ML20035C621
| ML20035C621 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 03/30/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20035C620 | List: |
| References | |
| NUDOCS 9304080197 | |
| Download: ML20035C621 (9) | |
Text
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 61 TO FACILITY OPERATING LICENSE NO. NPF-42 WOLF CREEK NUCLEAR OPERATING CORPORATION DOCKET NO. 50-482 WOLF CREEK GENERATING STATION
1.0 INTRODUCTION
By application dated October 28, 1992, and supplemented by letters dated January 28, 1993 and March 8,1993, Wolf Creek Nuclear Operating Corporation (the licensee) requested changes to the Technical Specifications (Appendix A to Facility Operating License No. NPF-42) for the Wolf Creek Generating Station. The proposed changes reflect the use of VANTAGE SH fuel with intermediate flow mixers, the licensee's performance of nuclear, thermal-hydraulic and safety analyses using a revised methodology and incorporating some changes to key assumptions, and the relocation of cycle specific i
parameters to a Core Operating Limits Report (COLR).
The most significant changes include increases in allowable core peaking based on revised thermal-hydraulic and accident analyses; introduction of an allowable positive moderator temperature coefficient; a decrease in the reactor coolant thermal design flow; an increase in the main steam safety valve setpoint tolerance; an increase in required shutdown margin in Mode 5, cold shutdown, to address boron dilution concerns; and the adoption of the Core Operating Limits Report.
The supporting analyses were performed assuming an increase in rated thermal power and a range of reactor coolant system temperatures. The technical specifications and related plant changes related to these parameters are not included in the proposed changes addressed by this Safety Evaluation.
l The January 28 and March 8,1993 submittals provided clarifying information and did not change the initial no significant hazards considerations determination published in the Federal Reaister on January 6, 1993.
2.0 EVALUATION The fuel assembly design being introduced for Cycle 7 operation is a standard Westinghouse design which has been utilized at other facilities and has been reviewed and approved by the staff (Ref.1). The analyses supporting the proposed changes to the technical specifications were performed using the NRC approved methodologies which are listed U the proposed revision to Technical Specification 6.9.1.9, which would add the requirement to submit a Core i
Operating Limits Report.
9304000197 930330 PDR ADOCK 05000482 P
k An exception to the use of approved methodologies is the Wolf Creek Nuclear Operating Corporation's transient analysis methodology topical report. The staff's review of the transient analysis topical report is nearing completion and no significant issues have been identified. The staff's review of the transient analysis topical report is sufficiently complete to conclude that the analysis performed in support of Cycle 7 operation was performed using an acceptable methodology. Technical Specification 6.9.1.9 has been modified from the licensee's amendment request to reflect the pending approval of the licensee's " Transient Analysis Methodology for the Wolf Creek Generating Station." This change was discussed with the licensee and was determined to be acceptable.
The only significant change from the specifics of an NRC reviewed methodology is the core thermal-hydraulics performed for the Cycle 7 reload design. The evaluation of the Cycle 7 core thermal-hydraulics model is provided below and this Safety Evaluation has been added to the list of approved methodologies provided by Technical Specification 6.9.1.9.
The introduction of the intermediate flow mixing grids (IFMs) associated with the new fuel assemblies being inserted for Cycle 7 required the licensee to repeat the process described in Reference 2 including the modelling of the IFMs and use of the associated critical heat flux correlation (WRB-2). The methodology outlined in the approved topical report was followed but the specific DNBR limits changed as a result of the revised fuel design and use of the WRB-2 correlaticn. Upon the determination of the DNBR limit, performance of the statistical core design and determination of operating limits was performed in accordance with the approved methodology.
The licensee performed a qualification of the NRC approved WRB-2 correlation (Ref. 3) using the VIPRE-01 thermal-hydraulic analysis code. This qualification was performed in the same manner as was used for the WRB-1 correlation in Reference 2.
The licensee's analyses resulted in the determination of a DNBR correlation limit (to provide 95% probability at 95%
confidence) of 1.14 for the WRB-2 correlation using the VIPRE-01 computer code.
In accordance with the staff's Safety Evaluation for the VIPRE-01 code (Ref. 4), the vendor's correlation limit of 1.17 [WRB-2 and the THINC code, (Ref. 3)] will be utilized since the licensee's limit was determined to be less than the original correlation limit using the vendor's computer code.
The staff has reviewed the licensee's analyses and has determined that a correlation limit of 1.17 is appropriate for use by WCNOC with the WRB-2 correlation and VIPRE-01 computer code.
Following the determination of the DNBR correlation limit, the licensee reperformed the statistical core design analyses which had been-presented in Reference 2.
