ML20211N261
| ML20211N261 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 09/08/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20211N256 | List: |
| References | |
| NUDOCS 9909100129 | |
| Download: ML20211N261 (6) | |
Text
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[ % UNITED S'TATES j
o 's j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055M001 4 90 9....
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO NORTH ANNA POWER STATION UNITS 1 AND 2 PROPOSED IRRADIATION OF FUEL RODS BEYOND CURRENT LEAD ROD BURNUP LIMIT VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 DOCKET NOS. 50-338 AND 50-339
1.0 INTRODUCTION
By letter dated April 16,1999 (Reference 1), Virginia Electric and Power Company (Virginia Power) informed the NRC of plans to irradiate a small number (8) of fuel rods to end-of-life !
average burnups ranging from 65,000 to 73,000 MWD /MTU to provide data on fuel and material performance to support the industry goal of extending the current fuel burnup limits.
Since a licensing basis commitment limits lead rod burnup to 60,000 MWD /MTU, Virginia Power ,
requested NRC approval prior to implementation of this program. Virginia Power also i requested a clarification of the terminology with regard to reconstituted fuel assemblies at North Anna, Units 1 and 2. During a June 24,1999 meeting, additional information pertaining to the proposed irradiation program was presented and a letter dated July 28,1999 (Reference 2),
documents that information.
2.0 EVALUATION 2.1 Use of High Burnup Fuel Rods A burnup restriction (batch average of 50,000 MWD /MTU or above, provided the maximum rod average burnup of any rod is no greater than 60,000 MWD /MTU) was documented in Refere,nces 3 and 4 and currently remains applicable to both North Anna and Surry although it is not explicitly stated in the License Conditions or Technical Specifications for either power station. This limit is consistent with the NRC Safety Evaluation Report (SER) on the Westinghouse topical report WCAP-10125, entitled " Extended Burnup Evaluation of Westinghouse Fuel." Lead test assemblies (LTAs) irradiated to higher burnups are needed in order to obtain data on fuel and material performance, help confirm the applicability of nuclear design and fuel performance models at high burnups, and address NRC questions related to fuel performance behavior at high burnups.
l Virginia Power plans to irradiate a small number of ZlRLO fuel rods in Assembly 3A4 to high l burnups in the North Anna 2 Cycle 14 core. The use of these fuel rods is being fully evaluated Enclosure
( 9909100129 990908 PDR ADOCK 05000338 p PDR Y_
f as part of the reload design process and the licensee has stated that all design criteria will be satisfied or another fuel assembly wah no lead fuel rods will replace fuel Assembly 3A4.
Additional conservatisms are being used in the evaluation of the high burnup fuel rods, and the analysis of Assembly 3A4 will be documented separately.
All the high burnup rods were well characterized when they were first fabricated for use in the original North Anna advanced material demonstration program. Examinations have been performed on several of the fuel rods after each operating cycle; in particular, the oxide thickness was measured to be low (35-40 micron range) after the third cycle. The high burnup fuel rods were visually inspected during insertion intc Fuel Assembly 3A4 and no indications of j unusual oxidation or crud buildup was observed. Examinations of the high burnup rods at the I end of North Anna 2 Cycle 14 will include fuel rod length, profilometry, and oxide measurements in addition, hot cell examinations to measure fission gas release, clad hydrogen pickup, cladding metallography and fuel microstructure are being considered.
Fuel rod design criteria that become more limiting for high burnup fuel rods include fuel rod growth, clad fatigue, rod internal pressure and cladding corrosion. Evaluations are being performed using the same NRC-approved models that are normally used for fuel rod design.
