ML15113A099

Safety Evaluation Supporting Amends 129,129 & 126 to Licenses DPR-38,DPR-47 & DPR-55,respectively

ML15113A099
Person / Time
Site:	Oconee
Issue date:	05/15/1984
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML15113A097	List: ML15113A096 ML15113A099
References
NUDOCS 8405240291
Download: ML15113A099 (9)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.12910 FACILITY OPERATING LICENSE NO.

DPR-38 AMENDMENT NO.129TO FACILITY OPERATING LICENSE NO. DPR-47 AMENDMENT NO.126TO 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

1.0 INTRODUCTION

By letter dated February 13, 1984, (Ref.' 1) 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.

These amendments would authorize proposed changes to the Oconee Nuclear Station (ONS) Technical Specifications (TSs) which are required to support the oper ation of Oconee Unit 3 at full rated power during Cycle 8. The proposed changes include the core protection safety limits (TS section 2.1), the protective system maximum allowable setpoints (TS section 2.3), and the rod position limits (TS section 3.5.2), as well as the administrative renumbering of the figures in.the TS section 3.5.2. To support the application, the licensee submitted report DPC-RD-2003, "Oconee Unit 3, Cycle 8 Reload Report" (Ref. 2) as an attachment to Reference 1.

The Cycle 8 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 8 is to have a length of approximately 400 effective full power days (EFPD) of operation. As has been the case for Cycle 7, Cycle 8 will be operated in a rods-out, feed-and-bleed mode.

Specific aspects of the Oconee Unit 3 Cycle 8 reload are discussed in the following sections.

2.0 EVALUATION OF FUEL SYSTEM DESIGN The analytical methods used in the safety analysis of the proposed eighth cycle of operation at Oconee 3 are described in the Duke Power Company Oconee Nuclear Station Reload Design Methodology Report (Ref. 3) which has been reviewed and approved by the NRC staff (Ref. 4).

Although the methodology report continues to rely on a number of analytical methods developed by the fuel Vendor, Babcock and Wilcox, this is not always the case. Where methods used in the Cycle 8 analysis are unchanged from those described in the Methodology Report, we have concluded that additional review is.unnecessary for Cycle 8 operation. Also, where conditions are identical or limited by the analysis of a previous-cycle of operation, the evaluation of that cycle continues to apply.

P40A5240291084052 PDR ADOCK& 05000269 P

PDR

2.1 Fuel Assembly Design Although all batches in the Oconee 3 Cycle 8 core will utilize the same Babcock and Wilcox 15x15 fuel design, the Batch 9 and 10 assemblies will be of the Mark B5 fuel design. The Mark B5 fuel assembly is identical to the Mark B4 except its upper end fitting has been redesigned to provide a positive holddown of fixed control components such as burnable poison rod assemblies (BPRAs), neutron source rod assemblies, and orifice rod assemblies.

Two regenerative neutron sources will be used in Mark B4 and BS assemblies.

The Mark B4 and B5 assembly designs were reviewed and found acceptable for previous B&W 177 FA reloads and are therefore acceptable for Oconee 3 Cycle 8.

2.2 Fuel Rod Design The Oconee 3 Cycle 8 core contains both,Mark B4 and Mark B5 fuel assemblies and the fuel rods used in both of these assemblies are virtually identical.

While the results of the linear-heat-rate-to-melt analysis (Table 4.2 of Reference 2) indicate some variation in densification characteristics for the different batches, the resulting linear heat rate values are the same for all batches in the Cycle 8 core.

2.2.1 Cladding Collapse The licensee has stated that the cladding collapse time for the most limiting Cycle 8 assembly was conservatively determined to be greater than the maximum projected residence time for any Cycle 8 assembly. The creep collapse analysis used the CROV computer code (Ref. 5) using input conditions from TACO-2 (Ref. 6) in a manner described in the Reload Methodology Report (Ref. 3).

All of these methods have been reviewed and approved by the NRC staff.

