ML15113A130

Safety Evaluation Supporting Amends 132,132 & 129 to Licenses DPR-38,DPR-47 & DPR-55,respectively

ML15113A130
Person / Time
Site:	Oconee
Issue date:	11/23/1984
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML15113A127	List: ML15113A126 ML15113A130
References
NUDOCS 8412060554
Download: ML15113A130 (16)

Text

C, pjt REGUZ1 0

UNITED STATES ILEAR REGULATORY COMMISSI WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 132 TO FACILITY OPERATING LICENSE NO. DPR-38 AMENDMENT NO.132 TO FACILITY OPERATING LICENSE NO. DPR-47 AMENDMENT NO.129 TO FACILITY OPERATING LICENSE NO. DPR-55 DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS NOS. 1, 2, AND 3 DOCKETS NOS. 50-269, 50-270 AND 50-287 INTRODUCTION By letter dated September 11, 1984, as supplemented on October 22, 26, and November 1, 1984 (Ref. 1, 6, 17, and 18), Duke Power Company (the licensee) proposed changes to the Technical Specifications (TSs) of Facility Operating Licenses Nos. DPR-38, DPR-47, and DPR-55 for the Oconee Nuclear Station, Units Nos. 1, 2 and 3. These amendments would consist of changes to the Station's common TSs. Oconee Unit 1 is currently completing a refueling outage and was originally scheduled for plant restart on November 26, 1984 (Ref. 1).

References 25 and 26 state that Oconee Unit 1 shutdown earlier than scheduled and startup is scheduled for November 24, 1984.

These amendments would authorize proposed changes to the Oconee Nuclear Station TSs which are required to support the operation of Oconee Unit 1 at full rated power during the upcoming Cycle 9. The proposed amendments would change the following areas:

1.

Core Protection Safety Limits (TS 2.1);

2.

Protective System Maximum Allowable Setpoints (TS 2.3);

3.

Rod Position Limits (TS 3.5.2); and

4.

Power Imbalance Limits (TS 3.5.2).

To support the license amendment application, the licensee submitted a Babcock and Wilcox (B&W) report, BAW-1841 (Ref. 2), "Oconee Unit 1, Cycle 9 Reload Report," as an attachment to Reference 1. A summary of the Cycle 9 operating parameters is included in the report, along with safety analyses.

The Cycle 9 core consists of 177 fuel assemblies, each of which is a 15 by 15 array containing 208 fuel rods, 16 control rod guide tubes, and -one incore instrument guide tube. Cycle 9 is to have a length of approximately 410 effective full power days (EFPD) of operation.

As has been the case for Cycle 8, Cycle 9 will be operated in a rods-out, feed-and-bleed mode with core reactivity control supplied mainly by soluble boron in the reactor coolant and supplemented by 61 full length control rod assemblies (CRAs) and 60 burnable poison rod assemblies (BPRAs). In addition, eight axial power shaping rods (APSRs) are provided for additional control of the axial power distribution. The licensed core full power level remains at 2568 MWt.

8412060554 841123 PDR ADOCK 05000269 P

PDR

-2 Du ring the refueling outage, 113 fuel assemblies will be reinserted similar to those previously used and 64 fuel assemblies will be discharged and replaced by new but substantially similar assemblies of the Mark BZ type.

The Mark BZ fuel assemblies are the same as previously approved and used assemblies in terms of fuel rods, end grid, end fittings, and guide tubes and differ only slightly from previously approved assemblies in the use of Zircaloy spacer grids rather than Inconel Intermediate spacer grids. The Mark BZ fuel assemblies are discussed further in Section 7.0.

EVALUATION 1.0 Evaluation.of the Fuel System Design 1.1 Fuel Assembly Mechanical Design Cycle 9 will contain 45 Mark B and four Mark BZ fuel assemblies in Batch 9B, 60 Mark B and four Mark GdB assemblies in Batch 10C and 10B, respectively, and 64 Mark BZ assemblies in Batch 11.

All of these fuel assemblies are mechanically interchangeable. The Mark BZ design is similar to the Mark B fuel assembly except that the six intermediate Inconel spacer grids have been replaced with Zircaloy grids. Four Mark BZ assemblies have been previously approved as demonstration assemblies in the last two cycles of Unit 1.

For Cycle 9, a significant portion of the core will contain Mark BZ fuel.

The design (Ref. 3) of these assemblies has been reviewed and approved by the NRC staff (Ref. 4).

However, in our Safety Evaluation Report on Asymmetric LOCA Loads for Oconge Units 1, 2 and 3 (Ref. 5),

we identified a measured critical load ( crit) versus maximum spacer grid load an PeTesulting temperature rise of 120F in peak cladding temperature due to a fully collapsed grid (41% of flow area reduction) in the core peripheral assemblies for standard fuel.

At our request, the licensee has submitted a corresponding analysis for an Oconee mixed core and a pure Mark BZ fuel core considering the reduction in strength due to the use of Zircaloy grids (Ref. 6).

