ML17156B165
| ML17156B165 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 05/15/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17156B164 | List: |
| References | |
| NUDOCS 8905250436 | |
| Download: ML17156B165 (12) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR PEACTOR REGULATION SUPPORTING AMENDMENT NO.
90 TO FACILITY OPERATING LICENSE NO. NPF-14 PENNSYLVANIA POWER AND LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.
DOCKET NO. 50-387 SUSQUEHANNA STEAM ELECTRIC STATION UNIT I
1.0 INTRODUCTION
By letter dated February 2, 1989, the Pennsylvania Power and Light Company (the licensee) requested an amendment to Facility Operating License No. NPF-14 for the Susquehanna Steam Electric Station (SSES),
Unit 1.
The proposed amendment would support Cycle 5 operation of the SSES Unit I with 9x9 reload fuel supplied by Advanced Nuclear Fuels Corporation (ANF).
The SSES Unit I Cycle 5 (hereinafter referred to as S1C5) reload will consist of 228 new ANF 9x9 (ANF-4) fuel assemblies 340 once burned ANF 9x9 (XN-3) fuel assemblies, 292 twice burned ANf 8x8 (XN-2I fue1 assemb1ies, and 4 thrice burned ANF 8x8 (XN-I) fuel assemblies.
S1C5 will contain no General Electric Company (GE) fuel assemblies.
The new 9x9 bundles are comprised of 79 fuel rods and two water rods.
The new 9x9 ANF fuel has similar operating characteristics (thermal-hydraulic and nuclear) to the previously used ANF 9x9 reload fuel.
In support of the S1C5 reload the licensee submitted reports which sumarize the reload scope (Ref. 2), the plant transient analyses (Ref. 3),
and the design and safety analyses (Ref. 4).
2.0 EVALUATION 2.1 Fuel Mechanical Desi n
The S1C5 core reload will include 228 ANF 9x9 fuel bundles with the designation ANF-4.
These reload bundles contain 79 fuel rods and 2 water rods.
The 228 fuel bundles will have a bundle average enrichment of 3.33 weight percent uranium-235.
The fuel design and safety analysis are described in the SSES Unit I specific report PL-NF-89-001 (Ref. 2) and the generic mechanical design report XN-NF-85-67(P)(A), Revision 1 (Ref. 5).
The NRC has approved the latter report and issued a Safety Evaluation Report on July 23, 1986 (Ref. 6).
Table 2.1 of XN-NF-85-67(P)(A), Revision I gives the pertinent design data for ANF 9x9 fuel.
Neutronic values specific to the SIC5 reload are given in Table 4.1 of ANF-88-169 (Ref. 4).
The burnable poison rods contain 4.0 weight percent gadolinia blended with 3.27 weight percent uranium-235.
The ANF-4 fuel bundles are designed to fit into existing GE channel boxes.
The analyses for SIC5 support fuel bundle discharge exposures of 35,000 MWd/MTU for ANF 8x8 fuel and 8905250436 890515 PDR ADOCK 05000387 P
PDC 40,000 MWd/MTU for ANF 9x9 fuel.
The discharge exposures for these fuel types are based on the approved ANF topical report XN-NF-82-06(P)(A), Supplement 1,
Revision 2 (Ref. 7).
Based on our review of the information presented, we find the mechanical design of the ANF 9x9 fuel for the S1C5 reload to be acceptable.
2.2 Rod Pressure For the SlC5 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 RODEX2A computer code which has been reviewed and approved by the staff (Ref. 8).
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 SlC5 is expected to result in Bx8 and 9x9 ANF fuel bundle exposures less than the assembly maximum allowed burnups.
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 ANF 9x9 fuel design (Ref. 10).
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 and anticipated operational occurrences.
The results of an evaluation reported in the S1C5 reload analysis report were based on the approved RODEX2A code (Ref. 11).
The staff has concluded that the generic methodology for the ANF 9x9 fuel is acceptable for the S1C5 reload fuel.
