ML19329D640
| ML19329D640 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/20/1977 |
| From: | Brandimore S FLORIDA POWER CORP. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8003160265 | |
| Download: ML19329D640 (7) | |
Text
NRCpCRM 195 U.C. NUCLEAR REGULATOnv COMMisSICN OOCKET NU*A RJ O
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PC DISTRIBUTION ron t,.RT 50 DOCKET MATERIAL F R O M'.
DATE OF DOCUMENT Florida Power Corp.
5/20/77
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Mr. John F. Stolz ST. Petersburg, Fla.
o47 Receiveo
- 5. A. Brandimore 5/2377
.N.ETTER
[NOTORIZE D PROP INPUT FORM NUMBER OF COPtEs RECEIVED
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, _ _$ { hh DESCRtPTION E NCLOSU RE Consists of requested additional info.
concerning.the effects of increased fission gas releases on the safety analysis.........
notorized 5/20/77......
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PLANT NAME:
Crystal River Unit No. 3-RJL 40 e.m SAFETY FOR ACTION /INFORMATION FNUTun ASSIGNED AD:
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3 May 20,1977 Ci e
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O Mr. John Stolz N8 t
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Division of Project Management U.S. Nuclear Regulatory Comission Washington, DC 20555
SUBJECT:
Florida Power Corporation Crystal River Unit 3 Docket N o. 50-302
Dear Mr. Stolz:
In your letter of February 11, 1977, you requested Florida Power Corporation to evaluate the effects of increased fission gas releases on the safety analysis for Crystal River Unit 3.
Since Crystal River Unit 3 will not reach a local exposure (burnup) of 20,000 megawatt-days per metric ton of uranium prior to June 1, 1977, we were to furnish the requested information to the Commission within 90 days of receipt of your letter.
Attached for your staff's review are 3 signed originals and 40 copies of our response to the information requested in Items (a) through (d) of your February 11, 1977, letter.
If you have any questions regarding this matter, please contact us.
Y ruly yo t'
g S.A. Brandimore Senior Vice President and General Counsel Attachment SAB:ECS: hic 5/4 General Office 3201 Tniny-founn street soutn. P o Box 14042. st Peterscurg. Fionca 33733 813-866-515I o
IN WITNESS WHERE0F, the applicant has caused its name to be hereunto signed by S.A. Brandimore, Senior Vice President and General Counsel, and its corporate seal to lee hereunto affixed by Betty M. Clayton, Assistant Secretary, thereunto duly authorized the 20th day of May,1977.
FLORIDA POWER CORPORATION
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gr sc:D M.A. Brandimore Senior Vice President and General Counsel ATTEST Betty M. Clayton Assistant Secretary (CORPORATE SEAL)
Sworn to and subscribed before ine this 20th day of May,1977.
Notary Public My Cearnission Expires:
Notary Public State of Florida at Large My Connission Expires July 9,1978 (NOTARIAL SEAL) 585-P
RESPONSE TO NRC QUESTIONS CONCERNING NEW NRC FISSION GAS RELEASE MODEL Response to Item a)
It is estimated at the present time that a maximum local exposure (burnup) e# 20,000 megawatt-days oer metric ton of uranium (MWD /tu) will be reached by any fuel rod in Crystal River Unit 3 during the second fuel cycle.
The second fuel cycle for Crystal River Unit 3 will not begin until late fall of 1978.
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Response to Item b)
The results of the evaluation of the revised fission gas release nodel for the TAFY fuel pin analysis are provided in the attached two tables. Table 1 defines the input parameters used for the evaluation and Table 2 provides the -
comparison of internal pin pressure for the two fission gas release models as a function of burnup. At a maximum burnup of 38,000 MWD /MTU, the informal pin pressure, based on the NRC staff model reflects a 26% increase when compared with the results of tha current model; however, in all instances, the internal pin pressure remains below system pressure.
The TAFY calculations have demonstrated that the average fuel temperature at BOL conditions and 17 kw/ft is equal to 2990 F for both the NRC and B&W models.
These calculations of pin pressure and temperature, based on utilizing the TAFY Code with and without the NRC fission cas release (FGR) equation, were performed for Three Mile Island, Arkansas and the Oconee reactors as a class of power reactors. The Crystal River Unit 3 reactor belongs to the above-mentioned class of reactors and therefore the results of these calculations provided in Table 2 are equally applicable to Crystal River Unit 3.
The TAFY Analysis without the NRC FGR equation was taken from the Oconee II Cycle 1 licensing analysis. Since the input data for these analyses are j
identical for both the NRC and B&W FGR models, the differences reported in Table 2 can be directly attributed to the difference in the FGR models.
Response to Item c)
The TAFY Code without the NRC FGR equation is the code used in the current LOCA and safety analyses for CR #3. Use of the NRC FGR equation in TAFY will impact the LOCA analysis as the worst pin pressure would now occur at earlier burnups. Since initial inside and outside cladding surface oxide layers would be thinner at earlier burnups, the zircaloy-water (metal-water) reaction would be larger than that previously calculated.
The increased energy generation in the cladding would raise the peak cladding temperatur'e and would probably result in the present LOCA limits violating the criteria of 10 CFR 50.46.
Requalification of the LOCA limits at CR #3 would then be necessary, ultimately resulting in the issuance of revised Technical Specifications.
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A survey of the non-LOCA-related safety analysis accidents was performed and it was concluded that the NRC FGR model would not affect the results.
The average fuel temperatures used in the transient cladding temperature calculations are not changed by the new NRC FGR model.
Further, the higher pin pressures at E0L would not result in cladding rupture during these transients.
Hence, the non-LOCA safety analysis accidents for CR #3 are
- not' impacted-by the NRC FGR'model.
Response to Item d)
As stated above, the internal fuel rod pressure does not exceed the nominal system pressure and therefore, no discussion is given for operations with fuel cladding in tension.
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PIN PRESSURE ANALYSIS INPUT
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(OCONEE ll NSS-4)
FUEL INITIAL MEAN DENSITY - % TD 92 5 0.370 INITIAL MEAN DIAMETER - IN INITIAL LTL DENSITY - % TD 92.0 FINAL DESNITY - % TD 96.5 0.150 DISH RADIUS - IN 0.0170 DISH FACTOR 144 INITIAL STACK LENGTH - IN CLAD CLAD 10 - IN 0 377 0.430 CLAD OD - IN CLAD LENGTH - IN 153 3
INITIAL PLENUM VOLUME - IN 0.75 INITIAL BACKFULL PRESSURE - PSIA 375 0
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RESTRICTIONS 25% REDUCTION Of! " GAP NO RESTRUCTURING
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SORSED GAS CONTENT - 0.01 CC/GM USE BOL TEMPERATURES FOR ACCIDENT ANALYS IS e
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TABLE 2
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OCONEE 1, 2, & 3; TMI-1: ANO-1 OATA Peak Rod Burnup TAFY B&W FGR Model TAFY NRC FGR Model (MWD /MTU)
Pin Pressure Pin Pressure (psi)
(psi) 20,000 1210 1230 22,000 1235 1240 25,000 1295 1320 27,000 1J4c 1410 30,000 1400 1550 32,000 1430 1615 35,000 1470 1745 37,000 1510 1865 38,000 1525 1925 O
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