ML15112A839
| ML15112A839 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 12/15/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML15112A836 | List: |
| References | |
| NUDOCS 7812280264 | |
| Download: ML15112A839 (13) | |
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p.F REGjj UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, 0. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 66 TO FACILITY OPERATING LICENSE NO. DPR-38, AMENDMENT NO. 66 TO FACILITY OPERATING LICENSE NO. DPR-47, AND AMENDMENT NO. 63 TO FACILITY OPERATING LICENSE NO. DPR-55 DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS NOS. 1, 2 AND 3 1.0 Introduction By letters dated September 18, 1978 and September 25, 1978 (Refer ences 1 and 2 respectively) Duke Power Company (DPC) has proposed changes to the Oconee Nuclear Station (0NS) Technical Specifica tions. Table 1 summarizes the proposed changes and indicates the applicability of each to changes to the three Oconee Units, ONS-1, ONS-2, or ONS-3.
Most of the proposed Technical Specification modifications are asso ciated with the refueling of ONS-2 for Cycle 4 operation. The in formation submitted by DPC in connection with this refueling is presented in References 3 and 4 which describe the fuel system design, nuclear design, thermal-hydraulic design, accident analyses, and startup test program.
The refueling of ONS-2 for Cycle 4 will result in a core loading consisting of 56 fresh Mark B-4 assemblies, 108 previously burned Mark 8-4 assemblies, nine previously burned Mark B-2, and four demon stration Mark C or Mark CR assemblies. In addition, the remaining (70) orifice rod assemblies will be removed from the core during the refueling outage. This will leave 106 vacant fuel positions which originally contained such orifice rod assemblies. The changes in the core loading and the removal of the orifice rod assemblies are the only physical modifications associated with the refueling.
The evaluation of DPC's proposed modifications to the Technical Specifications of ONS-1, 2, and 3 is presented in the following sections. For ONS-2, this evaluation has taken into consideration the proposed refueling of the core as described in Reference.3 and subsequent operation for the targeted 292 effective full power days (EFPDs) during Cycle 4.
-2 Table 1. Proposed Technical Specification Changes For Oconee Nuclear Station For Unit 2 Only
- 1. Modification to Core Protection Safety Limits (Figure 2.1-2B)*
- 2. Modification to Protective System Maximum Allowable Setpoints (Figure 2.3-2B)
- 3. Modifications to Rod Position Limits (Figures 3.5.2-1B1, 182, 2B1, and 2B2)
- 4. Modifications to Operational Power Imbalance Envelope (Figures 3.5.2-381 and 382)
- 5. Modifications to APSR Position Limits (Figures 3.5.2-4B1 and 4B2)
- 6. Reduction in FAH from 1.78 to 1.71
- 7. Increase in the allowed steady state quadrant tilt to 5%
and in the linear heat rate peaking increase associated with positive tilt to 7.50%.
For Units 2 and 3
- 8. Extension to Units 2 and 3 an allowance for operating above the power level cutoff associated with the rod position limits, provided the reactor has operated within 5% of the cutoff for more than two hours.
-3 Table 1. Proposed Technical Specification Changes For Oconee Nuclear Station (Contd)
For Units 1, 2, and 3
- 10.
Increase in the volume of boric acid solution in the Boric Acid Storage Tank from 980 Ft3 to 995 Ft3.
- 11. Modifications to Control Rod Operability and Surveillance Requirements
- All figures are in Reference 3.
-4 2.0 Evaluation of Modifications to ONS-2 Core Design 2.1 Fuel System Design We have evaluated the implications of introducing the 56 fresh Mark B-4 fuel assemblies and the nine once-burned Mark 8-2 fuel assemblies into the ONS-2 core and the subsequent operation at rated power for the intended 292 effective full power days.
Tables 4-1 and 4-2 of Reference 3 summarize the design character istics of the Mark B-4, Mark B-2, Mark C and Mark CR fuel types.
