ML20215E076

Safety Evaluation Supporting Amend 31 to License NPF-22

ML20215E076
Person / Time
Site:	Susquehanna
Issue date:	10/03/1986
From:	Office of Nuclear Reactor Regulation
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S UNITED STATES 8

i NUCLEAR REGULATORY COMMISSION k

WASHINGTON, D. C. 20555 g,*****,/

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 31 TO FACILITY OPERATING LICENSE NO. NPF-22 PENNSYLVANIA POWER & LIGHT COMPANY SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 2 DOCKET NO. 50-388

1.0 INTRODUCTION

By letter dated June 19,1986, (Ref.1) Pennsylvania Power and Light Company (PPLCo or the licensee) proposed to amend Appendix A of Facility Operating License No. NPF-22. The requested amendment furnished information to sup-port authorization for Susquehanna 2 Cycle 2 operation with 9X9 fuel supplied by Exxon Nuclear Company, and revised single loop operation (SLO) provisions in the body of the Technical Specifications.

The Susquehanna 2 Cycle 2 (S2C2) reload will consist of 324 fuel bundles fabricated by Exxon Nuclear Company (ENC). These 9X9 bundles are comprised of 79 active fuel rods and two inert water rods. During Cycle 2' operation, the 9X9 fuel will reside with 440 General Electric P8x8R fuel assemblies presently in the core.

In support of the S2C2 reload PPLCo submitted tootcal reports which describe the design and safety analysis (Ref. 2), the plant transient analysis (Ref. 3), and the LOCA-ECCS analysis (Ref. 4) for the ENC 9X9 fuel. Additional information in response to NRC inquiries was provided by the licensee in References 5 and 15.

To evaluate the single loop operation (SLO) provisions in the Technical Specifications, PPLCo submitted a core stability assessment of ENC 9X9 fuel at Susquehanna 2 in Appendix A of XN-NF-86-60 (Ref. 2). However PPLCo is not requesting SLO approval at this time. The interim TS modifications are discussed in Section 3.3 of this Safety Evaluation.

2.0 EVALUATION OF FUEL DESIGN 2.1 Fuel Mechanical Design The S2C2 core reload will include 324 Exxon Nuclear Company (ENC) new 9X9 fuel assemblies with the designation XN-1.

These reload assemblies contain 79 fuel rods and two water rods. The 324 assemblies will have a bundle average enrichment of 3.31 percent. The fuel design and safety analysis for the 9X9 fuel are described in the Susquehanna 2 specific report XN-NF-86-60 (Ref. 2) and the generic mechanical design report XN-NF-85-67 Revision 1 (Ref. 6). The staff has approved the latter report and issued an SER on July 23, 1986 (Ref. 7).

Table 2.1 of XN-NF-85-67 Revision 1 gives the pertinent data for the XN-1 9X9 fuel. Neutronic values specific to the S2C2 reload are given in Table 4.1 of XN-NF-86-60 (Ref.2). The burnable poison rods contain 4.00 weight 8610100019 861003 PDR ADOCK OD0003PO P

PDR

i

' percent Cd 0 blended with 3.27 weight percent U-235 to reduce the initial.

reactivity. 3The ENC XN-1 fuel is designed to fit into the existing channel 2

boxes. A more detailed description can be found in Table 2.1 of XN-NF-85-67.

Based on our review of the information in Table 2.1, we find the mechanical design of the Exxon 9X9 fuel for the 52C2 reload is acceptable. However, approval of extended exposure limits for future operating cycles is con-t tingent on our approval of XN-NF-82-06(P) Supplement 1 (Ref. 8).

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2.2 Rod Pressure For the 52C2 ENC 9X9 reload fuel, calculation of the fuel rod internal pres-1 sure was done in accordance with acceptance criteria cited by ENC in XN-NF-85-67, Revision 1 (Ref. 6). The evaluation was performed with RODEX 2A which is a revision of the RODEX2 code (revised fission gas release model) used in the analysis of previous ENC fuel designs. Our review of the RODEX 2A topical report is complete and the staff Safety Evaluation Report has been a

issued (Ref. 9). The staff has concluded that the acceptance criteria for l

rod internal pressure can be fully met throughout the entire expected irradia-tior life of the 9X9 fuel, f

2.3 Fuel Rod Bow Our review of XN-NF-85-67, Revision 1 (Ref. 6) has been completed (Ref. 7) so that we may conclude that Exxon has demonstrated conformance to approved rod bow design limits for minimum gap spacing to a fuel assembly average exposure of 23,000 MWD /MTU for the 9x9 fuel. Projected peak assembly burnups l

for the S2C2 reload are in the range of 11,000-13,000 MWD /MTU for the 9x9 fuel.

