ML19312C342
| ML19312C342 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 07/06/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19312C330 | List: |
| References | |
| NUDOCS 7912130874 | |
| Download: ML19312C342 (7) | |
Text
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4r UNITED STATES k*
NUCLEAR REGULATORY COMMIS$10N I,,
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j CASHINGTON, C. C. 20665 o, h7 f.!
s,.....f SAFETY EVALUATION BY THE OFFICE OF flUCLEAR REACTOR RcGULATI0f!
SUPPORTIflG AMEllDMENT NO. 63 TO FACILITY OPERATING LICENSE NO. DPR-38 AMENDf1ENT N0. 63 TO FACILITY OPERATIflG LICEf!SE N0. OPR-47 AMENDf1ENT tl0. 60 TO FACILITY OPERATING LICENSE N0. DPR-55 DUKE POWER C0fiPANY OCONEE NUCLEAR STATION, UNITS NOS.1, 2 and 3 DOCKETS NOS. 50-269, 50-270 and 50-287 1.b Introduction UI By letter dated May 30, 1978 supplemented by letters dated June 14, 23 and 28,1978(2,3&4), Duke Power Company (the licensee) requested amendments to Facility Operating Licenses Nos. DPR-38, DPR-47 and DPR-55.
The amendments would modify the comrron Technical Specifications for the Oconee Nuclear Station, Units Nos.
1, 2 and 3 for Cycle 4 operation of Unit No. 3.
2.0 Evaluation The Oconee 3 reactor core consists of 177 fuel assemblies.
The refueling for Cycle 4 will involve the removal of all 60 Batch 3 fuel assemblies, 5 Batch 1 fuel assemblies and relocation of the residual Batch 4, 4A and 5 fuel assemblies. The removed fuel will be replaced by 29 once-burned Batch 1 fuel assemblies, loaded in the control portion of the core and 36 new Batch 6 fuel assemblies occupying primarily the core periphery.
2.1 Fuel Mechanical Design The Batch 6 fresh fuel uses the flark B4 fuel assembly design re-viewed and accepted by us for use during Cycle 3.
Also, these types of fuel assemblies are currently operating in Oconee 1 and Arkansas Nuclear One, Unit No.1 (ANO-1) 7912130f7(/
m
, Creep collapse Mme was calculated to be in excess of 30,000 effective full ;,ower hours (EFPH) which is longer than the maximum three cycle design exposures. The calculation of creep collapse time was performed using the power history of the limiting fuel assembly.
As was done in Cycle 3 the CROV computercodewp})usedtopredictthecollapsetime(5).
The licensee statedl0 and we agree that the CROV code conservati'aly predicts cladding collapse.
Additional conservatisms used in the CROV calculations were ti at no credit was taken for fission gas release; the cladding thic.:-
ness used in CROV was the lower tolerance limit (LTL) of the as-built measurements; and the lowest as-fabricated pellet densities were assumed to be located in the worst case power re-gion of the core.
The fuel cladding strain analysis was performed usina a number of conservative assumptions:
maximum allowable fuel pellet diameter and density; lowest permitted tolerance for the cladding inner diameter; conservatively high local pellet burnup; and conser-vatively high heat generation rate.
This insures that the 1.0".
limit on cladding plastic circumferential strain is not violated.
The Batch 6 fuel assembly design is based upon established con-cepts and utilizes standard component materials.
Therefore, on the bases of the analyses presented and previous successful operations with equ.alent fuel, we conclude that the fuel mechanical design for Cycle 4 operations is acceptable and does not decrease the safety margin.
2.2 Fuel Thermal Design The Batch 6 fuel produces no significant differences in fuel thermal performance relative to the other fuel remaining in the core. As was done in the Cycle 3 reload calculations, the TAFY-3 computer code (7) capability was calculated using the linear heat rate (LHR)
The nominal LHR for Cycle 4 is 5.80 kw/ft and the LHR capability is 20.15 kw/ft.
During the last several years, data have become available that indicate the fission gas release rate from LWR fuel pellets increases with burnup.
This enhanced release at high burnup affects the fuel rod internal pressure and the pellet volumetric average temperature which are important inputs to the LOCA analyses.
m e
, The Oconee 3 reload was analyzed using the TAFY-3 fuel performance code that was approved prior to identification of the enhanced fission gas release at high buruup.
Anot her f uel perf ornwh r t ode for Oconee, TACO, was approved af ter enhanced release was identified and includes its effects.
The fuel supplier toe Ocenee 3 Cycle 4 Babcock A Wilcox Cont
- (B&W). statad that both the rod pressure and volumetric average fuel temperature calculated by TAFY-3 conser-vatively envelope those calculated by TAC 0.
Thus, the licensee's use of TAFY-3 to calculate the fuel rod pressure ana volumetric average temperature inputs for the LOCA analyses, have conservatively included the effects of enhanced fission gas release.
