ML16138A586

Safety Evaluation Supporting Amends 155,155 & 152 to Licenses DPR-38,DPR-47 & DPR-55,respectively

ML16138A586
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Site:	Oconee
Issue date:	03/19/1987
From:	Office of Nuclear Reactor Regulation
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NUDOCS 8703300194
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qUNITED STATES 2

NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 155 TO FACILITY OPERATING LICENSE NO. DPR-38 AMENDMENT NO.155 TO FACILITY OPERATING LICENSE NO. DPR-47 AMENDMENT NO. 152 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 December 12, 1986 (Ref. 1), as revised on January 29, 1987 (Ref. 2) and supplemented on February 11, 1987 (Ref. 6), 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 3 is currently completing a refueling outage.

These amendments would authorize changes to the Oconee Nuclear Station TSs which are required to support the operation of Oconee Unit 3 at full rated power during the upcoming Cycle 10.

The amendments would change Figure 3.5.2-12, the Unit 3 Operational Power Imbalance Envelope curve. The Figure would be updated to reflect current cycle operating characteristics.

These amendments would also provide a more conservative curve for Oconee Unit 1 Operational Power Imbalance Envelope (Figure 3.5.2-10) to allow 10 CFR Part 50.59 reviews of future core reloads; update TS 3.5.2.4.b.2 (quadrant power tilt) to reflect the fact that power level cutoffs (other than 100%)

are no longer applicable to Oconee; delete TS 3.5.2.6 (xenon reactivity) because operating restrictions resulting from transient xenon power peaking are implicitly included in the limits of TS 3.5.2.5 (control rod positions) and proposed TS 3.5.2.6 (reactor power imbalance) and note this in the bases of TS 3.5; and change TSs 3.5.2.7, 3.5.2.8 and 3.5.2.9 to reflect the deletion of TS 3.5.2.6 (xenon reactivity).

For Oconee Unit 3 only, these amendments would raise the minimum boron concentration in the borated water storage tank (BWST) from 1835 parts per million (ppm) to 2010 ppm to ensure that the core is at one percent delta k over k, 1% A k/k or shutdown margin, at 700 F without any control rods in the core. Other administrative type changes requested in the February 11, 1987 application are being handled separately.

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-2 To support the license amendment application, the licensee submitted "Oconee Unit 3, Cycle 10 Reload Report" as an attachment to the December 12, 1986 application. A summary of the Cycle 10 operating parameters is included in the report, along with safety analyses. On January 29, 1987, the licensee revised the reload report because Oconee Unit 3 was shutdown on December 17, 1986--earlier than scheduled because of possible wear indications in the 3B2 reactor coolant pump. The Oconee Unit 3 Cycle 10 core was then redesigned based on the shortened Cycle 9 length of 349 effective full power days. Results of this redesign indicated that to ensure the core will be shutdown in conformance with applicable criteria, the beginning of cycle, all rods out, 700F 1% A k/k shutdown boron concentration should be increased from the present 1835 to 2010 ppm. In a letter dated February 11, 1987, the licensee proposed revisions to the TSs to raise the minimum boron concentration in the BWST.

The Cycle 10 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. The fuel consists of dished-end, cylindrical pellets of uranium dioxide clad in cold-worked Zircaloy-4. The fuel assemblies in all batches have an average nominal fuel loading of 463.6 kilograms (kg) uranium. The undensified nominal active fuel lengths, theoretical densities, fuel and fuel rod dimensions, and other related fuel parameters are given in Table 4-1 (Ref. 3).

The Cycle 10 core loading diagram, enrichments, control rods and burnable poison rod assembly (BPRA) locations and enrichments are also given in Reference 3.

Cycle 10 will operate in a rods-out, boron feed-and-bleed mode. Core reactivity control is supplied mainly by soluble boron and supplemented by 61 full-length Ag-In-Cd control rods and 60 BPRAs. In addition to the full-length control rods, eight Inconel gray axial power shaping rods (APSRs) are provided for additional control of axial power distribution. Since gray APSRs are being utilized, there are eight control rods in group seven and twelve in group five to reduce the negative offset response to the group seven rod movement.

