IR 05000302/1977003
| ML19308D224 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 03/11/1977 |
| From: | Jape F, Robert Lewis, Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19308D195 | List: |
| References | |
| 50-302-77-03, 50-302-77-3, NUDOCS 8002270684 | |
| Download: ML19308D224 (11) | |
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UNITED 51 ATES
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230 PEACHTREE STREET, N.W. SulTE 818
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IE Inspection Report No. 50-302/77-3 Licensee: Florida Power Corporation 3201 34th Street, South P. O. Box 14042 St. Petersburg, Florida 33733 Facility Name: Crystal River 3 Docket No.:
50-302 License No.:
DPR-72 Category:
i Location:
Crystal River, Florida Type of License:
B&W, PWR, 2452, Mwt Type of Inspection:
Routine, Unant.aunced Dates of Inspection:
February 22-25, 1977 V
Dates of Previous Inspection: January 10-14, 1977 Principal Inspector:
F. Jape, Reactor Inspector Accompanying Inspectors:
T. N. Epps, Reactor Inspector J. D. Martin, Reactor Inspector Other Accompanying Personnel: None Principal Inspector:
[.C M b 8////77 F. Jape, R'eactor Inspe6for Date Reactor Project Section No. 2 Reactor Operations and Nuclear Support Branch Reviewed by:
[. 6.
8////[7 R. C. Lewis, Chief Date Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch O
8002270
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IE Rpt. No. 50-302/77-3-2-
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SUMMARY OF FINDINGS I.
Enforcement Matters Infraction Contrary to Technical Specification 6.9.1, " Routine Reports and Reportable Occurrence Reports" the notification of a deficiency by Babcock and Wilcox Company to Florida Power Corporation on December 30, 1976, regarding a Westinghouse Pressure Transmitter, Model No. 59PH, was not reported to the NRC.
(Details I, paragraph 2)
II.
Licensee Action on Previously Identified Enforcement Matters Deficiency Licensee's response dated February 21, 1977, concerning failure to maintain revision records current has been received. Followup on this item will be conducted on a future inspection.
III. New Unresolved Items 77-3/1 Offsite Organization Plans are underway to change the offsite organization.
An amendment to Fig. 6.2-1 of the Technical Specifications has not been submitted.
(Details I, paragraph 3)
IV.
Status of Previously Identified Unresolved Items UN 77-2/1 Plant Review Committee Followup revealed that this item is partially resolved.
Additional folicwup required to completely resolve.
Item remains open.
(Details II, paragraph 2)
UN 77-2/2 Modification Approval Record Followup revealed that the licensee has revised the modification approval record form and now requires testing to be specifically identified on the MAR.
Item is closed.
(Details I, paragraph 4)
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IE Rpt. No. 50-302/77-3-3-
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77-2/3 Equipment Status and Clearance Records
Followup inspection revealed that the' control center status board and the equipment clearance log index are being kept current. This item is closed.
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(Details II, paragraph 3)
V.
Unusual Occurrences l
None VI.
Other Significant Findings j
l None VII. Management Interview A management interview was held February 25, 1977, with G. P. Beatty, Jr.
and J. Alberdi. The inspection findings, which included the item of noncompliance on failure to report a reportable occurrence and the new unresolved item, were discussed.
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IE Rpt. No. 50-302/77-3 I-l
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DETAILS I Prepared by: [, d.
b 8[#/77 F. Jape,' Reactor Inspector Date Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch Dates of Inspection: February 22-25, 1977 Reviewed by: [. 6.
J////77 R. C. Lewis, Chief Date
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Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch 1.
Individuals Contacted Florida Power Corporation (FPC)
J. Alberdi - Project Manager G. P. Beatty, Jr. - Nuclear Plant Superintendent J. C. Hobbs, Jr. - Manager Generation Testing
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D. W. Pedrick, IV - Compliance Engineer
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W. R. Nichols - Operations Supervisor G. R. Westafer - Technical Support Engineer D. H. Ruzic - Results Engineer S. Woods - System Test Engineer H. E. Dumas, Jr. - Testing Superintendent 2.
Failure to Report a Reportable Occurrence The B&W Company notified FPC, by letter dated December 30, 1976, of the lack of adequate test qualification data under post LOCA condi-tions for the Westinghouse 59 PH pressure transmitter used for input to the Decay Heat Interlock, (DHI).
The DEI prevents opening of the decay heat system isolation valves within the reactor building when the reactor coolant pressure exceeds the operating pressure of the decay heat system, and ensures automatic closure of these valves if these pressure conditions exist.
The 59 PH transmitter is required to operate in an accident environ-ment for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the DEI application. The B&W letter states that the test results on this transmitter do not support this requirement. The corrective action recommended in the B&W report regarding this problem, which was enclosed with the December 30, 1976, letter is to replace the transmitter with one that is
qualified for the application.