We con clude that cladding collapse has been appropriately considered for Cycle 8 operation.

2.2.2 Claddina Strain The licensee has performed a cladding strain calculation using TACO-1 (Ref. 9) in accordance with methods and limits given in the Reload Methodology Report (Ref. 3)..It is concluded that the analysis of cladding strain has.been appropriately considered for Cycle 8 operation.

2.2.3 Rod Internal Pressure Section 4.2 of the Standard Review Plan (Ref. 8) was previously cited as a source of acceptance criteria used to establish the design bases and evaluation of the fuel system. Among those criteria which may affect the operation of the fuel rod is the internal pressure limit. The pressure criterion (SRP 4.2,Section II.S.1(f)) is a requirement that the fuel rod internal gas pressure should remain below normal system pressure during normal operation unless otherwise justified. Based on a TACO-1 analysis, the licensee has stated that fuel rod internal pressure will not exceed system pressure during normal operation for Cycle 8. We find this acceptable and conclude

-3 that the rod internal pressure limits have been adequately considered for Cycle 8 operation.

2.3 Fuel Thermal Design There are no major changes in the physical characteristics of the Cycle 8 core which would result in altered thermal conditions.

As pointed out in Section 2.2 of this report, the linear-heat-rate-to-melt for all batches in the Cycle 8 core is the same. The linear-heat-rate-to-melt capability was determined separately for Batches 8B and 9 using TACO-1 and for Batch 10 using TACO-2. The centerline melt limits are generated at both low and high burnup conditions and the linear heat rate capability is both batch and burnup dependent. All values given in Table 4.2'of the reload report (Ref. 2) are higher (less limiting) than those used in the Oconee 3 Cycle 7 reload. These values have been incorporated into the proposed Technical Specifications and we find them acceptable.

2.3.1 LOCA Initial Conditions A combination of TAFY and TACO-2 analyses were used to generate the LOCA limits as described in Tables 7-2 and 7-3 of Reference 2. Three sets of boundinc values for allowable LOCA peak linear heat rates are given as a function of core height. These limits apply during the periods 0-25 EFPD, 25-65 EFPD and 65 EFPD to end-of-cycle. These limits have been incorporated into the Technical Specifications for Cycle 8 through the operating limits on rod index and axial power imbalance.

It is concluded that the initial thermal conditions for LOCA analysis have been appropriately considered for Cycle 8 operation.

2.3.2 Fuel Rod Bowing The licensee has determined a fuel rod bowing gap closure correlation for use in the calculation of the rod bowing penalty as described in Reference 10.

It is concluded that this correlation adequately accounts for gap closure as a function of burnup in the Mark B fuel design. The rod bowing penalty is discussed in the Thermal Hydraulic Design section cf this report.

2.4 Operating Experience I

Babcock and Wilcox has accumulated operating experience with the Mark B4 15x15 fuel assembly at all of the eight operating B&W 177-fuel assembly plants and Mark B5 experience during Cycle 7 of Oconee 3. A summary of this operating experience is provided as part of our fuel operating experience report (Ref.. 7).

2.4.1 Holddown Spring Failures It has been noted during previous Oconee reload reviews (e.g., Ref. 11) that a small number of holddown spring failures are continuing to cccur at the Oconee station. These springs are contained in the upper end fitting of the Mark B4 fuel assembly and are used to accommodate length changes due to thermal expansion and irradiation growth while providing a positive holddown

-4 force for the assembly. On May 14, 1980, a failed holddown spring was dis covered by remote video inspection at Davis-Besse Unit 1. Further examination ultimately identified a total of 19 failed springs at that plant. Subsequent examination of spent fuel assemblies at other B&W reactors, including the Oconee station, revealed a small number of similar failures.

An inspection (Ref. 12) of all Oconee Unit 3 Cycle 6 assemblies revealed broken holddown springs in two assemblies due to be discharged. Another inspection (Ref. 13) revealed one broken holddown spring in Unit 1 Batch 4 fuel and three broken holddown springs in Oconee Unit 2 Batch 7 fuel.