The fresh Mark B-4 assemblies are identical to the previously burned Mark B-4 fuel with regard to assembly mechanical design, fuel rod design and thermal design. The fuel designs of Mark B-4, Mark C and Mark CR fuel types have been evaluated for ONS-2 in associa tion with earlier refuelings and found acceptable (References 5 and 6).
The Mark 8-2, which fuel has been analyzed in the ONS-2 Densification Report (Reference 10), was part of several earlier ONS core loadings.
2.1.1 Cladding Creep Collapse Fuel rod cladding creep collapse analyses have been performed for the most limiting (i.e., most highly exposed) Mark B and Mark C assemblies to be included for Cycle 4. The analyses were performed according to the conservative methods and assumptions described in References 7 and 8 and approved by the NRC staff in Reference 9. These analyses show that the time to rod cladding collapse will be in excess of 30,000 effective full power hours. Because no Mark B or Mark C assembly will reach a total exposure as high as 30,000 EFPH during Cycle 4 (Table 4-1 of Reference 3), we conclude that cladding creep collapse will not occur during the cycle.
2.1.2 Cladding Stress and Strain With regard to cladding stress and strain,the Mark B-2 fuel is most limiting for Cycle 4 because of its low prepressurization and density.
For this fuel, the cladding stress due to differential pressure, temperature gradient or axial loads and restraints will not exceed the yield stress or ultimate strength of the material during Cycle 4 (Reference 10).
In Reference 7, the anticipated cladding strain for Mark B-2 fuel was shown to be less than the 1% plastic cladding strain limit for up to 55,000 MWd/MTU, well below the exposure to be accumulated by the end of Cycle 4.
We previously accepted these conclusions regarding cladding stress and strain for ONS-2 Cycle 3 (Reference 6) and we conclude that they are valid for Cycle 4 also.
g
-5 2.1.3 Fuel Thermal Design The thermal linear heat rate (LHR) limits have been established for the Cycle 4 fuel using the TAFY code (Reference 11) and assumed fuel densification to 96.5% of theoretical density. These limits are stated in Table 4-2 of Reference 3. The thermal LHR limits which ensure that fuel center melting does not occur are less re strictive than the LOCA LHR limits.
Because the LOCA LHR limits will be met by operating within the limiting conditions for opera tions contained in the ONS-2 Technical Specifications, the thermal LHR limits will also be met.
We conclude that the indicated thermal LHR limits are acceptable for preventing center melting of the Cycle 4 fuel and that the limits will not be exceeded.
2.2 Nuclear Design Figure 3-1 of Reference 3 indicates the core loading arrangement for ONS-2 Cycle 4; the initial enrichments and burnup distributions are given in Figure 3-2. Most of the fresh Mark B-4 assemblies will be loaded into locations on the edge of the core and-will be below fuel thermal limits. Similarly, the Mark C and Mark CR demonstra tion assemblies will be in non-limiting locations.
Reactivity control and power distribution control will be maintained by control rods, axial power shaping rods and boron shim.- The rod locations are given in Figure 3-3 of Reference 3.
The projected Cycle 4 length is 292 effective full power days with a cycle burnup of 9138 MWd/MTU.
Cycle 4 nuclear parameters including critical boron concentrations, control rod worths, Doppler coefficients, moderator coefficients, xenon worth and effective delayed neutron fractions have been calcu lated using the approved PDQ07 code (Reference 12). These are pre sented in Table 5-1 of Reference 3 and compared to the Cycle 3 values.
Shutdown margins have been calculated for beginning of cycle (BOC) and end of cycle (EOC) (Table 5-2 of Reference 3).
The calculated minimum shutdown margin during Cycle 4 is 1.45% Ak/k which is larger than the required value of 1% Ak/k by an adequate margin.
e-6 We conclude that the Cycle 4 nuclear design does not differ in a significant way from earlier cycles, that the nuclear parameters of Cycle 4 have been calculated by acceptable methods and are within the range of values expected for a cycle approaching an equilibrium cycle, and that the nuclear design has resulted in an adequate shut down margin. The nuclear design for ONS-2 Cycle 4 is, therefore, acceptable.