We find the 52C2 core acceptable with respect to rod bow considerations.

r-However, since the rod bow criteria are only supported for two cycles of operation, additional justification with regard to fuel rod bowing must be provided for Susquehanna 2 operation beyond an average burnup of 23,000 MWD /MTU for the 9x9 fuel.

2.4 Fuel Centerline Melting l

The design basis for the ENC fuel centerline temperature is that no fuel centerline melting should result from normal operation including transient occurrences. The results of an evaluation reported in the S2C2 reload i

analysis report XN-NF-86-60 (Ref. 2) were based on R00EX2A. RODEX2A has been reviewed and approved (Ref. 9) and the staff has concluded that the generic methodology for the ENC 9X9 fuel is acceptable for the 52C2 reload fuel.

2.5 Cladding Swelling and Rupture 4

i The cladding swelling and rupture models in XN-NF-82-07 (Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model) have been approved for use j

in the ENC ECCS Evaluation Model and have been incorporated in the approved l

ENC EXEM/BWR ECCS model. This model was used in the ECCS analysis for the 52C2. The staff has verified that ENC is using the approved model for the 9X9 fuel ECCS analysis, and we find the application to be acceptable.

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2.6 Linear Heat Generation Rate (LHGR)

Limit for tm; 9K9 Fuel Pennsylvania Power and Light Company has provided a figure of Linear Heat Generation Rate Limit vs Planar Exposure for the ENC 9X9 fuel type to be incorporated in the Susquehanna 2 Technical Specifications (Ref. 4). This figure was approved in Reference 7 to reflect the design values which have been previously reviewed and approved for the ENC 9X9 fuel in connection with our review of XN-NF-85-67, Revision 1 (Ref. 6).

2.7 LOCA-Seismic Mechanical Response The licensee has discussed the mechanical response of the ENC 9x9 fuel assembly design during LOCA-seismic events in Appendix B of Reference 2.

The discussion included a comparison of the physical and structural prop-erties of the new 9x9 fuel and the prior GE 8x8 fuel and a reference to an ENC Topical Report XN-NF-84-97 (Ref. 10). The staff SER on Reference 10 has been issued (Ref.11); the conclusion in the SER stated that confomance to the acceptance criteria of Standard Re/iew Plan Section 4.2, Appendix A can be demonstrated by referencing XN-NF-84-97 (P) and submitting justification that the analyses in the topical report bounds the particular application under review. However, since an analysis specific to Susquehanna 2 has not been performed by ENC, the licensee has chosen to perform comparisons between the ENC 9x9 assembly and the GE 8x8 assembly currently loaded in the Susquehanna Unit 2 reactor to show that the results of the prior GE studies would still apply. The Seismic-LOCA analysis for the GE fuel has been vided in connection with a prior review on Susquehanna Unit 1 (Ref. 28) pro-The licensee has provided a comparison of the fuel assemblies in Table B1 of Reference 2.

Results of the comparison and data on the natural frequencies of the GE fuel were provided separately by the licensee in Reference 29, to support.the licensing of reload cores for PPLCo owned reactors only. Included in Reference 29 was a discussion of the differences in the determination of j

the natural frequencies of the ENC and GE fuel types. The comparisons are j

for ENC 8x8 and 9x9 and GE 8x8 fuel types. The staff has confimed that the physical and structural characteristics of the ENC 9x9 and GE 8x8 fuel assemblies are sufficiently similar so that the mechanical response to design Seismic-LOCA events is essentially the same. Based on the considera-tions discussed above, we conclude that the original analysis is still i

applicable to Susquehanna 2 and the analysis indicating that the design I

limits are not exceeded is acceptable.

l 3.0 THERMAL HYDRAULIC DESIGN s

l The review of the thermal-hydraulic aspects of the S2C2 reload consisted (a) the compatibility of the ENC 9X9 and prior GE 8X8 of the following): the fuel cladding integrity safety limit; (c) the oper-i fuel bundles; (b I

l ating limit minimum critical power ratio (OLMCPR); (d) the amount of

bypass flow associated with the different fuel designs; (e) thermal-hydraulic 4

stability, and (f) the proposed technical specifications.

