IR 05000302/1979030
| ML19254E603 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 09/13/1979 |
| From: | Burnett P, Sauer R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19254E597 | List: |
| References | |
| 50-302-79-30, NUDOCS 7911010560 | |
| Download: ML19254E603 (8) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION o
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E REGION 11
101 MARIETTA ST N.W., SUITE 3100 ATLANTA, GEORGIA 30303
Report No. 50-302/79-30 Licensee: Florida Power Corporation P. O. Box 14042, Mail Stop C-4 St. Petersburg, Florida 33733 Facility Name:
Crystal River 3 Docket No. 50-302 License No. PPR-72 Inspection at Crystal Rive site near Cry al River, Florida.
Inspected by: /b >
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h-809 T C. 4auer"' '
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Date Signed Approved by:
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Q--gf /f P.-E Burnett, Acting Section Chief Date Signed RONS Branch SUMMARY Inspection on August 5 - 9, 1979 Areas Inspected This routine, unannounced inspection involved 43 inspector hours onsite in the areas of physics test procedure review, surveillance and verification that modifications to the high pressure injection system as part of the ECCS Small Break Analysis were completed by August 6, 1979, as required by NRC Exemption letter dated July 3, 1979.
Results Of the three areas inspected, no apparent items of noncompliance or deviations were identified in two areas; two apparent items of noncompliance were found in one area (Infraction - failure to ensure rod drop times met Technical Specificatica requirements prior to reactor startup - Paragraph 5.a; infraction - failure to follow procedure -
Pcragraph 5.c).
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DETAILS 1.
Persons Contacted Licensee Employees
- G. P. Beatty, Jr., Nuclear Plant Manager
- J. Cooper, Compliance Engineer
- G. L. Boldt, Performance Engineering Supervisor
- F. W. Pleubel, Electrical Supervisor
- W. R. Nichols, Operations Superintendent
- P. F. McKee, Technical Services Superintendent
- G. M. Williams, Compliance Plant Engineer L. B. Tittle, Results Engineer M. E. Collins, Plant Engineer L. A. Hill, Compliance Auditor R. P. Cunningham, Nuclear Planning Coordinator Other licensee employees contacted included shift supervisors and reactor operators.
Other Organizations D. Downtain, Babcock and Wilcox R. McAndrew, Babcock and Wilcox
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summarized on Augt st 9,1979, with those persons indicated in Paragraph 1 above, and discussed in a subsequent telephone conversation on Augant 13, 1979. The licensee acknowledged the two identified items of noncompliance:
failure to ensure rod drop times met technical specification requirements prior to reactor startup (50-302/79-30-01) and failure to follow procedure (50-302/79-30-05). See Paragraphs 5.4 and 5.c for discussion.
Regarding the reference of the power escalation controlling procedure (PT-120) to the use of an outdated prerequisite test (SP-103),
the licensee agreed to revise PT-120 as discussed in Paragraph 5.b (0 pen Item 50-302/79-30-04).
Regarding the high pressure injection modification, Paragraph 7,.
as a result of the ECCS small break analysis, the inspector informed the licensee the item will be left open pending additional inspector review (0 pen Item 50-302/79 1n-06).
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s-2-3.
Licensee Action on Previous Inspection Findings Not inspected.
4.
Unresolved Items Ucresolved items were not identified during this inspection.
5.
Physics Test Procedure Review ( 75% FP plateau)
The following procedures were reviewed to verify test results met stated acceptance criteria and to ensure the data and procedure were technically correct and in accordance with the facility's Technical Specifications:
SP-102 " Control Rod Drop Tirae Test,"
SP-103 " Moderator Temperature Coefficient Determination at
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Startap after Refueling,"
SP-312 " Heat Balance Calculations,"
SP-400 " Shutdown Margin Physics Testing,"
SP-100 " Controlling Precedure for Precritical Testing,"
PT-110 " Controlling Procedure for Zero Power Phys _ics,"
PT-114 " Moderator Temperature Coefficient Determination at Hot Zero Power"and PT-120 " Controlling Procedure for Power Escalation Testing."
The review identified three procedures of concern:
SP-102 " Control Rod Drop Time Test".
Review of this procedure a.
identified that a timing-signal reference trace, used to verify time-interval-increment spacing from the point of power interruption to the control rod drive breakers to the 3/4 control rod insertion point (25% withdrawn), was not present on the Honeywell Model 1508A visicorder oscillograph paper. Without a time reference the time interval spacing, therefore rod drop time, cannot be adequately verified.