The PCT increase for Mark BZ fuel was estimated on the basis of calculations performed for Mark B fuel.

The Mark B PCT analysis assumed the maximum flow area reduction of 41% along the entire assembly. Since dynamic response analyses showed that the maximum horizontal impact forces and maximum flow area reduction occur on the two mid-height spacer grids for both the Mark B and Mark BZ assemblies and the calculated maximum flow area reduction for Mark BZ fuel was 37%, we conclude that the assumptions used in the Mark B analysis are conservative. Therefore, we believe that these conservatisms as well as the similarities in the grid geometry justify the estimation of the PCT increase for Mark BZ fuel on the basis of calculations performed for Mark B fuel.

The PCT increase was evaluated for the racking failure mode only since this case resulted in a larger flow area reduction than the crushing failure mode. Since the maximum horizontal impact forces occur in a peripheral fuel assembly, any racking or crushing failures would also be observed there. For these reasons, the PCT analysis for the Mark B fuel is expected to bound the analysis for the Mark BZ fuel.

The Mark.BZ assemblies are, therefore, acceptable for use in Cycle 9 and future cycles.

-3 The 62 retainer assemblies used on the two fuel assemblies that contain regenerative neutron source assemblies and on the 60 assemblies that contain BPRAs will undergo a fourth cycle of irradiation during Cycle 9.

Based on the results of an examination of retainers which have undergone irradiation during the last three cycles, we conclude that a fourth cycle of irradiation is acceptable.

1.2 Fuel Rod Design In addition to the Mark B and the Mark BZ fuel assemblies, four assemblies (Batch 10B) will contain fuel pellets containing both urania (UQ ) and gadolinia (Gd 03 ) as described in Reference 7. These four Mark GdB lead test assemblibs (LTAs) are part of a joint Duke Power/Babcock and Wilcox/

Department of Energy program to develop and demonstrate an advanced fuel design incorporating urania-gadolinia for extended burnup in pressurized water reactors. Since the addition of four assemblies containing gadolinia to the Oconee 1 Cycle 9 core does not affect the operating limits in the Technical Specifications nor adversely affect the existing safety analyses, we approve the continued use and irradiation of the LTAs in Cycle 9.

However, this should not be construed as an approval of the urania-gadolinia design for full scale applications.

1.2.1 Cladding Collapse The licensee has stated that the cladding collapse time for the most limiting Cycle 9 assembly (including the gadolinia-bearing LTAs) was conservatively determined to be greater than the maximum projected residence time for any Cycle 9 assembly. The creep collapse analysis used the CROV computer code (Ref. 8).

These methods and procedures have been reviewed and approved by the NRC staff. We conclude that cladding collapse has been appropriately considered and will not occur for Cycle 9 operation.

1.2.2 Cladding Stress and Strain The cladding stress and strain analyses for the Cycle 9 fuel designs, including the gadolinia LTAs and the gray APSRs, are either bounded by conditions previously analyzed for Oconee Unit 1 or were analyzed specifically for Cycle 9 using methods and limits previously reviewed and approved by the NRC staff. We conclude that the analysis of cladding stress and strain has been appropriately considered for Cycle 9 operation.

00

-4 1.3 Fuel Thermal Design The thermal behavior of all fuel in the Cycle 9 core, with the exception of the gadolinia-bearing LTAs, is virtually identical.

The thermal analysis for reinserted Batches 9B, 10B, 10C and feed Batch 11 fuel was performed with the approved TACO2 (Ref. 9) code using the approved meth odology described in Reference 10. The centerline fuel melt (CFM) limits of 20.5 kw/ft for UO fuel and 17.6 kw/ft for UO Gd 20 fuel were predicted using TACO2. These 3atter Mark GdB LTAs are loaded i the core in such a manner so as to ensure that there is sufficient margin to offset any negative impact on the loss of coolant accident (LOCA) kw/ft limits discussed in Section 4.0 of this Safety Evaluation (SE).

In addition, these LTAs will be limited to a design peak of 1.67 to ensure that they are not thermally limiting. These CFM limiting values are incorporated into the TSs, and we find them acceptable.

Standard Review Plan (SRP) 4.2,Section II.S.1(f), contains the requirement that the fuel rod internal gas pressure should remain below normal system pressure during normal operation unless otherwise justified. Based on TACO2 analyses, the licensee has stated that the internal pressure in the highest burnup rod will not reach the nominal Reactor Coolant System (RCS) pressure of 2200 psia. We find this acceptable and conclude that the fuel rod internal pressure limits have been adequately considered for Cycle 9 operation.

2.0 Evaluation of Nuclear Design The nuclear design parameters characterizing the Oconee Unit 1 Cycle 9 core have been computed by methods previously used and approved for B&W reactors. Comparisons are made between the physics parameters for Cycle 8 and 9. Changes in the radial flux and burnup distributions as well as changes in rod groupings and the gray APSRs (described below) between cycles account for the differences in control rod worths, including ejected and stuck rod worths. All safety criteria are still met.