More recently, four additional broken holddown springs were found in Unit 1 (Ref. 14).

In all cases, the fuel was due to be discharged or the holddown springs were replaced prior to reinsertion. It has been concluded that a continuing program of detection and discharge/replacement of failed holddown springs is necessary to minimize the probability of operating with broken holddown springs for the Mark B fuel design.

2.5 Conclusions We have reviewed those sections of the reload report for Oconee Unit 3 Cycle 8 dealing with the fuel system design and find those portions of the appli cation acceptable.

3.0 EVALUATION OF NUCLEAR DESIGN The nuclear design parameters characterizing the operation of Oconee Unit 3 Cycle 8 have been obtained with the Duke Power physics calculational methods (Ref. 3).

These methods have been approved for use in reload design calcu lations (Ref. 4) and were used previously in deriving the Cycle 7 nuclear design parameters. The Cycle 8 core will contain 68 fresh assemblies with a U-235 initial enrichment of 3.28%. In addition to the 68 fresh assemblies, there are two batches of exposed assemblies: a batch of 37 assemblies havina an initial U-235 enrichment of 3.07% and a batch of 68 assemblies having an initial enrichment of 3.18%. Four fresh assemblies are located in the central core region with the remaining fresh assemblies distributed in a checkerboard pattern in the surrounding annular region. No fresh assemblies are loaded in the outermost peripheral ring. This is characteristic of all current extended burnup PWR reloads. The excess reactivity is controlled by soluble boron which is supplemented by 61 full-length Ag-In-Cd control rods and 60 BPRAs. Furthermore, eight partial length axial power shaping rods (APSRs) are provided for additional control of axial power distribution. All safety criteria are met. Shutdown margin values at beginning and end of cycle are 4.14% and 2.73% Ak/k, respectively, compared to the minimum required value of 1.0 percent. Beginning of cycle radial power distributions show acceptable margins to limits.

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 Cycle 8 is acceptable.

-5.

4.0 EVALUATION OF THERMAL-HYDRAULIC DESIGN The objective of this review is to confirm that the thermal-hydraulic design of the reload core has been accomplished using acceptable.methods and provi-des an acceptable margin of safety from conditions which could lead to fuel damage during normal and anticipated operational transients. The reload design methodology is described in Reference 3 and has been approved (Ref. 4).

Discussion of the main differences between Cycle 8 and Cycle 7 follows.

4.1 Core Bypass Flow The incoming Batch 10 fuel is hydraulically and geometrically similar to the fuel remaining in the core from the previous cycles.

For Cycle 8 operation, 60 BPRAs will be inserted, and two assemblies will contain regenerative neutron sources, leaving 46 open assemblies, resulting in an increase in calculated maximum core bypass flow of 7.9 percent compared with 7.6 percent for Cycle 7. The bypass flow of 7.9 percent is less than the 8.2 percent assumed in the generic thermal-hydraulic design analysis (Ref. 2), and the consequent increase in Cycle 8 core flow relative to the generic analysis value establishes the generic analysis as conservative for Cycle 8 operation.

4.2 DNBR Penalty Due to Rod Bow A B&W topical report (Ref. 10) discussing the mechani8ms and resulting effects of bowing in B&W fuel has been reviewed and approved (Ref. 15).

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.

Therefore, no rod bow penalty needed to be considered for Cycle 8 operation.

We conclude that the available margin for Cycle 8 has been demonstrated and that the thermal-hydraulic design is, therefore, acceptable.