2.3 Thermal Hydraulic Design The thermal-hydraulic design conditions for ONS-2 Cycle 4 are in cluded in Table 6-1 of Reference 3. Only the reference design radial local power peaking factor and anticipated minimum departure from nucleate boiling ratio (DNBR) differ from the Cycle 3 values. The first of these differences is discussed below. The second is acceptable in that the minimum departure from nucleate boiling ratio, with densifi cation penalty, increases from 1.91 in-Cycle 3 to 1.98 for Cycle 4; 1.30 is the safety limit, thus the current Cycle 4 in this regard represents a slight increase in margin to the safety limit.
The effect of the demonstration Mark C and Mark CR assemblies on the ONS-2 thermal hydraulic design have been evaluated for earlier cycles (References 5 and 6). The continued use of the demonstration assem blies does not involve any physical effect not previously considered and is acceptable.
2.3.1 Removal of Orifice Rod Assemblies The most significant difference between the thermal hydraulic design for Cycle 4 and that for Cycle 3 is the removal of the 70 orifice rod assemblies (ORAs).
This will leave a total of 106 vacant fuel assem blies and will result in an increase in bypass flow from 8.34% for Cycle 3 to 10.4% for Cycle 4. The increased bypass flow also involves a decreased flow to fuel assemblies, and DPC has re-evaluated the effect of this modification on the reactor core DNBR safety limit.
The re-evaluation indicated that a decrease in the reference design radial-local peaking factor (FAH) from 1.78 to 1.71 compensates for the larger bypass flow so that no change in the DNBR safety limit will be necessary.
The DNBR safety limit was derived using the BAW-2 critical heat flux correlation (Reference 14).
Based on the sensitivity of the heat flux correlations, such as BAW-2, to small changes in flow, we have concluded that the proposed reduction in FH to 1.71 is adequate to offset the increased bypass flow.
2.3.2 Effect of Rod Bow on Thermal Design The effect of fuel rod bow has been reviewed generically in Reference
- 13.
Based on the rod bow model approved by the NRC staff, DPC has applied a rod bow DNBR penalty of 11.2% to all analyses that define plant operating limits and to design transients (Reference 3).
-7 The 11.2% penalty which has been applied includes a 1% contribution associated with pitch reduction due to fabrication tolerances and initial rod bow, and a 10.2% contribution from burnup dependent bowing. The 11.2% penalty is valid for a maximum burnup of 33,000 MWd/MTU and, therefore, bounds -the burnup expected for Cycle 4.
Based on the use of an approved model and a bounding assumed burnup, we conclude that DPC has adequately taken fuel rod bowing into account for the thermal design of ONS-2 Cycle 4.
3.0 Evaluation of Accidents and Transients The refueling of ONS-2 for Cycle 4 will not involve a change in the DNBR safety limit (Section 2.3 of this Report).
Plant operating limits, as proposed in Reference 2, have been established to compen sate for the effect of fuel rod bowing on DNBR.
The two pump coastdown, which is the limiting event with regard to reduction in DNBR,has been analyzed from an initial power level of 102% with a flux/flow trip set-point of 1.055.
The combined re duction in DNBR due to the transient and fuel rod bowing would not result in a DNBR below the safety limit value of 1.30 Other transients are discussed below.
As discussed in Section 2.2 of this Report, the nuclear parameters, which comprise a portion of the input to the accident and transient analyses, have been evaluated using acceptable methods. Of the transients and accidents considered in the ONS-2 FSAR (Reference 16),
the loss of electric power, steam generator tube failure, fuel handling accident, waste gas tank rupture, maximum hypothetical accident, and LOCA do not depend on nuclear parameters. Rod withdrawal accidents, the cold water accident, stuck or dropped rod accidents, steamline failure, and the rod ejection accident do depend on nuclear parameters.
We conclude, based on Tables 6-1 and 7-1 of Reference 3, that the Cycle 4 nuclear parameters are bounded by values assumed for accident analyses in the FSAR (Reference 16) and the ONS-2 Densification Report (Reference 10).