The highly-absorbing (black) APSRs utilized during the previous cycles have been replaced by less absorbing (gray) APSRs in Cycle 9. These gray APSRs have a greater absorber length than the previously employed ones and utilize an Inconel absorber instead of the previous silver indium-cadmium (Ag-In-Cd) alloy. Since gray APSRs are being utilized, there are now eight control rods in group 7 and twelve in group 5 to reduce the negative offset response to the group 7 rod movement.

Previous cycles utilizing black APSRs contained twelve rods in group 7 and eight in group 5. Calculations with approved B&W models were used to verify that these gray APSRs provide adequate axial power distribution control and will not adversely affect Cycle 9 operation.

Revisions to the TSs to account for these changes for Cycle 9 operation were made in accordance with methods and procedures found acceptable in connection with previous reloads. The replacement of the black APSRs by gray APSRs and the changes in control rod groupings for Cycle 9 are, therefore, acceptable.

-5 Shutdown margin calculations for Cycle 9 include the effects of poison material depletion, a 10% calculational uncertainty, and flux redistribution as well as a maximum worth stuck rod. Beginning and end-of-cycle shutdown margins show adequate reactivity worth exists above the total required worth during the cycle. The required shutdown margin is 1.00%A k/k, the shutdown margins at the beginning and end-of-cycle are 3.15%Ak/k and 2.35%A k/k, respectively.

The four Mark GdB LTAs which were inserted at the beginning of the last cycle and had an initial enrichment of 4.0 weight percent U-235 will undergo a second cycle of irradiation during Cycle 9. Based on a reduction in U-235 enrichment due to the previous cycle burnup of over 15,000 MWD/MTU for these four LTAs, B&W has calculated the Mark GdB assemblies average enrichment at beginning of Cycle 9 to be 2.48 weight percent U-235 with the highest enriched pin being approximately 2.57 weight percent U-235.

The effect of these fuel assemblies on the nuclear design continues to meet all criteria including those applicable to radial power peaking, ejected rod worths, moderator temperature coefficient (MTC), and shutdown margin.

Based on our review, we conclude that approved methods have been used, that the nuclear design parameters meet applicable criteria and that the nuclear design of Oconee 1 Cycle 9 is acceptable.

3.0 Evaluation of Thermal-Hydraulic Design Cycle 9 fuel includes 64 fresh Mark BZ fuel assemblies, four irradiated Mark GdB LTAs, and four irradiated Mark BZ demonstration assemblies, all of which incorporate Mark BZ spacer grids. The effect of the higher pressure drop caused by these grids, and by the BPRA retainers, is a slightly lower flow in the Mark 87 assemblies. Therefore, the departure from nucleate boiling (DNB) margin for these assemblies is reduced. In order to preserve the DNB margin, the radial - local design power peaking has been reduced to 1.67 for the Mark-BZ and Mark GdB assemblies for Cycle 9 only. All other Mark B assemblies continue to have a 1.71 design radial times local peak. The maximum expected peaking during Cycle 9 is 1.416. We concur that the reduction in allowable power peaking limits compensates for the reduced thermal margin so that required safety margins are maintained.

The safety limits presented for Cycle 9 have been generated with the BAW-2 (Ref. 11) used in the previous cycle, and 8WC (Ref. 12) critical heat flux (CHF) correlations. The BWC correlation was used to predict the DNB behavior of the Mark BZ fuel assemblies with a departure-from nucleate boiling ratio (DNBR) of 1.18 corresponding to a 95 percent probability at a 95 percent confidence level that DNB will not occur.

The subchannel analysis for Cycle 9 used the CHATA and TEMP core thermal hydraulic codes. B&W has previously provided data comparisons to the NRC which the staff concluded justified the use of the BWC correlation with these codes.

-6 A B&W topical report (Ref. 13) discussing the mechanisms and resulting effects of bowing in B&W fuel has been reviewed and approved (Ref. 14).

The report concludes that the DNBR penalty due to rod bow need not be imposed for those assemblies with significant bow because the power production capability of the fuel decreases sufficiently with irradiation to offset the effects of bowing. Post irradiation measurements on Mark BZ lead demonstration assemblies verified that the methodology of Reference 13 conservatively predicts the rod bow in Mark BZ assemblies also.

Therefore, we conclude that no rod bow penalty need be considered for Cycle 9 operation.

4.0 Safety Analyses The important kinetics parameters for Cycle 9 have been compared to the values used in the Final Safety Analysis Report (FSAR) and/or the densification report. The licensee has shown that the Cycle 9 values are bounded by those previously used. The licensee has also determined that the initial conditions of the transients in Cycle 9 are bounded by either the FSAR, the fuel densification report, or previous reload analyses. These analyses have been previously accepted by the NRC.

B&W has performed a generic LOCA analysis for the B&W 177-FA, lowered loop nuclear steam supply system (NSSS) using the final acceptance criteria (FAC) Emergency Core Cooling System (ECCS)evaluation model (Ref. 15).