The present reload involves no significant changes in core fuel design or methodology. Proposed revisions to the TSs required for Cycle 10 operation were made in accordance with methods and procedures found acceptable in connection with previous reloads (Ref. 4) and are the result of minor cycle-to-cycle fuel changes.

EVALUATION Evaluation of Fuel System Design The types of fuel assemblies and pertinent fuel design parameters for Oconee 3 Cycle 10 are listed in Table 4-1 (Ref. 3).

All fuel assemblies are mechanically interchangeable. Two regenerative neutron sources will be used in the Mark BZ fuel assemblies. The Cycle 10 core contains only fuel designs which have been previously loaded in the Oconee Unit 3 reactor and have been previously approved by the NRC staff. The fuel rod design, cladding collapse, cladding stress and strain, and the thermal design fuel analyses for Cycle 10 fuel designs, including the gray APSRs, are either bounded by conditions previously analyzed for Oconee 3 or were analyzed specifically for Cycle 10 using methods and limits previously reviewed and approved by the NRC staff.

Therefore, we conclude that the overall fuel system design for Oconee 3 Cycle 10 is acceptable.

-3 Nuclear Design Table 5-1 (Ref. 3) compares the core physics parameters of Cycle 10 with those of the reference Cycle 9. The values for Cycle 9 and Cycle 10 were generated by Duke Power Company using the reload design methods described in Reference 5 which have been reviewed and approved by the NRC staff.

We have determined that approved methods have been used, and the nuclear design parameters meet the acceptance criteria of Standard Review Plan, Section 4.3, Part II, and, therefore, conclude that the nuclear design of Oconee 3 Cycle 10 is acceptable.

Evaluation of Thermal-Hydraulic Design The generic Mark B and Mark BZ thermal-hydraulic design analyses supporting Cycle 10 operation were performed by Duke Power Company using the methods described in Reference 5. The Cycle 9 and Cycle 10 thermal-hydraulic design conditions are summarized in Table 6-1 (Ref. 3).

The Cycle 10 core will include 60 fresh Mark BZ Batch 12 fuel assemblies, all of which will contain-BPRAs. This results in a core bypass flow of 7.9% of the total system flow, which is less than the bypass flow assumed in the generic thermal-hydraulic analyses.

The Mark BZ fuel assembly has a slightly higher pressure drop than the Mark B assembly as a result of the increased flow resistance of the Zircaloy spacer grids. The presence of Mark BZ and Mark B assemblies in a core results in less coolant flow in the Mark BZ fuel than would occur in an all Mark BZ core. The generic Mark BZ analyses conservatively account for this transition core effect.

In a Mark BZ transition core, the limiting Mark B hot channel will receive more coolant and yield better departure from nucleate boiling (DNB) performance than would be predicted for a full Mark B core. Thus, the generic Mark B analyses, based on the B&W-2 critical heat flux (CHF) correlation, are bounding and are applicable to the Cycle 10 transition core.

We have determined that approved methods have been used, and the thermal-hydraulic design parameters meet the departure from nucleate boiling ratio (DNBR) safety limit using approved CHF correlations and, therefore, conclude that the thermal hydraulic design of Oconee 3 Cycle 10 is acceptable.

Safety Analyses The important kinetics parameters for Cycle 10 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 10 values are bounded by those previously used. The licensee has also determined that the initial conditions of the transients in Cycle 10 are bounded by either the FSAR, the fuel densification report, previous reload analyses, or analyses using approved methods.

-4 Babcock & Wilcox (B&W) has performed a generic loss of coolant accident (LOCA) analysis for the B&W 177-FA, lowered-loop nuclear steam supply system using the final acceptance criteria Emergency Core Cooling System evaluation model.

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. These results are based upon a bounding analytical assessment of NUREG-0630 on LOCA and operating LHR limits performed by B&W. The B&W analyses have been approved by the NRC staff and the LHR limits are satisfactorily incorporated into the TSs for Cycle 10 through the operating limits on rod index and axial power imbalance and, therefore, are acceptable.