The objective of the review was to confim that the thermal-hydraulic design j

of the reload core was ac::omplished using acceptable analytical methods, i

provides an acceptable margin of safety from conditions which would lead to fuel damage during nomal operation and anticipated operational occurrences and ensures that the core is not susceptible to themal-hydraulic instability.

3.1 Hydraulic Compatibility 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 bundles will be equal. However, differences in the hydraulic resistances of the fuel designs may cause variations in axial pressure drop profiles across the bundles.

4 Component hydraulic resistances for the proposed constituent fuel types in the S2C2 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 S2C2 reload report (Ref. 2). The 4

results of these analyses showed that the 9X9 hydraulic performance is equivalent to the GE 8X8 fuel. Based on our review of the information pro-vided in the Cycle 2 reload report we conclude that the GE and ENC fuel types are hydraulically compatible.

3.2 Thermal-Hydraulic Stability The thermal-hydraulic stability of the Susquehanna 2 core was analyzed using the methods identified in References 18 and 19. Reference 19 describes the i

use of the COTRAN model for use in the analysis of core thermal-hydraulic stability. The NRC has concluded that the use of COTRAN is acceptable in i

accordance with the restrictions cited in the applicable SER (Ref. 22).

For S2C2 ooeration, the licensee has provided additional stability analyses I

for the ENC 9X9 fuel using ENC's advanced system stability model COTRANSA 2 documented in XN-NF-84-67(P) (Ref. 23), which is under review by the staff.

The results of these analyses and comparison with results from the approved

)

COTRAN code are provided in Reference 5.

NRC Generic Letter 86-02 (Ref. 20) l provided acceptance criteria to be applied to all core reloads and other j

design or operating modifications relating to thermal-hydraulic stability.

An acceptable margin for ENC analysis of stability is a decay ratio of 0.75, which is a result of the estimated uncertainty of 25 percent in the calcula-l tion of the thermal-hydraulic stability decay ratio with the COTRAN code.

Permanent approval of the COTRANSA 2 analytical methodology and results for Susquehanna 2 stability analyses is subject to benchmark tests to demonstrate that COTRANSA2 can adequately predict the decay ratio for reactor cores with Exxon 9x9 fuel as it approaches the limit value of 1.0.

The licensee has i

comitted to a stability test during startup of Susquehanna 2, with post test analysis perfomed by Exxon Nuclear Company using the current analytical methodology.

i

The stability tests are to be performed in conjunction with a cooperative program between the NRC, Oak Ridge National Laboratory (ORNL), Pennsylvania Power and Light Company and Exxon Nuclear Company (ENC). The proposed test program involves the collection of neutron noise measurements at pertinent operating states followed by independent calculations by ORNL using the LAPUR computer code and ENC using the COTRAN and COTRANSA2 stability method-ology. The stability test proposal was evaluated by ORNL and is discussed in Reference 24; the evaluation included a review of the relevant documenta-tion and a sumary of two meetings between the parties involved. As a result of the discussions, it was agreed that noise level data would be collected at two measurement points on the Power / Flow map for the Susquehanna 2 reactor. The basis for the selection of the two points was the identifica-tion of one point close to the baseline noise level used as a reference for all fuel reloads and a second point within the " detect and suppress" region as defined by the General Electric Service Information Letter 380, Revision 1(Ref.25). Measurements collected at the second point will be made during single loop operation (SLO) tests instead of startup. The NRC was involved in the detennination of the proposed data points, and we concur in their selection.