The combination of average fuel temperature as a function of linear heat rate (LHR) and the lifetime pin pressure data used is conservative relative to those calculated for this cycle. Three sets of bounding values for allowable LOCA peak LHRs for Cycle 9 are given as a function of core height. These limits apply during the periods 0 to 30 EFPD, 30 to 250 EFPO, and for the balance of the cycle. These results are based upon a bounding analytical assessment of NUREG-0630 on LOCA and operating LHR limits performed by B&W (Ref. 16). The B&W analyses have been approved by the NRC staff and the LHR limits are satisfactorily incorporated into the TSs for Cycle 9 through the operating limits on rod index and axial power imbalance.

5.0 Technical Specification Modifications Oconee Unit 1 Cycle 9 TSs have been modified to account for changes in power peaking and control rod worths, the replacement of black APSRs by gray APSRs, use of the BWC CHF correlation, and elimination of the DNBR rod bow penalty factor. We have reviewed the proposed specification re visions for Cycle 9. These changes concern the (1) Core Protection Safety Limits of Specification 2.1, (2) Protective System Maximum Allowable Setpoints of Specification 2.3, and (3) Rod Position Limits and Operational Power Imbalance Envelope of Specification 3.5.2. On the basis that approved methods were used to obtain these limits, we find these TS modifications acceptable.

-7 In Reference 18, the licensee provided clarification to Reference 1, which transmitted the TS amendment request to support the operation of Oconee Unit 1 during Cycle 9. Specifically, Reference 18 included a minor revision to the text comprising "Bases-Unit 1" of TS 2.1 and changed the words "22 percent" to "the CHF correlation quality limit".

This revision makes the TSs consistent with transition cores containing both Mark B and Mark BZ type fuel assemblies. As discussed in Reference 1, the BAW-2 and the BWC CHF correlations are applicable to the Mark B and the Mark BZ type fuel assemblies, respectively. The present wording is applicable only to a full Mark B core using the BAW-2 correlation.

Therefore, the change in wording reflects the use of both the BAW-2 (for.Mark B) and the BWC (for Mark BZ) correlations in the transition Cycle 9 core, and thus merely provides the required generality for the description. The licensee states that the omission of the change in the Reference 1 submittal constitutes an administrative oversight and this revision possesses no significance with respect to the safety analyses included in the original submittal (Ref. 1).

6.0 Evaluation Findings

We have reviewed the fuels, physics, thermal-hydraulic, and accident information presented in the Oconee Unit 1 Cycle 9 reload report. We find the proposed reload and the associated modified TSs acceptable.

7.0 Mark BZ Fuel Assembly Design Review - Introduction By letter dated October 7, 1983 (Ref. 19) as supplemented on October 22, 1984 (Ref. 6), the licensee submitted topical report BAW-1781P, "Rancho Seco Cycle 7 Reload Report, Volume 1, Mark BZ Assembly Design Report",

for NRC staff review. The same report was submitted by Sacramento Municipal Utility District (SMUD, Ref. 20).

The Babcock & Wilcox Mark BZ fuel design is based on the approved Mark B fuel assembly design.

The Mark BZ fuel assembly has an array of 15x15 fuel rods with six Zircaloy intermediate spacer grids, replacing the six Inconel Intermediate spacer grids of the Mark B fuel assembly. The other components such as fuel rods, end grids, end fittings, and guide tubes are the same for both designs.

Since the Mark B fuel design has been approved, this Safety Evaluation will address only the adequacy of the Zircaloy grids of the Mark BZ fuel design. Therefore, the NRC staff has reviewed the Rancho Seco Topical and has found acceptable the Mark BZ fuel design for all lowered-loop B&W designed 177 fuel assembly plants (Ref. 4).

7.1 Fuel Mechanical Design In order to maintain similar mechanical strength, the Zircaloy grid is made wider and thicker than the Inconel grid. However, the total assembly weight is about the same for both designs because fuel and cladding contribute much more in weight. Although most aspects of.the Mark BZ fuel performance will not differ from those of the Mark B fuel, we raised several questions about possible deviations from previous Mark B fuel analysis. In a letter dated July 13, 1984 (Ref. 21),

SMUD provided responses to address our questions. We will discuss these differences in the following sections.

-8 7.2 Design Bases In response to our question concerning design bases, the licensee states that the design bases for Mark BZ fuel are identical to those approved for Mark B fuel.

In addition, the design evaluations to verify the adequacy of the Mark BZ fuel design were performed according to SRP Section 4.2.II.C. As far as mechanical design is concerned, our evaluation has determined that there is little difference between the Mark B and Mark BZ fuel designs with respect to the criteria and limits of this section of the Standard Review Plan. We therefore con clude that design bases of the Mark BZ fuel assembly are acceptable.

