power Distribution
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Approved test procedure TP-7-I-800-11-0 was reviewed. 'One purposia of this test is to compare measured and c.1,ntated
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j Power distributions. The test procedure, however, had no
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acceptance criteria for avial y=We factors. Site personnel
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stated that this subject would be reviewed. This item will be
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reviewed during a. future inspection.
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c.
Loss of Electrical Lead Approved test procedure TP-7-1-800-13-0 was reviewed. It has
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been observed at other operating B W units that an unanticipated loss of electrical load above certain power levels will cause a reactor trip. Acceptance criteria in this test procedure
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cover cases if the reactor trips and if it does not. The
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licensee agreed to reference the operating procedure to use, j
if the reactor trips, to restore the plant to normal conditions.
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The izfspisctor had no further questions on this procedure.
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d.
Loss of Offsite Power Approved test procedure TP-7-1-800-26 " Loss of Offsite Power,"
we.a reviewed. One purpose of this test is to verify proper operation of the emergency turbine driven feedwater pump. The
inspector stated that the test procedure addressed operation
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,l of the emergency feedwater pump, but the acceptance criteria
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did not include pump operating requirements. Licensee personnel stated that anergency feedwater pump operating requirements
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would be added to acceptance criteria. Followup will be l
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conducted during a future inspection.
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Reactor Prote:rion System ij The inspector participated in a tour of the facility. Part of the
.I cour involved observations of reactor protection system (RPS)
equipment and discussions with plant personnel.
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Figure 15-4 of the PSAR indicates g design value for reactor coolant systemt (RCS) flow rate of 131 X 10 pounds per hour. Licensee personnel stated that hot functional tests were conducted with a
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i flow restricting device in the reactor vessel, to simulate core pressure drop, and reactor coolant flow was significantly higher
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than th? design flow rate. The inspector stated that other operating e
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IE Rpt. No. 50-302/76-13 II_3
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B&W units had experienced flow rates in excess of design values and that this caused the nuclear overpower based on RCS flow and axial power imbalance trip setpoint to be in error in the nonconservative direction, due to reactor coolant sgstem pressure transmitters indicating 100 e
percent flow at 131 X 10 pounds per hour rather than,the actual value.
Licensee personnel agreed to r3 view RCS pressure transmitter calibration to determine if a similar trip setpoint error could occur at Crystal River. Folloup on this matter will be conducted on a future inspection.
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