ML20207J572
| ML20207J572 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 09/16/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20207J557 | List: |
| References | |
| NUDOCS 8809280043 | |
| Download: ML20207J572 (5) | |
Text
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UNITED STATES 8
NUCLEAR REGULATORY COMMISSION n
E
-l WASH WGTON. D. C. 20666
%*.../
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENTS NOS. 84 AND 77 TO FACILITY OPERATING LICENSES NOS. OPR-42 AND DPR-60 NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNITS NOS. 1 AND 2 DOCKETS NOS. 50-282 AND 50-306
1.0 INTRODUCTION
By letter dated July 5, 1988 (Ref. 1), Northern States Power Company (NSP or the licensee) requested amendments to the Technical Specifications (TSs) appended to Facility Operating Licenses Nos. DPR-42 and DPR-60 for the Prairie Island Nuclear Generating Plant, Units Nos. I and 2.
The proposed amendments would change TS 3.10. "Control Rod and Power Distribution Limits."
Specifically, the licensee proposed to increase the hot channel enthalpy riso respectiv$Yy,,andhotchanneltotalpeakingfactor,F,to1.70and2.50, factor, F andtodeleteTSFigure3.10-8,"Accep9ableValuesofF ThisproposedTSchangewillstartwiththeoperationofP9a(Fine)and I$g(F)."
F n
7and Unit 1 Cycle 13 which is scheduled to begin on September 28, 1988.
The Unit 1 Cycle 13 will contain a full core of Westinghouse 14x14 optimized fuel assemblies (OFA).
Unit 2 Cycle 12 will be operating with two-thirds of 0FA fuel and one-third of Exxon TOPROD fuel until March 1989.
The proposed TS change is an extension of the previous TS change requested in April 1987 for Unit 1 Cycle 12.
For Unit 1 Cycle 12 operation, the safety analyses for the anticipated operational transients and accidents other than respectively, and the safety limits in Figure 2.1-1 N Tt 2.9 were.70 and 2.50 loss-of-coolant accident (LOCA) were analyzed with F and F of 3 determined HoweverFigure3.10-8waspreposk$andF.of1.70and2.50respectively.
to be conservative relative to F todeYineF function of F with the Thiswasbasedonthelargeb9eakLON(etat 9.asa'0and1.66,resp raximum allowable values of F and F s
LBLOCA) analysis performed with the Wesunghouse 1981 Emergency Core Cooling System (ECCS) evaluation model (EM).
Starting with Unit 1 Cycle 13 operation, the LBLOCA analysis is performed with a new Westinghouse ECCS evaluation model described in WCAP-10924-P, "Westinghouse Large-Break LOCA Best Estimate Methodolgy" (Ref. 2).
This analysis indicated that the proposed F of 1.70 and F of 2.50 would not violate the acceptance criteria of 10 N R 50.46, g
2.0 DISCUSSION AND EVALVATION 2.1 LBLOCA Analysis The Prairie Island units are Westinghouse designed two-loop plants equipped with low pressure upper plenum injection (UPI) systems as part of ECCS.
The previous ECCS evaluation model assumed the UPI water fell directly into the 8909280043 8G0916 DR ADOCK 0000
lower plenum without interaction with the core and could therefore be treated as if the plants were cold leg injection plants.
In support of the proposed TS change to increase the power peaking factors, the licensee provided a new LBLOCA analysis in Exhibit E of the July 5, 1988 submittal (Ref. 1).
This analysis uses a new Westinghouse ECCS evaluation model which is developed for application to t:.e two-loop UPI plants and is described in Westinghouse topical report WCAP-10924-P (Ref. 2).
This ECCS evaluation model used a l
best-estimate thermal-hydraulic code WCOBRA/ TRAC and the approach described in SECY 83-472 (Ref. 3).
Though the methodology described in WCAP-10924 is generic to all Westinghouse-designed two-loop UPI PWR plants, the analysis uses Prairie Island plant-specific data as a lead plant to demonstrate compliance to the Appendix K requirements and the SECY 83-472 guidelines.
Therefore, the results of WCAP-10924 are directly applicable to the Prairie Island units.
t The NRC staff review (Ref. 4) has concluded that WCAP-10924 is acceptable for e
licensing application to Westinghouse two-loop UPI plants with conditions that the UPI-licensees would apply for exemptions to Items I.D.3 and I.D.5 of Appendix K to 10 CFR 50.