7.3 Holddown Spring In the past, the B&W Mark B fuel has experienced significant holddown spring failures. We questioned the adequacy of the Mark BZ fuel design regarding the holddown spring failure. SMUD stated that Mark BZ fuel has incorporated several major changes to improve holddown spring performance. They include increasing wire diameter, changing to a more fatigue-resisting material, and tighter fabrication control.

SMUD concluded that these procedures will reduce the possibility of fatigue induced failures. Inasmuch as SMUD has provided assurance that the holddown spring is redesigned to alleviate the past problem, we conclude that holddown spring failure is adequately addressed in the Mark BZ fuel design, subject to acceptable results from the holddown spring surveillance program for Mark B fuel described in Reference 24.

7.4 Assembly Liftoff The function of the holddown spring is to maintain the fuel assemblies seated in the lower core plate during the worst-case hydraulic load.

The new holddown spring design raised a concern as to whether holddown capability remains intact. SMUD's analysis showed that there is enough positive holddown margins for Mark BZ fuel to prevent assembly liftoff in the most limiting condition of maximum hydraulic lift force.

We thus conclude that the Mark BZ fuel assembly will maintain its holddown capability during the worst-case hydraulic load.

7.5 Seismic and LOCA Loads Appendix A to SRP 4.2 describes a fuel assembly structural response analysis under combined seismic and LOCA loads including asymmetric blow down loads. Although the Mark BZ fuel assembly has mechanical strength similar to the Mark B fuel assembly, we will require a plant-specific analysis of combined seismic-and-LOCA loads for mixed core and a pure Mark BZ fueled core. It is permissible to perform a bounding analysis or to make a comparison with the previous approved Mark B fuel results using approved methods to demonstrate conformance to Appendix A of Standard Review Plan Section 4.2.

-9 7.6 Post-Irradiation Surveillance A lead test assembly of Mark BZ fuel was irradiated earlier in Oconee 2 and was examined non-destructively after discharge. The result showed that the fuel performed adequately.

In response to our questions, SMUD indicated that B&W has a surveillance plan of visual examinations on selected assemblies for a total of nine demonstration assemblies with Zircaloy spacer grids, which are currently being irradiated in Oconee 1. Visual examination includes water channel, holddown spring, and length measurement at the end of each cycle. Considering.that Mark BZ fuel has only one significant change (spacer grids), we conclude that the post-irradiation surveillance, though minimal, is acceptable per SRP 4.2 guidelines.

7.7 Thermal Hydraulic Design The acceptance criterion specified in Section 4.4 of the Standard Review Plan for the thermal hydraulic design requires that there is at least a 95 percent probability at a 95 percent confidence level that the hot fuel rod in the core does not experience a DNB during normal operation or anticipated operational occurrences. The safety analysis of the Mark BZ fuel design must demonstrate that this criterion is met.

7.7.1 Hydraulic Characteristics Since the Zircaloy spacer grid in the Mark BZ fuel design has a grid height and grid strip thickness larger than.the Inconel grid used in the Mark B fuel, and since the outer grid strips of the Zircaloy grid lead-in taps, the Mark BZ fuel has higher hydraulic resistance than than the Mark B fuel.

B&W has performed flow tests of a full length Mark BZ prototype fuel assembly using the Control Rod Drive Line facility. A fuel bundle flow distribution test was also performed using laser Doppler Velocimeter. These tests provided data for development of grid form loss coefficients on both an assembly and a subchannel basis.

The pressure drop across the Mark BZ assembly is found to increase by less than 3 percent over the Mark B assembly and therefore, its impact on the reactor system flow rate is insignificant.

7.7.2 Thermal Margin Analysis In the evaluation of the effect of Mark BZ fuel design on thermal marain, B&W performed analyses with regard to the variable pressure-temperature (P-T) limit envelope and the maximum allowable peaking (MAP) limits which are used as bases for the reactor protection system setpoints for the low DNBR trip. The steady state analysis for the Mark BZ fuel assembly was performed with a two-pass method where the closed channel thermal hydraulic code, CHATA, was used for the core-wide analysis, and the subchannel TEMP code was used for the hot assembly/hot channel analysis. The two-pass method and the two thermal hydraulic codes have been approved for licensing analysis and have been used extensively in many B&W reactors. In contrast to the B&W-2 correlation used for the Mark B fuel with a DNBR limit of 1.3,

-10 critical heat flux was calculated using the BWC correlation which has been approved (Ref. 22) for the Mark BZ fuel with a DNBR limit of 1.18. The results of analyses show that the Mark BZ fuel has less restrictive P-T limits than Mark B fuel and that the MAP limits for Mark BZ and Mark B have only a small difference.

An analysis was also performed to determine the effect of Mark BZ fuel on thermal margin for reactor transients. Since the partial loss of coolant flow transient is the limiting anticipated operational occurrence event, this event was analyzed for Mark BZ fuel.