These models have been used to perform similar evaluations for other high burnup LTAs. In i particular, the cycle-specific assessment of end-of-life corrosion thickness for the high burnup fuel rods will be performed using the NRC-approved model normally used for fuel rod design of I ZlRLO-clad fuel. In addition, a developmental corrosion model for ZlRLO will be used as an assessment tool. Since this developmental model is based primarily on data from fuel irradiated at the V. C. Summer and North Anna plants (including measurements from these high burnup fuel rods af ter previous cycles), there is high confidence that it will provide accurate end-of-life estimates of the oxidation for the high burnup fuel rods.
1 Reactor operations, including core operating limits and setpoints, are not affected by the presence of a small number of fuel rods operating to high burnups. The high bumup fuel rods will not be in the highest fuel rod power density locations in the core and will not be limiting with respect to any safety limit. The average assembly power for Assembly 3A4 over the length of Cycle 14 will be approximately equal to the core average, but the high burnup rods are less reactive and will operate at approximately 0.8 times the core average power. Virginia Power will track and model the different burnups and slightly different enrichments of the high burnup fuel rods and the remaining rods in the assembly.
Virginia Power has evaluated the impact of the higher burnup fuel rods on the spent fuel pool and determined that the analyses of record will remain conservatively bounding for the high l burnup fuel rods in Assembly 3A4. To determine the impact on the loss-of-coolant accident (LOCA) analysis, Virginia Power examined the limiting conditions for input to the LOCA analysis and concluded that they remained unchanged since the fuel design criteria are met. Therefore, operation with the high burnup fuel rods will not impact the existing LOCA analysis for North Anna Unit 2 Cycle 14. The NRC staff has reviewed the justification provided by Virginia Power and agrees with the conclusion as stated.
The potentialimpact of the high burnup fuel rods on the non-LOCA safety analyses is addressed by the licensee as part of the North Anna Unit 2 Cycle 14 reload safety analysis.
Again, since the fuel design criteria are satisfied for the high burnup fuel rods, the limiting fuel temperatures inputs to the safety analysis remain unchanged. The cycle-specific reload t
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analyses for North Anna Unit 2 Cycle 14 will c>nfirm that operation vith the high burnup fuel rods will not increase the probability of occurrence or consequences for any postulated accidents. Wri have reviewed the licensee's submittal with respect to non-LOCA safety analyses and agree with the sbove conclusion. Therefore, it is acceptable to irradiate a small J
number (8) of lead fuel rods to end-of-life average burnups ranging from 65,000 to 73,000 l MWD /MTU. '
2.2 Reconstitution Definition i in 1994, a Technical Specification change was approved for North Anna, Units 1 and 2 to allow substitution of solid stainless steel for zirconium alloy filler rods for a limited number of failed fuel rods in fuel assemblies. Both the Virginia Power submittal and the NRC approval stated ;
that substitutions are permitted for failed fuel rods. The wording of the North Anna Technical '
Specification states that limited substitutions of filler rods "for fuel rods" may be made. Since that time, changes in reconstitution technology have occurred. It may now be necessary to remove one or more intact rods from an assembly. Replacing an intact, or non-failed, rod with a filler rod was not explicitly considered in the November 19,1993, submittal. However, in l evaluating the reuse of fuel assemblies with solid filler rods, it is the final configuration of the I assembly that is important, rather than the integrity of the fuel rods that were replaced. The licensee desired to clarify the term " reconstituted fuel assembly" such that it could be applied to any North Anna fuel assembly in which fuel rods have been replaced with solid filler rods, regardless of whetiicr the replaced rod was failed or intact. Furthermore, the licensee reiteidad tbs fact that reconstituted assemblies will continue to be explicitly modeled neutronical,y, and thermal hydraulic calculations will continue to evaluate the exact configuratian of any reconstituted assembly, to confirm that the critical heat flux correlation used for departure from nucleate boiling predictions remains applicable. We have reviewed the licensee's justification and agree that compliance with NRC requi ements on the use of ,
reconstituted fuel will not be impacted by the interpretation of " reconstituted fuel assembly" to ,
mean that the replaced fuel rods may be failed or intact rods. The afore, this interpretation is j acceptable.
