IR 05000302/1977018
| ML19317G155 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 11/09/1977 |
| From: | Dance H, Long F, Wessmman R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19317G150 | List: |
| References | |
| 50-302-77-18, NUDOCS 8002280743 | |
| Download: ML19317G155 (9) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION g
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Report No.: 50-302/77-18 Docket No.: 50-302 License No.: DPR-72 Licensee: Florida Power Corporation 3201 34th Street, South P. O. Box 14042 St. Petersburg, Florida 33733 Facility Name: Crystal River 3 Inspection at: Crystal River site, Crystal River, Florida Inspection conducted: October 4-7, 1977 Inspector:
R. H. Wessman Accompanying Personnel:
H. C. Dance (October 6-7, 1977)
Reviewed by:
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F. J. Long, Chief
/ Date Reactor Operations and Nuclear Support Branch
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Inspection Sununary Inspection on October 4-7, 1977:
(Report No. 50-302/77-18)
Areas Inspected: Routine, unannounced inspection of plant operations for the July-September quarter; followup on reportable events and review of jtssper controls. The inspection involved 40 inspector-hours on site by two NRC personnel.
Results: Of the three areas inspected no items of noncompliance were found.
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RII Rp t. No. 30-302/77-18 I-1
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DETAILS I Prepared by:
AL h il f, R.' H. Wessman, Reacfdr Inspector lat'e Reactor Projects Section No.1 Reactor Operations and Nuclear Support Branch (
W II & 11 H. C. Dance, Chief
/Date Reactor Projects Section No.1 Reactor Operations and Nuclear Support Branch
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Dates of Inspection: October 4-7, 1977 i
Reviewed b b
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F. J. Long, Chief
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~ Date Reactor Operations and Nuclear j
Support Branch 1.
Persons Contacted
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Florida Power Corporation (FPC)
- G. P. Beatty, Jr., Nuclear Plant Manager
- W. R. Nichols, Operations Supervisor
- P. F. McKee, Assistant Nuclear Plant Manager
- C. F. Westafer, Technical Support Engineer
- D. W. Pedrick, IV, Compliance Engineer
- J. L. Harrison, Assistant Chem / Rad Protection Engineer
- J.
Cooper, Compliance Auditor T. H. Wayble, Nuclear Shift Supervisor U, M. Embach, Nuclear Shift Supervisor H. E. Reeder, Nuclear Shift Supervisor R. N. Stuart, Nuclear Operator W. W. Surrency, Chief Nuclear Operator E. R. Sessoms, Nuclear Operator C. G. Goering, Compliance Auditor
- Denotes those attending the exit interview.
2.
Licensee Action on Previous Inspection Findings Not inspected.
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RII Rpt. No. 50-302/77-18 I-2
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3.
Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, items of noncompliance, or deviations. Unresolved items disclosed during the inspection are discussed in Paragraphs 5 and 6.
4.
Exit Interview The inspectors met with Mr. G. P. Beatty, Jr., Nuclear Plant Manager, and members of his staff (denoted in Paragraph 1) on October 7, 1977. Inspection findings were discussed including the unresolved items (Paragraphs 5 and 6).
Additionally discussed were licensee commitments relating to Makeup Tank Venting (Paragraph 11) and the Reactor Building Purge Monitor Modification (Paragraph 12).
5.
Heat Balance Calculations and Nuclear Instrumentation Calibration In the course of reviewing a Nonconforming Operations Report (NCOR),
the inspector identified a possible discrepancy in the licensee's
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processing of that NCOR and in the execution of Surveillance Proce-
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dure SP-312 (Heat Balance Calculations).
NCOR 77-203 indicated that on July 5, 1977, while performing SP-312, Nuclear Instrumentation Channels 5, 6, 7, and 8 were found to be ranging from 96.6 to 97.2% reactor power at a heat balance reactor power of 99.35%. The inspector's review indicated that the
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instrument setpoints (as shown by this data) would yield a reac' tor trip with the core above the 105.5% Racer Thermal Power limit required by Technical Specification 2.2.1.
The licensee stated that Technical Specification 2.2.1 requirements pertain to instru-mentation trip setpoint values; and that at the time of conduct of SP-312 the NI Channel trip setpoints were ragning between 104.56 and 104.59%.
The nuclear instrumentation channels were recalibrated within four hours of the initial discovery of the out-of-calibration channels.
The heat balance /NI correlation following calibration was satisfactory.
The inspector reviewed SP-113, Power Range Nuclear Instrumentation Calibration. SP-113 limits the NI setpoint to a maximum of 104.7%
reactor power. This value does not recognize the variance between the NI and thermal power calculation. The inspector's review also established that the acceptance criteria in SP-312, Heat Balance Calculations, specified that Heat Balance and NI Power must be within two percent. This tolerance will not always assure that the reactor core Rated Thermal Power will be below the 105.5% required trip setpoint.
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RII Rpt. No. 50-302/77-18 I-3
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The discrepancy between heat. balance power and limiting acceptable values for NI-indicated power is designated as Unresolved Item 77-18/I-1.