The analysis was performed by evaluating the flux/flow setpoint which is designed for the DNBR protection for partial. pump flow operation. The flux/flow limit is determined from a thermal hydraulic analysis for the pump coastdown transient using the approved RADAR code to ensure that the hot channel DNBR will not exceed the minimum DNBR limit. A comparison was made of the flux/flow limits for a full core Mark BZ and a full core Mark B. The results shown in Figure 5-2 of topical report BAW-1781P show almost identical flux/flow limits for both Mark B and Mark BZ fuel.

Since the flux/flow trip setpoint is to ensure that the minimum DNBR limit will not be violated during partial loss of flow transient if a reactor trip is initiated as soon as the ratio of re actor power to RC flow reaches the flux/flow limit, and since the flux/

flow limits for Mark BZ fuel are very close to those for the Mark B fuel, we conclude that a full core of Mark BZ fuel has no significant impact on thermal margin.

7.7.3 Transitional Mixed Core Incompatibility in the hydraulic characteristics has an additional effect on thermal margin during transitional mixed core cycles when both Mark BZ and Mark B fuel assemblies co-exist in the core. Since Mark BZ fuel has higher hydraulic resistance, the presence of Mark BZ fuel tends to force more flow into the Mark B fuel.

Therefore, if a Mark B fuel assembly is the limiting assembly, the hot channel will receive more coolant and yield better DNB performance comoared to a whole core of Mark B assemblies. As a result, the existing safety analysis for Mark B fuel is bounding and applicable to a transitional mixed core. For the cases where a Mark BZ assembly is limiting, a transitional mixed core will result in worse DNB performance than a whole core of Mark BZ assemblies.

In response (Ref. 21) to an NRC staff question, SMUD provided a de scription on how the hydraulic incompatibility between the Mark BZ and Mark B fuel is accounted for. The limiting assembly is assumed to be a Mark BZ fuel assembly. In the core-wide analysis, the core is modeled with 177 parallel channels and the number of the Mark BZ assemblies is conservatively chosen to be less than or equal to the actual number of Mark BZ assemblies in the cycle being analyzed. This approach is con servative because more flow would be diverted to the Mark B assemblies having less hydraulic resistance and, therefore, ensures that the lowest flow rate is used in the highest powered fuel assembly. This conservative assembly flow rate is then input into the subchannel TEMP code for the hot assembly/hot channel DNBR calculation. For transient analysis using the RADAR code, the hot channel flow rate and DNBR are benchmarked against the steady state TEMP results. Therefore, the approach is also conser-

-11 vative. The fact that the Mark BZ assembly has less flow in a mixed core will result in lower maximum allowable peaking and a lower enthalpy rise factor F in order to maintain the same DNBR limit compared to a whole core of Mark BZ fuel.

As indicated in the report, the amount of peaking reduction necessary during the transitional mixed core will be identified in the cycle specific section of a licensee's reload report, and the F H reduction will be determined by an analysis of the flux/flow setpoint. We, therefore, conclude that the transitional mixed core has been adequately addressed.

7.7.4 ECCS Analysis The ECCS analysis has shown that the increase in the core pressure drop due to the higher hydraulic resistance of the Mark BZ fuel design has no adverse effect on the core thermal hydraulic conditions and thus the LOCA limits. However, since the BWC correlation is used for the Mark BZ fuel CHF calculation, the analysis shows that use of RWC results in the pre diction of earlier inception of DNB at higher (above 6 feet) core elevation relative to the generic 177 FA lowered loop ECCS evaluation using the B&W 2 correlation. This earlier inception of DNB results in a reduction of less than 1 kw/ft in the LOCA limit at the 6-foot core elevation, but does not effect the plant operational limits since the LOCA limits at lower elevations are not affected, and previous analysis has shown that the LOCA limits at the 2-and 4-foot elevation are the controlling parameters.

The use of Zircaloy spacer grid for the Mark B7 also increases the amount of Zircaloy material for metal-water reaction. However, by conservatively assuming the Zircaloy grid temperature equal to that of the hottest point of the cladding, the maximum local oxidation of the Zircaloy grid is 6.2%, the same as the Zircaloy cladding at the 10-foot elevation. This local oxidation is below the limit of 17% specified in 10 CFR 50.46.

In response (Ref. 4) to an NRC staff question, B&W also performed an ECCS analysis for the transitional mixed core at the 2-and 6-foot elevations to determine the effects on peak cladding temperature and LOCA limit. In the analysis, Mark BZ fuel assemblies were assumed in the hot channel and surrounding bundle locations and the Mark B fuel in all remaining locations.

The results of analyses are compared to the results of a full core of Mark BZ fuel and show negligible impact on the peak cladding temperature and LOCA limit.

7.7.5 Rod Bowing Althouah the licensee used a Mark B fuel rod bowing correlation from the approved report BAW-10147, there is only one datum point of Mark BZ fuel presented in the data base of Mark B fuel.

We questioned whether one datum point was sufficient to represent Mark BZ fuel bowing magnitude.

In response to our question, SMUD stated that the change from the Inconel to Zircaloy grids may actually improve the rod bowing performance because Zircaloy grids reduce axial compression on Mark BZ fuel rods.