2.5 Linear Heat Generation Rate (LHGR) Limit for ANF 9x9 Fuel A figure of LHGR limit ver sus planar exposure (MWd/MTU) for,the ANF 9x9 fuel is incorporated into the Susquehanna 1 Technical Specifications (Figure 3.2.4-2 attached to Reference 1).
This Figure was approved in Reference 6 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(P),
Revision 1 (Ref. 5).
Based on the results of the generic review, the staff finds the LHGR limits for the 9x9 fuel to be acceptable.
2.6 LOCA-Seismic Mechanical Res onse The licensee has discussed the mechanical response of the ANF 9x9 fuel assembly design during LOCA-seismic events in Appendix B of Reference 4.
The discussion includes a comparison of the physical and structural properties of the ANF 9x9 fuel and the GE Bx8 fuel.
The staff has reviewed this information in connection with a previous review (see Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No.
31 to Facility Operating License No. NPF-22 dated October 3, 1986).
The staff has confirmed that the physical and structural characteristics of the ANF and GE fuel assemblies are sufficiently similar so that the mechanical response to design LOCA-seismic events is essentially the same.
Based on the considerations discussed
- above, we conclude that the original analysis is applicable to Susquehanna 1 and the analysis indicating that the design limits are not exceeded is acceptable.
2.
The ANF nuclear design methodology for SlC5 is that presented in XN-NF-80-19(A),
Volume 1 and Volume 1 Supplements 1 and 2 (Ref. 12), which were reviewed and approved by the staff for generic application to BWR core reloads.
The S1C5 licensing analysis employed a new cross section-void history correlation different from that used in Reference 12.
The new correlation provides an increased accuracy in cross section determination at exposures greater than 30,000 MWd/MTU.
ANF has submitted a letter (Ref. 13) describing this new cross section-void history correlation to the NRC.
Use of the correlation by ANF has been approved by the NRC.
The beginning of cycle shutdown margin is calculated to be 1.82 percent delta-k/k, and the R factor is zero.
Thus the cycle minimum shutdown margin is well in excess of the 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 fuel assembly.
Based on ANF calculations of 9x9 fuel, an average lattice enrichment of less than 4.0 weight percent uranium-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'ry or flooded conditions.
Since the zone average enrichment of the new fuel is 3.44 weight percent uranium-235 and the maximum cold, uncontrolled, beginning-of-life k-infinity for the ANF fuel bundle enriched zones is 1.10349, the ANF calculations show that the staff's acceptance criterion is met for the new fuel storage vault under dry and flooded conditions.
ANF also performed analyses for 9x9 fuel stored in the spent fuel pool.
A maximum enriched zone of less than 4.0 weight percent uranium-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 ANF-4 9x9 fuel has a zone average enrichment of 3.44 weight percent uranium-235 and a maximum k-infinity of 1.2206 at peak reactivity exposure, the staff's acceptance criterion for spent fuel storage is met for the ANF-4 9x9 fuel.
Susquehanna will continue to use the ANF POWERPLEX core monitoring system to monitor core parameters.
The system has been in use for a number of cycles for both Susquehanna Unit 1 and Unit 2 and has provided acceptable monitoring and predictive results.
2.8 Thermal-H draulic Desi n
The review of the thermal-hydraulic aspects of the SlC5 reload consisted of the following:
(a) the compatibility of the ANF 9x9 and prior ANF 8x8 fuel assemblies, (b) the fuel cladding integrity safety limit, (c) the bypass flow characteristics, and (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 susceptib'ie to thermal-hydraulic instability.
2.8.1 H draulic Com atibilit Since a
BWR core is a series of parallel flow channels connected to a cottmon lower and upper plenum, the total pressure drop across the assemblies will 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 constituent fuel types in the SlC5 core have been determined in single phase flow tests of full scale assemblies.
Analyses of the effects of hydraulic compatibility on thermal margin have been considered in the S1C5 reload.
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 ANF 9x9 and ANF 8x8 fuel for Susquehanna Unit 1, and on the hydraulic equivalence of the ANF 9x9 and ANF 8x8 fuel
- designs, the staff concludes that the two ANF fuel types are hydraulically compatible.