ENC has provided the results of decay ratio computations using COTRAN and COTRANSA2 in Reference 27. The computations were made for operating points comparable to the points selected for the test measurements; i.e., points within or at the boundaries of the detect and suppress region established by Reference 25 (GE SIL-380, Revision 1). Additional calculational results were provided in Reference 27 for benchmark tests perfonned at Peach Bottom Atomic Power Station. The results of calculations with COTRAN, COTRANSA2 (original version) and COTRANSA2 (current version) were compared by ENC in Reference 27. The conclusion drawn by ENC was that the refinements incorp-orated in the version of COTRANSA2 presently under staff review do not 1

l alter the basic methodology and the current version of COTRANSA2 is accept-able for stability calculations. For the purpose of judging the adequacy of the proposed Susquehanna 2 stability test program, the staff has reviewed the reference data and comparative calculations with COTRAN and COTRANSA2.

Since the scope of the comparisons includes the test range and encompasses the detect and suppress region of the Power / Flow map, we conclude that the i

commitment by PPLCo to the proposed stability test program is acceptable.

Based on the favorable predictions (0.59 using COTRAN and 0.70 using current ENC stability methodology), we also conclude that the proposed one-third core reload with the ENC 9x9 fuel in S2C? is acceptable. However, we require that additional analysis and evaluation of the test results be performed for subsequent Susquehanna 2 reloads up to and including a full core loading 4

with 9x9 fuel.

I As part of its review of the relevant documentation, ORNL considered the l

reference data and discussion provided by ENC in Reference 26.

In Reference 24, ORNL had two coments on the parameters affecting thermal-hydraulic stability. These coments are noted here for the record. The first coment 4

dealt with the effect of burnup on stability. ENC drew the conclusion that I

the results of stability demonstration tests at two KRB-II reactors in Germany showed that the effect of the 9x9 fuel assemblies is a trend toward more stable conditions.

In drawing the conclusion, the effect of burnup should be given more consideration. The second comment dealt with the need to differentiate between fuel time constant and attenuation factor in their effect on stability. These connents do not change the conclusions in this Safety Evaluation.

4 3.3 Single loop Operation The Pennsylvania Power and Light Company presently has Technical Specifica-tions to pemit extended reactor operation of Susquehanna Unit 2 with one recirculation loop out of service. Prior staff approval was based on a Company (GE) peration (SLO) analysis performed by the General Electric single loop o to determine SLO operating limits with GE fuel in the core.

The introduction of the ENC 9x9 reload fuel requires a separate analysis for S2C2. At this time, sufficient analysis has not been completed to support extended SLO of Susquehanna 2 with ENC 9x9 fuel. As an interim measure, the licensee has proposed a modification to the present Technical i

Specification Limiting Condition of Operation (LCO) for the SLO mode. The 1

proposed change consists of setting the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit multiplier to 0.0 for extended SLO. The 4

effect of this change is to preclude SLO for more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The staff finds this proposal acceptable. Revised analyses with current approved methodology applicable to the 9X9 fuel are to be provided in a future sub-mittal and should include a specific analysis of the one-pump seizure acci-dent.

l 4.0 TRANSIENT AND ACCIDENT ANALYSIS 4.1 Minimum and Operating Limit Critical Power Ratio The minimum critical power ratio (MCPR) safety limit for the Cycle 2 reload i

I was determined by the licensee to be 1.06 for all fuel types. A safety i

limit of 1.06 for GE fuel types was approved for the previous Susquehanna 2 l

operating cycle. The methodology for Cycle 2 is based on ENC's revised critical power methodology in XN-NF-524. Revisinn 1 (Ref. 12) which incor-porates a constant flow MCPR fomulation for BWR applications. The staff has completed its generic review of XN-NF-524 (Ref.13) and has concluded that the methodology for arriving at an MCPR safety limit is acceptable.

The XN-3 correlation used to develop the MCPR safety limit has been approved for the new 9X9 fuel type (Ref. 14). The methodology of XN-NF-524 Revision 1 was applied generically for the upcoming Cycle 2 and is considered applic-able to the resident GE 8x8 fuel types as well as the ENC fuel. The staff has verified through its review of the S2C2 transient analysis report XN-NF-86-55 (Ref. 3) that the methodology for determining uncertainties and the application in determining the MCPR safety limit is in accordance with l

NRC approved methodology and is acceptable.

l

1 1

4.2 Operational Transients i

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 critical power ratio. The transients which resulted in the largest delta-CPR are the Load Rejection Without Bypass and Feedwater Con-troller Failure. The results of an updated anal sis using current i

methodology are provided in Reference 15. The staff review is discussed j

below.