4.3 Conclusions The pertinent thermal-hydraulic parameters summarized in Table 6-1 of the submittal are identical except for the core bypass flow of 7.9 percent of total flow for Cycle 8 as compared to 7.6 percent for Cycle 7 and 8.2 percent for the generic analysis. The decrease of bypass flow, relative to the generic analysis value,.resultina in a net increase in core flow indicates that, with other parameters unchanged, the safety margin for Cycle 8 is comparable to that of the generic analysis. The reload design methodology for Cycle 8 included in Reference 3 has been approved as indicated in Reference 4. We conclude from the examination of the Cycle 8 core thermal-hydraulic design, with respect to the FSAR values, that the core reload will not adversely affect the capability to operate Oconee Unit 3 safely during Cycle 8 and that the proposed changes to Technical S.pecifications discussed in Section 6.0 of the submittal are acceptable.

-6 5.0 ACCIDENT ANALYSIS The important kinetics parameters for Cycle 8 are compared to the values used in the FSAR in Table 7.1 of.the reload submittal (Ref. 2).

For the parameters quoted, the Cycle 8 values are bounded by those used previously. The licensee has also determined that the initial conditions of the transients in Cycle 8 are bounded by the FSAR and/or the fuel densification report (Ref. 16).

Since the Batch 10 reload fuel contains rods with a theoretical density higher than that considered in the densification report, the conclusions in Ref. 16 are still valid. These analyses have been previously accepted by the NRC.

The licensee's Reload Methodology Technical Report (Ref. 3), which has been accepted by the NRC staff (Ref. 4), was examined vis-a-vis the Accident Analysis Review process.

Of the items contained in the "Key Safety Parameter Checklist" (Table 8-1, Ref. 3), virtually all are addressed in Table 7-1 and other tables in the submittal.

It should be noted, however, that the "Minimum Tripped Rod Worth" available in case of a steamline break is not given and therefore cannot be compared to the value assumed in the FSAR analysis.

However, the total available worth and the shutdown margin for Cycle 8 presented in Table 5-2 of Ref. 2 are larger, and hence more conservative, than the corresponding values for Cycle 7. In addition, the values in Table 5-1 for the effective delayed neutron fraction are lower than the nominal values assumed in the FSAR analysis of the rod ejection accident (REA).

While this would tend to increase the maximum fuel enthalpy associated with a postulated REA event, the maximum ejected rod worth at HFP for Cycle 8 is so much lower than that assumed in the FSAR analysis that it offsets this non-conservatism.

Three sets of bounding values for allowable LOCA peak linear heat rates are given as a function of core height. These limits apply during the periods 0 -

1000 MWD/MTU, 1000 -

2600 MWD/MTU, and for the balance of the cycle.

These results are based upon a bounding analytical assessment of NUREG-0630 on LOCA and operating linear heat rate limits performed by Babcock & Wilcox (Ref. 17).

The B&W analyses have been approved by the NRC staff and the three sets of limits were accepted in conjunction with the review of the Oconee Unit 2 Cycle 7 reload submittal (Refs. 18 and 11).

New dose calculations were not performed for Oconee 3 Cycle 8. The licensee has determined that the dose considerations for Oconee 1 Cycle 8 (Ref. 19) are characteristic for Oconee 3 Cycle 8 based on comparisons of key parameters which determine radionuclide inventories. Therefore, it is acceptable.

6.0 TECHNICAL SPECIFICATION MODIFICATIONS Oconee Unit 3 Cycle 8 Technical Specifications have been modified to account for (i) minor changes in power peaking and control rod worths during Cycle 8 operation, (ii) incorporation of NUREG-0630 data (Ref. 17) in the LOCA analysis and (iii) employment of a Monte Carlo simulation technique in determining instrument string errors (Refs. 18 and 11).

We have reviewed the proposed Specification revisions for Cycle 8. 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 of

-7 Specification 3.5.2. The limiting safety system settings and the limitina conditions for operation have been established by approved methods. Changes which reflect the core thermal-hydraulic response continue to maintain the safety limit DNBR criterion of 1.30. The control rod withdrawal limits for the various pump combinations and times in core life are presented as well as part length axial power shaping rod position limits.

On the basis that previously approved methods were used to obtain the limits, we find these Technical Specification modifications acceptable.