The failure of the licensee to properly ensure that the times depicted on the oscillograph were correct and within the Technical Specifications 3.1.3.4 requirement of 1.66 seconds prior to reactor startup is considered an apparent item of noncompliance (79-30-01).
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t-3-To adequately ensure that the licensee was not operating outside technical specification limits, the inspector requested and received the following information from the licensee:
(1) A ten hour run of the visicorder was performed, with a 60 herte reference timing trace taken heurly, to ascertain varying operating characteristics. The results yielded an interval variance of 1.2 intervals (95.8 intervals per 60 hz sweep to 97 intervals, resulting in a maximum error of 4.2%).
(2) The visicorder tolerance spread based on ambient temper-ature and operating voltage ranges had a maximum variance of 3%. Applying the sum of these variancis to the control rod with the longest rod drop time (safety rod 1-3 at 1.240 seconds), it can be shown that an additional 27% error would have had to exist in the recording apparatus in order to exceed the 1.66 second specification. Based on this information, the inspector considers the rod-drop times to be within the Technical Specification requirement 3.1.3.4 and continued operation of the facility acceptable. Though resolved prior to the issuance of this report, the licensee should identify the corrective action he intends to take to ensure recurrence of this type of problem does not re-occur.
Other contributing discrepancies of improper procedure performance / review noted in inspector's review of SP-102 were:
(a) Inadequacy of the procedure to properly measure the complete time interval of a control rod drop.
Procedural step 6.9 specifies the length of time of each rod drop commences from the point where the 120 VAC control rod drive breaker drops to 0 VAC vice when the breaker first indicates opening. The difference between the two methods increases each rod drop time by.005 seconds or a.5% nonconservatism in each rod drop time.
The licensee intends to correct this deficiency (IFI 79-30-02).
(b) Use of a non-calibrated piece of equipment in measuring rod drop times. The normal method of verifying the width of each time interval produced by the visicorder is by internally generating a 60 hertz timing signal through an oven-controlled relaxation oscillator. The frequency range is based on the RC (resistance -
capacitance) network selected on the visicorder
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timer scaler. The network is variable and can drift I283 I87
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with age /use (e.g., in the case of (1) above, a maximum error existed of 4.2%). Questioning as to whether the visicorder was on the current calibration schedule resulted in a negative response from the electrical supervisor. Further, he reported that the instrument appeared not to ever have been calibrated by the facility.
The licensee indicated the visicorder will be placed on the calibration schedule (IFI 79-30-03).
b.
SP-103 " Moderator Temperature Coefficient Determination at Startup af ter Refueling." Inspection of the SP-103 test results disclosed a handwritten message on the cover sheet indicating the procedure had not been performed because performance of PT-114 satisfied the requirements of SP-103.
Since SP-103 is identified as a prerequisite test to PT-120, a closer inspection of the SP-103 contents revealed that the initial conditions to this procedure were not compatible with those of PT-114. The licensee representative stated that SP-103 was written specifically for Cycle 1 (approved December 22, 1976) and that PT-114 is the proper procedure for performance.
The licensee intends to upgrade PT-120 to reflect PT-114 as the proper orerequisite test for the moderator temperature coefficient determination. Retirement of SP-103 will be considered.
This item is considered open pending licensee corrective ac'. ion (79-30-04).
SP-312 " Heat Balance Calcolations". After reviewing the c.
75% full power (IP) heat talance calculation obtained from the Batley 855 computer r,roup 21 routine, the inspector requested the reactor engineer to perform a similar calculation with the Auto Data 8/ IBM 5100 basic heat balance routine. This routine is used primarily when the Bailey 855 computer is not available or as a check of the computer calculation.
The IBM 5100 results indicated that the core heat balance, as determined from reactor coolant loop RTD's, differed from the average power rr.nge nuclear instruments (NI) by as much as 5.9% (70.7% vs. 76.6% respectively).
Since a similar comparison of heat balance calculations between the Bailey 855 computer groups 17 and 21 and the Auto Data 8/ IBM 5100 were previously performed at 15% FP, with the group 21 routine being selected as the most valid heat balance calculation by the reactor engineer, (in accordance with Step 8.2 of PT-120), a quick review of the July 31, 1979 procedure performance indicated an imbalance of 2.8% (avg. NI-15.3%
to core wT heat balance - 12.5%) existed.
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-5-Additional review ident ified that the current Cycle 2 reactor coolant loop RTD normalization constants (Cn), as calculated by the IBM-5100 RTD Normalization Routine (Step 11, Enlcosure 1, PT-100), were not input ic.; the IBM-5100 Heat Balance II routine as required by Step 12, Enclosure 1 of PT-100. The oversight was made by the reactor engineer thinking the new Cn values were to be input via the computer terminal rather than physically changing the constants stored in the computer software. After correcting the constants, the routine was rerun at 75% FP. The results yielded an average NI reading of 76.6% with a core wT heat balance of 76.3%.
