ML18040A878
| ML18040A878 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 04/25/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18040A879 | List: |
| References | |
| NUDOCS 8805030493 | |
| Download: ML18040A878 (8) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLFAR REACTOR REGUILATION SUPPORTING AMENDMENT NO. 45 TO FACILITY OPERATING LICENSE NO.
NPF-22 PENNSYLVANIA POMER 5 LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE, INC.
DOCKET NO. 50-388 SUS UEHANNA STEAM ELECTRIC STATION, UNIT 2
1.0 INTRODUCTION
By letter dated December 23, 1987, Pennsylvania Power 5 Light Company (PPSL) requested an amendment to Facility Operating License No.
NPF-22 for the Susquehanna Steam Electric Station, Unit 2.
The proposed amendment would change the Technical Specifications to support authorization for Susquehanna Steam Electric Station (SSES) Unit 2 operation with 9X9 Cycle 3 (S2C3) reload fuel supplied by Advanced Nuclear Fuels (ANF) Corporation.
The Susquehanna 2
S2C3 reload will consist of 236 fresh ANF fuel assemblies (XN-2) intermixed with 324 ANF 9X9 assemblies (XN-1) and 204 initial core General Electric (GE)
P8XBR assemblies.
In support of the S2C3 reload, PPAL submitted topical reports which describe the reload scope, the proposed startup physics tests, the plant transient analysis, and the design and safety analyses.
2.0 EVALUATION 2.1 Fuel Mechanical Desi n
The S2C3 core reload will include 236 ANF 9X9 fuel assemblies with the designation XN-2.
These reload assemblies contain 79 fuel rods and two water rods.
The XN-2 reload fuel consists of 140 assemblies which contain nine burnable poison rods and 96 assemblies which contain 10 burnable poison rods.
These 236 assemblies will have an assembly average enrichment of 3.33 weight percent (w/o) U-235.
The fuel design and safety analysis for the XN-2 fuel are the same as those for the previous cycle XN-1 fuel and are described in the Susquehanna 2 specific report ANF-87-126 and the generic mechanical design report XN-NF-85-67(P)(A), Revision 1.
The staff has approved the latter report and issued an SER on July 23, 1986.
Table 2.1 of XN-NF-85-67, Revision 1 gives the pertinent data for the ANF 9X9 fuel.
Neutronic values specific to the S2C3 reload are given in Table 4. 1 of ANF-87-126, Revision 1.
The ANF XN-2 fuel is designed to fit into the existing GE channel boxes.
Based on the staff's review of the information presented, the mechanical design of the ANF 9X9 fuel for the S2C3 reload is acceptable.
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2.2 Rod Pressure For the S2C3 ANF 9X9 reload fuel, calculation of the fuel rod internal pressure was done in accordance with acceptance criteria cited by ANF in Reference 6.
The evaluation was performed with the RODEX?A computer code which has been reviewed and approved by the staff.
The staff has concluded that the acceptance criteria for rod internal pressure can be fully met throughout the entire expected irradiation life of the 9X9 fuel.
2.3 Fuel Rod Bow S2C3 is expected to result in a peak XN-1 assembly exposure of less than 30,000 MWD/MTU at end-of-cycle.
The staff has reviewed Reference 9 which provides additional rod bow measurements on 9X9 Lead Test Assemblies and has concluded that assembly discharge exposures of 40,000 MWD/MTU are acceptable for the XN-I and XN-2 fuel designs.
2.4 Fuel Centerline Meltin The design basis for the ANF fuel centerline temperature is that no fuel centerline melting should result from normal operation including anticipated transient occurrences.
The results of an evaluation reported in the S2C3 reload analysis were based on the approved RODEX2A code and the staff has concluded that the generic methodology for the ANF 9X9 fuel is acceptable for the S2C3 reload fuel.
?.5 Linear Heat Generation Rate
'LHGR)
Limit or 9X ue A figure of LHGR limit versus Planar Exposure (MWD/MT) for the ANF 9X9 fuel type is incorporated into the Susquehanna 2 Technical Specifications.