The exemptions are necessary because Item I.D.3, which requires the use of a carry-over fraction to calculate the reflood core exit fluid flow, and Item I.D.5 setting specific requirements for refill and reflood heat transfer calculation were intended for the conventional cold leg injection plants and are not applicable to UPI plants.
The licensee's letter of July 28, 1988 (Ref. 5) requested exemption to these two requirements and the exemption request has been granted (Ref. 6).
t The Appendix X calculation provided in WCAP-10924, Volume 2, Revision 1, was made prior tn a proper implementation of Appendix K, Item I.C.4, which prohibits a return to nucleate boiling during the blowdown phase.
Therefore the results provideo in Exhibit E of the licenu e's submittal, which was obtained from WCAP-10924, Volume 2, Revision 1, do not comply with the Appendix K requirement and are not acceptable.
By letter dated August 5, 1988 (Ref. 7), the licensee submitted a reanalysis of the Appendix K calculation using the corrected version of WCOBRA/ TRAC with proper implementation of a code logic to block the return to nucleate boiling during the blowdown phase as required by Appendix K.
The analysis was performed with a full core of 0FA fuel and appropriate power peaking factors of 1.70 and 2.50 for F and F,
respectively.
SinceUnit1Cyc1g12stillcontainsone-thiruofNCTOPROD fuol, a mixed core penalty of 10 F, which was based on the previous EM model calculation, was added to the resulting peak cladding temperature (PCT).
The resultsofgnalysis,showninTable2oftheAugust5,1988 submittal,showa PCT of 2041 F, the maximm local cladding oxidation of 11.65%, and the total i
cladding oxidation or hydrogen generation of less than 0.3%.
These results are below the acceptance criteria set forth in 10 CFR 50.40 and are, j
therefore, acceptable.
l t
2.2 Safety Analysis In Exhibit 0 of the July 5, 1988 submittal, the licensee provided a revised I
safety analysis report to support the increased F and F i
previous safety analysis for Unit 1 Cycle I? was k rforme9.
Even though the with the same FAH and Fn of 1.70 and 2.50, respectively, as proposed for this amendment, the limitTng transients and accidents are reanalyzed.
This is because several changes have been made in the analysis meth,ralogy.
These changes include the following items:
~
l i
(1) An error in the surface heat trane.fer coefficients in the film boiling region of the analysis of the rod ejection event is corrected.
(2) The reliability factor applied to the Doppler coefficient is
' creased fror' 10% to 25%.
This change reflects the increase of Dopple coefficient uncertainty and 25% reliability factor bounds the uies.ertainty.
i (3) The low setpoint for the high neutron flux trip is increased from 25% of
{
i the rated power to 40%.
This change reflects a previously approved j
i change (Ref. 8) in the Prairie Island TSs for the neutron flux low j
setpoint in the power range.
i l
(4) The fuel rod bow penalty is reduced to 2.6%.
This change is a result of a previously approved (Ref. 9) reduction from 33,000 MWD /MTV to 24,000 MWD /MTU as the cutoff burnup for the maximum rod bow penalty calculation.
1 (5) The scram reactivity insertion curve is revised to insure conservatism in l
all cases.
The change reflects a revised model that takes into account the drag, buoyancy and gravity forces of the rod drop.
The new j
calculation tends to delay the scram reactivity insertion comparsd to the previous analysis of assuming a constant control Md velccity and is j
therefore acceptable.
i All limiting transients and potential accidents that are affected by these 4
i changes are reanalyzed.
These include fast and slow control rod withdrawal, I
loss of power to both reactor coolant pumps (i.e., loss uf flow), locked
[
rotor of one reactor coolant pump, loss of electric load, rod cluster control
}
l assembly (RCCA) ejection and fuel handling accidents that assumed a radial peaking factor (Fn) 1.70 as required by Regulatory Guide 1.25.
The mein i
(
steamline brsak iY not significantly affected by these char.ges, and the licensee has determined that it is bounded by the analysts of Unit 2 Cycle 10.
[
i The rest.31ts of analyses for the control rod withdrawal \\ransients, loss of flow and loss of load transients indicate that the minimum doparture from nucleate l
Doiling ratios (DN8R) calculated for these transients do not fall below the l
minimum ONBR limit of 1,17 for the WRB-1 critical heat flux correlation used and that the system pressures do not exceed the accepta1ce criterion of 110%
of the design pressure of 2500 psia.
For the Class IV locked rotar event, the ONBR falls below t'e 1.17 limit.