2.8.2 Hinimum Critical Power Ratio Safet Limit The minimum critical power ratio (HCPR) safety limit for the S1C5 reload was determined by the licensee to be 1.06 for all fuel types.
The methodology for SlC5 is based on the ANF critical power methodology in XN-NF-524, Revision 1
(Ref. 14), which has been approved by the staff (Ref. 15).
The XN-3 correlation used to develop the MCPR safety limit has been approved for the ANF 9x9 fuel (Ref. 16).
The methodology of XN-NF-524, Revision 1 was applied generically for S1C5 and is considered applicable to both ANF 9x9 and ANF 8x8 fuel types.
The staff has verified through its review of the S1C5 transient analysis report (Ref. 3) that the methodology for determining uncertainties and the application in determining the NCPR safety limit is in accordance with NRC approved methodology and is acceptable.
The core bypass flow fraction has been calculated to be 9.9 percent of total core flow using the approved methodology described in XN-NF-524(P)(A),
Revision l (Ref. 14).
This is used in the HCPR safety limit calculations and as input to the S1C5 transient analyses and is acceptable.
2.8.4 Thermal-H draulic Stabilit The licensee's response to the recently issued NRCB 88-07, Supplement 1 (Ref.
- 17) is under development.
The licensee has not completed preparation of revised Technical Specifications to implement the interim recommendations of Reference 17.
The licensee is also an active participant in the BWR Owners Group which is attempting to develop a final resolution of the stability issue.
The licensee performed stability analyses using the COTRAN code.
The ana'lysis predicted stable reactor operation at the boundary of the region defined by the area above and to the left of the 80 percent rod line and the 45 percent constant flow line.
Operation outside or on the boundary of this region is supported by these COTRAN calculations and result in decay ratios of less than or equal to 0.75 as required by the NRC Safety Evaluation Report (SER) on COTRAN (Ref. 18).
This region is the same as previously specified for Susquehanna Unit 1.
The licensee performed stability tests in Susquehanna Unit 2 during initial startup of Cycles 2 and 3.
The test results for S2C2 (Ref.
19) show very low decay ratios with a core containing 324 ANF 9x9 fuel assemblies.
The ANF "ANNA" software was used to analyze APRM signals from the S2C3 startup and gave a "measured" decay ratio of about 0.37 at a core condition of 60 percent power and 46 percent flow.
The S2C3 core contains 556 ANF 9x9 fuel assemblies while the S1C5 core will contain 468 ANF 9x9 fuel assemblies.
Therefore, about the same stability behavior is expected for SlC5 as was obtained for S2C3.
Based on the above discussion, we conclude that the licensee has acceptably addressed the thermal-hydraulic stability issue for SlC5.
In addition, PPKL has completed and implemented the NRC requested actions specified in Supplement 1 to NRC Bulletin 88-07, "Power Oscillations in Boiling Reactors."
2.9 Transient and Accident Anal ses 2.9.1 0 eratio'nal Transients Various operational transients could reduce MCPR below the safety limit.
The most limiting transients have been analyzed to determine which event could potentially result in the largest reduction in the initial CPR, that is, the delta CPR.
The core wide transient which resulted in the largest delta CPR from a 104 percent power and a 100 percent flow condition is the generator load rejection without bypass event (LRWOB).
The delta CPR for this event is 0.27 for ANF 9x9 fuel, which is the most limiting fuel type.
The most limiting local transient, the control rod withdrawal error (RWE), was analyzed to support a rod block monitor (RBM) setpoint of 108 percent and resulted in a delta CPR of 0.27.
The LRWOB and the RWE events were the most limiting events for S1C5 at rated power and flow conditions.
At less than rated power, the feedwater controller failure (FWCF) event is limiting and a curve of MCPR versus power, which is 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 controller failure is limiting and MCPR Operating limits for manual flow control reduced flow operation for SIC5 based on the analysis of this event are provided as a
Technical Specification figure of MCPR versus core flow.
The calculations of the thermal margin were performed with approved methodology (Ref. 20) 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 MCPR operating limits with either of these systems inoperable.
This resulted in increased MCPR limits which are also proposed for SIC5.