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IE Rpt. No. 50-302/77-3 I-2
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Technical Specification 6.9.1, " Routine Reports and Reportable Occurrence Reports," states that RG 1.16, Rev. 4, shall be used to determine reportability of events. Review of the problem reveals that part 2.a.(9) of RG 1.16 requires the licensee to submit a prompt notification with written followup in 14 days. Failure to report the problem is in noncompliance with TS 6.9.1.
3.
Offsite Organization Discussions with licensee management revealed that changes to the offsite organization are currently being made.
Plans for reorgani-zation are not firm and licensee management stated that an ammendment to the Technical Specifications (TS) to change Figure 6.2-1 has not yet been submitted to NRR. The inspector commented that changes to the organization as depicted in Figure 6.2-1 of TS 6.2 should be coordinated with a TS ammendment authorizing the new organizational s tructure. Followup on this item will be conducted on a future inspection.
The item has been identified as unresolt ed item 77-3/1.
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4.
Modification Approval Record During inspection 50-302/77-2, a discrepancy was identified involving checkout and testing following completion of work. Followup revealed that procedure CP-ll4, " Procedure for Control of Permanent Modifi-cations. Temporary Modifications and Deviations," has been extensively revised to resolve this item. The test requirements have been expanded, and the required testing will be explicitly stated on the modification approval record. Unresolved item 77/2-2 is clos 2d.
5.
Followup on Licensee Event Reports (LER)
The following LER's were reviewed for accuracy, safety significance, completeness of corrective action and to verify that the Plant Review Committee (PRC) had reviewed the event as required by TS.
The findings are sunnarized below for each event:
LER 77-7 Engineered Safeguards Activation This event was reviewed by the PRC on January 6, 1977 and the report was as delineated by RG 1.16.
The corrective action stated in the licensee's report dated February 2,1977, is to correct the error on the drawing. The drawing change has not been completed; therefore this item will be reviewed on a subsequent inspection.
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IE Rpt. No. 50-302/77-3 I-3 LER 77-10 Failure to Place Pressurizer Pilot Operated Relief t
Valve in Low Range During Cooldown
- The PRC reviewed this event on February 11, 1977, and the report was f ound to be as required by RG 1.16.
Corrective action, given in the licensee's report dated February 11, 1977 was verified to be complete.
50.55(e) Report on Snubbers on the RC Pressure Sensing Lines This event is described in the licensee's letter dated December 21, 1976. Corrective actions were verified as complete.
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IE Rpt. No. 50-302/77-3 II-1 DETAILS II Prepared by:
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- [ */ 7 J. D. Martin, Reactor inspector
/Date Nuclear Support Section Reactor Operations and Nuclear Support Branch
Dates of Inspectio : February 22-25, 1977 C
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Reviewed by:
H. C. Dance, Acting Chief
'Date Nuclear Support Section Reactor Operations and Nuclear Support Branch 1.
Personnel Contacted Florida Power Corporation (FPC)
G. P. Beatty, Jr. - Nuclear Plant Superintendent P. F. McKee - Assistant Nuclear Plant Superintendent
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W. R. Nichols - Operations Supervisor D. W. Pedrick, IV - Compliance Engineer
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P. E. Griffith - Training Coordinator 2.
Plant Review Committee
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Followup inspection of previously identified unresolved item 77-2/1 was accomplished by reviewing administrative procedure AI-300, reviewing minutes of seven recent Plant Review Committee (PRC)
Meetings and through discussion with the plant operations staff.
The (PRC) minutes were reviewed for agreement with the requirements of Technical Specification (TS) 6.5.1.
The following (PRC) meeting minutes were reviewed:
Number Date of Meeting 77-1 1/6/77 77-2 1/13/77 77-3 1/20/77 77-4 1/27/77 77-5 1/28/77 77-6 2/3/77 77-7 2/11/77
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IE Rpt. No. 50-302/77-3 II-2
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Based on the review of the above documents the current status of the four previously identified issues is as follows:
a.
Alternate Appointment
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The method now being used to designate alternate appointments
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to the (PRC) is in agreement with the requirements of TS 6.5.1.3.
This item is closed.
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Members in Attendance Via Telephone
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The licensee has commi:ted to a limited use only in emergency situations with a conference call (speaker phone) arrangement for a member not physically present at a PRC Meeting. This iten is closed.
c.
PRC Responsibilities and Authorities The lack of complete agreement between AI-300 and T.S. 6.5.1.6 was corrected with a procedure change to AI-300, item i part 6.0.
This item is closed.
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Items 2. and 3. remain unresolved until a T.S. change is made.
(New unresolved item 77-3/1)
d.
