ML19310A407
| ML19310A407 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 06/13/1980 |
| From: | Richard Bright FLORIDA POWER CORP. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8006170488 | |
| Download: ML19310A407 (6) | |
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t NQP Florida Power C O R PO M A 180 N June 1?., 1980 File:
3-0-3-a-3 Director Office of Nuclear Reactor Regulation Division of Operating Reactors U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72
Dear Sir:
On April 30, 1980, Florida Power Corporation submitted the Crystal River Cycle 3 Reload Report, BAW-1607, Revision 1.
Since that time we have had several discussions with members of your staff, concerning the anal-ysis in that report, and the Startup Physics Test Program.
We now wish to amend our earlier submittal and provide additional information for clarification.
Enclosed please find our response to questions on the analysis in BAW-1607, and the Startup Physics Test Program.
Sincerely, FLORIDA POWER CORPORATION b
R. M. Bright Acting Manager Nuclear Support Services RMBekcF01(D70)
General Office 3201 Tnirty-fourin street soutn. P O Box 14o42. st. Petersburg. Florida 33733 e 813-866-5151 8006170 N p
STATE OF FLORIDA COUNTY OF PINELLAS R. M. Bright states that he is the Acting Manager, Nuclear Support Ser-vices Department of Florida Power Corporation; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information and belief.
A*l R. 'M. Aright Subscribed and sworn to before me, a Notary Public in and for the State and County above named, this 13th day of June,1980.
Notary Public Notary Public, State of Florida at Large, My Commission Expires:. August 24, 1983 RMBekcF01(D70)
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RESPONSES TO NRC QUESTIONS ON BAW-1607 l
, 1.
Concerning' Fuel Rod Internal Pressure:
During Cycle 3, fuel rod internal pressure will not exceed system pressure at any. time in the cycle.
-2.
Concerning Control Rod Guide Tube Wear:
We have reviewed the final draf t of the B&W Contol Rod Guide Tube Wear-Generic Report, BAW-1623, and are in agreement with the con-clusions presented in the report. - This report will be submitted to NRC by.B&W on or before July 15, 1980.
3.
Concerning the Locked Rotor Analysis:
Supplementary information on the Locked Rotor Analysis is included as Attachment A_ to this response.
The information in Attachment A
. applies to the hot channel.
4.
.Concerning the Startup Physiys Test Program:
The Startup Physics Test Program for Cycle 3 will be as in Cycle 2
(
Reference:
letter, W. P. Stdwart to R. W. Reid, 6/8/79) with the exception of a revised review criterion applied to the power dis-tribution measurements.
The review criterion for the radial power dis'.ributions requires-that the root mean square of the difference between the calculated and measured power distribution must be less than or equal to.072.
This number is defined as the upper toler-ance limit for acasured radial power distribution in the B&W Topi-cal Report (BAW-10119P-A), " Power Peaking Nuclear Reliability Fac-tors".
The root mean square of the radial power distribution is defined.as:
177 (Cj - Mj)2 R MS = '
E i=1 177 Where:
Cj = calculated RPD Mi = measured RPD
~
177 = number of assemblies in the core
'The_ review ~ criterion of RMS <.072 will be applied to all three power levels-(40%, 75%, and 1003).
At 40% full power, the core is not in equilibrium, and nay not meet this review criterion.
If the review criterion is not met at any: power level, a review of plant conditions which could cause disagreement will-be performed.
This review-will bel docunented in the Startup Report.
The. acceptance criteria for:the radial and-total power distribution L
neasurements 'will~ be the same as 'those described in BAW-1607.
I
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' ATTACHMENT A BACKGROUND In order to obtain accurate reactor coolant flow measurements, pressure drop taps have been installed in the hot leg of all 177 FA plants. To dampen noise, hydraulic enubbers were installed in the AP measurement lines.
It was determined that the snubbers cause a time delay in the AP signal by as much as 0.75 seconds with respect to the unsnubbered signal. The conse-quences of the increased' delay time were investigated in a locked rotor analysis performed for Crystal River -3 cycle 1 in March, 1974.
ANALYSIS The hot channel was examined during the transient at 102 percent initial rated power, and the limitations of temperature and pressure errors, hot channel factors, and fuel densification effects were included in the analysis.
The reactor scram is initiated by a flux / flow trip (consistent with the F.P.C.
Technical Specificiations), witht the inclusion of the increased control rod delay time of 1.4 seconds.
The transient inputs are presented in Figure 1, and the results of this revised analysis and those of BAW-1397 are compared in Figure 2.
In com-paring the hot channel DNB ratios, it should be remembered that the original analysis was comrleted using the W-3 CHF correlation; this revision is based on the B&W-2 cocrelation. Similarly the original analysis assumed that film boiling was initiated at a hot channel LNBR (W-3) = 1.30; whereas the revision assumes that film boiling begins when the hot channel DNBR (B&W-2) reaches 1.32.
CONCLUSION In comparing the cladding surface temperatures it can be seen that the re-vised analysis is less limiting than the original analysis presented in BAW-1397, " Crystal River - 3 Fuel Densification Report." Since the fuel densification report remains the bounding analyses, it is referenced in following reload fuel cycle reports.
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