The applicable LOCA analyses for ONS-2 have been presented in Refer ence 17 which has been accepted by the NRC staff for generic application to B&W plants of the ONS-2 class (177-FA Lowered Loop Plants). The fuel densification report (Reference 10) describes the effect of densification on LOCA analyses and the use of the TAFY code (Refer ence 11) to calculate fuel rod internal pressure and pellet-volumetric average temperature. The latter parameters, which are part of the LOCA input, are also affected by enhanced fission gas release, but the original TAFY calculations did not include the effect. Calculations
4.0
-8 using the B&W code, TACO, (Reference 21) have shown that the internal pressures and average temperatures calculated using TAFY adequately bound the effects of enhanced fission gas release for up to 42,000 MWd/MTU fuel rod burrup, a higher burnup than will be attained during Cycle 4 operation.
Technical Specification proposals associated with LOCA LHR limits were presented in Reference 1. These proposed limits include a statistical combination of nuclear uncertainty factor, engineering hot channel factor and rod bow peaking penalty amounting to a 9%
net peaking penalty (Reference 15).
B&W has demonstrated that power spikes caused by densification need not be considered in LOCA or DNB analyses. These tests and analyses show that for LOCA the radiant heat transfer to the cool cladding surrounding the gap, where the peaking occurs, more than offsets the heat generated by the power spike. For DNB, which is a function of critical heat flux, B&W has shown that heat flux power spikes have a negligibly small effect on critical heat flux thus the effect on DNB is negligible. The staff has accepted these demonstrations and analyses in Reference 22.
We conclude that the refueling of ONS-2 for Cycle 4 will not result in kinetics parameters outside the bounds assumed for the FSAR analysis, and that no change in the DNBR safety limit is required. Furthermore, the effects of fuel row bowing, fuel densification, and enhanced fission gas release on safety limits and on all transients and acci dents, including LOCA, have adequately been taken into account.
Fuel misloadings for Cycle 4 which could result in departure from nucleate boiling (DNB) will be detected during the physics startup testing to be performed at the 8OC.
These tests have been described in References 3 and 4 and evaluated in Section 4.0 for this report.
Based on these conclusions, and the fact that the dose calculations of the FSAR assumed maximum peakings and burnups which bound all reloads, we further conclude that the consequences of transients or accidents during Cycle 4 will be no greater than previously evaluated.
There will be no increase in the probability of occurrence of any accident or transient, and no new type of accident or transient will be introduced as a result of the refueling. We, therefore, accept the transient and accident analyses presented for ONS-2 Cycle 4.
4.0 Startup Tests Startup tests have been proposed by DPC to provide assurance that ONS-2 has been loaded as intended.
The tests are described in Refer ences 3 and 4 and are consistent with the startup tests performed in association with recent B&W reloads.
We have reviewed the tests and consider them acceptable.
5.0 Evaluation of Technical Specification Changes Proposed modifications to the ONS-1, ONS-2 and ONS-3 Technical Specifications are listed in Table 1.
-9 The changes indicated in Items I through 5 of Table 1 are based on FLAME code calculations (Reference 18) applied accordinf to the descriptions in References 19 and 20.
For these calculations, the statistical cm7bination of nuclear uncertainty, engineering un certainty and rod bow peaking, as approved in Reference 15, was applied to the linear heat rate peaking. Change lo.
6 is discussed and justified in Section 2.3.1 of this Safety Evaluation.
The relation belteen linear heat rate peaking increase and quadrant tilt implied in Item 7 of Table 1 is based on information in Refer ences 19 and 23. Reference 23 was provided in connection with the review of the Unit 1 quadrant tilt technical specification.
We believe that the information in References 19 and 21, which shows that the quad rant tilt linear heat rate peaking increase is related to the quadrant tilt by a multiplication factor of 1.495, includes a sufficiently broad data base to apply to ONS-2. The licensee has proposed to increase the current quadrant tilt Technical Specification limit to 5% from 3.4%.