The failure to input the new Ca values into the IBM-5100 Heat Balance II routine as required by Step 12, enclosure 1 of PT-100 constitutes an apparent noncompliance with Technical Specification 6.8.1 (79-30-05).
6.
Post Refueling Startup Testing Problems The inspector observed the following reactor-related problems which occurred during startup testing for Crystal River 3 (CR-3), beginning of cycle 2, from 75% FP to 100% FP. All problems were verified not to be outside Technical Specification requirements prior to test continuance.
a.
Peaking Factors At approximately 0630 on August 6, 1979, while at 75% FP, the highest measured radial and total peaking factors as compared to the highest predicted radial and total peaking factors for a symmetrical eighth core representation were found to have differed by - 7.9% and -8.4% respectively.
The allowed deviation limits of 5% and 7.5% respectively are given by Step 10.2.6 of PT-120 for this comparison.
Babcock and Wilcox (B&W) was informed by the licensee.
Since the deviation criteria is B&W derived, and not technical specification related, B&W subsequently recommended lowering the nuclear overpower trip setpoint by 8% (104.7% to 96.7%) for added conservatism in core protection; to continue power escalation to the power level cutoff point of 92% (required by Technical Specification 3.1.3.8 until equilibrium xenon is reached) and thereupon repeat the peaking factor divergence test.
These reccmmendations were approved by the Plant Review Committee (PRC) and power escalation commenced at approximately 1130.
On August 7, 1979, the deviation of measured to predicted radial and total peaking factors were again calculated at 92% FP.
The total peaking factor was within the allowable 7.5% with a 1283 189
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-6-s deviation of -6.18%.
However the radial peaking factor was again outside the tolerance of 5% with a deviation of -6.57%.
Using the actual end of cycle IB burnup history of 440 EFPD and final control rod positions in lieu of the projected end of cycle values in their computer calculations for the 75% FP and 92% FP plateaus, B&W concluded that the data met the new acceptance criteria. Subsequently, the PRC approved the recommendations to continue with power escalation to 100%. The escalation commenced at 1350 after the nuclear overpower trip
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setpoints were returned to 104.7%.
On August 8, 1979, using the newly acquired B&W calculated total and radial peaking factor for 100% FP the deviations were computed to be.20% and.19% respectively.
b.
Erratic Reed Switch Operation At 0445 on August 7, 1979, with reactor power at 92% FP, the rod position indicator (RPI) for regulating rod 7-6 became erratic. When rod group 7, in bank overlap, was withdrawn greater than 13%, the RPI for rod 7-6 would jump to approximately 40% wd.
To correct the problem a reactivity change was made to bring the group 7 rod bank in to less than 13% wd.
Reactor power dropped by approximately 1%, and was restored by diluting the borated reactor coolant. Repairs to the RPI were made made by 0830 through use of an electrical tickler to open any stuck reed switch contacts in the 7-6 control rod position indicating tube.
No items of noncompliance or deviations were identified in this area.
7.
ECCS Small Break Analysis Ref (a): Letter of Exemption in the matter of Florida Power Corporation, et al, Crystal River Unit No. 3 Nuclear Generating Plant, by the Division of Opacating Reactors, Office of Nuclear Reactor Regulation, dated July 3, 1979.
The inspector reviewed modifications made to the high pressure injection (HPI) system as part of a permanent solution to the small break analysis problem identified in April 1978. The modifications involved leaving the HPI pump cross connect valves open and providing a means of supplying electrical power to the motor operators of the HPI valves from the engineered safeguards (ES)
electrical busses of both channels. The modifications were to be complete by August 6, 1979, as discussed in Ref(a) in order for the facility to continue operation and to achieve full compliance with the requirements of 10 CFR Part 50 Paragraph 50.46(a).
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The inspector reviewed Maintenance Action Request MAR M-79-5-63 1283 190
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-7-s documentation which specified the equipment and procedure necessary to ensure normal and backup power supplies to the HPI valves (MUV -23, -24, -25, -26) were properly installed. The valves were satisfactorily time-stroke tested with each power supply in accordance with SP-355 " Operations ES Monthly Functional Tests" and SP-370 " Quarterly Cycling of Valves" August 5,1979.
From the material inspected, the inspector considers the facility appears to be in compliance with Ref(a). This item is open pending completion of inspector review in this area (79-30-06).
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