Technical Specification Modifications Oconee Unit 3 Cycle 10 TSs have been modified to account for normal cycle-to-cycle fuel changes in power peaking and control rod worths. We have reviewed the proposed specification revisions for Cycle 10. These changes concern the Operational Power Imbalance Envelope (Figure 3.5.2-12). In addition, the licensee has provided a more conservative curve for the Unit 1 Operational Power Imbalance Envelope (Figure 3.5.2-10) in order to reduce future TS changes and to allow more of their future reload cores to be reviewed under 10 CFR 50.59. On the basis that approved methodology was used to obtain these limits which assure that general design criteria 10 and 12 are satisfied, we find these TS modifications acceptable.

The licensee also proposed administrative changes to TSs 3.5.2.4.b.2, 3.5.2.6, 3.5.2.7, 3.5.2.8 and 3.5.2.9 which are common to all three Oconee units.

These changes reflect the fact that power level cutoff is no longer applicable to Oconee and operating restrictions resulting from transient xenon power peaking are implicitly included in the control rod.position and reactor power imbalance limits.

Therefore, we find these changes acceptable.

Increase in Boron Concentration in the Borated Water Storage Tank As a result of a shortened Cycle 9 of Oconee Unit 3 the design of the Cycle 10 core will require an increase in the BWST boron concentration to ensure the core will be shutdown in conformance with TS 3.8.4 and TS 3.3 criteria. By letter dated February 11, 1987 (Ref. 3), as supplemented on February 27, 1987 (Ref. 6), Duke Power Company presented the results of its analysis which indicates that the beginning of cycle, all rods out, 700 F, 1 percent delta k over k shutdown boron concentration should be increased from the present 1835 ppm to 1873 ppm in order to meet the 1 percent subcritical acceptance criteria. Duke has requested TS changes which will conservatively increase the minimum concentration in the Oconee Unit 3 BWST to 2010 ppm for Cycle 10.

We have determined that approved methods have been used to insure that the 1 percent subcritical acceptance criteria are conservatively met, and that the plant will remain bounded by the FSAR safety analyses. Therefore, we conclude that the increase in the BWST boron concentration to 2010 ppm for Oconee Unit 3 Cycle 10 is acceptable.

-5 EMERGENCY CIPCUMSTANCES In its February 11, 1987 letter, the licensee reouested that these amendments be treated as an emergency because insufficient time exists for the Commission's usual 30-day notice without extendina the current outane.

Because of the early shutdown of Oconee Unit 3, the licensee determined that emergency circumstances exist for approval of these proposed revisions to support startup of Oconee Unit 3, Cycle 10.

The licensee revised the reload report because Oconee Unit 3 was shutdown on December 17, 1986 - earlier than scheduled because of possible wear indications on a reactor coolant pump. The Oconee Unit 3 Cycle 10 core was then redesigned based on the shortened cycle. Results of this redesign indicated that to ensure the core shutdown margin, the boron concentration in the BWST would need to be increased from the present 1835 to 2010 ppm. In its February 11, 1987 letter, the licensee proposed revisions to the TSs to raise the minimum boron concentration in the PWST.

The Commission has determined that emergency circumstances exist in that swift action is necessary to avoid a delay in startup not related to safety and finds that, for the reason stated above, emergency circumstances exist.

In connection with a request indicating an emergency, 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 a 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 HA7ARPS CONSIDERATION DETERMINATION The Commission's reaulations in 10 CFR 50.92 state that the Commission may make a final determination that a license amendment involves no sionificant 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.

We have confirmed the basis of the no significant hazards consideration findings described in the notice published in the FEDERAL PFGISTER on March 6, 1987 (52 FR 7050).

The amendments change the TSs to reflect new operating limits based on the fresh fuel to be inserted into the core. These parameters are based on the new physics of the core and fall within the acceptance criteria.

There are no significant changes in the fuel being used, or the fuel assembly design. We have previously reviewed postulated fuel-related transients and accidents. As part of these analyses, bounding parameters were used, for example, power peaking limits and reactor system pressure. Accident analyses previously submitted by the licensee and approved by the NRC staff for Oconee 3

utilized input values of physics parameters which are designed to be boundinq for various operating cycles and operating conditions. The power imbalance limit curve for Cycle 10 was derived by the licensee so that the previous analyses for the postulated accidents would remain valid for Cycle 10.