The quadrant tilt Technical Specification in conjunction with the control rod insertion limit and power imbalance limit Technical Specifications ensure that plant limiting conditions for operation are not exceeded.
These conditions ensure that limiting values of linear heat generation rate and peak enthalpy rise assumed in the safety analysis are not exceeded. These limiting values are not altered by the proposed Technical Specification change. The margin to safety and operating limits have not been altered. Hence, Change No. 7 is acceptable.
The increased tilt limit permits greater operating flexibility with no decrease in safety margin.
Based on our acceptance of the 1.075 peaking increase for quadrant tilt and the maximum allowed quadrant tilt of 5% just discussed, the acceptance of the 8% transient xenon peaking increase discussed below, the previous acceptance of other peaking factors, and the use of the approved FLAME code to derive the limits associated with Items 1 through 5, we conclude that these proposed Technical Specification changes are acceptable.
Items 8 and 9 in Table 1 are related to analyses of the design basis maximum xenon transient described in Reference 19 and performed using the FLAME code (Reference 18).
Reactor power levels, except for physics tests, are not permitted by Technical Specification 3.5.2.6 to be increased above the power level cutoff curves of the Rod Position Limits of Figures 3.5.2-18 and 1B2 of the Technical Specifications, unless xenon reactivity transients and the associated change in power distribution during power operation is limited by restricting the nonequilibrium xenon. The Reference 19 calculations show that if the provisions of Technical Specification 3.5.2.6 (including modification 8 of Table 1) are met, the transient xenon peaking increase needbe no greater than 80 to assure that linear heat rate limits are not exceeded.
The transient xenon peaking factor of 1.08 was used in deriving the limits associated with Items 1 through 5.
Based on the use of the accepted design basis maximum xenon transient and the application of an accepted calculational method,.we conclude that the modifications proposed in Items 8 and 9 are acceptable.
-10 Modification 10 of Table 1 applies to ONS-1, 2, and 3. The increase in the volume of boric acid in the boric acid storage tank has been proposed to assure that an adequate cold shutdown capability will be maintained. The PDQ07 code (Reference 12) was used to evaluate the negative reactivity effects of the boric acid for this purpose.
PDQ07 has been accepted by the NRC staff for calculations of this type and we consider it acceptable for the current application. We, therefore, conclude that modification 10 should be adopted.
Proposed modification 11 of Table 1 applies to ONS-1, 2 and 3. The control rod drive operability history, with one exception, has been favorable at the Oconee Station. The drive system has not experienced any binding or frictional problems nor has it failed to perform its in tended trip (scram) function. An electrical component of the drive system, the stator coil, has failed in the past due to an electrical short in the coil.
Stator failures have not prevented the affected rod from performing its required safety function, namely the trip function. A shorted stator makes it difficult to move a rod and occasionally an attempt to move such a rod causes it to drop into the core. Control rod drop events have been analyzed in the FSAR (Reference 16). They do not result in fuel damage. The stator is coupled to the rod only by a magnetic field. The licensee proposes to extend the periodic rod exercise interval from two weeks to one month, thus avoiding a situation where the rod must be exercised possibly causing the rod to drop into the core and at a time of possible high power demand from the electrical distribution system.
In a previous NRC staff evaluation of this problem regarding Rod 6 of Group 4 in Oconee Unit No. 2, issued with the July 6, 1978 License Amendment,we stated, "...., we agree with the licensee's conclusion that the circuit fault (i.e., stator short) discovered in Rod 6 would not prevent the rod from performing its assigned safety function." In our letter transmitting the License Amendment, we noted that the request for that amendment could have been avoided if the licensee had previously adopted the Standard Technical Specifications of Babcock & Wilcox designed reactors. The requested change puts the test interval for control rod movement in parallel with the Standard Technical Specifications.