Fuel rod failure is assuned to occur when a rod exper' oce* NS, i.e., the DNBR is bulow the 1.17 limit, The number of failed r.t: eeo, is calculated to be 1.54%, well below the ecceptance criterion of less, than 4% fuel failure.
The peak system pressure also does not exceed the accep'e co criterion of 2750 psia.
For the RCCA rod ejection events initiated from hot full power and hot zero power conditions, the results indicate that the average hot spot fuel enthalpy remains below the acceptance criterion of 280 cal / gram and that the system pressure does not exceed the acceptance criterion of 2750 psia.
The fuel handling accidents have been reanalyzed considering the proposed change in the peaking factor, Fn from 1.J5 to 1.70 pursuant to 10 CFR Part 100, i
The results of this reanalysis Indicate that the existing fuel handling accident
. analysis is still tvunding with the high peaking factor.
In addition, our Safety Evaluation (Ref 10) at the time of plant licensing showed that with a peaking factor of 1.72, the analysis of a fuel handling accident conservatively estimated the following doses:
Two Hour Dose at Course of Accident Exclusion Boundary Dose at low (714 Meters)
Population Zone (2410 Meters)
Thyroid 33 rems 55 rems Whole Body 4 rems 6 rems These doses are within the guidance values in the NRC Standard Review Plan (NUREG-0800) for fuel handling accidents, and are well within the values of 10 CFR Part 100 guidelines which consist of 300 rem (thyroid) and 25 rem (whole body).
2.3
SUMMARY
The licensee proposed i's changes dealing with the values of F and F.
These peaking factors would De changed from the existing requiremenN where the k
assignec values are based on a function of each other to a constant value for F and F f 2.50 and 1.70, respectively.
The NRC staff concludes, based on theabovhHevaluation, that the proposed changes to the peaking factors are acceptable.
3.0 ENVIRONMENTAL CONSIDERATION
An Environmental Assessment and Finding of No Significant Impact has issued for these amendments (53 FR 35134, September 9, 1988).
4.0 CONCLUSION
We have concluded, based en the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; and (2) such activities will be conducted in compliance with the Commissions's regulations and issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the publ',c.
5.0 REFERENCES
1.
Letter from David Musolf (NSP) to Director of Nuclear Reactor Regulation, USNRC, "Prairie Island Nuclear Genersting Plant, Docket Nos.
50-282/50-306, License Nos. OPR-42/DPR-60, License Amendment Request Dated July 5, 19A8, Best Estimate LOCA Analysis," July 5, 1988.
2.
WCAP-10924-P, "Westinghouse Large-Break LOCA Best Estimate Methodology,"
l Volume 1, Jur.J 1986 Volume 2, Revision 1. April 1988.
i 3.
Information Report from William J. Dircks to the Commissioners, "Emergency Core Cooling System Analysis Methods, "SECY 82-472, November 17, 1983.
4.
Letter from Ashok C. Thadani to W. J. Johnson (Westinghouse), "Acceptance for Referencing of Licensing Topical Report, WCAP-10924 Westinghouse Large Break LOCA Best Estimate F>.thodology," August 29, 1988.
5.
Letter from David Musolf (NSP) to Director of Nuclear Reactor Regulation, USNRC, "Request for Exemption to Selected 10 CFR 50, Appendix K Requirements," July 28, 1988.
6.
Letter from D. C. Dilanni to D. W. Musolf (NSP), "Acceptance of Request for Exemption to Selected 10 CFR 50, Appendix K Requirements,"
September 16, 1988 7.
LetterfromDavidMusolf(NSP)toDirectorofNuclearReactorRegulation, USNRC, "Supplemental Information for License Amendment Request Dated July 5,1988, Best Estimate LOCA Analysis", August 5,1988.
8.
Letter from D. C. Dilanni to D. W. Musolf (NSP), "Amendment Nos. 83 and 76 to Facility Operating License Nos. DPR-42 and DRP-60: High Flux, Power Range (Low Setpoint) (TACS Nos. 66865 and 66866)," Nay 31, 1988.
9.
Letter from Carl Berlinger (USNRC) to E. P. Rahe, Jr. (Westinghouse),
"Request for Reduction in Fuel Assembly Burnup Limit for Calculation of Maximum Rod Bow Penalty," June 18, 1986.
10.
Safety Evaluation of the Prairie Island Nuclear Generating Plant Units 1 and 2 dated September 28, 1972, Page 15-4 and Table 15.1-1.
Principal Contributors:
Y. Hsii l
James A. Martin
~
Date:
Septe M r 16, 1988