This included the incorporation of uncertainties as discussed in Reference 2 with the exception of the axial peaking factors. The presence of the IFNs and their effect on predicted DNBR was stated to result in difficulties in obtaining an acceptable fit for the response surface model.
The licensee has therefore treated the uncertainties associated with axial peaking in the traditional manner of a conservative DNBR penalty. The incorporation of the remaining parameters was consistent with the approved p
6 methodology of Reference 2.
The resultant DNBR statistical design limit for the k'RB-2/VIPRE-01 analysis is 1.31.
The accounting for the axial peaking uncertainties, transition core penalty, lower plenum flow anomaly, rod bow l
penalty, and design margin results in a DNBR thermal design limit (TDL) of 1.80.
The TDL value of 1.80 is used in subsequent thermal-hydraulic analyses to ensure that 95% protection at 95% confidence is provided by DNBR related operating limits and protection systems. The staff's review of the licensee's analyses has determined that the analyses are consistent with the methodology presented in Reference 2 and can be used for Cycle 7 and subsequent reload analyses.
The specific technical specification changes proposed by the licensee can be divided into four major categories. These are:
1)
Changes resulting from the licensee's reload design methodologies; 2)
Changes in key analysis assumptions incorporated into the licensee's analyses and loss of coolant accident analysis; 3)
Adoption of the Core Operating Limits Report; and 4)
Editorial changes resulting from the above major changes.
Each of the major changes is discussed below. The editorial changes required due to the referencing of technical specifications which have been revised or otherwise necessary for clarification have also been reviewed and determined to be acceptable.
TS 1-2: Core ODeratinQ limits Definition The proposed change introduces a definition for the Core Operating Limits Report. The staff's evaluation determined that the proposed change is consistent with the guidance in Generic Letter 88-16, " Removal of Cycle-Specific Parameter Limits from Technical Specifications." The staff finds the i
proposed change acceptable.
TS 2-1: Fiaure 2.1-1 The proposed change in the figure results from changes in the assumptions related to parameters such as reactor coolant system flow rate and maximum allowable peaking factors. The changes also result from physical differences in the fuel design and the associated critical heat flux correlation used to predict the occurrence of departure from nuclear boiling (DNB).
In addition, the change in the limits reflect the use of the licensee's methodology, including statistical core design, instead of the analytical methodology -
associated with the existing technical specifications. The limits were determined using the licensee's approved methodology (Ref. 2) as discussed above for the specific application to Cycle 7.
The assumptions were
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. adequately justified and, where appropriate, were incorporated in other proposed technical specification changes. The staff finds the proposed change acceptable.
TS 2.2: Table 2.2-1 Several terms associated with the Overtemperature Delta-T (OTDT) and Overpower Delta-T (0PDT) setpoints were changed to reflect changes incorporated into the transient analyses, the revised safety limits provided in TS Figure 2.1-1, and revised setpoint calculations performed by the licensee. The methodology used to determine the setpoints is in accordance with the licensee's approved methodology (Ref. 2) as well as established procedures and practices used in the determination of protection system setpoints. The revised OTDT and OPDT time constants were incorporated into the reanalysis of Updated Safety Analysis Report (USAR) Chapter 15 transients and were determined to provide adequate protection against violation of core safety limits or other applicable acceptance criteria. The f(delta-1) penalties were determined in accordance with the approved methodology addressed by Reference 5.
The staff finds the proposed changes acceptable.
i TS 3/4 1.1.1 and TS 3/4 1.1.2: Boration Control The proposed change increases the required shutdown margin in Mode 5, cold shutdown, to address concerns regarding the boron dilution transient and the boron dilution mitigation system. The reanalysis of the boron dilution transients is presented in the supporting documentation. The staff determined that the increase in required shutdown margin for Mode 5 operation was appropriate to cddress identified problems with the previous boron dilution analysis. The inclusion of Mode 5 in TS 3/41.1.1 and deletion of TS 3/4 1.1.2 is considered appropriate given that the shutdown margin requirements have become the same for the various modes of operation.
TS 3/4 1.1.3: Moderator Temperature Coefficient (MTC)
The MTC technical specification increases the beginning of cycle (most positive) limit from a constant 0 pcm/deg F to a limit which is a function of power level with the most positive limit being 6 pcm/deg F from 0% rated thermal power (RTP) to 70% RTP. The limit decreases linearly from 6 pcm/deg F at 70% RTP to O pcm/deg F at 100% RTP. This positive limit is introduced to the technical specifications as Figure 3.1-1.
The value of the end of cycle (most negative) MTC limit (-41 pcm/deg F) has been transferred to the Core Operating Limits Report in accordance with guidance of Generic Letter.88-16.