These calculations follow standard procedures and operation within the proposed HCPR operating limits with either the main turbine bypass system inoperable or the end-of-cycle RPT inoperable is acceptable for SIC5.
Compliance with overpressurization criteria was demonstrated by analysis of the main steam isolation valve (MSIV) closure event assuming HSIV position switch scram failure and an HSIV closure time of 2.0 seconds.
Six safety-relief valves were assumed to be out-of-service.
Maximum pressure was 105 percent of vessel design pressure, well within the 110 percent criterion.
The calculation was done with approved methodology and the results are acceptable.
2.9.2 Postulated Accidents The LOCA analyses for the Susquehanna plants (Ref. 21) was performed for a full core of ANF 9x9 fuel and is applicable for the S1C5 residual and reload ANF fuel.
These analyses have covered an acceptable range of conditions, have been performed with approved methodology and the resulting Technical Specification HAPLHGR values for the ANF fuel remain acceptable.
The control rod drop accident (CRDA) was analyzed with approved ANF methodology (Ref. 12).
The maximum fuel rod enthalpy was 219 cal/gm, which is well below the design limit of 280 cal/gm, and less than 560 fuel rods exceed 170 cal/gm, which is less than the 770 rods assumed in the Susquehanna FSAR analysis.
To ensure compliance with the CRDA analysis assumptions, control rod sequencing below 20 percent core thermal power must comply with GE's banked position withdrawal sequencing constraints (Ref. 22).
The staff concludes that the analysis and results for the SIC5 CRDA are acceptable.
2.9.3 Sin le Loo 0 eration SLO)
Current Technical Specifications for Susquehanna Unit I permit plant operation with a single recirculation loop out-of-service for an extended period of time.
Analyses for S1C5 (Ref. 4) show that the HCPR Safety Limit must be increased by 0.01 because of the increased measurement uncertainties.
The pump seizure event is more severe under SLO than under two-loop operation, assuming pump seizure of the operating loop.
This is the limiting event over most of the power and flow operating region for SLO.
The delta CPR for the SLO pump seizure event was determined to 0.35.
Previous analyses reported by
the licensee (Refs.
23 and 24) have shown that other events which could be affected by SLO were non-limiting when analyzed under SLO conditions.
SLO for SIC5 must maintain the 80 percent recirculation pump speed restriction because of the previous GE vessel internal vibration analysis, as discussed in Reference 23.
One of the stabi lity tests performed for Susquehanna Unit 2 Cycle 2 was performed under SLO conditions.
The measured decay ratio was 0.30 with a standard deviation of 0.064 at core conditions of 55 percent power and 44 percent flow.
ANF analysis of this test with the computer code COTRAN gave a
calculated decay ratio of 0.29.
The licensee states that this data, the stability results presented in Reference 4,
and S1C5 Technical Specification stability surveillance requirements support SLO in S1C5.
The original GE SLO analysis required adjustment of the APRM scram, APRM rod block, and the Rod Block Monitor setpoints in SLO to bound the changes in the assumed recirculation drive flow to core flow relationship between two loop operation and SLO.
The GE analysis indicated that the two loop to single loop change is less than 7 percent drive flow for a given core flow.
The licensee's data for Susquehanna indicate that a value of 8.5 percent would bound differences between two loop operation and SLO.
This value of 8.5 percent will be incorporated into the Susquehanna Technical Specifications and is acceptable.
2.10 Technical S ecification Chan es The following Technical Specification (TS) changes have been proposed for operation of SIC5:
1.
S ecification 2.2.1 Reactor Protection S stem Instrumentation Set pints The change to this Specification is to Functional Unit 2.b of Table 2.2.1-1, where a footnote has been added to the allowab'1e value.
This change is an editorial change and is, therefore, acceptable.
2.
Basis 2.0 Introduction Editorial changes are made to this Basis.
The references to GE and EXXON fuel are deleted and replaced with a reference to ANF fuel.
SIC5 will not contain GE fuel and ANF has replaced EXXON as a corporate entity.
Revision 1 is appended after report XN-NF-524(A).
These changes are consistent with the proposed changes and are, therefore, acceptable.
3.