2.3 Nuclear Analysis The reactor core physics parameters for Oconee 3 Cycle 4 operatien were calculated using a PDQ0T computer code.
Since the core has not yet reached an equilibrium cycle, there were minor differences in the physics parameters between the Cycle 4 and Cycle 3 cores.
The licensee proposed a change in the plant Technical Specifications increasing the allowable steady state quadrant tilt from 3.41%
to 5.00%.
The additional peaking allowed is a result of the statistical combination of the nuclear uncertainty factor, the hot channel factor and the rod bow peaking penalty.
The licensee has also proposed to change the quadrant tilt, that if exceeded will require bringing the plant to hot shutdown within four hours, from 9.44% to 20.0%.
To compensate for this increase the proposed Technical Specifications reduce the core power and trip setpoints as quadrant tilt increases, insuring that the initial conditions for accident analyses are preserved.
Similar quadrant tilt limits are in the standard Technical Specifications for B&W plants, and in the Technical Specifications for Three Mile Island Nuclear Station, Unit No.1 and ANO-1 (Maximum quadrant tilt for AN0-1 is 25%.) We find the Technical Specification changes for quadrant tilt are acceptable and do not decrease the safety margin.
The licensee has proposed a Technical Specification change of the axial power shaping rod (APSR) position limits.
The APSR position limits would provide added control of power peaking to insure that peak power limits for loss of Coolant Accident (LOCA) conditions would not be violated.
a
m We find that, based on our review of the licensee's nuclear analysis techniques and their commitment to perform acceptable physics startup testing, the Oconee 3 nuclear analysis is accepta ble.
We also find the proposed Technical Specifications of APSR position limits and the usual regulating control rod and imbalance limits, which assure that the LOCA LHR limits are not exceeded, are acceptable.
2.4 Thermal-Hydraulic Analyses The licensee is proposing to remove all the Orifice Rod Assembli9s (0RA) and has revised the thermal-hydraulic analysis accordinglyl2),
The core bypass flow has increased to 10.4% (106 ORA's removed) from the 8.34% value used for Cycle 3 analysis (44 ORA's removed).
To offset the increase in core bypass flow, the reference design radial times local peaking factor (Fah) has been reduced from 1.78 to 1.71.
The most limiting transient, the loss of two reactor coolant pumps, has been reanalyzed with an Fah of 1.71 and the minimum Departure from Nucleate Boiling Ratio (DNBR) remains
~
above 1.3.
The licensee has also reviewed the pressure-temperature limits and power imbalance and has stated that they remain acceptable.
The Technical Specifications previously proposed for Cycle 4 (before removing the ORA's) are more conservative than those that could % proposed with the lower Fah.
Therefore, the licensee has not aposed a change to the Technical Specifications as a result of removing the ORA's.
The licensee has applied an 11.2% rod bow penalty to all analyses that define plant operating limits and to design transients.
A 1% credit for the flow area (pitch) reduction factor has been applied to offset the rod bow penalty.
We have reviewad the licensee's analyses and conclude that the thermal hydraulic analyses for Oconee 3 Cycle 4 are acceptable.
2.5 Accident and Tran:ient Analysis The accident and transient analysis provided b.y the licensee demon-strates that the Oconee Final Safety Analysis Report (FSAR) analyses conservatively bound the predicted conditions of tne Oconee Unit 3 Cycle 4 core and are, therefore, acceptable.
Each FSAR accident
m 5-analysis has been examined, with respect to changes in Cycle 4 parameters, to determine the effects of the reload and to insure that performance is not degraded during hypothetical transients.
The core thermal parameters used in the FSAR accident analysis:
Were design operating values based on calculated values plus uncertainties.
FSAR values of core thermal parameters were com-pared with those used in the Cycle 4 analysis.
The effects of fuel densification on the FSAR accident results have been eval '
uated and are reported in the Oconee Unit 3 fuel densification report (8).
Since Cycle 4 reload fuel assemblies contain fuel rods with theoretical density higher than those considered there, the conclusions derived in that report are valid for Oconee Unit 3 Cycle 4.
The limited conditions of the analyses for transientsin Cycle 4 are bounded by the initial conditions for previous analyses performed in either the FSAR, the fuel densification report or previous reload submittals.
Calcula-tional techniques and methods for Cycle 4 analyses remain con-sistent with those used for the FSAR.
No new dose calculations were perfonned for Cycle 4 operation.
The dose considerations in the FSAR are based on maximum peaking and burnup for all core cycles; therefore, the dose considerations are independent of the reload batch.
2.6 ECCS Analysis This matter has been separately considered by the staff and is i
discussed in the NRC's Order in the captioned matter dated April 26,1978, and in the NRC's Exemption in the captioned matter dated July 6,1978, which accompanies this Safety Evalu-ation Report.
s 2.7 Physics Startup Tests The physics startup test program for Cycle 4 as stated in Section i
9 of the reload submittal has been reviewed.
Additional information was requested and supplied by letters dated June 23 and 28,1978.