2.3 Radiological Assessbient l The licensee assessed the impact on both normal effluents and releases from postulated accidents of operation with lead fuel rods with burnup exceeding 60,000 MWD /MTU. The following sections provide the staff's evaluation of that assessment.
l 2.3.1 Normal Effluents in the licensee's July 28,1999 letter, it was indicated that even assuming that high burnup fuel rods have some increased contribution to the core inventory of long-lived isotopes, since only eight fuel rods (0.02% of the fuel rods in the core) are operating to high burnup, there would be no measurable increase in the levels of isotopes in reactor coolant. Consequently, there would be no effect on normal operating plant releases. The staff has reviewed this assessment and is in agreement with its conclusion.
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2.3.2 Accident Assessment 2.3.2.1 Fuel Handling Accident The July 28,1999, letter noted that the North Anna fuel handling accident involves a single assembly and follows the guidance of Regulatory Guide 1.25. Consecuently, the analysis is based upon an assembly operating at 1.65 times the core average power. In addition, the North Anna analysis assumes the accident occurs 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> following reactor shutdown even though the North Anna Technical Specifications prohibit fuel movement until 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> following shutdown. For Fuel Assembly 3A4, the assembly average power during North Anna 2 Cycle 14 is approximately the core average power which is much lower than the assumed power for analysis of the fuel handling accident. The licensee indicated that the high burnup rods have a limited amount of reactivity remaining and will operate at stilllower powers of about 0.8 times the core average power. Therefore, the licensee concluded that in the event that the fuel rods of Assembley 3A4 were involved in a fuel handling accident, the activity releases that would result from damage to the rods would be considerably lower than those determined for the existing fuel handling accident for North Anna. Consequently, the existing analysis would remain bounding. The staff has reviewed this assessment and is in agreement with the licensee's conclusion.
2.3.2.2 Steam Generator Tube Rupture and Main Steamline Break The licensee stated that the steam generator tube rupture (SGTR) and the main steamline break (MSLB) accidents do not involve fuel failures and that the consequences of these accidents are based upon fuel failures in existence at the time of the accident. The staff has reviewed this assessment and has concluded that the utilization of Assembly 3A4 would not result in fuel failures as a result of an SGTR or an MSLB accident. In addition, given no fuel i failures, these accidents are limited by the technical specification values for dose equivalent *l in primary coolant. Even if the rods in Assembly 3A4 would develop leaks, the technical specification limits on dose equivalent *l remain to limit primary coolant activity levels and assure that the existing analysis remained the bounding analysis. The staff is in agreement with the licensee's conclusion with respect to the SGTR and the MSLB accidents. '
2.3.2.3 Locked Rotor Analysis North Anna analyses of loss of flow accidents show that the minimum depF tore from nucleate boiling ratio (DNBR) dm? not decrease below the limit, so no cladding faae or release of fission products is expected. The current North Anna locked rotor accident predicts that no fuel rods will experience DNB. However, the North Anna analysis assumes failure and gap release for 13% of the fuel rods. The licensee's July 28,1999, letter indicated that the 13% value was based upon a locked rotor analysis that has now been superseded. The licensee stated that fuel failures associated with a locked rotor accident would be expected to occur in high power locations because the high power rods are more likely to enter a boiling regime during a transient. The high burnap fuel rods in Assembly 3A4, which will be operating in an area at I about 0.8 times the core average power, would not be expected to fail in a locked rotor accident I scenario. The licensee further noted that, since the current analysis predicts no other fuel rod
} failures during a locked rotor event, even if the eight high burnup fuel rods should fail, there would be no impact upon the current dose calculations for the locked rotor. The staff has
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4 5-reviewed the licensee's assessment of the consequences of a locked rotor accident and is in agreement with their conclusions.