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6.
Jumper Controls (Unresolved Item 77-18/I-2)
The inspector reviewed the licensee's program for control of tempo-
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rary modifications and jumpers for consistency with the requirements of the Technical Specifications and 10 CFR 50.59.
As part of that review, the inspector svamined CP-114, Procedure for Control of
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Permanent Modifications, Temporary Modifications, and Deviations;
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and reviewed the Jumper and Lif ted Wire Log maintained in the control room.
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Technical Specification 6.5.1.6(d) requires that the Plant Review
Committee review all "... proposed changes or modifications to i
plant systems or equipment that affect nuclear safety." 10 CFR 50.59 allows the licensee to make changes to the facility, as described in the FSAR, provided that such changes do not involve a change to the Technical Specifications or an unreviewed safety question.
However, 10 CFR 50.59 requires a written safety evaluation of a proposed change which includes the bases for the determination that an unreviewed safety question is not created by the change.
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CP-114 does not provide assurance that proposed jumpers will be evaluated consistent with the requirements of Technical Specification 6.5.1.6(d) and 10 CFR 50.59 prior to their installation. and use in operational safety-related systems. This insufficiency in jumper controls is designated as Daresolved Item 77-18/I-2.
The inspector reviewed the Jumper and Lifted Lead Log maintained in
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the control room and discussed two findings with licensee personnel:
a.
The description and purpose entry for any particular jumper is
terse. This brevity makes it difficult to determine the i
purpose of the jumper without consulting the control room operators.
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b.
The inspector noted -10-15 jumpers that related to annunciators.
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The licensee stated that modification requests were pending on
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many annunciator-related jumpers.
-The licensee acknowledged the inspector's comments. The above two
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items will be reinspected during a future inspection.-
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RII Rpt. No. 50-302/77-18 I-4
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7.
Review of Ioas and Short Term Instructions The inspector reviewed selected plant logs and Short Term Instruc-tions for the July-September 1977 period.
These logs were reviewed for consistency with the requirements of Technical Specification 6.10.1 and the procedural guidance of A.I.500, Conduct of Operations.
l Documents reviewed included:
- Control Room Nuclear Operators Log - 9/10-19/77, 10/1-6/77
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- Shift Supervisors Log - September 1977
- Operating Daily Surveillance Lot
- Control Complex and Auxiliary Building Log
- Short Term Instructions - September, October 1977 The inspector determined that log reviews were being conducted by appropriate members of the plant staff and that log entries were consistent with Technical Specification limits. No discrepancies were identified.
8.
Review of Nonconforming Operations Reports The inspector reviewed approximately 200 Nonconforming Operations Reports (NCOR's) generated during the July-September period. These NCOR's were reviewed for consistency with the requirements of the Plant Technical Specifications and CP-111, Procedure for Documenting the Reporting and Review of Nonconforming Operations. Various questions of a minor nature were resolved in discussion with licensee personnel. No discrepancies were identified.
9.
Review of Licensee Event Reports (LER's)
The inspector reviewed selected LER's for consistency with the requirements of the Technical Specifications and Regulatory Guide 1.16 (Reporting of Operating Information - Appendix A Technical Specifications). The inspector examined the licensee's analysis of the event and the corrective action and discussed the LER with licensee representatives. The followin', LER's were reviewed:
a.
LER 77-90. DHV-4 Containment Isolation Inoperative. The licensee's engineering evaluation determined the inoperability to be due to a high differential pressure across the valve.
Since normal operation, as well as future valve surveillance tests (per Section II of the ASME Code), will be at low differ-ential pressures this event should not reoccur. This LER is closed.
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LER 77-101. DHV-4 Containment Isolation Inoperative. This LER
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was due to a broken wire in the valve's motor control center and is unrelated to LER 77-90.
LER 77-101 is closed.
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LER 77-109. CFV-12 Inoperative. This LER was due to a torque switch drift. The licensee stated that they have experienced i
few torque switch-related problems. This LER is closed.
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LER 77-112. Two Absolute Position Indication (API) Reed Switches Intermittently Inoperable.
This event has not reoccurred, and is believed to be attributed to a sticking reed switch.
This LER is closed.
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LER 77-114. Containment Isolation CFV-12 Failed to Close.
The
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licensee has verified operability of this valve and this event has not reoccurred. Additionally, the motor control center starter interlocks have had preventative maintenance (September 9,
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l 1977). This LER is closed.
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LER 77-115. 120 Volt AC Vital Bus 3D Inoperable Due to Failed Inverter. The inspector reviewed the system electrical schematic and verified compliance with Technical Specification 3.8.2.1.
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This LER is closed.
10.
In-Office Review of Licensee Event Reports (LER's)
The inspector reviewed various LER's for consistency with Technical
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Specification and Regulatory Guide 1.16 requirements.