This Figure was approved in the staff's safety evaluation dated July 23, 1986 to reflect the design values which have been previously reviewed and approved for the ANF 9X9 fuel in connection with the Staff's review of XN-NF-85-67, Revision 1.
Based on the results of the generic review the staff finds the LHGR limits for the 9X9 fuel acceptable.
3.0 NUCLEAR DESIGN The ANF nuclear design methodology for S2C3 is that presented in XN-NF-80-19(A),
Volume 1, and its Supplements 1 and 2, which were reviewed and approved by the staff for generic application to BWR reloads.
The beginning-of-cycle shutdown margin is calculated to be 1.50 percent delta k/k, and the R factor is zero.
Thus the cycle minimum shutdown margin is well in excess of required 0.38 percent delta k/k.
The Standby Liquid Control System also fully meets shutdown requirements.
The existing new fuel storage calculations are based on k-infinity of the assembly.
Based on ANF calculations of 9X9 fuel, an average enrichment of less than 4.0 w/o U-235 and a k-infinity of less than or equal to 1.388 will meet the acceptance criterion of k-effective no greater than 0.95 under dry or flooded conditions.
Since the maximum enrichment of the new fuel is 3.4P w/o U-235 and the maximum cold, uncontrolled, k-infinity is 1. 10349, the calculations show the staff's acceptance criterion is met for the new fuel storage vault under all normal and postulated abnormal conditions.
ANF also performed analyses for 9X9 fuel stored in the SSES spent fuel pool.
A maximum enriched zone of, less than 4.0 w/o U-235 with. an uncontrolled, zero void, cold, k-infinity of less than or equal to 1.457 meets the staff acceptance criterion of k-effective no greater than 0.95.
Since the XN-2 9X9 fuel has an enrichment of 3.44 w/o U-235 and a maximum k-infinity of 1.2206 at peak reactivity, the staff's acceptance criterion for spent fuel storage is also met for ANF 9X9 fuel.
Susquehanna will continue to use the ANF POWERPLEX core monitoring system to monitor reactor parameters.
The system has been in use since the previous operating cycle and during Unit 1 Cycles 2, 3, and 4 and has provided suitable monitoring and predictive results.
4.0 THERMAL-HYDRAULICDESIGN The review of the thermal-hydraulic aspects of the S2C3 reload consisted of the following:
(a) the compatibility of the ANF 9X9 and prior GE 8X8 fuel assemblies; (b) the fuel cladding integrity safety limit; (c) the bypass flow characteristics; (d) thermal-hydraulic stability.
The objective of the review was to confirm that the thermal-hydraulic design of the reload core was accomplished using acceptable analytical
- methods, provided an acceptable margin of safety from conditions which would lead to fuel damage during normal operation and anticipated operational occurrences and ensured that the core is not susceptible to thermal-hydraulic instability.
4.1 H draulic-Com atibilit Since a
BWR core is a series of parallel flow channels connected to a
common lower and upper plenum, the total pressure drop across the assemblies wi 11 be equal.
However, differences in the hydraulic resistances of the fuel designs may cause variations in axial pressure drop profiles across the assemblies.
Component hydraulic resistances for the proposed constituent fuel types in the S2C3 core have been determined in single phase flow tests of full scale assemblies.
Additional analyses of the effects of hydraulic compatibility on thermal margin were presented in the S2C3 reload reports.
Based on the staff's review of the information provided in the pertinent documentation and on the fact that the staff has previously approved coresidence of GE PBX8R and ANF 9X9 fuel for Unit 2, and on the hydraulic equivalence of the XN-2 9X9 design and the XN-I 9X9 design, the staff concludes that the ANF and GE fuel types in S2C3 are hydraulically compatible.
4 4.2 Minimum Critical Power Ratio Safety Limit The minimum critical power ratio (MCPR) safety limit for the Cycle 3 reload was determined by the licensee to be 1.06 for all fuel types.