2.3.2.4 Loss-of-Coolant Accidents The licensee's assessments of the consequences of a LOCA encompass both a large break and small breaks. The July 28,1999, letter indicated that the dose calculations for the LOCAs are bounded by the large break which assumes damage to the entire core. North Anna's large break LOCA analysis follows the guidelines of Regulatory Guide 1.4, which incorporates dose calculations which are based on specific distribution of the core inventory of fission products and not just the material present in the fuel-to-clad gap. Since the doses for a LOCA are primarily due to short-lived iodine and noble gas isotopes and their core inventory is a function of operating power rather than burnup, the core inventory for a LOCA will be unaffected by the utilization of the eight high burnup rods. Consequently, the calculated releases for a LOCA would remain unchanged. The staff has reviewed the licensee's assessment and is in agreement.
2.3.2.5 Rod Ejection in the July 28,1999 submittal, the licensee stated that for the rod ejection accident, limited melting (<10% of the pellet)is predicted to occur at the hot spot in the core. The analysis of this accident is based on conditions in the peak rod in the core, which is typically a fresh or once-burned fuel rod. However, because the melting point of UO2 decreases with burnup, the licensee's cycle-specific evaluations currently in progress for North Anna 2 Cycle 14 will need to l confirm that the presence of the high burnup rods does not affect the current analysis of the rod ejection transient. The licensee has indicated that they do not expect this to present any difficulty because the high burnup fuel rods in Assembly 3A4 will only be operating at about 0.8 times the core average power. In addition, the licensee's July 28,1999, letter stated that a l Westinghouse position letter to the NRC in May 1995 indicated that the predicted effect of an
- ejected rod is very localized. Since Fuel Assembly 3A4 is not going to be placed in a rodded location and will also be located well away from the core locations that have historically been i limiting for the control rod ejection accident at North Anna, the eight rods of Fuel Assembly 3A4 l are not anticipated to be affected by a rod ejection accident. However, in the July 28,1999 letter, the licensee stated that if the cycle-specific evaluation for North Anna 2 Cycle 14 l ' det, ermines the current rod ejection analysis is affected by the presence of the high burnup fuel, l consistent with the requirements of 10 CFR 50.59 either (1) a new analysis will be performed,
! as appropriate, and NRC approval obtained or (2) the assembly will be removed from the core loading pattern. The staff has reviewed the licensee's assessment and has concluded that the licensee's approach is acceptable.
3.0 CONCLUSION
Based on the evaluation in Section 2.0 above, the staff concludes that the proposed irradiation of fuel rods beyond the current lead rod burnup limit and the clarification of terminology with respect to reconstituted fuel assemblies are acceptable. We conclude that operation of the high burnup fuel rods will provide much needed data on fuel performance behavior at high burnups while maintaining a high standard of safety. We have no objection to the requested use of such rods in a reconstituted assembly.
Principal Contributors: M. Chatterton J. Hayes Date: September 8,1999
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4.0 REFERENCES
- 1. . Letter from D. A. Christian, Virginia Electric and Power Company to NRC Document Control Desk, dated April 16,1999.
- 2. ' Letter from D. A. Christian, Virginia Electric and Power Company to NRC Document Control Desk, dated July 28,1999.
- 3. Letter from Leon B. Engle and Bart C. Buckley, (U.S. Nuclear Regulatory Commission) to W. L. Stewart (Virginia Electric and Power Company), "Surry, Units 1 and 2, and North Anna Units 1 and 2 -Removal of 45,000 MWD /MTU Batch Average Burnup Restriction (TAC Nos. M87767, M87768, M87812 and M87813)," December 13,1993.
- 4. Letter from Leon B. Engle and Bart C. Buckley, (U.S. Nuclear Regulatory Commission) to W. L. Stewart (Virginia Electric and Power Company), "Surry, Units 1 and 2, and North Anna Units 1 and 2 -Removal of 45,000 MWD /MTU Batch Average Burnup Restriction (TAC Nos. M87767, M87768, M87812 and M87813)," April 20,1994.
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