Use of Review Consideration Form The licensee has decided to eliminate the use of this form.
This item is closed.
3.
Equipment Status and Clearance Records Followup inspection of previously identified unresolved item 77-2/3 was accomplished by reviewing administrative procedure Al-500 and inspecting the control center status board and equipment clearance log index located in the control room. The status board and log index were updated and therefore this item is closed.
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4.
Followup on Licensee Event Reports (LER's)
The following LER's were reviewed to ascertain that the report of the event and associated conditions are adequate and in conformance with regulatory requirements. The LER's were reviewed for a.: curacy, safety significance, reporting requirements and verification that appropriate corrective action was taken.
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IE Rpt. No. 50-302/77-3 II-3
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LER 77-2 Makeup Pump 3C Started with Suction Valve Closed
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This event was reviewed by discussions with the Operations Supervisor and observation of the pump and piping configuration. The reported corrective action was considered appropriate and was verified to be complete.
- LER 77-3 Makeup Pump 3C Suction Valve Spuriously Tripped Closed This event was reviewed by discussions with the Operations Supervisor and review of Nonconforming Operations Report No. 76-33.
The results of this event required replacing the rotating element of the pump. The corrective action and subsequent retesting was considered appropriate and was verified to be complete.
LER 77-4 Valve DHV-42 Inadvertently Opened The control switch for DHV-42 was placed in the open position permitting a flow path from the Reactor Coolant S; stem to the Reactor Building Sump.
The valve was closed upon discovery.
This event was discussed with the Operation Supervisor and the corrective action taken was considered appropriate and complete.
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IE Rpt. No. 77-3 III-l
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DETAILS III Prepared by:
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T. N. Ipps, Rey 6tof Inspector Date Reactor Projects Section No.1 Reactor Operations and Nuclear Support Branch Dates of Inspection: February 22-25, 1977 J!^'bi Reviewed by:
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H. C. Dance, Chief Date Reactor Projects Section No. 1
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Reactor Operations and Nuclear Support Branch
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1.
Individuals Contacted G. P. Beatty, Jr. - Nuclear Plant Superintendent P. F. McKee - Assistant Nuclear Plant Superintendent W. R. Nichols - Operations Supervisor D. W. Pedrick, IV - Compliance Engineer J. R. Wright - Chem / Rad. Protection Engineer J. L. Hatrison - Assistant Chem / Rad Protection Engineer J
P. E. Griffith - Training Coordinator 2.
Plant Operations A plant tour was conducted on February 23, 1977, including the auxiliary building, turbine building and control room during unit heatup.
Monitoring instrumentation for RCS temperature and pressure indi-cated that unit heatup was being conducted in accordance with Technical Specification (TS) Table 3.4-2, and the controlling procedure for unit heatup was being followed.
No radiation control problems were observed and no adverse house-keeping, fluid leak or pipe vibration conditions were identified.
The inspector verified that the unborated makeup and purification
demineralizer was properly tagged to prevent inadvertent operation.
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The inspector also verified that control room manning was in con-formance with Technical Specification Table 6.2-1.
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IE Rpt. No. 77-3 III-2
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The inspector observed the control room status board, on February 25, 1977, to verify that core flood tank boron concentrations were within the limits of Technical Specification 3.5.1.
Concentrations were within acceptable limits. An error was observed in the date of CFT surveillance on boron concentration but the concentrations were verified to be correct.
Main steam line hydraulic restraint oil levels were also observed to be adequate indicating apparent operability per T.S. 3.7.9.1.
3.
Licensee Event Reports (LER) Review LER 76-1 This event involved failure to have an intermediate range monitor operable when control rod drive breakers were closed, during control rod verification tests on December 22, 1976. The event was contrary to Technical Specification 3.3.1.1.
The licensee caution-ed operations personnel to perform adequate research of Technical Specification requirements including surveillance requirements and revised operating procedure OP-502, on December 23, 1977, to clarify the specific Technical Specification requirement. The event was reported and reviewed by the licensee.
The inspector verified the above corrective action and had no further questions.
LER 77-5 This event involved instrument failure which caused the Boric Acid Storage Tank "A" (BAST) level instrument to read 100% (full) when l
the tank was empty. Boron crystals had formed inside the level sensing tube in the tank. Corrective action was to connect a i
temporary air line to the sensing tube to periodically purge air
through the tube thereby clearing the tube. At the time of the inspection final corrective actions were not complete. This item
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I will be the subject of review during a future inspection.
LER 77-8 This event involved inadvertent primary system deboration of 50 ppm boron on January 22, 1977. The cause was inadvertent restoration of an unborated makeup and purification desireraliser after loss of a 4160 volt buss. Corrective action stated in the licensee's report was verified and there were no further questions.
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