The methodology for Cycle 3 is based on the ANF critical power methodology in XN-NF.-524, Revision I, which has been approved by the staff.
The XN-3 correlation used to develop the MCPR safety limit has been approved for the ANF 9X9 fuel.
The methodology of XN-NF-524, Revision I was applied generically for the upcoming Cycle 3 and is considered applicable to the resident GE 8X8 fuel as well as the ANF fuel.
The staff has verified through its review of the S2C3 transient analysis report that the methodology for determining uncertainties and the application in determining the MCPR safety limit is in accordance with NRC approved methodology and is acceptable.
4.3 ~C..
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The core bypass flow fraction has been calculated as 10.1'A of total core flow using the approved methodology described in XN-NF-504(A), Revision I.
This is used in the MCPR safety limit calculation and as input to the Cycle 3 transient analyses and is acceptable.
4.2 Thermal-H draulic Stability The thermal-hydraulic stability of the SSES 2 core was analyzed using the methods identified in Exxon Reports related to Nutronic design and analysis
- methods, and stability evaluation methods.
The licensee has also performed a
stability startup test in Unit 2 during initial startup of Cycle 2 to demonstrate stable reactor operation with ANF 9X9 fuel.
In addition, the Unit 2 Technical Specifications have implemented survei llances for detecting and suppressing power osci llations.
The acceptability of these surveillance requirements as well as the tests and analyses mentioned above have been evaluated for S2C3 in a separate safety evaluation for Unit 2 s'ingle loop operation.
5.0 TRANSIENT AND ACCIDENT ANALYSES
- 5. 1 0 erational Transients Various operational transients could reduce the MCPR below the intended safety limit.
The most limiting transients have been analyzed to determine which event could potentially induce the largest reduction {delta CPR) in the initial CPR.
The core-wide transients which resulted in the largest delta CPR from rated conditions (104$
power/100%%u flow) are the load rejection without bypass (LRWOB) and the feedwater controller failure (FWCF).
These resulted in delta CPRs of 0.24 and 0.23, respectively.
The most limiting local transient, the rod withdrawal error (RWE), was analyzed to support a rod block monitor (RBM) setting of 108% and resulted in a delta CPR of 0.26, requiring a
MCPR operating limit of 1.32.
This was the most limiting event for S2C3 at rated power and flow conditions.
At less than rated power, the FWCF event is limiting and a curve of MCPR versus power based on the FWCF results is included in the Technical Specifications as a power dependent MCPR operating limit.
At reduced flow conditions, the recirculation flow control failure is limiting and MCPR operating limits for manual flow control reduce flow operation for Cycle 3 based on the analysis of this event are provided as a Technical Specification figure of MCPR versus core flow.
These calculations were performed with approved methodology and the resulting Technical Specification limiting curves are acceptable.
It was assumed for the above analyses that the turbine bypass system and the end-of-cycle recirculation pump trip (RPT) were operable.
Analyses were also performed to determine the MCPR operating limits with either of these systems inoperable.
This resulted in increased MCPR limits which are also proposed for S2C3.
These calculations follow standard procedures and operation within the proposed MCPR operating limits with either the main turbine bypass inoperable or the end-of-cycle RPT inoperable is acceptable for S2C3.
Compliance with overpressurization criteria was demonstrated by analysis of main steam isolation valve (MSIV) closure with MSIV position switch scram failure.
Six safety relief valves were assumed out of service.
Maximum pressure was 104~< of vessel design pressure, well within the 110$
criterion.
The calculation was done with approved methodology and the results are acceptable.
5.2 Postulated Accidents The GE loss of coolant accident (LOCA) analysis and maximum average planar linear heat generation rate (MAPLHGR) limits for the GE 8X8 fuel remain applicable for Cycle 3 although an additional exposure point at 40,675 MWD/MTV is added to the GE Tvpe III MAPLHGR limit curve.
The staff has previously approved this new curve for S1C3 operation and finds it acceptable for GE Type III fuel in S2C3 as well.