The supporting analyses have shown that the limits associated with both the most positive and most negative values of MTC are acceptable in that all FSAR Chapter 15 transient analyses continue to meet the applicable acceptanco criteria. The staff finds the proposed increase in the positive HTC limit and the relocation of the most negative MTC limit to the COLR to be acceptable.
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. TS 3/4 1.3.5/1.3.6: Shutdown and Control Rod Insertion Limits The only change to TS 3/4 1.3.5 and TS 3/4 1.3.6 is that the actual insertion limits are relocated to the COLR. No other changes to the limiting conditions for operation, required actions, or surveillance requirements are introduced.
This relocation is consistent with the guidance of Generic Letter 88-16 and is acceptable to the staff.
TS 3/4 2.1: Axial Flux Difference (AFD)
The only change to TS 3/4 2.1 is that Figure 3.2-1, " Axial Flux Difference Limits as a Function of Rated Thermal Power," is relocated to the COLR. No other changes to the limiting conditions for operation, required actions, or surveillance requirements are introduced. This relocation is consistent with i
the guidance of Generic Letter 88-16 and is acceptable to the staff.
TS 3/4 2.2: Heat Flux Hot Channel Factor (F/X.Y.7)
TS 3/4 2.2 is changed to reflect the licensee's performance of the nuclear analysis and related surveillances in accordance with the approved methodology described in Reference 5.
The change is similar to the Babcock and Wilcox methodology described in Reference 6.
In addition to the methodology related changes, the relocation of the F, limits and related parameters to the COLR is also proposed. The staff's review of the proposed changes concluded that, with the exception of several minor terminology and plant specific implementation differences, the proposed changes are consistent with the model technical specification included in Reference 6.
The staff finds the proposed changes acceptable.
It is noted that the licensee has utilized a more recent loss of coolant accident evaluation model (Ref. 7) in order to increase the F, limit from 2.32 to 2.50.
The evaluation model has been approved by the staff and is included in the list of approved methodologies provided in.TS 6.9.1.9.
TS 3/4 2.3: Nuclear Enthalov Rise Hot Channel Factor F (X.Y)
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TS 3/4 2.3 is changed to reflect the licensee's performance of the nuclear analysis and related surveillances in accordance with the approved methodology J
described in Reference 5.
The change is similar to the Babcock and Wilcox methodology described in Reference 6.
In addition to the methodology related changes, the relocation of the F limits and related parameters to the COLR a
is also proposed. The staff's review of the proposed changes concluded that, with the exception of several minor terminology and plant specific implementation differences, the proposed changes are consistent with the model technical specification included in Reference 6.
As part of the changes, requirements associated with reactor coolant system flow rate are moved to TS 3/4 2.5, DNB parameters. The staff finds the proposed changes acceptable.
It is noted that the licensee has included increased nuclear enthalpy rise assumptions in the analysis for Cycle 7.
The F limit has been increased a
from 1.55 to 1.65.
The increase has been justified by the supporting analysis
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. and related NRC approved methodologies. The related topical reports as well as this Safety Evaluation have been included in the list of approved methodologies provided by TS 6.9.1.9.
TS 3/4 2.5: DNB Parameters The basic change to TS 3/4 2.5 is the addition of reactor coolant flow rate requirements to the specification. The required flow rate has been reduced from the current value of 384,400 gpm to 384,000 gpm.- The reduced flow was assumed in the supporting analyses, including the core thermal-hydraulic analyses which detertnined the core safety limits and protection system setpoints.
In addition to adding the reactor coolant flow requirement to the limiting condition for operation, the existir!g action and surveillance requirements associated with reactor coolant system flow (existing TS 3/4 2.4) have been added to TS 3/4 2.5.
The staff determined that the existing action and surveillance requirements remain applicable and therefore finds the proposed change acceptable.
i TS 3/4 7.1: Table 3.7.2. Steam Line Safety Valves Per Looo The proposed change involves a revision to the required as found lift setting of the main steam safety valves from +/- 1% of the setpoint to +/- 3% of the setpoint. The increased tolerance was included in the supporting analyses and it was determined that adequate protection from secondary system over-pressurization was maintained..In order to prevent exceeding the setpoint by an excessive amount over the course of several. cycles, the proposed change includes a requirement to leave the main steam safety valves within +/- 1% of the setpoint following inservice testing. The testing requirements, including i
the proposed as found and as left tolerances, will continue to meet the requirements of the licensee's ASME Section XI inservice testing program. The staff finds the proposed changes acceptable.
TS 3/4 9.1: Boron Concentrations (Refuelino Operations)
The only change to TS 3/4 9.1 is that required boron concentration for the reactor coolant system and refueling canal during refueling operations is relocated to the COLR. No other changes to the limiting conditions for operation, required actions, or surveillance requirements are introduced.