The physics st6rtup test program includes zero power measurements of critical baron concentration, temperature coefficients, ejected control rod worth and control rod group reactivity worth.
Power distribution, temperature coefficient and power coefficient measurements will be made at higher powers.
The acceptance criteria and the actions to be taken if the acceptance criteria l
are not met were reviewed as well as the tests. The licensee has stated that the action to be taken if the sum of the worth of groups 5, 6 & 7 differs from the predicted by more than +10%, is to measure group 4 and that if the sum of the worths of grolips 4, 5, 6 and 7 differs from the predicted by more than +10%,
addition measurements, as well as evaluation of the discrepancy, will be made.
,.i.
1 A summary of the results of this test program will be submitted to the flRC within 45 days after completion of the program.
This entire program has been reviewed by the flRC staff and found to be acceptable.
3.0 Conclusions Based on our evaluation of the reload application and available information, we conclude that it is acceptable for the licensee to proceed with Cycle 4 operation of Oconee 3 in the manner proposed.
We have reviewed the proposed changes to the Technical Specifications and find them acceptable.
We have determined that the amendments do not authorize a change in effluent types or total amounts nor an increase in power level Having and will not result in any significant environmental impact.
made this dettrmination, we have further concluded that the amendments involve an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR s51.5(d)(4), that an environmental impact statement, or negative declaration and environ-mental impact-appraisal need not be prepared in connection with the issuance of this amendments.
On the' basis of the foregoing, we have concluded, based on the consider-ations discussed above, that: (1) because the amendments do not involve a significant increase in the probability or consequences of accidents pre-viously considered and do not involve a significant decrease in a safety margin, the amendments do not involve a significant hazards con-sideration; (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
Dated: July 6,1973 i
I-l
m REFERENCES 1.
Ltr. from William 0. Parker, Jr., Duke Power Company (DPC) to R. Reid, U. S. Nuclear Pegulatory Commission (NRC), 5/30/78, forwarding the Oconee Nuclear Station, Unit No. 3, Cycle 4 Reload Report, BAW-1486.
2.
Ltr..from William 0. Parker, Jr. (DPC) to R. Reid (NRC), 6/14/78, forwarding Amendment 1 to BAW-1486.
3.
Ltr. from William 0. Parker, Jr. (DPC) to R. Reid (NRC), 6/23/78, forwarding additional information.
4.
Ltr. from Willian O. Parker, Jr. (DPC) to R. Reid (NRC), 6/28/78, forwarding additional information.
5.
BAW-1453, "0conee Unit 3, Cycle 3, Reload Report," 8/77.
j 6.
BAW-10084, Rev.1, " Program to Determine In-Reactor Performance of B&W Fuels - Cladding Creep Collapse," 11/76.
7.
BAW-1004, "TAFY-Fuel Pin Temperature and Gas Pressure Analysis,"
5/72.
8.
BAW-1399, "0conee 3 Fuel Densification Report," 11/73.
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n 755.-01 UNITED STATES NUCLEAR REGULATORY C0PMISSION DOCKETS N05. 50-269, 50-270 AND 50-287 DUKE POWER COMPANY NOTICE OF ISSUANCE OF AMENDMENTS TO FACILITY OPERATING LICENSES The U. S. Nuclear Regulatory Commission (the Commission) has issued Amendments Nos. 63, 63, and 60 to Facility Operating Licenses Nos. DPR-38, DPR-47 and DPR-55, respectively, issued to Duke Power Company for operation of the Oconee Nuclear Station, Units Nos.1, 2 and 3, located in Qconee County, South Carolina.
The amendments are effective as of the date of issuance.
These amendments revise the Technical Specifications to support the operation of Oconee Unit No. 3 at full rated power during Cycle 4 after core reload and removal of the orifice rod assemblies from the core.
The application for the amendments complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations.
The Commission has made appro-priate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendments.
Prior public notice of these amendments was not required since the amendments do not involve a significant hazards consideration.
The Comission has determined that the issuance of these amendmants
- vill not result in any significant environmental impact and that pursuant to 10 CFR 651.5(d)(4) an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of these amendments.
7?aRFo g
For further details with respect to this action, see (1) the appli-cation for amencments dated May 30, 1978, as supplemented June 14, 23 and28,1978,(2) Amendments Nos. 63, 63 and 60 to Licenses Nos. DPR-38, DPR-47 and DPR-55, respectively, and (3) the Commission's related Safety Evaluation.
All of these items are available for public inspection at the Commission's Public Document Room,1717 H Street, N. W., Washington, D. C. and at the Oconee County Library, 201 South Spring Street, Walhalla, South Carolina 29691.
A copy of items (2) and (3) may be obtained upon request addressed to the U. S. Nuclear Regulatory Commission, Washington, D. C.
20555, Attention:
Director, Division of Operating Reactors.
Dated at Bethesda, Maryland this 6th day of July 1978.
FOR THE NUCLEAR REGULATORY COMMISSION A
Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors
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