Selected Technical Specification changes for Oconee Unit 1 and 2 were also included in the Oconee Unit 3 Cycle 8 submittal'. These changes are admin istrative only, i.e., figure and page numbering changes, and are, therefore, acceptable.

7.0 START-UP TESTING The startup testing program for Oconee Unit 3 Cycle 8 will be carried out in accordance with approved methods and procedures.

8.0 EVALUATION FINDINGS We have reviewed the fuels pnvsics, thermal-hydraulic and tr'sien information presented in the Oconee 3 Cycle 8 reload report. We find the proposed reload and the associated modified Technical Specifications acceptable.

9.0 ENVIRONMENTAL CONSIDERATION

We have determined that the amendments do not authorize a chance 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 the amendments involve an action which is insignificant from the standpoint of environmental impact, and pursuant to 10 CFR §51.5(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.

10.0 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:

May 15, 1984 Principal Contributors.: M. Dunenfeld, M. Todosow, J. Carew, D. Cokinos, P. Neogy

-8

11.0 REFERENCES

1. Letter from H. B. Tucker (DPC) to H. R. Denton (NRC) "Oconee Nuclear Station, Units 1, 2, and 3", February 13, 1984.

2.

"Oconee Unit 3 Cycle 8 Reload Report," DPC-RD-2003, Duke.Power -Company, February 1984.

3.

"Oconee Nuclear Station Reload Design Methodology," Technical Report, NFS-1001, Revision 4, Duke Power Company, Charlotte, North Carolina, April 1979.

4. Letter from P. C. Wagner (NRC) to W. B. Parker, Jr. (DPC), "Safety Evaluation by the Office of Nuclear Reactor Regulation of the Reload Design Methodology Technical Report NFS-1001," July 29, 1981.

5. A. F. J. Eckert, H. W. Wilson, K. E. Yoon, "Program to Determine In-Reactor Performance of B&W Fuels:

Cladding Creep Collapse,"

Babcock and Wilcox Company Report BAW-10084P-A, Revision 2, October 1978.

6.

"TACO2 -Fuel Pin Performance Analysis", BAW-10141P-A, Rev. 1, Babcock & Wilcox, Lynchburg, Virginia, June 1983.

7. W. J. Bailey and M. Tokar, Fuel Performance Annual Report for 1982, NUREG/CR-3608 (PNL-4817), March 1984.

8. U. S. Nuclear Regulatory Commission Standard Review Plan, Section 4.2, "Fuel System Design," U. S. Nuclear Regulatory Commission Report NUREG-0800 (Formerly NUREG-75/087), Revision 2, July 1981.

9.

"TACO-Fuel Pin Performance Analysis," Babcock and Wilcox Company Report BAW-10087P-A, Revision 2, August 1977.

10.

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

11.

L. S. Rubenstein (NRC) memorandum for G. Lainas, "SER-Oconee Unit 2 Reload For Cycle 7 (TACS 52311 and 52447)," November 17, 1983.

12. W. 0. Parker (Duke) letter to J. P. O'Reilly (NRC) dated July 23, 1982.

13. W. 0. Parker (Duke) letter to J. P. O'Reilly (NRC) dated February 16, 1982.

14.

H. B. Tucker (Duke) letter to J. P. O'Reilly (NRC) dated July 21, 1983.

15.

Letter from Cecil 0. Thomas (NRC) to James H. Taylor (B&W), February 15, 1983, "Acceptance for the Referencing of Licensing Topical Report BAW 10147(P)."

-9

16.

"Oconee 3 Fuel Densification Report," BAW-1399, Babcock & Wilcox, November 1983.

17.

"Bounding Analytical Assessment of NUREG-0630 on LOCA and Operating kw/ft Limits," B&W Document No. 77-1141256-00, Babcock & Wilcox.

18.

"Oconee Unit 2, Cycle 7 Reload Report," DPC-RD-2002, Duke Power Company, September 1983.

19.

"Oconee Unit 1, Cycle 8 Reload Report," BAW-1774, February, 1983.