The LER's were examined to determine the cause of the event and the licensee *s corrective action. The following LER's were reviewed and closed without conmient in the NRC Regional Office:
LER Number Subj ect 77-25 High Pressure Injsetion Loop "B" Inoperable 77-26 Makeup Valve MUV-24 Inoperable
77-33 Inoperable Absolute Position Indication 77-36 Inoperable Containment Isolation Valves CAV-2 and
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77-39 Boric Acid Storage Tank Suction Valves Found Closed 77-41 Primary Containment Temperature Exceeded Limit 77-43 Regulating Rod Group 7 Insertion Limits Exceeded 77-44 Regulatory Rod Group 6 Insertion Limits Exceeded 77-46 Hydraulic Snubber MUH-44 Inoperable 77-47 Group 7 Control Rods Dropped on Ratchet Trip
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RII Rpt. No. 50-302/77-18 I-6
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77-48 Wind Direction Monitor Inoperable 77-53 Chemical Addition Pumps Inoperable 77-54 Control Rod Drive System Automatic Programmer
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Inoperable 77-55 Diesel Generator "A" Inoperable 77-57 Two-Out-of-Three RCS Leakage Detection Systems Inoperable 77-63 Reactor Building Cooling Fan Inoperable 77-64 Auxiliary Building Exhaust Fans Inoperable 77-65 Steam Driven Emergency Feed Pump Inoperable
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77-71 ARPI Inoperative for One Control Rod Drive 77-75 Loss of All Absolute Position Indicatior Due to Voltage Spike
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77-76 Auxiliary Building Fan Inoperative 77-83 Failure to Have Two Channels of BWST Level Indication Operable 77-87 Containment Isolation Valve WDV-94 Inoperable 77-88 Containment Isolation Valve SWV-47 Inoperable 77-89 Containment Isolation Valve DHV-41 Inoperable 77-113 Missed Weekly Battery Surveillance 11.
Unplanned Releases While Venting the Makeup Tank
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The inspector reviewed with the licensee the Crystal River 3 proce-dural controls effected to reduce the risk of an unplanned gas release via waste gas system loop seals when venting the makeup tank. The following procedural changes are (or will be prior to the next makeup tank venting) promulgated vial Short Term Instruc-tions:
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- Sample the makeup tank for noble gases prior to venting,
- Maintain makeup tank overpressure less than 40 psig,
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- Check all waste disposal system loop seals twice per shift to assure the water seal has not been lost, l
- Check the existance of loop seals on the concentrated waste storage tank, concentrated boric acid storage tank, and the waste system evaporators prior to venting the makeup tank.
- Periodically vent the concentrated waste storage tanks and the concentrated boric acid storage tank.
The licensee stated that proposed permanent system design changes
- have completed engineering review (telecon between Region II and W. P.~ Stewart - Director, Power Production on September 29, 1977)
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and that implementation is awaiting materials procurement. As a temporary measure, the loop seals on the waste disposal system evaporators have been extended from four to ten feet (approximately).
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The inspector will confirm implementation of procedural controls during a future inspection.
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Reactor Building Purge Monitt.r Modification J
The inspector reviewed the status of a proposed modification to the reactor building ; urge monitor. This modification will trip the reactor building purge valves in the event of a atB Purge Monitor
(RM-A1) failure, thus preventing an unmonitored radioactive gas i
release. The licensee has issued MAR-77-9-24 to implement this change. Implementation will be completed by November 1, 1977 and will he inspected during a future inspection.
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Facility Tour The inspector toured various areas of the plant to observe operations and activities in progress and to ascertain the general state of
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cleanliness and housekeeping. Of particular interest were house-
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keeping items identified in IE Reports 50-302/77-8 and 50-302/77-17.
The licensee has been responsive to housekeeping and cleanliness
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items identified by the inspector in these reports.
The inspector.
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encouraged licensee supervisory personnel to maintain a practice of frequent plant tours and continue efforts to reduce the number of contaminated areas and auxiliary systems water leaks.
14.
Sodium Thiosulfate Tank The sodium thiosulfate addition tank is isolated in accordance with Amendment No. 1 to Facility Operating License No. DPR-72.
Although this tank is isolated the surveillance requirements of Technical Specification 4.6.2.2.b still apply for this tank. The licensee has sought release from this requirement (Technical Specification Change Request No. 5, submitted to the Office bf Nuclear Reactor Regulation on July 15, 1977). The licensee has sampled this tank on October 4, 1977 and has found it to meet the Technical Specifica-tion requirements.
The inspector has recommended that the proposed Technical Specifica-tion requirements concerning surveillance of this tank be resolved prior to the next scheduled chemical analysis and volume verification required by the existing Technical Specification 4.6.2.2.b.
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RII Rpt. No. 50-302/77-18 I-8
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15.
Review of IE Circulars The inspector reviewed IE Circular 77-11 (Leakage of Containment i
Isolation Valves with Resilient Seats) with licensee personnel to 8l ascertain that it had been received on site and was reviewed by cognizant supervisory personnel. Site personnel have initiated a request for a corporate egnieering review of this circular to ascertain applicablity and need for action. The inspector has no further questions relating to IE Circular 77-11.
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