The original Susquehanna 2 proposed Amendment 39 contained analysis results for the core-wide transients Load Rejection Without Bypass and Feedwater Controller Failure which were based on methodology described in XN-NF-79-71(A)

(Ref. 16). This analysis was revised using an updated methodology based on XCOBRA-T (Ref. 17) and documented in Reference 15 for S2C2.

(The XCOBRA-T Topical Report is currently under NRC review and the staff SER is in final processing.) The evaluation for S2C2 is based on the information provided by the licensee in Reference 15. This information includes the calculated delta-CPRs for the overpressurization transients and the MCPR operating limits for 52C2 operation. The staff review of the XCOBRA-T methodology has been completed to the point that we may conclude the approach to the i

calculation of delta-CPRs is acceptable. The calculated delta-CPRs for the Feedwater Controller Failure and Load Rejection Without Bypass are both equal to 0.24. The resulting MCPR operating limit of 1.30 is acceptable for incorporation into the S2C2 Technical Specifications for all fuel types.

l 4.3 Reactivity Insertion Transients The control rod withdrawal error, the fuel loading error and the rod drop accident were evaluated for Cycle 2.

The licensee used methods described i

in XN-NF-80-19. Volume 4 (Ref. 21 with staff SER included). The use of the Single Sequence Control strategy (in which rods inserted during power opera-tion have low worth) assures that the control rod withdrawal error will not be limiting. Using a Rod Block Monitor setting of 108 percent of full power results in a delta-CPR of 0.21 for the control rod withdrawal error i

transient for 9X9 fuel. The change in CPR due to a fuel loading error is i

0.19. These values are comparable to previous reloads and are not limiting.

The control rod drop accident evaluation yields a value of 109 cal /gm for the.naximum deposited fuel enthalpy. This is well below the staff's l

criterion of 280 cal /gm, and is therefore acceptable, j

5.0 LOSS OF COOLANT ANALYSIS (MAplHGR LIMIT) 1 The MAPLHGR limits for the GE 8X8 fuel as given in the Susquehanna 2 plant i

Technical Specification remain applicable for Cycle 2.

The licensee has proposed additional MAPLHGR limits for the ENC 9X9 fuel based on the l

analysis results provided in XN-NF-86-65 (Ref. 4). The limiting LOCA break l

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'e calculations were performed for the Susquehanna 2 reactor with a full core of ENC 9X9 fuel. The approved EXEM/BWR ECCS Evaluation Model codes were used for the LOCA calculations with array dimensions increased to accomodate the 9X9 array. The resulting Peak Cladding Temperature (PCT) was 2147* F at a burnup of 20 GWD/MTU allowing a 53* F margin to the 10 CFR 50.46 limit.

Metal-water reaction also peaks at 5.14 percent at a burnup of 20 GWD/MTU remaining well below the 17 percent limit required by 10 CFR 50.46. The MAPLHGR limits from this analysis are proposed for the Susquehanna i Tech-nical Specifications for the ENC 9X9 fuel design. Since analysis of the LOCA was perfonned with reviewed and accepted codes, and the results are well within the limits of 10 CFR 50.46, the staff finds the proposed MAPLHGR limits for 52C2 acceptable.

6.0 TECHNICAL SPECIFICATION CHANGES The Technical Specification Changes for S2C2 involve three general areas and are summarized below:

(1)

Incorporation of Linear Heat Generation Rate (LHGR) limits for ENC 9X9 fuel as a Limiting Condition for Operation (LCO).

The additional information on LHGR limits discussed in Section 2.6 of this SER is provided in the addition of Figures 3.2.2-1 and 3.2.4.2-1 in the Susquehanna 2 Technical Specifications and the identification of the LCO TH TS Section 3.2.4.2 (page 3/4-10a).

(2) Addition of ENC 9X9 fuel type.

MAPLHGR values for the new ENC 9X9 fuel type were added and burnup limits were adjusted for the earlier GE fuel types. MCPR safety limits were added for the new fuel and retained for the previous 8X8 4

fuel types. Additional MCPR operating limits were specified for Manual Flow Control and Automatic Flow Control for all fuel types.

i l

The previous TS Figures 3.2.3-1 and 3.2.3-2 were updated to reflect the new information.

l (3) RestrictiononSingleLoopOperation(SLO) Provisions j.