This relocation is consistent with the guidance of Generic Letter 88-16 and is acceptable to the staff.
TS 3/4 9.12: Fioure 3.9-1. Burnuo Vs. Enrichment i
TS Figure 3.9-1 provides the required fuel assembly burnup as a function of i
initial enrichment for storage of a spent fuel assembly in Region 2 of the spent fuel storage pool. The figure needed to be reanalyzed due to the replacement of inconel spacer grids with zircaloy spacer grids when going from the Westinghouse Standard to VANTAGE SH fuel assembly design. The previous figure had included credit for the neutron absorption characteristics for the inconel spacer grids. The maximum initial enrichment assumption was reduced from 4.50 weight percent to 4.45 weight percent when the spent fuel pool
. criticality analyses were performed for the transition to VANTAGE SH fuel assemblies. The net effect of the change in spacer grid material and the reduction in the maximum initial enrichment is a minor change to the burnup versus enrichment limits provided in Figure 3.9-1.
The staff finds the proposed change acceptable.
TS 5.3: Desion Features - Fuel Assemblies The proposed change deletes the maximum enrichment value f,om 15 5.3.
However, the maximum enrichment value is retained in TS 5.6, Fuel Storage.
The proposed changs is a deletion of a redundant specification and is acceptable to the staff.
TS 5.6: Desion Features - Fuel Storace The proposed change deletes the specific value for the uncertainty allowance in the calculation of spent fuel pool subcriticality but retains the reference to the applicable FSAR section.
In addition, Figure 5.5-1, which is a duplication of TS Figure 3.9-1, is deleted and the TS references Figure 3.9-1.
The maximum enrichment limit is reduced from 4.50 weight percent to 4.45 weight percent to account for insertion of VANTAGE SH fuel assemblies into the spent fuel pool. The reduction in the enrichment limit as well as the editorial changes to this specification are acceptable to the staff.
TS 6.9.1.9: Core Operatino Limits Report The proposed change deletes the requirement for the submittal of a radial peaking factor limit report and introduces the requirement for the submittal of a Core Operating Limits Report (COLR), which documents specific values of cycle-specific parameters using NRC-approved methodologies. The licensee's proposed change is in accordance with the guidance of Generic Letter 88-16 and lists the associated technical specifications and methodologies approved by the staff for the determination of cycle-specific parameters. The requirement that only the listed NRC-approved methodologies may be used to determine core operating limits in the COLR ensures safe operation of the facility within approved acceptance criteria. Therefore, removal _ of cycle-specific parameters, as proposed, is acceptable.
Based on the above, the staff finds the proposed TS changes acceptable.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Kansas State official was notified of the proposed issuance of the amendment. The state official had no comments.
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4.0 ENVIRONMENTAL CONSIDERATION
t The amendment changes a requirement with respect to installation or use of a facility component located with the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant changes in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative i
occupational radiation exposure. The Commission has previously issued a L
proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (58 FR 601 dated January 6, 1993).
In addition, the amendment changes recordkeeping ur reporting requirements. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (c)(10).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or
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environmental assessment need be prepared in connection with the issuance of the amendment.
5.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the l
public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
REFERENCES 1.
WCAP-10444-P-A, Addendum 2A, " VANTAGE SH Fuel Assembly,"
Westinghouse Electric Corporation, February 1989.
i 2.
Safety Evaluation by the Office of Nuclear Reactor Regulation Relating to Topical Report TR-90-0025 WOI, Core Thermal-Hydraulic Analysis Methodology for the Wolf Creek Generating Station, October 29, 1992.
3.
WCAP-10444-P-A, " VANTAGE 5 Fuel Assembly Reference Core Report,"
Westinghouse Electric Corporation, September 1985.
4.
Letter from Rossi, C.E. (NRC) to Blaisdell, J.A. (UREA), " Acceptance for Referencing Licensing Topical Report, EPRI NP-2511-CCM, "VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores," May 1,1986.
5.
NSAG-007, " Wolf Creek Nuclear Operating Corporation Reload Safety
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Evaluation Methodology for Wolf Creek Generating Station Wolf Creek l
Nuclear Operating Corporation," March II, 1992, as supplemented by Forrest T. Rhodes' letter dated February 3, 1993.
6.
BAW-10163P-A, " Core Operating Limits Methodology for Westinghouse Designed PWRs," Babcock & Wilcox, June 1989.
i
1 7.
WCAP-10266-P-A, Rev. 2, "The 1981 Version of Westinghouse Evaluation Model Using BASH Code," March 1987.
Principal Contributor: William D. Reckley, NRR/DRPW Date:
March 30, 1993 6
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