Therefore, it was unnecessary to analyze any accident for Cycle 10 of Oconee 3.

Since the postulated accidents previously analyzed remain applicable to the new core (i.e., continue to be boundino', the probability or consequences of an accident previously evaluated have not increased. Recause of the fundamental identity of the new fuel in terms of its nuclear and fuel assembly desinnr the possibility of a new or different kind of accident *rom any accident oreviously evaluated has not been created.

Finally, the power imbalance curve ensures that the licensed margin of safety has not been reduced.

To ensure that the core shutdown margin is 1 percent delta k over k, at 700%

without any control rods in the core, the minimum boron concentration in the PWST will have to be raised from 1835 ppm to 2010 ppm. We have confirmed that approved methods have been used to ensure that the 1 percent subcritical acceptance criteria are conservatively met, and that the plant will remain bounded by the FSAP safety analyses. Therefore, the probability of any Design Basis Accident (DBA) is not affected by this change, nor are the consequences of a PRA affected by this change. The key physics parameter affected by the Oconee Unit 3 Cycle 10 redesign is the BOC boron concentration.

The limiting FSAR transient with respect to changes in the boron concentration is the moderator dilution transient at power. Only the non-LOCA boron dilution transient was found to have a more potentially severe result due to increased boron concentration. This event is bounded by the values assumed in the FSAR.

Therefore, the moderator dilution transient presented in the FSAR remains conservative for Oconee Unit 3, Cycle 10.

Analysis of the increase in the Oconee Unit 3, Cycle 10, minimum BWST boron concentration has indicated that the 2010 ppm concentration is well within all acceptance criteria.

For refueling and LOCA conditions, the proposed concentration is sufficient to maintain the core 1 percent subcritical at 70'F with all control rods removed; this chance affects only previously evaluated accidents, discussed above, and does not create the possibility of a new or different kind of accident from any accident previously evaluated. The predicted boron concentration reouired to maintain the core 1 percent subcritical at 700F with all rods out of the core during refueling or a LOCA has been compared to the current TS value for the BWST. The predicted 8OC, all rods out, 70'F, 1 percent subcritical boron concentration of 1873 ppm has necessitated a change in the required boron concentration for the BWST from 1835 ppm. To provide additional shutdown margin during refueling or a LOCA, a more conservative BWST boron concentration of 2010 ppm will be used.

For the non-LOCA events, the moderator dilution transient has been shown to be bounded by the FSAR analysis and involves no significant reduction in a margin of safety.

Therefore, 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.

-7 (2)

Operation of the facilities in accordance with the amendments would not create the possibility of a new or different kind of accident

.fron 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.

Accordingly, we conclude that the amendments to Facility Operating Licensps Nos. DPR-38, DPP-47 and DPR-55 to support operation of Oconee Unit 3 at full rated power during the upcoming Cycle 10, 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 significart hazards consideration determination.

ENVIRONMENTAL CONSIDERATTON 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 previously issued a proposed finding that these amendments involve no sianificant hazards consideration and there has been no public comment on such findinq.

Accordingly, these amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFP 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 requlations 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:

March 19, 1987 Principal Contributor:

G. Schwenk

REFERENCES

1. Letter, H. B. Tucker (Duke) to H. R. Denton (NPC), "Oconee Nuclear Station Unit 3," December 1?, 1986.

2. Letter, H. B. Tucker (Duke) to H. P. Denton (NRC), "Oconee Nuclear Station," January 29, 1987.

3. Report, "Oconee Unit 3, Cycle if Reload Report," DPC-PP-2008, Duke Power Company, January 1987.

4.

Letter, H. Nicolaras (NRC) to H. B. Tucker (Duke), September 19, 1985.

5. Report, "Oconee Nuclear Station Reload Design Methodology II,"

DPC-NF-1002, Duke Power Company, Charlotte, North Carolina, March 1985.

6. Letter, H. B. Tucker (Duke) to H. R. Denton (NRC) "Oconee Nuclear Station", February 27, 1987.