The Limiting Conditions of Operation for Single Loop Operation in TS Section 3.4.1.1.2 (page 3/4 4-Ic) were revised to set the MAPLHGR limit multiplier equal to 0.0 and to delete the RBM/APRM Control Rod Block Setpoints for one loop operation.

f Administrative changes were also made to relevant definitions and core design information to reflect the addition of the new 9x9 fuel, 7.0 BASES FOR CONCLUSIONS i

We have reviewed the infonnation furnished by Pennsylvania Power and Light in References 1, 5, 15, and 29 and Supplementary ENC reports (Ref. 2, 3 and 4) i l

I

b relative to the proposed License Amendment to allow operation'of Cycle 2 of Susquehanna 2.

Based on the results of our review, we find that suf-ficient basis has been provided to allow the addition of 324 ENC 9X9 fuel bundles in the Susquehanna 2 core and interim restrictions on operation in the single loop operation (SLO) mode. The proposed TS changes are therefore approved for $2C2.

Our review as discussed in the Evaluation Sections above has identified certain restrictions relating to our incomplete. review of the ENC 9x9 fuel and thermal-hydraulic stability considerations which limit approval to the upcoming Cycle 2 only. Specifically:

i (1) SE Section 2.0: Approval of extended exposure limits for the ENC 9x9 fuel for future operating cycles is contingent upon our approval of XN-NF-82-06(P), Supplement 1 (Ref. 2).

In addition, justification l

with regard to rod bow is required for operation beyond the projected S2C2 exposure levels.

(2) SE Section 3.2: The staff will reevaluate the thermal-hydraulic stability (THS) for Susquehanna 2 at the next reload cycle. The evalu-ation will consider permanent approval of the COTRANSA 2 analytical methodology including the results of benchmarking tests in the high decay ratio area.

8.0 ENVIRONMENTAL CONSIDERATION

This amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes to the surveillance requirements. The staff has detennined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released i

offsite.and that there is no significant increase in individual or cumula-tive 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.

9.0 CONCLUSION

The Commission made a proposed determination that the amendment involves no significant hazards consideration which was published in the Federal Register l

(51FR29009)onAugust 13, 1986.

It should be noted that additional infor-mation for the purpose of clarification was provided to the staff after noticing of the proposed amendment. The staff consulted with the state of i

Pennsylvania. No public comments were received, and the state of Pennsylvania did not have any coments.

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The staff has concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the will not be endangered by operation in the proposed manner, and (2) public 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 Contributors: Mari-Josette Campagnone, BWD-3, DBL Larry Phillips, RSB, DBL Mike McCoy, RSB, DBL Dated:

October 3,1986 i

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6 REFERENCES 1.

Letter, B. D. Kenyon (PPLCo) to Director (ONRR), Susquehanna Steam Electric Station Proposed Amendment 39 to License No. NPF-22, dated June 19, 1986 (withattachments).

2.

XN-NF-86-60, "Susquehanna Unit 2 Cycle 2 Reload Analysis", Exxon Nuclear Company, May 1986.

3.

XN-NF-86-55, "Susquehanna Unit 2 Cycle 2 Plant Transient Analysis", Exxon Nuclear Company, May 1986.

4.

XN-NF-86-65, "Susquehanna LOCA-ECCS Analysis MAPLHGR Results for 9X9 Fuel".

Exxon Nuclear Company, May 1986.

5.

Lettsr B. D. Kenyon (PPLC) to Director (NRR), "Susquehanna Steam Electric Station - Response to Request for Additional Information Regarding 9X9 Fuel Stability", dated July 10, 1986.

6.

XN-NF-85-67, Revision 1, Generic Mechanical Design Report for Exxon Nuclear Jet Pump BWR Reload Fuel, April 1986.

7.

Letter 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 for Exxon Nuclear Jet Pump BWR Reload Fuel", dated July 23, 1986.

8.

XN-NF-82-06(P) " Qualification of Exxon Nuclear Fuel for Extended Burnup",

i March 2, 1982, and Supplements 1, 2, 4 and 5.