Basis 2.1.1 Thermal Power Low Pressure or Low Flow The reference to GE fuel is deleted since SIC5 will not contain GE fuel.
This is consistent with the proposed changes and is, therefore, acceptable.
4.
Basis 2.1.2 Thermal Power Hi h Pressure and Hi h Flow This cha'nge involves adding Revision I to report XN-NF-524(A).
This is an editorial change and is, therefore, acceptable.
5.
S ecification 3/4.2.1 Avera e Planar Linear Heat Generation Rate The references to GE fuel and the MAPLHGR limits are deleted because S1C5 will not contain GE fuel.
This change is consistent with the proposed changes and is, therefore, acceptable.
6.
S ecification 3/4.2.2 APRM Set pints The reference to GE fuel is deleted because SIC5 will not contain any GE fuel.
This change is consistent with the proposed changes and is, therefore, acceptable.
5 7.
S ecification 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 criteria.
Therefore, the revised MCPRs are acceptable for SIC5 as discussed in this SER.
8.
S ecification 3/4.2.4 Linear Heat Generation Rate The limit and reference to GE fuel are being deleted because S1C5 will not contain any GE fuel.
These changes are consistent with the proposed changes and are, therefore, acceptable.
9.
S ecification 3/4.3.6 Control Rod Block Instrumentation This change corrects an administrative error by applying a footnote to trip functions la and 2a in Table 3.3.6-2.
The footnote refers to Specification 3/4.1.1.2.a for single loop operation.
This change is editorial and is, therefore, acceptable.
10.
S ecification 3.4.1.1.2 Sin le Loo 0 eration These changes are required to suppor t single loop operation (SLO) for S1C5.
Specifications 3.4.1.1.2a.2, 4 and 6 incorporate setpoint adjustments to account for the change in the previous GE value of 7 percent less drive flow for a given core flow, when going from two<<loop to SLO, to a new value of 8.5 percent determined by PP8L.
Specification 3.4.1.1.2.a.3 is revised to delete reference to GE fuel because S1C5 will not contain GE fuel.
New Specification 3.4.1.1.2.a.5 is needed to provide new MCPR limits for ANF fuel.
These changes are acceptable because they propose new limits for SLO for S1C5 based on approved methodologies and results.
ll.
Ba'sis 3/4.2.1 Avera e Planar Linear Heat Generation Rate The reference to GE fuel is deleted.
Other minor editorial changes are also made.
These changes are acceptable.
12.
Basis 3/4.2.2 APRM Set pints The reference to GE fuel is deleted.
This change is acceptable because S1C5 will not contain any GE fuel.
13.
Basis 3/4.2.3 Minimum Critical Power Ratio A reference to EXXON is changed to a reference to ANF.
This change is acceptable.
14.
Basis 3/4.4.1 Recirculation S stem The Basis is changed to reflect the MAPLHGR multiplier for ANF fuel for SLO.
The Basis is also changed by replacing the 7 percent for the decrease in recirculation drive flow for a given core flow by the new PPAL determined value of 8.5 percent.
These changes are acceptable because the analysis for SLO for SIC5 was performed using approved methodologies which gave acceptable results.
- 3. 0 ENVIRONMENTAL CONSIDERATION This amendment involves changes 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 and/or changes to the surveillance requirements.
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 Comission 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 (54 FR 13767) on April 5, 1989 and consulted with the State o7Dennsy van a.
No public comnents were received, and the State of Pennsylvania did not have any comments.
The staff has,concluded, based on the considerations discussed above, that:
(I) 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 comon defense and security nor to the health and safety of the pub'lic.
Principal Contributor:
Daniel Fieno Dated:
May 15, 1989
5.0 REFERENCES
Letter (PLA-3141) from H.W. Keiser (PP8L) to W.R. Butler (NRC), "Proposed Amendment 119 to License No. NPF-14:
Unit 1 Cycle 5 Reload," dated February 2, 1989.
2.
3.
5.
6.
7.
8.
9.
PL-NF-89-001, "Susquehanna SES.Unit 1-Cycle 5:
Reload Su@nary Report,"
dated January 1989.