In addition, two more data points for the Mark RZ fuel bundle were added

-12 to the overall analysis. These two points are also below the predicted value of rod bow; thus the rod bow correlation appears to conservatively predict Mark BZ fuel behavior using the current rod bowing correlation of Mark B fuel.

This is consistent with observations of other PWR fuel assembly designs with Zircaloy grids and fuel rods. We therefore con clude that the Mark BZ fuel rod bowing analysis is adequate and the effect of rod bow on DNBR as discussed in BAW-10147 is applicable to the Mark BZ fuel.

7.8 Summary and Conclusion We have reviewed the topical report B&W-1781P and conclude that the Mark BZ fuel is acceptable for reload application and the report is referenc able for all lowered-loop B&W designed 177 fuel assembly plants. A licensee referencing the Mark BZ fuel design is required to submit-a plant specific analysis of combined seismic and LOCA loads according to Appendix A to SRP 4.2. The thermal margin reduction, i.e. the reduction of the maximum allowable peaking and F, during transitional mixed core having both the Mark B and Mark BZ fuel'assemblies must be addressed in the re load licensing reports for the reload cycles having a mixed core. The NRC staff has reviewed Duke Power Company's submittals (Refs 19 and 6) and the topical report, "Rancho Seco Cycle 7 Reload Report, Volume 1, Mark BZ Assembly Design Report" (Ref. 20) and has concluded that the Mark BZ fuel design is acceptable for the Oconee Unit 1 Cycle 9 and for all lowered-loop B&W designed 177 fuel assembly plants.

EXIGENT CIRCUMSTANCES The following reasons describe the exiqent circumstances:

1. In Ref. 26, the licensee states that originally, the refueling outage was scheduled to begin on October 12, 1984. However, the October 8th date was referred to within the September 11th letter since this was the date that the nominal cycle length would fall based on a 100% capacity factor. The decision to begin the Oconee 1 refueling outage on October 5th was based on the following considerations:

a. In order to assure the maximum electrical output from the Oconee Nuclear Station during the colder winter months, an earlier shutdown date was desirable.

b. The planned outage duration for Oconee I was 49 days. The McGuire Nuclear Station Unit 1 was scheduled to begin a maintenance outage on November 24, 1984. In order to avoid having these two outages overlap, an October 5th shutdown date for Oconee 1 was chosen.

The Commission has determined that exigent circumstances exist in that swift action is necessary to avoid a delay in startup not related to safety and finds that for the reasons stated above, these circumstances caused the outage to start earlier than scheduled.

-13 2.' NRC regulation 10 CFR 50.91 describes the procedures that will be followed on applications received after May 6, 1983, requesting license amendment. These procedures require that, in addition to other requirements, a 30-day comment period will be provided to allow for public comment on the Commission's proposed no significant hazards consideration determination.

The notice of such determination related to these amendments was published in the Federal Register on October 24, 1984, and, therefore, the 30-day comment period should have expired on November 23, 1984. The expiration date, however, given in the Federal Register was November 26, 1984 at 49 FR 42814.

In connection with requests indicating an exigency, the Commission expects its licensees to apply for license amendments in a timely fashion. However, with this consideration in mind, it has been determined that an extraordinary circumstance has arisen where the licensee and the Commission must act quickly, and the licensee has made a good effort to make a timely application.

FINAL NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATTON The Commission's regulations in 10 CFR 50.9? state that the Commission may make a final determination that a license amendment involves no significant hazards considerations if operation of the facility in accordance with the amendment would not:

(1)

Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2)

Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3)

Involve a significant reduction in a margin of safety.

The information in this SE provides the basis for evaluating these license amendments against these criteria. The request for amendment changes the TSs to reflect new operating limits based on the fuel to be inserted into the core. These parameters are based on the new physics of the core and fall within the acceptance criteria. Since the requested change does not affect the original design basis, plant operating conditions, the physical status of the plants, and dose consequences of potential accidents, we conclude that:

(1)

Operation of the facilities in accordance with the amendments would not significantly increase the probability or consequences of an accident previously evaluated.

(2)

Operation of the facilities in accordance with the amendments would not create the possibility of a new or different kind of accident from any accident previously evaluated.

(3)

Operation of the facilities in accordance with the amendments would not involve a significant reduction in a margin of safety.

-14 Accordingly, we conclude that the amendments to Facility Operating Licenses DPR-38, DPR-47, and DPR-55 to support operation of Oconee Unit 1 at full rated power during the upcoming Cycle 9, involve no significant hazards considerations.

STATE CONSULTATION In accordance with the Commission's regulations, consultation was held with the State of South Carolina by telephone. The State expressed no concern either from the standpoint of safety or of our no significant hazards consideration determination.

ENVIRONMENTAL CONSIDERATION These amendments involve a change in the installation or use of a facility.

component located within the restricted area as defined in 10 CFR Part 20.