9.

Letter, G. C. Lainas (NRC) to G. N. Ward (ENC), Acceptance for Referencing of Licensing Topical Report XN-NF-85-74(P), "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model", dated June 24, 1986.

i

10. XN-NF-84-97(P), "LOCA-Seismic Structural Response of an ENC 9X9 BWR Jet i

Pump Fuel Assembly", dated January 3, 1985,

11. Letter, G. C. Lainas (NRC) to G. N. Ward (ENC). Acceptance for Referencing of Licensing Topical Report XN-NF-84-97(P), "LOCA-Seismic Structural i

Response of an ENC 9X9 BWR Jet Pump Fuel Assembly", dated August 4, 1986.

12. XN-NF-524(P)(A), Revision 1, " Exxon Nuclear Critical Power Methodology for BWRs" November, 1983.

[

13. Letter, C. O. Thomas (NRC) to J. C. Chandler (ENC) October 31, 1983,

" Acceptance for Referencing of Licensing Topical Report XN-NF-524(P)".

14. Letter, C. O. Thomas (NRC) to J. C. Chandler (ENC), February 1, 1985,

" Acceptance for Referencing of Licensing Topical Report XN-NF-734, Confim-i l

ation of the XN-3 Critical Power Correlation for 9X9 Fuel Assemblies."

1

i.,

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15. Letter,B.D.Kenyon(SSES)toDirector,NRR(NRC)datedJuly 25, 1986,

" Supplement to Proposed Amendment 39 to License No. NPF-22".

16. XN-NF-79-71(A), Rev. 2. " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors" November 1981.

17. XN-NF-84-105(P)Rev.1, Supplements 1and2,"XCOBRA-T: A Computer Code for BWR Transient Thermal Hydraulic Core Analysis", March 1986.

18. XN-NF-80-19(A) Volume 1, Supplements 1 and 2 " Exxon Nuclear Methodology for Boiling Water Reactors - Neutronics Methods for Design and Analysis",

March 1983.

19. XN-NF-691(A), and Supplement 1, " Stability Evaluation of Boiling Water Reactor Cores", Exxon Nuclear Company, August 1984.

20. Technical Resolution of Generic Issue B Thennal Hydraulic Stability (GenericletterNo.86-02), January 23, 1986.

21. XN-NF-80-19(P)(A) Volume 4. " Exxon Nuclear Methodology for Boiling Waters:

Application of the ENC Methodology to BWR Reloads", dated September 1983.

22. Letter,G.Lainas(NRC) tog.N. Ward (ENC)datedApril 30, 1986,

" Acceptance for Referencing of Licensing Topical ~ Report XN-NF-80-19(P),

Volume 4, Revision 1. Exxon Nuclear Methodology for Boiling Water Reactors, Application of the ENC Methodology for BWR Reloads."

23. XN-NF-84-67(P), " Stability Evaluation Methodology for BWR Cores: The COTRANSA 2 Advanced BWR Stability Model and Application to Analysis of Anticipated Operation, Exxon Nuclear Company, Inc., June 1984,

24. Letter,*P. J. Otaduy (ORNL) to T. Huang (NRC) " Review of Susquehanna 2 Reload Stability Test Proposals" dated September 13, 1986.

25. General Electric Service Information letter No. 380, Revision 1 February 10 1984.

26. XN-NF-86-90(P), Rev.1, " Boiling Water Reactor 9x9 Fuel Operating Experience" dated July, 1986.

1

27. " Supplemental COTRAN and COTRANSA2 Stability Calculations " attachment to July 10, 1986 letter, B. D. Kenyon (PP&L) to E. Adensam (NRC).

28. PLA-1263, " Response to License Condition No. 14", Pennsylvania Power and Light Company letter dated August 27, 1982.

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29. Letter, H. W. Keiser (PP&L) to E. G. Adensam (NRC) " Response to NRC Question: Seismic /LOCA Analysis of U2C2 Reload" dated September 25, 1986.

30. Letter, H. W. Keiser (PP&L) to E. G. Adensam (NRC) " Revised Unit 2 Cycle 2 Stability Test Program" dated September 16, 1986.

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