ANF-88-168, "Susquehanna Unit 1 Cycle 5:
Plant Transient Analysis,"
dated January 1989.
ANF-88-169, "Susquehanna Unit 1 Cycle 5 Reload Analysis:
Design and Safety Analyses,"
dated January 1989.
XN-NF-85-67(P)(A), Revision 1, "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel," Exxon Nuclear Company, Inc., September 1986.
Letter from G.C. Lainas (NRC) to G.N. Ward (ENC), "Acceptance for Referencing of Licensing Topical Report XN-NF-85-67(P), Revision 1,
'Generic Mechanical Design Report for Exxon Nuclear Jet Pump BWR Reload Fuel,'" dated July 23, 1986.
XN-NF-82-06(P)(A), Supplement 1, Revision 2, "gualification of Exxon Nuclear Fuel for Extended Burnup - Supplement 1 Extended Burnup gualification of ENC 9x9 BWR Fuel," dated May 1988.
Letter from G.C. Lainas (NRC) to G.N. Ward (ANF), "Acceptance for Referencing of Licensing Topical Report XN-NF-85-74(P)," dated June 24, 1986.
Letter from G.N.
Wood (ANF) to G.C.'Lainas (NRC), "Additional Information on Rod Bowing," dated March 11, 1987.
10.
Letter from A.C. Thadani (NRC) to D.A. Adkisson (ANF), "Acceptance for Referencing of Licensing Topical Report XN-NF-82-06, Supplement 1,
Revision 2."
11.
Letter from G.C. Lainas (NRC) to G.N. Ward (ANF), "Acceptance for Referencing of Licensing Topical Report XN-NF-85-74(P),
'RODEX2A (BWR)
Fuel Rod Thermal-Mechanical Evaluation,'" dated June 24, 1986.
12.
XN-NF-80-19(A), Volume 1 and Volume 1 Supplements 1 and 2, "Exxon Nuclear Methodology for Boiling Water Reactors:
Neutronic Methods for Design and Analysis," dated March 1983.
13.
Letter (RAC:058:88) from R.A. Copeland (ANF) to M.W. Hodges (NRC), "Void History Correlation," dated September 13, 1988.
14.
XN-NF-524(A), Revision 1, "Exxon Nuclear Critical Power Methodology for BWRs," dated November 1983.
15.
Letter from C.O.
Thomas (NRC) to J.C. Chandler (ANF), "Acceptance for Referencing of Licensing Topical Report XN-NF-524(P)," dated October 31, 1983.
16.
Letter fz om C.O.
Thomas (NRC) to J.C.
Chandler (ANF), "Acceptance for Referencing of Licensing Topical Report XN-NF-734, 'Confirmation of the XN-3 Critical Power Correlation for 9x9 Fuel Assemblies,'"
dated February 1, 1985.
17.
NRCB-88-07, Supplement 1, "Power Osci llations in Boi ling Water Reactors,"
USNRC Bulletin, dated Oecember 30, 1988.
18.
XN-NF-691(P)(A) and Supplement 1, "Stability Evaluation of Boiling Water Reactor Cores - Sensitivity Analyses and Benchmark Analysis," dated August 1984.
19.
XN-NF-86-90, Supplement 1, "Susquehanna Unit 2 Cycle 2 Stability Test Results,"
dated January 1987.
20.
XN-NF-84-105(A), Volume 1 and Volume 1 Supplements 1 and 2, "XCOBRA-T:
A computer Code for BWR Transient Thermal-Hydraulic Core Analysis," dated February 1987.
21.
XN-NF-86-65, "Susquehanna LOCA-ECCS Analysis MAPLHGR Results for 9x9 Fuel," dated May 1986.
22.
NED0-21231, "Banked Position Withdrawal Sequence,"
- Company, dated January 1977.
23.
Letter (PLA-2885) from PPEL to NRC, "Proposed Amendment 52 to License No.
NPF-22," dated June 30, 1987.
24.
Letter (PLA-2935) from PPKL to NRC, "Additional Information on Proposed Amendment 52 to License No. NPF-22," dated October 30, 1987.
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