The licensee has also presented MAPLHGR limits for the ANF 9X9 fuel*based on the analysis results provided in XN-NF-86-65.
The LOCA analyses have covered an acceptable range of conditions, have been performed with approved methodology, and the results meet the limits specified in 10 CFR 50.46.
Therefore, the staff finds the proposed MAPLHGR limits for SZC3 acceptable.
The control rod drop accident (CRDA) was analyzed with approved ANF methodology.
The maximum fuel rod enthalpy was 205 cal/gm, which is well below the design limit of 280 cal/gm, and less than 250 fuel rods exceeded 170 cal/gm, which is less than the 770 rods assumed in the SSES FSAR analysis.
To ensure compliance with the CRDA analysis assumptions, control rod sequencing below 205 core thermal power must comply with GE's banked position withdrawal sequencing constraints.
The staff finds the analysis and results of the CRDA for Cycle 3 acceptable.
6.0 TECHNICAL SPECIFICATION CHANGES The following Technical Specification (TS) changes have been proposed for SSES for operation during reload Cycle 3:
)
t'
(I)
TS 3/4.2.1, Avera e Planar Linear Heat Generation Rate The allowed exposure for GE 2.33 w/o enriched fuel has been increased to 40,675 MllD/MTU.
In addition, editorial changes to correct misarranged wording and the vendor reference are made.
The change to the GE limit is based on an approved GE LOCA analysis and is acceptable as discussed in Section 4.3 of this SER.
The editorial changes include the replacement of references to "Exxon" with "ANF" to reflect the corporate name change.
These editorial changes are administrative only with no safety significance and are, therefore, acceptable.
(2)
TS.3/4.2.2, APRM Set pints An editorial change corrects
'the vendor reference from "Exxon" to "ANF".
This change is administrative only with no safety significance and is acceptable.
(3)
TS 3/4.2.3, Minimum Critical Power Ratio Operating limit MCPRs have been revised to reflect the results of the cycle specific transient analyses.
The methodology used to evaluate the limiting transients and accidents is consistent with previously approved methods and meets all the appropriate NRC criteria.
Therefore, the revised MCPRs are acceptable for Cycle 3 as discussed in Section 5.0 of this SER.
(4)
TS 3/4.2.4, Linear Heat Generation Rate An editorial change corrects the vendor reference from "Exxon" to "ANF."
This change is administrative only with no safety significance and is acceptable.
(5)
TS 3/4.3.6, Control Rod Block Instrumentation Footnote "¹¹" to trip function 2a has been added to refer to TS 3.4. 1.1.2.a for single loop operation requirements.
This change is editoriaT in nature and since single loop operation has been approved for S2C3 in the staff's safety evaluation for single loop operation, it is acceptable.
(6)
TS 3/4.4. 1, Recirculation System Changes have been made to the single loop and two loop operation requirements.
These changes have been reviewed and evaluated under a separate licensing action.
a Based on the above considerations the staff concludes that appropriate material was submitted and that the fuel design, nuclear design, thermal-hydraulic design and transient and accident analyses are acceptable.
The Technical Specification changes submitted for this reload suitably reflect the necessary modifications for operation in this cycle and are acceptable.
- 3. 0 ENVIRONMENTAL CONS IDFRATION This amendment involves a change to a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The staff has determined that the amendment involves 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 this amendment involves no significant hazards consideration and there has been no public comment on such finding.
Accordingly, this amendment meets 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 nor environmental assessment need be prepared in connection with the issuance of this amendment.
4.0 CONCLUSION
The Commission made a proposed determination that the amendment involves no significant hazards consideration which was published in the Federal Re ister (53 FR 2322) on January 27, 1988 and consulted with the~tate of ennsy vania.
No public comments were received, and the State of Pennsylvania did not have any comments.
The staff has 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 this amendment will not be inimical to the common defense and security nor to the health and safety of the public.
Principal Contributor:
L. Kopp Dated:
April 25f 1988