We have determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has made a final no significant hazards consideration finding with respect to these amendments.

Accordingly, these amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of these amendments.

CONCLUSION We have concluded, based on 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 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: November 23, 1984 Principal Contributors: L. Kopp, Y. Hsii, S. Wu

REFERENCES 1.' Letter from H. B. Tucker (DPC) to H. R. Denton (NRC), "Oconee Nuclear Station, Unit 1", September 11, 1984.

2. "Oconee Unit 1, Cycle 9 Reload Report", BAW-1841, Babcock & Wilcox, August 1984.

3. "Rancho Seco Cycle 7 Reload Report, Volume 1, Mark BZ Fuel Assembly Design Report", BAW-1781P, Babcock & Wilcox, April 1983.

4. Memorandum from L. S. Rubenstein to G. C. Lainas, "Safety Evaluation of Mark BZ Fuel Assembly Design", USNRC, November 2, 1984.

5. Letter from J. F. Stolz (NRC) to H. B. Tucker (DPC), "Safety Evaluation Report on Asymmetric LOCA Loads - Oconee Unit 1, 2, and 3", May 20, 1983.

6. Letter from H. B. Tucker (DPC) to J. Stolz (NRC), "Mark BZ Fuel Assembly Supplemental Information", October 22, 1984.

7. "Gadolinia - Bearing Lead Test Assemblies Design Report", BAW-1772-P, Babcock & Wilcox, June 1983.

8. A. F. J. Eckert, H. W. Wilson, K. E. Yoon, "Program to Determine In Reactor Performance of B&W Fuels: Cladding Creep Collapse", BAW-10084P-A, Revision 2, Babcock & Wilcox, October 1978.

9. Y. H. Hsii, et al., "TACO2 - Fuel Pin Performance Analysis", BAW-10084A, Revision 1, Babcock & Wilcox, June 1983.

10. Letter from J. H. Taylor (B&W) to J. S. Berggren (NRC), "B&W's Responses to TACO 2 Questions", April 8, 1982.

11. "Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water", BAW-10000, Babcock & Wilcox, March 1970.

12. "Correlation of 15x15 Geometry Zircaloy Grid Rod Bundle CHF Data with the BWC Correlation", BAW-10143P, Part 2, Babcock & Wilcox, August 1981.

13. "Fuel Rod Bowing in Babcock & Wilcox Fuel Designs", BAW-10147P-A, Rev. 1, Babcock & Wilcox, May 1983.

14. Letter form C. 0. Thomas (NPC) to J. H. Taylor (B&W), "Acceptance for the Referencing of Licensing Topical Report BAW-10147 (P)", February 15, 1983.

15. R. C. Jones, J. B. Biller, and B. M. Dunn, "ECCS Analysis of-B&W's 177-FA Lowered-Loop NSS," BAW-10103, Rev. 1, Babcock & Wilcox, September 1975.

16. "Bounding Analytical Assessment of NUREG-0630 on LOCA and Operating kW/ft Limits", 77-1141256-00, Babcock & Wilcox, May 1983.

17.

Letter from H. Br-Tucker (DPC) to H. R..Denton (NRC), "Oconee Nuclear Station, Unit 1", October 26, 1984.

18.

Letter from H. B. Tucker (DPC) to H. R. Denton (NRC), Oconee Nuclear Station, Unit 1", November 1, 1984.

19. Letter form H. B. Tucker (DPC) to H. R. Denton (NRC), "Oconee Nuclear Station", October 7, 1983.

20.

Letter from J. J. Mattimore (SMUD) to D. G. Eisenhut (NRC), "Docket 50 312, Rancho Seco Nuclear Generating Station, Unit No. 1, Rancho Seco Cycle 7 Reload Report - Volume 1", October 18, 1983.

21.

Letter from R. J. Rodriguez (SMUD) to J. F. Stolz (NRC), "Docket 50-312, Rancho Seco Nuclear Generating Station, Unit No. 1, Rancho Seco Cycle 7 Reload Report - Volume 1, Request for Additional Information", July 13, 1984.

22. Memorandum from L. S. Rubenstein to F. J. Miraglia, "Staff Evaluation of Babcock & Wilcox Topical Report B&W-10143P, Parts 1 and 2", May 10, 1984.

23.

Letter from R. J. Rodriguez (SMUD) to J. F. Stolz (NRC), "Docket 50-312, Rancho Seco Nuclear Generating Station, Unit No. 1, Rancho Seco Cycle 7 Reload Report - Volume 1, Request for Additional Information", August 28, 1984.

24.

Bailey, W., et.al., "Fuel Performance Annual Report for 1981", NUREG/CR 3001 Battelle Pacific Northwest Laboratory, December 1982.

25. Letter from H. B. Tucker (DPC) to H. R. Denton (NRC), "Oconee Nuclear Station", November 6, 1984.

26. Letter from H. B. Tucker (DPC) to H. R. Denton (NRC), "Oconee Nuclear Station", November 16, 1984.