As the previous history of rod motion has been favorable, as dis cussed above, we find the change in surveillance of rod motion from two weeks to one month to be acceptable. The remainder of the
11 Technical Specification changes in this section are of an editorial nature and since they clarify the meaning of Section 3.5.2, we find the changed wording acceptable. The definition of shutdown margin, and the accompanying limiting condition of operation, are unaffected by the changes.
6.0 Environmental Consideration We have determined that these amendments do not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that these amendments involve an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of these amendments.
7.0 Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendments do not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the amendments do not involve a significant hazards consideration, (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 amendmentswill not be inimical to the common defense and security or to.the health and safety of the public.
Dated:
December 15, 1978
-12 References
- 1. Letter from William 0. Parker, Jr., Duke Power Company (DPC), to Harold R. Denton, NRC, September 18, 1978.
- 2. Letter from W. 0. Parker, Jr., DPC, to H. R. Denton, NRC, September 25, 1978.
- 3. Oconee Unit 2 Cycle 4 Reload Report, BAW 1491, August 1978.
- 4. Letter from W. 0. Parker, Jr., DPC, to H. R. Denton, NRC, November 1, 1978.
- 5. Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendments Nos. 27, 27 and 23 to Facilty License Nos. DPR-38, DPR-47, and DPR-55,Duke Power Company, Oconee Nuclear Station 2, June 30, 1976.
- 6. Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendments Nos. 45, 45 and 42 to Facility License Nos. DPR-38, DPR-47, and DPR-5 Duke Power Company, Oconee Nuclear Station 2.
- 7. Oconee 2 Cycle 3 Reload Report, BAW-1452, Babcock & Wilcox, Lynchburg, Va., April 1977.
- 8.
Program to Determine In-Reactor Performance of B&W Fuels Cladding Creep Collapse, BAW-10084, Rev. 1, Babcock & Wilcox, Lynchburg, Va., November 1976.
- 9. Letter from A. Schwencer (NRC) to J. F. Mallary (B&W) dated January 29, 1975.
- 10. Oconee 2 Fuel Densification Report, BAW-1395, Babcock & Wilcox, Lynchburg, Va., June 1973.
- 11. C. D. Morgan and H. S. Kao, TAFY - Fuel Pin Temperature and Gas Pressure Analysis, BAW-10044, Babcock & Wilcox, Lynchburg, Va.,
May 1972.
- 12. H. A. Hassan, et al., Babcock & Wilcox's Version of PDQ07 User's Manual, BAW-10117, Babcock I Wilcox, Lynchburg, Va.,
June 1976.
-13
- 13.
Memo to D. B. Vassallo (NRC) from 0. F. Ross (NRC), Revised Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing on Thermal Margin Calculations for Light Water Reactors, February 16, 1977.
- 14. Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, BAW-10000A, May 1976.
- 15. Letter, S. A. Varga (NRC) to J. H. Taylor (B&W), "Comments on B&W's Submittal on Combination of Peaking Factors," May 13, 1977.
- 16.
Oconee Nuclear Station, Units 1, 2, and 3, Final Safety Analysis Report, Docket Nos. 50-269, 50-270, 50-287, Duke Power Co.
- 17. R. C. Jones, J. R. Biller, and B. M. Dunn, ECCS Analysis of B&W's 177-FA Lowered Loop NSS, BAW-10103A, Rev. 3, Babcock & Wilcox, Lynchburg, Va.
- 18.
C. W. Mays, FLAME 3 - A Three-Dimensional Nodal Code for Calculating Core Reactivity and Power Distributions, BAW-10124, Babcock & Wilcox, Lynchburg, Va., May 1976.
- 19.
Operational Parameters for B&W Rodded Plants, BAW-10078, September 1973.
- 20. Normal Operating Controls, BAW-10122, July 1978.
- 21. TACO - Fuel Pin Performance Analysis, BAW-10087.
- 22. Letter, S. A. Varga (NRC) to J. H. Taylor (B&W), "Update of BAW-10055 Fuel Densification Report," December 5, 1977.
- 23. Letter from W. 0. Parker, Jr., DPC, to E. G. Case, (NRC),
March 16, 1978.